Text

Initial Exam 2015-301 Draft Simulator Scenarios
	ML15119A483
Person / Time
Site:	Turkey Point
Issue date:	04/29/2015
From:	; Division of Reactor Safety II
To:	; Florida Power & Light Co
References
	Download: ML15119A483 (190)
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This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Facility:

Turkey Point Nuclear (PTN) - Units 3 and 4 Scenario No.:

1 Op Test No.:

2015-301 Examiners:

Operators:

(SRO)

(RCO)

(BOP)

Initial Conditions:

The plant is at 100% power (MOL). Online risk is green. B train is protected on both units. Maintain 100% power for the shift. Weather forecast is overcast skies with scattered pockets of severe rain.

Turnover:

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event No.

Malf. No.

Event Type*

Event Description 1

TFE2Z93S C-BOP & SRO 3D Feeder Breaker 3AB19 overcurrent and fails open - Loss of 3D 4KV Bus. (TS) 2 TVHHSGC C-RO & SRO 3C SG Tube Leakage. (TS) 3 TFS1MAML I-RO & SRO Turbine Inlet Pressure Channel, PT-3-447 fails low. (TS) 4 N/A R-RO & SRO N-BOP 3C SGTL Fast Load Reduction from 100% power.

5 TFF1M87M I-BOP & SRO 3B Feedwater Channel FT-3-487 fails as is. (TS) 6 TVH1TX45 TFHV456O I-RO & SRO Pressurizer Control Channel PT-3-445 fails high.

7 TVHHSGC TFSVV50A M-RO,BOP,

& SRO 3C SGTR starts and RV-3-1411, 3C Steamline Safety Valve fails open during the trip.

8 TFL3F50 TFL3F60 TFFVV980 C-BOP & SRO Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

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

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

L-15-1 NRC Scenario #1 The plant is at 100% power (MOL). Online risk is green. B train is protected on both units. Maintain 100% power for the shift. Weather forecast is overcast skies with scattered pockets of severe rain.

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A CWP is ready for start for a PMT.

Event 1 Shortly after taking the watch, the 3D Feeder Breaker 3AB19 faults and opens. The loss of the 3D 4KV Bus leads to a power loss to 3C CCW Pump and 3C ICW Pump. Loss of these pumps places the plant in a 14 day (ICW) and 30 day (CCW) TS LCO Action. The crew stabilizes the plant by starting an ICW Pump and ensuring CCW flow is adequate.

3-ARP-097.CR Annunciator response is used to stabilize the plant. 3-ONOP-004.5, Loss of 3D 4KV Bus, will be used for bus restoration.

Event 2 Once Event 1 crew actions are complete or enough time is allowed, then a 3C SG Tube leak occurs and ramps in over 5 minutes. The crew will enter 3-ONOP-071.2, Steam Generator Tube Leakage. They ensure the alarm is valid and estimate a leakage rate.

The leakrate exceeds Tech Spec limits for daily SG leakage. Once R-3-19, Steam Generator Blowdown Monitor, count rate is in alarm, then Steam Generator Blowdown Flow Control Valves (FCV-3-6278A/B/C) and Blowdown Tank to Canal Level Control Valve, LCV-3-6265B are closed.

Event 3 Once Event 2 has progressed, the crew will calculate and brief a downpower to respond to the tube leak. After the brief is complete prior to the downpower, then PT-3-447, Turbine Inlet Pressure Channel, fails low. This failure affects the Tref signal to Rod Control. Control Rods insert due to the Tavg and Tref mismatch. The crew uses diverse indications to ensure a runback is not in progress. Control rods are taken to manual to stop the movement. The crew enters 3-ONOP-028, Reactor Control System Malfunction, for follow-up after immediate action response, and then they use 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, to remove the channel from service.

2 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Event 4 The crew performs the downpower using the previous calculation and accommodates the response if a Tavg/Tref mismatch exists from the previous event. The fast load reduction is started with a boration of the primary and a noted RCS temperature decrease. The Unit Supervisor directs the BOP to unload the Turbine at the specified rate. The BOP sets up for the automatic load reduction by selecting desired MW target, reduction rate, and depresses GO to start the reduction.

Event 5 As Turbine load lowers, a slow failure occurs with auto feedwater control to 3B SG since FT-3-487 fails as is. With this failure, the actual flow to 3B SG actually lowers along with SG level. The BOP selects manual control for the 3B Feedwater Control Valve operation. Once the feedwater control channel is selected out, automatic control can be re-established. The US enters 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, to remove the channel from service. During this process, the US references Tech Specs for required actions.

Event 6 The crew demonstrates precise control during the Steam Generator Tube Leak load reduction. During this evolution, Pressurizer Control Channel PT-3-445 fails high. This channel causes PCV-3-456 to fail open. The RCO takes action to close the PORV manually which is unsuccessful. The RCO isolates the open PORV with its Block Valve, MOV-3-535. The US directs 3-ONOP-041.5, Pressurizer Pressure Control Malfunction, for response.

Event 7 After about 5 to 10% downpower, 3C SGTR starts on a 5 minute ramp. The crew takes actions to maximize Charging and to isolate Letdown. When the leakage exceeds the CVCS capacity, the US orders to trip the Reactor and Turbine, initiate Safety Injection and Phase A, and enters to 3-EOP-E-0, Reactor Trip or Safety Injection. The Reactor Trip triggers the RV-3-1411, 3C Steamline Safety Valve to fail open during the trip. The mitigation sequence is 3-EOP-E-0 to 3-EOP-E-3 to 3-EOP-ECA-3.1. The scenario is terminated once the crew has the opportunity to depressurize the RCS to refill the Pressurizer in 3-EOP-ECA-3.1 or at the Lead Evaluator's discretion.

Event 8 This post trip failure removes the auto isolation of feedwater flow to the 3C SG. The 3C SG Feedwater Control Valve FCV-3-498 fails open along with MOV-3-1409. MOV 1409 is manually closed to isolate this flowpath prior to the cooldown in 3-EOP-E-3.

3 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Critical Tasks:

CT1: Control SG Level After a failure of 3B Feedwater Channel FT-3-487, manually control 3B SG Water Level to prevent a Unit trip on low or high SG level.

Safety Significance: The failure to control 3B SG level demonstrates a lack of corrective actions which create a challenge to plant safety with an unintentional reactor protection system (RPS) actuation CT2: Isolate Faulted 3C SG After Unit 3 Reactor is tripped when a heat sink is established by total feedwater flow greater than 400 gpm or one narrow range SG level greater than 7%, manually isolate the faulted 3C SG including closing MOV-3-1409 3C SG Feedwater Isolation MOV prior to wide range SG level reaching 100%.

Safety Significance: The failure to isolate the faulted 3C SG challenges the Integrity CSF beyond the amount introduced by the postulated conditions. It also contributes to more off-site radiological release from the ruptured SG. This significant failure is the crews lack of ability to recognize a failure or an incorrect automatic actuation of an ESF system or component.

CT3: Control Initial RCS Cooldown When 3-EOP-E-3, Steam Generator Tube Rupture is entered, dump steam from 3A and 3B SGs at maximum rate using the Condenser Steam Dump Valves or Steam Dump To Atmosphere Valves to achieve Core Exit TCs less than required temperatures based on the lowest ruptured S/G pressure without causing a required transition to 3-EOP-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

Safety Significance: A SGTR mitigation strategy leading to a transition from 3-EOP-E-3 to a contingency procedure constitutes an incorrect performance requiring the crew to take additional compensatory actions that complicate the event mitigation strategy. With a SGTR, there exists a breach of the RCS fission-product and Containment barriers which allows radioactive RCS inventory to leak into the SG and associated piping. Without controlling the cooldown, the primary-to-secondary leakage is not stopped. This continued leakage results in a larger release of radioactivity to the environment affecting the safety of the public.

4 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112014)

Facility:

Turkey Point Nuclear (PTN) - Units 3 and 4 Scenario No.:

2 Op Test No.:

2015-301 Examiners:

Operators:

(SRO)

(RCO)

(BOP)

Initial Conditions:

The plant is at 50% power (MOL) following repairs to the 3B Steam Generator Feedwater Pump. Online risk is green. B train is protected on both units. Power escalation to 100% will begin next shift. Weather forecast is overcast skies with scattered pockets of severe rain.

Turnover:

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event No.

Malf. No.

Event Type*

Event Description 1

TVKD001X TFK3B11S C-BOP & SRO 3A TPCW Pump Seizes 2

TFCMSSF4 C-RO & SRO R-3-19 Failure - Isolate Steam Generator Blowdown 3

TFLII25S TFLII29S I-RO & SRO Loss of Two Individual Control Bank D Rod Positions (TS) 4 N/A R-RO & SRO N-BOP Shutdown the Unit from 50%

5 TFU7H1523B TAFV1523A TFC1523B C-BOP & SRO CV-3-1523A, MSR A Drain To HDT 3A, Fails As Is -

Dump Valve Operation Required 6

TVHP1ALK M-RO, BOP,

& SRO Pressurizer Steam Space Break From V-3-551A Pressurizer Safety Valve 7

TFL3SIA1 TFL3SIA2 TFQ634CF TFQ634DF C-RO & SRO Failure of U3 Auto SI and U4 HHSI Pump Start (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112014)

L-15-1 NRC Scenario #2 The plant is at 50% power (MOL) following repairs to the 3B Steam Generator Feedwater Pump. Online risk is green. B train is protected on both units. Power escalation to 100% will begin next shift. Weather forecast is overcast skies with scattered pockets of severe rain.

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event 1 The 3A TPCW Pump slowly seizes with a manual start of 3B TPCW Pump required. The crew refers to 3-ARP-097.CR for guidance. TPCW motor overload, pump trip, and low pressure alarms are received. The auto start signal has failed. 3B TPCW is manually started to establish flow to avoid further adverse conditions due to no turbine/generator cooling.

Event 2 Once Event 1 has progressed, a failure is inserted causing R-3-19 radiation monitor to fail. 3-ARP-097.CR is reference for the PRMS Channel Failure alarm. Since this is a failure of R-19 with a release in progress, Steam Generator Blowdown is isolated per H1/6 guidance. The crew recognizes isolation of Steam Generator Blowdown is a reactivity manipulation and discusses the effects of the manipulation.

Event 3 After Steam Generator Blowdown is isolated, a sequential failure of Control Bank D individual rod position (NARPI) occurs on two rods. With this failure, B7/1, NIS/RPI Rod Drop/Rod Stop, alarms. The crew enters the ARP response and verifies with diverse indications that no movement occurred. The rod bottom light for the rod is lit along with the NARPI indicating a failure. 3-ONOP-028.1, RCC misalignment, is entered to verify RCCs are aligned with the associated bank. Control Rods are placed in manual. Reactor Engineering confirms a RCC position indication malfunction exists. The crew transitions to 3-ONOP-028.2, RCC Position Indication Malfunction. TS 3.1.3.1 is referred to for this failure and LCO 3.0.3 is determined. Since two or more RPIs per bank are inoperable, then within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , a plant shutdown must commence to place the plant in MODE 3 and in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months MODE 4.

Event 4 Since two or more RPIs per bank are inoperable, then within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the crew commences a plant shutdown e to place the plant in MODE 3 and in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months MODE 4. The crew briefs the downpower and the RCO commences a boration to start lowering temperature. Once Tavg has decreased at least one degree, the BOP lowers Turbine load at an equivalent rate using the Turbine Control System (TCS).

2 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112014)

Event 5 After a 5 to 10% load decrease, the Moisture Separator Drain system experiences a perturbation in response to LT-3-1523B failing high and CV-3-1523A, MSR a Drain to HDT 3A, failing in a mid-position. In response to this failure, the operator has minutes to respond to the D3/1, MSR A Hi Level, alarm before a Turbine trip. The MSR A Overview screen is open on the plant computer (DCS) to select manual control and open the dump valve CV-3-1523B to re-establish level control.

Event 6 Next, V-3-551A Pressurizer Safety Valve fails open. The crew notes the lowering Pressurizer pressure and responds as time permits with maximum Charging. As pressure continues to lower, the US orders to trip the Reactor and perform 3-EOP-E-0, Reactor Trip or Safety Injection, immediate operator actions. The auto SI signal is blocked. The crew ensures SI actuation by manually pushing the SI Actuation pushbuttons, and then actuating Containment Isolation Phase A. The mitigation sequence is 3-EOP-E-0 to 3-EOP-E-1.The scenario is complete after the crew has terminated RHR Pump operation when evaluation of plant status for cold leg recirculation is determined.

Event 7 This post trip failure removes the Unit 3 auto SI actuation and the auto start of Unit 4 HHSI Pumps. The 3B HHSI Pump is OOS at the start of the scenario. 3-EOP-E-0 is performed to ensure equipment starts and alignments are proper for plant conditions. During this verification, the BOP ensures proper HHSI flow based on requirements.

The scenario is terminated prior to transitioning from 3-EOP-E-1 once the crew has terminated RHR recirculation flow or the time to terminate RHR recirculation flow has expired or at the Lead Evaluator's discretion.

3 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112014)

Critical Tasks:

CT1: When the automatic SI signal is failed, actuate SI manually with at least one train of SIS-actuated safeguards equipment prior to reaching a Red Path on Integrity.

Safety Significance: Failure to manually actuate SI under these postulated conditions constitutes the misoperation or incorrect crew performance. The crew does not prevent a degraded emergency core cooling system (ECCS) capacity, demonstrate the ability to direct/manipulate engineered safety feature (ESF) controls, or recognize an incorrect automatic actuation of an ESF system or component.

CT2: When a RHR Pump starts and is operating at shutoff head, limit the operating time at shutoff head with minimum flow recirculation to no more than 44 minutes.

Safety Significance: Failure to secure the RHR Pumps operating at shutoff head leads to pump overheating and adverse vibration which would constitutes mis-operation or incorrect crew performance in which the crew does not prevent a degradation of the emergency core cooling system (ECCS) capacity.

Note: Individual minimum flow recirculation lines are provided to allow enough flow (330 gpm) for pump operation for a limited time (44 minutes).

[0-ADM-232, Time Critical Operator Action Program-Attachment 1, page 9 of 24]

4 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Facility:

Turkey Point Nuclear (PTN) - Units 3 and 4 Scenario No.:

3 Op Test No.:

2015-301 Examiners:

Operators:

(SRO)

(RCO)

(BOP)

Initial Conditions:

The plant is at 75% power (MOL) following repairs to the 3B Steam Generator Feedwater Pump. Power escalation to 100% at 10%/hr will begin following turnover. Online risk is green. B train is protected on both units. Weather forecast is overcast skies with scattered pockets of severe rain.

Turnover:

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event No.

Malf. No.

Event Type*

Event Description 1

N/A R-RO & SRO N-BOP Raise Reactor Power to 100%.

2 TVKD001X I-RO & SRO N-44 fails high. (TS) 3 TVKB002X TFKCSMB C-BOP & SRO 3B ICW Pump seizes causing a shaft shear. (TS) 4 TVBMBFN C-RO & SRO Gradual loss of 3C Charging Pump flow.

5 TFSVV91A TFE2Z40S C-BOP & SRO 3C 4KV Bus trip initiates a fast Turbine runback.

Condenser Steam Dump CV-3-2827 fails open.

6 TVFALN2 TFL10101 TFL4AF TFL2XASE TFL2XBSE TFU1ASBF TFU1ASTF TCB1AMC TCB1BMC M-RO,BOP,

& SRO Feedwater Break Outside Containment with ATWS.

Main Turbine fails to trip. Auto rod insertion is blocked.

3A and 3B Boric Acid Pumps fail to run.

7 TCF5MTA TFFXC03 C-BOP & SRO A AFW Pump trips. MOV-3-1403 fails to auto open.

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

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

L-15-1 NRC Scenario #3 The plant is at 75% power (MOL) following repairs to the 3B Steam Generator Feedwater Pump. Power escalation to 100% at 10%/hr will begin following turnover.

Online risk is green. B train is protected on both units. Weather forecast is overcast skies with scattered pockets of severe rain.

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event 1 The RCO reviews the maneuvering guide and calculates the amount of water needed for a dilution to raise power to 100%. Either the BOP or US verifies the amount. The US or RCO hold a brief for the 25% power change. 3-GOP-301, Hot Standby to Power Operation, guidance is used for this maneuver. The crew performs the up-power using dilutions and rods as applicable to accommodate the response if a Tavg/Tref mismatch.

Event 2 After power is increased on the Turbine, N-44 Power Range NI fails high. This failure causes Control Rods to step in based on detected high power. The RCO uses diverse indications for the failure and stops inward rod motion due to the failed channel. 3-ONOP-028, Reactor Control System Malfunction, and/or 3-ONOP-059.8, Power Range Nuclear Instrumentation Malfunction, will be entered for initial response. 3-ONOP-059.8 is used to remove the NI from service. (TS 3.3.1)

Event 3 After response to Event 2, the 3B ICW Pump fails which requires an additional pump to be started. The US enters a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO Action for this condition. 3-ARP-097 gives appropriate direction to start the standby pump. (TS 3.7.3.b)

Event 4 After time for the review of 3B ICW Pump Technical Specifications, the 3C Charging Pump flow is gradually lost due to a pump failure. 3-ARP-097 and/or 3-ONOP-047.1 is entered to re-establish flow with either 3A or 3B Charging Pump.

Event 5 After the recovery of Charging, a protective relay trips the 3C Bus which leads to a fast Turbine Runback since 3B SGFP is lost. The crew enters 3-ONOP-089, Turbine Runback for response. As Turbine load lowers, Condenser Steam Dump CV-3-2827 fails open. The RCO will ensure Control Rod insertion. When Condenser Steam Dumps are disabled by taking their handswitch to OFF, then CV-3-2827 will close.

2 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Event 6 After the plant is stabilized from the runback, 3B Feedwater line breaks outside Containment combined with an ATWS. Control Rods fail to auto insert during the ATWS.

The RCO ensures negative reactivity is added by manually inserting Control Rods and emergency boration. The RCO attempts to use 3A and 3B Boric Acid Pumps, but they fail to run which leaves the RWST as the source of water for emergency boration. Also, the Main Turbine fails to auto trip. The BOP initiates a manual Turbine trip. After the Turbine trip A AFW Pump trips. The BOP manually opens MOV-3-1403 to re-initiate AFW flow. The mitigation sequence is from 3-EOP-E-0 to 3-EOP-FR-S.1 to 3-EOP-E-0.

The scenario is complete after the crew subsequently exits 3-EOP-E-0 after re-entry from 3-EOP-FR-S.1.

Event 7 This post trip failure prevents MOV-3-1403 from auto opening. When A AFW Pump trips, AFW flow is restored by manually opening 3A SG AFW Steam Supply to provide steam to the C AFW Pump.

Critical Tasks:

CT1: After an ATWS event and auto rod insertion is defeated, insert negative reactivity into the core by manually inserting Control Rods within one minute.

Safety Significance: The failure to insert negative reactivity, under the postulated plant conditions, results in a situation with the Reactor remaining critical when a protective signal is demanding subcriticality to provide sufficient shutdown margin protection to prevent any subsequent return to criticality. A failure to insert negative reactivity constitutes misoperation or incorrect crew performance associated with reactivity control.

[0-ADM-232, Attachment 2, Time Critical Operator Actions - PTN PSA Model -

page 4 of 5, NHFPMRODIN]

CT2: After an ATWS event without an auto Turbine trip, perform a manual trip of the Main Turbine by pushing the Manual Trip pushbutton or closing MSIVs within one minute.

Safety Significance: The failure to trip the Main Turbine or close MSIVs under the postulated plant conditions can lead to a violation of the RCS emergency stress limit. The analysis of the ERG Background Document for FR-S.1 demonstrates the importance of maintaining SG inventory and primary-to-secondary heat transfer. If the main turbine is not tripped, the SG U-tubes uncover sooner and faster. Thus, primary-to-secondary heat transfer starts to deteriorate earlier in the transient and deteriorates more rapidly. The result is that once tube uncovery 3

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114) begins, RCS temperature and pressure increase more rapidly and reach higher values.

[0-ADM-232, Attachment 2, Time Critical Operator Actions - PTN PSA Model -

page 2 of 5, FHFPTRBTRP]

CT3: After an ATWS event, manually align AFW Pumps by opening Auxiliary Feedwater Pump Steam Supply Valves to restore Auxiliary Feedwater Flow prior to the WR SG Levels reaching less than 9% on the non-faulted SGs.

Safety Significance: The failure to start at least the minimum required number of AFW Pumps under the postulated plant conditions can lead to a violation of the RCS emergency stress limit. The analyses presented in the ERG Background Document for FR-S.1 related to the RCS-overpressure ATWS events assume all AFW Pumps start at 60 seconds after the initiating event. In these reference cases, the RCS pressure does not exceed the RCS pressure emergency stress limit of 3,200 psia.

The failure to start at least the minimum required number of AFW Pumps when possible to do so leaves the plant in a condition outside the bounds of the ATWS analyses and can cause RCS pressure to exceed the emergency stress limit.

4 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Facility:

Turkey Point Nuclear (PTN) - Units 3 and 4 Scenario No.:

4 Op Test No.:

2015-301 Examiners:

Operators:

(SRO)

(RCO)

(BOP)

Initial Conditions:

The plant is at 25% power (MOL). Online risk is green. B train is protected on both units. Weather forecast is overcast skies with scattered pockets of severe rain.

Turnover:

Chemistry parameters are satisfied for increasing unit power. Raise Reactor Power to 35% power. There are no ramp rate fuel integrity restrictions for this power increase to 35%.

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event No.

Malf. No.

Event Type*

Event Description 1

TFB1PWT TFBVO19 R-RO & SRO N-BOP Raise power up 10% with dilution and rods and hold at 35%.

Makeup dilution valve CV-3-114A fails open with Makeup Counter failure.

2 TVF1MABD C-BOP & SRO 3B SG Level Transmitter LT-3-486 drifts low. (TS) 3 TFH244GH I-RO & SRO Pressurizer Spray Valve PCV-3-455A fails open. (TS) 4 TVKKB2 C-RO & SRO 3A RCP High Vibration 5

TVPGM03P TFP8SWYD TFE2Z50S TFQ5B51S TFQ5GBS M-RO,BOP,

& SRO A trip of 3A RCP causes a 4KV Bus fault leading to a bus lockout followed by a delayed loss of Switchyard and loss of 3B EDG.

6 TFP8D7BA TFP8D7MA C-BOP & SRO Mid and East Main Generator Output Breakers fail to open on Main Turbine trip.

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

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

L-15-1 NRC Scenario #4 The plant is at 25% power (BOL). Online risk is green. B train is protected on both units. Weather forecast is overcast skies with scattered pockets of severe rain.

Chemistry parameters are satisfied for increasing unit power. Raise Reactor Power to 35% power. There are no ramp rate fuel integrity restrictions for this power increase to 35%.

The following equipment is Out-Of-Service: 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

Event 1 The crew briefs the power increase. While the RO is performing a dilution, CV-3-114A fails open with the Makeup Counter failing. The dilution slightly increases with the audible noise also increasing. The RCS dilution continues to dilute the RCS until the RO notices the failure with diverse indications. The RO turns the RCS Makeup Control Switch to Off. The US enters 3-ARP-097.CR and/or 3-ONOP-046.4, Malfunction of Boron Concentration Control System.

Event 2 After the crew finishes with compensatory actions from Event 2, 3B Steam Generator Level Transmitter, LT-3-486 drifts low. As this slow failure occurs, 3B SG FW flow increases which causes a C6/2, SG B Level Deviation/Cntrl Trbl Alrm. The BOP takes manual control of 3B SG Level and restores to program level. 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, is reviewed for applicability.

PTN Technical Specifications are entered with 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to trip LT-3-486 bi-stables.

Event 3 After sufficient time for the crew to refer to Technical Specifications and regain control of SG Level, then the Pressurizer Spray Valve PCV-3-455A fails open. The immediate effect of this failure is a corresponding RCS pressure drop below TS DNB limits. The RO compensates for this auto failure by taking manual control of the Pressurizer Spray Valve. 3-ARP-097 and/or 3-ONOP-041.5, Pressurizer Pressure Control Malfunction are used for crew response.

Event 4 Once the crew has adequate time for the responding to the pressure control malfunction, 3A RCP develops an imbalance causing a high RCP vibration. The US enters 3-ONOP-041.1 to monitor the 3A RCP status. From indications provided, 3A RCP meets trip criteria. The US briefs the crew on tripping 3A RCP after the Reactor is verified trip. The RO trips the Reactor and performs immediate operator actions of 3-EOP-E-0, Reactor Trip or Safety Injection.

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Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Event 5 A trip of 3A RCP causes a 4KV Bus fault leading to a bus lockout followed by a delayed loss of Switchyard and loss of 3B EDG. On the trip, PORV PCV-3-455C fails open and manual closure is allowed. The crew starts with immediate operator actions of 3-EOP-E-0 and when power is verified lost, then they transition to 3-EOP-ECA-0.0. Power is restored to the 3B 4KV Bus after bus stripping through the SBO Crosstie using - 3B 4KV Bus Restoration. 3D 4KV Bus is aligned and Load Centers energized.

Event 6 After the Reactor is tripped, as the Main Turbine Generator Breakers do not auto open.

The BOP will carry out his actions to divorce the 4 KV Buses from the switchyard and manually open these breakers.

The scenario is complete after the crew has the opportunity to start an ICW Pump, CCW Pump, and Charging Pump/HHSI Pump for inventory control as applicable.

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Appendix D Scenario Outline Form ES-D-1 NRC Draft (Rev 112114)

Critical Tasks:

CT1: Control SG Level After a failure of 3B SG Level Transmitter LT-3-486, manually control 3B SG Water Level to prevent a Unit trip on low or high SG level.

Safety Significance: The failure to control 3B SG level demonstrates a lack of corrective actions which create a challenge to plant safety with an unintentional reactor protection system (RPS) actuation CT2: Re-energize 3B 4KV Bus After a loss of all AC Power and confirmation of an SBO condition, restore this units 4 KV Bus power through the SBO crosstie from the non-SBO unit EDG via the 4 KV Crosstie and the Swing Bus Switchgear within 10 minutes.

Safety Significance: The failure to energize an ac emergency bus constitutes a misoperation or incorrect crew performance in which the crew does not prevent a degraded emergency power capacity. The failure to perform this critical task also results in a needless degradation of any barrier to fission product release, specifically of the RCS barrier at the point of the RCP seals.

CT3: Manually Start An ICW Pump After Unit 3 power is restored to one 4 KV Bus, start at least one ICW Pump prior to allowing CCW Supply Header temperature to exceed 125°F.

Safety Significance: The failure to manually start at least the minimum required number of ICW (SW) Pumps in an operating safeguards train represents a failure by the crew to effectively direct or manipulate engineered safety feature (ESF) controls that would prevent a significant reduction of safety margin beyond that irreparably introduced by the scenario.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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References Revision Number 3-ARP-097.CR various 0-ADM-115 8B 0-ADM-211 3

0-ADM-232 4

0-EPIP-20101 16 0-EPIP-20134 3

3-EOP-E-0 10 3-EOP-F-0 3

3-EOP-FR-S.1 4

3-GOP-301 30 3-ONOP-004.5 03/25/10 3-ONOP-041.5 0A 3-ONOP-049.1 3A 3-ONOP-071.2 8

PTN Technical Specs 293

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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SIMULATOR EXERCISE GUIDE REQUIREMENTS Terminal Objective Given this simulator scenario and resources normally found in the Control Room, the operating crew will perform Control Room operations IAW approved plant procedures in order to maintain the integrity of the plant and the health and safety of the public.

Enabling Objectives:

Given this simulator scenario and resources normally found in the Control Room, operate in accordance with approved plant procedures, Operations Department Instructions, and management expectations:

1. (ALL) Demonstrate personnel SAFETY awareness in interactions with plant staff and outside agencies.

2. (ALL) Demonstrate ALARA awareness in interactions with plant staff and outside agencies.

3. (ALL) Exchange correct information using 3-point communication/Repeat-backs with Control Room personnel and other plant staff.

4. (ALL) Inform plant personnel and System of plant conditions, as needed.

5. (SM) Employ timely and concise crew briefs where appropriate.

6. (SM) Make timely and correct emergency plan classifications where appropriate.

7. (SM/COMM) Make timely and correct emergency plan protective action recommendations if General Emergency declared with SHIFT MANAGER and Communicator assigned.

8. (COMM) Make accurate and timely emergency plan notifications where appropriate.

9. (ALL) Maintain awareness of plant status and control board indication.

10. (ALL) Correctly diagnose plant situations.

11. (ALL) Solve operational problems as they arise.

12. (RCO/BOP) Manipulate plant controls properly and safely.

13. (ALL) Demonstrate self-checking using STAR and peer checks(when required)

14. (US) Demonstrate command and control of the crew.

15. (US) Coordinate the input of crew members and other plant staff.

16. (US) Utilize the input of crew members and other plant staff.

17. (ALL) Demonstrate conservative decision making.

18. (STA) Provide informed technical assistance to the operating crew.

19. (ALL) Demonstrate teamwork.

20. (ALL) Respond to plant events using procedural guidance (OPs/ONOPs/EOPs) as applicable in accordance with rules of usage.

21. (RCO/BOP) Implement any applicable procedural immediate operator actions without use of references.

22. (SRO) Maintain compliance with Tech Specs.

23. (ALL) Identify/enter applicable Tech Spec action statements.

24. (ALL) Respond to annunciators using ARPs (time permitting).

25. (ALL) Maintain written communication, logs, and documentation as needed to permit post-event reconstruction.

While addressing the following events:

1. 3D Feeder Breaker 3AB19 Overcurrent and Fails Open - Loss of 3D 4KV Bus

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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2. 3C SG Tube Leakage

3. Turbine Inlet Pressure Channel, PT-3-447 Fails Low

4. 3C SGTL Fast Load Reduction from 100% Power

5. 3B Feedwater Channel FT-3-487 Fails As Is

6. Pressurizer Control Channel PT-3-445 Fails High

7. 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip

8. Feedwater Valves, MOV-3-1409 and FCV-3-498, Fail to Isolate on a Safety Injection Prerequisites:

None Training Resources:

PTN Unit 3 Plant Simulator Development

References:

TR-AA-230,-1003, SAT Development TR-AA-230,-1007 Conduct of Simulator Training and Evaluation 0-ADM-232, Time Critical Action Program OP-AA-100,-1000 Conduct Of Operations OP-AA-103,-1000 Reactivity Management 0-ADM-200, Operations Management Manual 0-ADM-211, Emergency and Off-Normal Operating Procedure Usage WCAP-17711-NP, Pressurized Water Reactor Owners Group Westinghouse Emergency Response Guideline Revision 2-Based Critical Tasks Protected Content:

N/A Evaluation Method:

Performance Mode Operating Experience:

None Risk Significant Operator Actions:

1. After Unit 3 Reactor is tripped when a heat sink is established by total feedwater flow greater than 400 gpm or one narrow range SG level greater than 7%, manually isolate the ruptured/faulted 3C SG including closing MOV-3-1409 3C SG Feed-water Isolation MOV prior to wide range level reaching 100%.

2. When 3-EOP-E-3, Steam Generator Tube Rupture is entered, dump steam from 3A and 3B SGs at maximum rate using the Condenser Steam Dump Valves or Steam Dump To Atmosphere Valves to achieve Core Exit TCs less than required temperatures based on the lowest ruptured S/G pressure without causing a required transition to 3-EOP-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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TASKS ASSOCIATED WITH SIMULATOR EXERCISE GUIDE SRO Task #

Task Title 02200006300 INVESTIGATE AND CONTROL STEAM GENERATOR TUBE LEAK 02028026300 RESPOND TO CONTINUOUS ROD INSERTION 02089026300 AUTHORIZE FAST LOAD REDUCTION 02200044500 RESPOND TO HIGH STEAM GENERATOR LEVEL 02041013300 RESPOND TO A PORV MALFUNCTION 02028033500 AUTHORIZE UNIT TRIP 02200008500 RESPOND TO A STEAM GENERATOR TUBE RUPTURE 02200001500 RESPOND TO UNIT TRIP 02200022500 DIAGNOSE CAUSE OF SAFEGUARDS ACTUATION 02200002500 EVALUATE CRITICAL SAFETY FUNCTION (CSF) STATUS TREE OUTPUT 02200007500 RESPOND TO A STEAM LINE FAULT RO Task Task Title 01005030100 TRANSFER D 4KV BUS FROM B TO A 4KV 01200006300 INVESTIGATE AND CONTROL STEAM GENERATOR TUBE LEAK 01028026300 RESPOND TO CONTINUOUS ROD INSERTION 01089026300 RESPOND TO / ADJUST TURBINE DURING FAST LOAD REDUCTION 01200044500 RESPOND TO HIGH STEAM GENERATOR LEVEL 01041013100 RESPOND TO A PORV MALFUNCTION 01200001500 RESPOND TO UNIT TRIP 01200008500 RESPOND TO A STEAM GENERATOR TUBE RUPTURE 01200022500 DIAGNOSE CAUSE OF SAFEGUARDS ACTUATION 01200007500 RESPOND TO A STEAM LINE FAULT

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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UPDATE LOG:

NOTES:

Place this form with the working copies of lesson plans and/or other materials to document changes made between formal revisions. For fleet-wide training materials, keep electronic file of this form in same folder as approved materials. Refer to TR-AA-230-1003 SAT Development for specific directions regarding how and when this form shall be used.

Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

PREPARER DATE REVIEWER DATE 1-0 Initial Revision Revised for L-15-1 NRC Exam 01982463 5

5 5

5 1-1 1-2 1-3 1-4 1-5

1.

Individual updating lesson plan or training material shall complete the appropriate blocks on the Update Log.

2.

Describe the change to the lesson plan or training materials.

3.

State the reason for the change (e.g., reference has changed, typographical error, etc.)

4.

Preparer enters name/date on the Update Log and obtains Training Supervisor approval.

5.

Initial dates and site approval on cover page.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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OVERVIEW Initial Conditions The plant is at 100% power (MOL). Online risk is green. B train is protected on both units. Maintain 100% power for the shift. Weather forecast is overcast skies with scattered pockets of severe rain.

Equipment OOS 3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

SYNOPSIS Event 1 Shortly after taking the watch, the 3D Feeder Breaker 3AB19 faults and opens. The loss of the 3D 4KV Bus leads to a power loss to 3C CCW Pump and 3C ICW Pump. Loss of these pumps places the plant in a 14 day (ICW) and 30 day (CCW) TS LCO Action. The crew stabilizes the plant by starting an ICW Pump and ensuring CCW flow is adequate. 3-ARP-097.CR Annunciator response is used to stabilize the plant. 3-ONOP-004.5, Loss of 3D 4KV Bus, will be used for bus restoration.

Event 2 Once Event 1 crew actions are complete or enough time is allowed, then a 3C SG Tube leak occurs and ramps in over 5 minutes. The crew will enter 3-ONOP-071.2, Steam Generator Tube Leakage. They ensure the alarm is valid and estimate a leakage rate. The leakrate exceeds Tech Spec limits for daily SG leakage. Once R-3-19, Steam Generator Blowdown Monitor, count rate is in alarm, then Steam Generator Blowdown Flow Control Valves (FCV-3-6278A/B/C) and Blowdown Tank to Canal Level Control Valve, LCV-3-6265B are closed.

Event 3 Once Event 2 has progressed, the crew will calculate and brief a downpower to respond to the tube leak.

After the brief is complete prior to the downpower, then PT-3-447, Turbine Inlet Pressure Channel, fails low. This failure affects the Tref signal to Rod Control. Control Rods insert due to the Tavg and Tref mismatch. The crew uses diverse indications to ensure a runback is not in progress. Control rods are taken to manual to stop the movement. The crew enters 3-ONOP-028, Reactor Control System Malfunction, for follow-up after immediate action response, and then they use 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, to remove the channel from service.

Event 4 The crew performs the downpower using the previous calculation and accommodates the response if a Tavg/Tref mismatch exists from the previous event. The fast load reduction is started with a boration of the primary and a noted RCS temperature decrease. The Unit Supervisor directs the BOP to unload the Turbine at the specified rate. The BOP sets up for the automatic load reduction by selecting desired MW target, reduction rate, and depresses GO to start the reduction.

Event 5 As Turbine load lowers, a slow failure occurs with auto feedwater control to 3B SG since FT-3-487 fails as is. With this failure, the actual flow to 3B SG actually lowers along with SG level. The BOP selects manual control for the 3B Feedwater Control Valve operation. Once the feedwater control channel is selected out, automatic control can be re-established. The US enters 3-ONOP-049.1, Deviation or

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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Failure of Safety Related or Reactor Protection Channels, to remove the channel from service. During this process, the US references Tech Specs for required actions.

Event 6 The crew demonstrates precise control during the Steam Generator Tube Leak load reduction. During this evolution, Pressurizer Control Channel PT-3-445 fails high. This channel causes PCV-3-456 to fail open. The RCO takes action to close the PORV manually which is unsuccessful. The RCO isolates the open PORV with its Block Valve, MOV-3-535. The US directs 3-ONOP-041.5, Pressurizer Pressure Control Malfunction, for response.

Event 7 After about 5 to 10% downpower, 3C SGTR starts on a 5 minute ramp. The crew takes actions to maximize Charging and to isolate Letdown. When the leakage exceeds the CVCS capacity, the US orders to trip the Reactor and Turbine, initiate Safety Injection and Phase A, and enters to 3-EOP-E-0, Reactor Trip or Safety Injection. The Reactor Trip triggers the RV-3-1411, 3C Steamline Safety Valve to fail open during the trip. The mitigation sequence is 3-EOP-E-0 to 3-EOP-E-3 to 3-EOP-ECA-3.1.

Event 8 This post trip failure removes the auto isolation of feedwater flow to the 3C SG. The 3C SG Feedwater Control Valve FCV-3-498 fails open along with MOV-3-1409. MOV-3-1409 is manually closed to isolate this flowpath prior to the cooldown in 3-EOP-E-3.

The scenario is terminated once the crew has the opportunity to depressurize the RCS to refill the Pressurizer in 3-EOP-ECA-3.1 or at the Lead Evaluator's discretion.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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CREW CRITICAL TASKS Event # Description 5

CT1: Control SG Level After a failure of 3B Feedwater Channel FT-3-487, manually control 3B SG Water Level to prevent a Unit trip on low or high SG level.

7 CT2: Isolate Faulted 3C SG After Unit 3 Reactor is tripped when a heat sink is established by total feedwater flow greater than 400 gpm or one narrow range SG level greater than 7%, manually isolate the ruptured/faulted 3C SG including closing MOV-3-1409 3C SG Feed-water Isolation MOV prior to wide range level reaching 100%.

7 CT3: Control Initial RCS Cooldown When 3-EOP-E-3, Steam Generator Tube Rupture is entered, dump steam from 3A and 3B SGs at maximum rate using the Condenser Steam Dump Valves or Steam Dump To Atmosphere Valves to achieve Core Exit TCs less than required temperatures based on the lowest ruptured S/G pressure without causing a required transition to 3-EOP-FR-P.1, Response to Imminent Pressurized Thermal Shock Condition.

SEQUENCE OF EVENTS Event # Description

1.

3D Feeder Breaker 3AB19 Overcurrent and Fails Open - Loss of 3D 4KV Bus

2.

3C SG Tube Leakage

3.

Turbine Inlet Pressure Channel, PT-3-447 Fails Low

4.

3C SGTL Fast Load Reduction from 100% Power

5.

3B Feedwater Channel FT-3-487 Fails As Is

6.

Pressurizer Control Channel PT-3-445 Fails High

7.

3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip

8.

Feedwater Valves, MOV-3-1409 and FCV-3-498, Fail to Isolate on a Safety Injection

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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SIMULATOR SET UP INSTRUCTIONS Check Action Restore IC-1 (100% MOL) or equivalent IC.

Unfreeze the Simulator.

Stop 3A1 Circulating Water Pump.

Open & execute lesson file L-15-1 NRC SCEN 1 - 112114.lsn Ensure the following lesson steps are triggered:

SETUP - OOS EQUIPMENT (IMM23BT, TAMSD3BS, TAFF01B, TCF5MTB).

Place ECO tag on B AFW Pump Trip and Throttle Valve.

Remove velcro tags for B AFW alignment at RPM meter.

Place the 3B HHSI Pump in PTL and hang an ECO tag on it.

Ensure Rod Group Step Counters have completed stepping out.

Ensure PT-3-447 is selected for Turbine Inlet Pressure control.

Allow the plant to stabilize.

Acknowledge any alarms and freeze Simulator.

Ensure B train is protected train on VPA.

Perform the SIMULATOR OPERATOR CHECKLIST or equivalent.

Place TURNOVER SHEETS on ROs desk or give to the Lead Evaluator.

Ensure at least 4 copies of the maneuvering guide are available for shift turnover.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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BRIEFINGS Shift turnover information is attached to the back of this guide.

Ensure all applicants are prior briefed on Appendix E of NUREG 1021, Policies and Guidelines For Taking NRC Examinations.

Ensure all applicants are signed the on weekly Single Use Security Agreement, TR-AA-220-1002-F03.

Conduct a Crew Pre-brief to cover turnover information.

US:

RCO:

BOP:

SCENARIO NOTE 0-ADM-211 Prudent Operator Actions - If redundant stand-by equipment is available and ready, the operator is permitted to start the redundant equipment for failed or failing operating equipment. Immediate follow up of applicable ARPs and ONOPs (AOPs) shall occur as required.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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3-NOP-010, CIRCULATING WATER SYSTEM TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

NOTE This procedure may be done anytime during scenario.

BOOTH OPERATOR If asked, reply that pre-start checks for 3A1 CWP are complete.

If called as Chemistry, acknowledge starting 3A1 CWP.

BOP:

As directed, Uses 3-NOP-010, Circulating Water System, to start 3A1 CWP.

Notifies Chemistry of intent to start and records SG Blowdown rates.

Starts 3A1 CWP and ensures discharge valve MOV-3-1416 opens.

Informs US that 3A1 CWP is running.

EVENT 1 - 3D Feeder Breaker 3AB19 Overcurrent and Fails Open - Loss of 3D 4KV Bus.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

NOTE Ensure the Simulator is in RUN before the crew enters the Simulator.

US:

Conducts shift turnover.

BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 1 - 3AB19 OC RELAY FAILURE LOSS OF 3D BUS.

Acknowledges multiple annunciators for loss of 3D 4KV Bus, loss of CCW, and ICW.

Reviews ARPs and recommends entry into 3-ONOP-004.5.

3-ONOP-004.5, LOSS OF THE 3D 4KV BUS NOTE The US may direct starting an ICW and/or CCW Pump first prior to procedure entry.

US:

Directs ARP and 3-ONOP-004.5 response Understands 3C ICW TS LCO Action 3.7.3.a - 14 day Understands 3C CCW TS LCO Action 3.7.2.a - 30 day

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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EVENT 1 - 3D Feeder Breaker 3AB19 Overcurrent and Fails Open - Loss of 3D 4KV Bus.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR If dispatched, investigate loss of 3D 4KV Bus and report in 2 minutes that 3AB19 Bus feeder has a dropped flag on breaker with an 86 lockout actuated.

BOP:

Check 3D 4KV Bus Lockout Relay - RESET STEP 1 BOP:

Check 3A And 3B 4KV Buses - AT LEAST ONE ENERGIZED Both energized STEP 2 BOP:

Disconnect Loads From 3D 4KV Bus

a. Verify 3C Intake Cooling Water Pump Breaker, 3AD05 - OPEN

b. Verify 3C Component Cooling Water Pump Breaker, 3AD04 -

OPEN STEP 3 BOP:

Determine Source Of Power For 3D 4KV Bus

a. Check 3A and 3B 4KV buses - BOTH ENERGIZED

b. Consult with Nuclear Plant Supervisor to determine desired source of power for 3D 4KV bus:

3A 4KV bus OR

3B 4KV bus

c. Check desired source of power for 3D 4KV bus - 3A 4KV bus STEP 4 BOP:

Re-energize 3D 4KV Bus From 3A 4KV Bus:

a. Open Feeder To 4KV Bus 3D, 3AB19

b. Open Supply From 4KV Bus 3B, 3AD06

c. Close Supply From 4KV Bus 3A, 3AD01

d. Close Feeder To 4KV Bus 3D, 3AA17 Continues with Step 8 STEPS 5 & 6

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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EVENT 1 - 3D Feeder Breaker 3AB19 Overcurrent and Fails Open - Loss of 3D 4KV Bus.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Verify 3D 4KV Bus - ALIGNED TO AN ENERGIZED BUS

3A 4KV bus OR

3B 4KV bus STEP 8 BOP:

Locally Verify No Breaker Targets Exist On 3D 4KV Bus Breakers STEP 9 BOP:

Restart Components Supplied By 3D 4KV Bus As Directed By Shift Manager:

3C Intake Cooling Water Pump 3C Component Cooling Water Pump Return To Procedure And Step In Effect STEPS 10 & 11 BOOTH OPERATOR Respond as WCC/Electrical Maintenance.

BOP:

Notify the following:

Electrical Maintenance to make repairs WCC to generate of PWO.

NOTE:

Once power or loads are restored or at the Lead Evaluators discretion, proceed to Event 2.

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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EVENT 2 - 3C SG Tube Leakage TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 2 - 3C SG TUBE LEAK.

RCO Observes ANN H1/4, PRMS HIGH RAD.

Since an alarm is on R-15 OR R-19, refer to 3-ONOP-071.2, Steam Generator Tube Leakage.

Recommends entering 3-ONOP-071.2.

BOP Reviews ARP and confirms alarm.

Makes a plant announcement to warn personnel of the radiation hazard.

US:

Directs ANN response.

Refers to PCB to estimate leakage on graphs Refers to Tech Spec Action 3.4.6.2. a -

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to Hot Standby Primary to Secondary leakage > 150 gpd.

BOP:

Addresses ANN H1/4 response.

Verifies alarm is on R-3-15 Refers to 3-ONOP-071.2, Steam Generator Tube Leakage US:

Directs response per 3-ONOP-071.2.

Reviews Foldout Page with the crew.

RCO:

Monitor Affected Plant Parameters Checks PRZ Level - STABLE OR INCREASING

- Start Charging Pumps and isolate Letdown as required.

- IF PRZ level can NOT be maintained, THEN, manually trip the reactor.

STEP 1 BOP:

Check High Alarm ON For The Following PRMS Channels:

Check R-15 High Alarm light - ON STEP 2

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EVENT 2 - 3C SG Tube Leakage TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Check PRMS Channel R-15 Alarm Valid As Follows Check readout on affected channel - GREATER THAN OR EQUAL TO ALARM SETPOINT Check R-15 operability:

RDU green OPERATE LED - ON Plus sign (+) - ROTATING STEP 3 BOP:

Check PRMS Channels R-19 For Proper Operation:

Check R-19 High Alarm light - ON (NO)

STEP 4 BOP:

Check The Following Radiation Monitors:

Check RAD-3-6417 (SJAE SPING) -ALARM CLEAR Check RAD-6426 (DAM-1 Monitor) - ALARM CLEAR STEP 5 BOP:

Checks PRMS Channel Counts -INCREASING OR STABLE AT HIGHER VALUE R-3-15 (yes)

R-3-19 STEP 6 NOTE This also a foldout page item that may be manually initiates if R-19 Count Rate is increasing OR High Alarm.

BOOTH OPERATOR If asked, there is no flow at the S/G sample flow indicators.

BOP:

Check R-3-19 HI ALARM - CLEAR When the R-3-19 HI ALARM comes in verify the following:

Steam Generator Blowdown Flow Control Valves are Closed.

FCV-3-6278A FCV-3-6278B FCV-3-6278C Blowdown Tank to Canal Level Control Valve, LCV 6265B is Closed.

Verify NO FLOW on S/G Sample Flow Indicators at the Cold Chem Lab. (Ensures Sample Valves SV-3-2800, SV-3-2801, SV-3-2802 are Closed.)

STEP 7

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EVENT 2 - 3C SG Tube Leakage TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Approximate Tube Leakage:

Determines leakage to 3 gpm using any or all off the following: OSP-041.1, RCS LEAK RATE CALCULATION

- Unit 3 SJAE SPING Primary to Secondary Leak Rate Graph in the Plant Curve Book (Section 5, Figure 14)

AND record on ATTACHMENT 7

- Unit 3 R-15 Primary to Secondary Leak Rate Graph in the Plant Curve Book (Section 5, Figure 15) AND record on ATTACHMENT 7 STEP 8 BOOTH OPERATOR When RP is asked to check secondary radiation, wait 5-10 min and then report 3B MSL radiation > background with the SJAE discharge contaminated.

When Chemistry is asked to sample S/Gs for radiation, wait 5-10 min and the report counts highest on 3B SG sample.

BOP:

Identify Leaking S/G:

Monitor the following for S/G tube leak indications

- Unexplained increase in any S/G level High radiation detected on a S/G sample

- High radiation detected on a main steam line

- High radiation detected on AFW steam supply line (if running)

- High radiation detected from a S/G Blowdown line

- Unexplained difference between steam flow and feedwater flow

- Increasing radiation levels indicated on R-15, R-19, SPING, AND DAM-1 STEP 9.a COMM:

Directs Radiation Protection to:

Monitor radiation levels on Main Steam Lines.

Monitor radiation levels on AFW steam supply line (if running).

Monitor radiation levels on S/G Blowdown lines Monitor airborne activity at Steam Jet Air Ejectors.

STEP 9.b

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EVENT 2 - 3C SG Tube Leakage TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR Acknowledge directions COMM:

Directs Chemistry to Perform 0-NCAP-104, PRIMARY TO SECONDARY LEAK RATE CALCULATION.

Increase S/G sampling frequency as determined by Nuclear Chemistry.

Monitor DAM-1 and SJAE SPING readings.

STEP 9.c BOOTH OPERATOR Respond as Unit 3 TO if directed to de-energize MOV-3-1405 BKR 3D01-

27.

After 5 min.,

When directed by crew, trigger LOA - CLOSE DEENERGIZE MOV 1405.

BOP:

When the 3C Steam Generator is identified as the affected S/G, Verify Closed 3C Steam Generator AFW Steam Supply MOV.

Dispatch an Operator to de-energize BKR 3D01-27 for MOV-3-1405.

FOP - ATTACHMENT 4 BOOTH OPERATOR Acknowledge request for support US:

Determine If Operation May Continue:

Determines 3-ONOP-71.2 Attachment 3 Action level 3A applies:

Be in Mode 3 in the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Increase monitoring of PRMS channel to every 15 minutes.

Coordinate with RP to contain activity in secondary systems.

STEP 10 SM:

Determines SG Tube Leak > T.S. 3.4.6.2.a limit of 150 GPD, Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Directs Unit shutdown be completed within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months per 3-ONOP-071.2, Action 3B STEP 10

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EVENT 2 - 3C SG Tube Leakage TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR Respond as SM.

US:

Notifies the Shift Manager To Review The Following Procedures And Make Any Required Notifications 0-EPIP-20101, Duties of Emergency Coordinator.

0-ADM-115, Notification of Plant Events.

Verify NRC Resident notified of Fast Load Reduction and S/G Tube Leakage.

STEP 11 US/COMM:

Complete Attachment 1, Fast Load Reductions Maneuvering Plan.

STEP 12 NOTE Before the Control Room fast load reduction brief or at the Lead Evaluators discretion, continue to the next event.

EVENT 3 - Turbine Inlet Pressure Channel, PT-3-447 Fails Low TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 3 - PT-3-447 FAILS LOW.

US:

Directs the crew to check for a runback.

No runback, directs RCO to take the rods to manual.

3-ONOP-028, REACTOR CONTROL SYSTEM US:

Directs response of 3-ONOP-028, Reactor Control System Malfunction.

NOTE This is an Immediate Operator Action RCO:

Verify/Place the Rod Motion Control Selector switch to the MAN position.

RCO:

Adjust rods or reduce turbine load to restore Tavg equal to Tref RCO:

Place Channel Select Turbine Inlet Control to operable channel, PT-3-446.

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EVENT 3 - Turbine Inlet Pressure Channel, PT-3-447 Fails Low TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

RCO:

Compare rod position to control rod insertion limits If RILs are exceeded, borate at 16 gpm RCO:

Manually position the RCC control bank to restore steady state conditions.

Place Rod Motion Control Selector Switch to AUTO BOOTH OPERATOR Acknowledge notifications to WCC and I&C.

BOP:

Notify:

WCC and I&C 3-ONOP-049.1, DEVIATION OR FAILURE OF SAFETY OF SAFETY RELATED OR REACTOR PROTECTION CHANNELS US:

Enters and directs actions of 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, for response.

BOOTH OPERATOR Acknowledge the request to bypass PT-3-447.

Using Schema, go to AMSAC Panel and take both Power 2 (447) switches to bypass, then depress System Reset. In five minutes report when complete.

BOP:

Dispatches operator to place the Bypass Switches for PT-3-447 to Bypass position at the AMSAC panel using Attachment 5.

NOTE Once the crew has stabilized the plant and initiated actions to match Tave to Tref and /or Tech Specs exercised, then at the Lead Evaluators discretion move on the next event.

US:

Enters to Tech Spec Action 3.3.1 FU17.b - Action 7 for P within 1 hr verify the associated bi-stable is in the correct state Enters to Tech Spec Action 3.3.2 FU4.d - Action 15 for Steamline Isolation - within 6 hrs trip bi-stables

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EVENT 4 - 3C SGTL Fast Load Reduction From 100%

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Directs 3-ONOP-071.2 response Reviews Foldout page with the crew US:

Reviews Attachment 1, Fast Load Reductions Maneuvering Plan.

US:

Briefs Control Room Personnel Using ATTACHMENT 2 BOOTH OPERATOR Respond to notifications as required to acknowledge reports.

BOP:

Notify the Following:

System Dispatch Plant personnel using Page Boost.

Chemistry to perform RCS TS samples (>15% power change)

RCO:

Begins Boration For Initial Tave Effect Sets BA totalizer per Attachment 1 Place the Reactor Makeup Selector Switch to BORATE Place the RCS Makeup Control Switch to START Adjust FC-3-113A, Boric Acid Flow Controller to achieve 40 gpm boric acid flow as indicated on FR-3-113 WHEN Tavg begins to lower from the boration, THEN, adjust FC-3-113A, Boric Acid Flow Controller to the boration rate of ATTACHMENT 1 NOTE After Rx Power has been lowered about 5% or at the Lead Evaluators discretion go to the next event.

BOP:

Initiate Turbine Load reduction in MW CNTRL Select MW CNTRL Set TARGET power level - MW VALUE from Attachment 1.

Set RAMP RATE - MW/M VALUE FROM Check Tave 1 to 2° F lower than Tref.

Depress GO Ensure RCO FC-3-113A, Boric Acid Flow Controller, has been adjusted to the load reduction boration rate from ATTACHMENT 1

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EVENT 4 - 3C SGTL Fast Load Reduction From 100%

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Monitor Load reduction Monitors turbine load reduction and informs crew at each 50 MWe increment of power reduction Adjusts turbine load reduction rate to maintain Tavg/Tref within

+/- 4°F T.

Monitors S/G level control to ensure feed reg valves properly maintain level control in automatic BOOTH OPERATOR Respond as SNPO. If asked, idle Charging Pump ready for start.

RCO:

Monitor Load reduction Monitors Pressurizer Level to maintain on program.

Starts 2nd Chg Pp and places 2nd orifice in service.

US:

Refers to Enclosure 1 for expected alarms.

BOOTH OPERATOR When directed to realign Aux Stm Supply acknowledge request but take no action at this time.

BOP:

Isolate Miscellaneous Flow paths From Leaking S/G Verify Blowdown Isolation valve CV-3-6275C for 3C S/G CLOSED.

Direct Unit 3 Turbine Operator to realign Aux Stm Supply from U4 per Attachment 8.

RCO:

Monitor Boration Rate During Downpower Determine if Insertion Limit and Bank D position are converging at a rate that will cause rod insertion limit alarms

- Adjust power reduction rate as needed to control rod insertion

- Increase boration rate and/or total amount as necessary to limit control rod insertion Monitor Annunciator B8/1, ROD BANK LO LIMIT - CLEAR Monitor B8/2 ROD BANK A/B/C/D EXTRA LO LIMIT - CLEAR EVENT 5 - 3B Feedwater Channel FT-3-487 Fail As Is TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 5 - 3B FW CHANNEL FT-3-487 FAILS AS IS BOP:

Recognizes and reports FT-3-487 failure.

Takes manual control of 3B S/G level control valve FCV-3-488.

Restores 3B S/G level to normal.

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EVENT 5 - 3B Feedwater Channel FT-3-487 Fail As Is TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

RCO:

Addresses Alarm Response for SG B Level Deviation.

US:

Enters and directs actions of 3-ONOP-049.1, Deviation or Failure of Safety Related or Reactor Protection Channels, for response BOP:

Verify FT-3-487 failure by channel check comparison.

Verify no off-normal conditions exist on FT-3-487.

Place 3B S/G FW Control Transfer Switch to FT-3-486 and ensure 3B S/G level is returned to auto.

BOOTH OPERATOR WCC/I&C: Acknowledge the report. Wait to trip bi-stables for trouble-shooting.

BOP:

Notifies WCC to initiate PWO and I&C for troubleshooting.

US:

Enters to Tech Spec Action 3.3.1 FU12 - Action 6 within 6 hrs trip bi-stables NOTE:

After S/G level control is restored, at the Lead Evaluators discretion, proceed to Event 6.

US:

Conducts crew brief.

EVENT 6 - Pressurizer Control Channel PT-3-445 Fails High TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 6 - HIGH FAILURE OF PT-3-445 CAUSES PORV TO OPEN.

RCO:

Recognizes PORV 456 Open.

Verifies PZR pressure less than 2235.

Attempts to close PORV 456.

Closes the PORV Block Valve.

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EVENT 6 - Pressurizer Control Channel PT-3-445 Fails High TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

NOTE With the failure of the PORV to close, the stroke time of the PORV Block Valve may not be quick enough time to recovery the unit. This is not a critical step to prevent a unit trip. Proceed to the next event when a trip occurs or when conditions stabilize.

BOP:

Addresses Alarm Response for A9/2.

Checks if alarm is due to instrument failure, then refers to 3-ONOP-041.5, PZR Press Control Malfunction.

US:

Enters and directs actions of 3-ONOP-041.5, PZR Press Control Malfunction, for response.

Reviews 3-ONOP-041.5 Foldout Page IF any Pressurizer Pressure control Instrument Loop fails, THEN place applicable control switches to a position that isolates the failed instrument.

IF PZR pressure cannot be maintained greater than 2000 psig, THEN perform the following:

Continue efforts to restore PZR pressure and Trip the reactor and turbine and go to 3-EOP-E-0, REACTOR TRIP OR SAFETY INJECTION.

IF any PORV is OPEN OR Leaking AND pressure is less than 2235 psig, THEN CLOSE the applicable PORV and/or Block valve.

RCO:

Check PZR Pressure Control Instrument Loop Not Failed Check PT-3-444 - NOT FAILED by comparison with adjacent pressure channels and known plant parameters.(YES)

Check PT-3-445 - NOT FAILED by comparison with adjacent pressure channels and known plant parameters.

.(NO)

Verify PCV-3-456 OR MOV-3-535 is CLOSED.

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EVENT 6 - Pressurizer Control Channel PT-3-445 Fails High TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

RCO:

Checks PORVs closed Checks PZR Spray Valves Closed with PZR pressure normal or trending to normal.

Checks PZR Safety Valves closed.

Check PZR Pressure Stable or Increasing.

Check Pressurizer Pressure Above Normal Value (NO)

RCO:

Check Pressurizer Pressure Low Or Decreasing.

Maintain PZR Pressure Greater Than 2000 psig.

Check PZR Heater Operable Check PZR Heaters ON.

Check PZR Heaters capable of maintaining pressure.

BOOTH OPERATOR WCC/I&C: Acknowledge the report.

BOP:

Notifies WCC to initiate PWO and I&C for troubleshooting.

NOTE At the Lead Evaluators discretion go to the next event.

US:

Conducts crew brief.

EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip BOOTH OPERATOR When directed by Lead Evaluator or if Reactor trips, trigger MAJOR EVENT 7 - 3C SGTR.

RCO Notices loss of RCS pressure control as a result of either the PT-3-445 failure or the subsequent 3C SGTR.

Recommends entering 3-EOP-E-0, Rx Trip or Safety Injection.

NOTE The fault associated with the 3C SG is delayed and occurs after the 3-EOP-E-3 cooldown is started by a manual trigger.

US:

Directs 3-EOP-E-0 response after auto Reactor trip.

OR Directs RCO to manually trip the Reactor, then for operators to perform their IOAs.

Auto Trigger Event 7 auto triggers EVENT 8 - 3C FAILS TO ISOLATE ON SI.

RCO:

Manually trips Reactor. (N/A, if Auto Rx Trip occurred)

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EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip RO/BOP:

Perform IOAs.

RCO:

Verifies Reactor Trip Rod bottom lights - ON Reactor trip & bypass breakers - OPEN Rod position indicators -AT ZERO Neutron flux - DECREASING STEP 1 BOP:

All turbine stop or associated control valves - CLOSED Moisture Separator Reheater Steam Valves-CLOSED MSR Main Stm Sup Stop MOVs Reheater Timing Valves MSR Purge Stm Valves Checks Mid and East GCBs - OPEN STEP 2 BOP:

Verifies Power To Emergency 4 KV Buses

a. Checks the 3A and 3B 4 KV buses - MAINTAIN AT LEAST ONE ENERGIZED

b. Checks the 3A and 3B 4 KV buses - MAINTAIN BOTH ENERGIZED

c. Maintains the 3D 4 KV Bus energized - ALIGNED TO AN ENERGIZED 4 KV BUS STEP 3 NOTE The SGTR ramps in. The signs of the SI requirement may not be evident until the reading of the IOA actions.

RCO:

Checks If SI Is Actuated SI Annunciators ANY ON OR Safeguards equipment - AUTO STARTED STEP 4

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EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip RCO:

a. Checks if SI is required:

Low PZR pressure - 1730 psig OR High Containment pressure - 4 psig OR High steam line differential pressure - 100 psid OR High steam flow with low S/G pressure 614 psig OR low Tavg (543°F)

b. IF SI is required, THEN: (N/A, if Auto occurred)

Manually actuate SI.

Manually actuate Containment Isolation Phase A.

STEP 4 RNO RCO:

Verifies SI actuation when required.

US:

Directs 3-EOP-E-0 response and reviews the IOAs.

Reviews Foldout page with the crew CREW:

Reviews FOP for 3-EOP-E-0 Adverse Cntmt - Not met RCP Trip Criteria - Not met Faulted S/G Isolation - Not met initially. When met,

1) maintains AFW flow > 400 gpm until NR level is > 7%

2) isolate AFW flow

3) stabilize RCS Hot Leg temperatures using Steam Dumps when faulted SGs < 9% WR (may not be met due to SGTR)

RUPTURED S/G ISOLATION CRITERIA when 3A S/G narrow range level is greater than 7%[27%], then stop feed flow to 3A S/G (May not initially be found with higher 3B and 3C SG levels)

AFW Sys Operation Criteria - time monitored CST Makeup Water Criteria - N/A RHR System Operation Criteria - one started running at shutoff - time monitored Loss of Offsite Power or SI on the Other Unit-Not met Loss of Charging Criteria - Not met FOLDOUT PAGE

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EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip BOP:

Continues with ATTACHMENT 3 to complete The Prompt Action Verifications.

STEP 5 NOTE The RCO may notice that MOV-3-1409 and FCV 498 fail to isolate on Safety Injection.

RCO:

Check AFW Pumps - AT LEAST TWO RUNNING.

STEP 6 RCO:

Verify AFW Valve Alignment - PROPER EMERGENCY ALIGNMENT STEP 7 RCO:

Verify Proper AFW Flow:

a. Check Narrow Range Level in at least one S/G -

GREATER THAN 7%[27%]

b. Maintain feed flow to S/G until Narrow Range Levels between 21%[27%] and 50%

STEP 8 RCO:

Check RCP Seal Cooling:

a. All RCP Thermal Barrier Alarms - CLEAR:

A 1/1, RCP THERMAL BARR COOLING WATER HI FLOW A 1/2, RCP THERMAL BARR COOLING WATER HI TEMP A 1/3, RCP THERMAL BARR COOLING WATER LO FLOW STEP 9 RCO:

Check RCS Temperatures:

a. Check RCPs - ANY RUNNING (yes)

b. Check RCS Average temperatures -

STABLE BETWEEN 545°F AND 547°F OR

TRENDING DOWN TO 547°F STEP 10

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EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip RCO:

Check PRZ PORVs, Spray Valves And Excess Letdown Isolated:

a. PORVs - CLOSED

b. Normal PRZ Spray valves - CLOSED

c. CV-3-311, Auxiliary Spray Valve - CLOSED

d. Excess Letdown - NOT IN SERVICE STEP 11 RCO:

Check If RCPs Should Be Stopped:

a. RCPs - ANY RUNNING a. Go to Step 13.

b. High-Head SI Pump - AT LEAST ONE RUNNING, AND SI Flowpath - VERIFIED

c. RCS subcooling - LESS THAN 19°F[41°F] - met STEP 12 RCO:

Check If S/Gs Are Faulted: (NOT met at this time)

a. Check pressures in all S/Gs -

ANY S/G PRESSURE DECREASING IN AN UNCONTROLLED MANNER OR

ANY S/G COMPLETELY DEPRESSURIZED STEP 13

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EVENT 7 - 3C SGTR Starts and RV-3-1411, 3C Steamline Safety Valve Fails Open During The Trip BOOTH OPERATOR If Chemistry or RP is called, local secondary radiation readings indicate a 3C SGTR.

RCO:

Check If S/G Tubes Are Ruptured: (met for 3C SG)

a. Check levels in all S/Gs and secondary radiation levels:

Any S/G level - INCREASING IN AN UNCONTROLLED MANNER OR

Condenser Air Ejector Radiation R HIGHER THAN NORMAL OR

S/G Blowdown Radiation R HIGHER THAN NORMAL OR

DCS S/G OR secondary radiation readings - HIGHER THAN NORMAL OR

Local steamline radiation - HIGHER THAN NORMAL

b. Perform the following:

1) Monitor Critical Safety Functions using 3-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES

2) Go to 3-EOP-E-3, STEAM GENERATOR TUBE RUPTURE, Step 1 STEP 14 EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

3-EOP-E-3, STEAM GENERATOR TUBE RUPTURE US:

1) Monitors Critical Safety Functions using 3-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES.

2) Go to 3-EOP-E-3, STEAM GENERATOR TUBE RUPTURE, Step 1

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EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Directs 3-EOP-E-3 response.

Reviews Foldout Page Adverse Containment Setpoints RCP Trip Criteria SI Re-Initiation Criteria Secondary Integrity Criteria Cold Leg Recirculation Switchover Criteria CST Makeup Water Criteria Multiple Tube Rupture Criteria Loss Of Offsite Power Or SI On Other Unit.

If RHR flow is less than 1100 gpm, then the RHR Pumps shall be shut down within 44 minutes of the initial start signal.

RCO:

Checks If RCPs Should Be Stopped STEP 1 BOOTH OPERATOR If called as RP, report the radiation readings on 3C SG lines are higher than normal.

BOP:

Identify Ruptured S/G:

Identify 3C as the Ruptured S/G Directs RP to take rad readings on Main Steam and Blowdown Lines Evaluates DAM1 on DCS Determines ruptured SG by level increase or radiation STEP 2 BOP:

Isolate Flow From Ruptured S/G Adjusts 3C S/G Steam Dump To Atmosphere controller setpoint to 1060 psig Checks 3C S/G Steam Dump To Atmosphere Closed.

- WHEN ruptured S/G pressure is less than 1060 psig, THEN Verify S/G Steam Dump to Atmosphere is closed.

STEP 3.a

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EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by crew, trigger LOA - CLOSE DEENERGIZE MOV 1405 Report complete after 5 min.

BOP:

Close steam supply valves from ruptured S/G(s) to AFW Pumps using:

Attachment 19 for 3C S/G Close MOV-3-1405 by dispatching an Operator To Locally Perform The Following:

a. Place Breaker 4D01-27 in OFF

b. Verify MOV-3-1405 - CLOSED Notify Unit Supervisor That 3C S/G AFW Steam Supply Is Isolated and Attach 19 is complete.

ATTACHMENT 19 BOP:

Verify 3C S/G Blowdown Isolation Valve FCV-3-6278C is closed on 3C S/G.

Closes 3C MSIV.

STEP 3.d-e BOOTH OPERATOR When directed by Lead Evaluator, trigger EVENT 1 - 3AB19 OC RELAY FAILURE LOSS OF 3D BUS BOP:

Check Circulating Water Pumps - ANY RUNNING STEP 3.f BOOTH OPERATOR When directed by Lead Evaluator, trigger LOA -

ALIGN AUX STEAM SUPPLY FROM U4.

After 5 minutes, report complete BOP:

Dispatch Operator to align main steam auxiliaries using STEP 3.f BOP:

Isolate miscellaneous flow paths from 3A S/G using 6.

- Verify 3C S/G Blowdown Sample MOV MOV-3-1425-CLOSED

- Dispatches Operator to locally isolate 3C S/G Main Steamline Steam Trap 3-10-321 Steam Sample Valves 3-10-896 for S/G 3C Inform Unit Supervisor That Attachment 16 Is Complete.

STEP 3.f

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EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

NOTE The fault is delayed in 3C SG to establish a target for cooldown.

BOP:

Check Rupture S/G Level:

Check 3C S/G Level Narrow Range level - GREATER THAN 7%[27%]

Verify feed flow stopped to 3C S/G.

STEP 4 US:

Checks 3C S/G pressure greater than 450 psig. (yes)

STEP 5 US/BOP:

Initiate RCS Cooldown Determines required CET Temp for Cooldown.

Check feed sources to intact S/Gs - CAPABLE OF PROVIDING 400 GPM Check Condenser AVAILABLE (IF lost, uses Stm Dumps to Atmosphere)

Manually dump steam to atmosphere from 3A & 3B S/G(s) at maximum rate using Steam Dump to Atmosphere Valves.

Continue with step 7

- When Core Exit TCs - LESS THAN REQUIRED TEMPERATURE, then stops cooldown.

- Maintains core exit TCs - LESS THAN REQUIRED TEMPERATURE.

STEP 6 BOP:

Check Intact S/G Level:

Any Narrow Range Level - GREATER THAN 7% [27%].

Maintain 3A and 3B S/G narrow range level between 21%

[27%] and 50%.

Narrow Level - LESS THAN 50%.

STEP 7 RCO:

Verify SI - RESET STEP 8

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EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

RCO:

Establish Charging Flow:

Charging Pumps - AT LEAST ONE RUNNING Check offsite power - AVAILABLE Start all available Charging Pumps.

Adjust speed controller as necessary to establish maximum Charging flow from the running Charging Pump(s).

Place RCS Makeup Control in STOP.

Adjust HCV-3-121, Charging Flow To Regen Heat Exchanger, to maintain proper Seal Injection flow.

Verify Charging Pump Suction auto transfers to RWST.

STEP 9 NOTE May become an additional critical task if the RHR Pump operating time at shutoff head on minimum flow recirculation is NOT limited to 44 minutes or less.

RCO:

Check If RHR Pumps should Be Stopped:

Check RCS pressure - GREATER THAN 275 PSIG[575 PSIG]

Check RHR flow - LESS THAN 1100 GPM Stop RHR Pumps and place in standby.

STEP 10 RCO:

Check PRZ PORVs And Block Valves:

Check Power to block valves - AVAILABLE Check PORVs - CLOSED Check Block valves - AT LEAST ONE OPEN STEP 11 BOOTH OPERATOR When directed by Lead Evaluator, trigger POST RX TRIP RV-3-1411 3C STM SAFETY FAILS OPEN.

RCO:

Reset Containment Isolation Phase A & Phase B.

STEP 12 BOOTH OPERATOR If requested to check source of steam leak, wait one minute and report a 3C SG Safety is lifting.

RCO:

Verify Instrument Air To Containment:

Verify CV-3-2803, Instrument Air Containment Isolation -

OPEN Verify Instrument Air pressure, as indicated on PI-3-1444 -

GREATER THAN 95 PSIG STEP 13

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EVENT 7 - 3C SGTR Starts And RV-3-1411 3C Steamline Safety Fails Open During The Trip.

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Check If RCS Cooldown Should be Stopped:

Check CETs < REQUIRED WHEN core exit TCs are less than required temperature Stop RCS cooldown Maintain CETs < REQUIRED STEP 14 BOP:

Checks Ruptured S/G(s) Pressure STABLE or INCREASING (no)

STEP 15 NOTE BOP:

Checks RCS Pressure decreasing to < 250 psig from intact SGs STEP 15.RNO NOTE The crew may apply Secondary Integrity Criteria from Foldout Page prior to or in 3-EOP-ECA-3.1.

US:

1) Monitors Critical Safety Functions using 3-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES.

2) Go to 3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED RUPTURE, Step 1.

The scenario is terminated after RCS depressurization which may be done in 3-EOP-ECA-3.1 or 3-EOP-E-3 or at the discretion of the Lead Evaluator.

- - END OF SCENARIO ***

When Implemented, 3-EOP-E-2 Faulted Steam Generator Isolation (3-EOP-ECA-3.1 steps follow)

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Conducts EOP transition crew brief.

Directs 3-EOP-E-2 response.

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When Implemented, 3-EOP-E-2 Faulted Steam Generator Isolation (3-EOP-ECA-3.1 steps follow)

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Reviews FOP for 3-EOP-E-2 with the crew Adverse Containment Setpoints Containment atmosphere temperature greater than 180°F OR Containment radiation levels greater than 1.3x105 R/hr FOLDOUT PAGE BOP:

Checks Main Steamline Isolation And Bypass Valves On faulted SG(s) - Closed.

STEP 1 BOP:

Check If Any S/G Is Not Faulted:

3A and 3B are not faulted.

STEP 2 BOP:

Identify Faulted S/G(s)

Identifies 3C S/G to be faulted STEP 3 BOP:

Verifies Faulted S/G Isolated:

Verify Main Feedwater Line Isolation valve MOV-3-1409 or FCV-3-498 closed.

Verify closed feedwater bypass line Isolation valve POV 487 or FCV-3-489 closed.

- Verify AFW Flow Isolated for 3C SG with both Aux Feed Flow Controllers, in Manual and sets output to zero.

Dispatch operator to verify closed 3C Steam Generator AFW Steam Supply, MOV-3-1405.

Verify de-energized MOV-3-1405 BKR, 3D01-27 Verify closed 3C S/G Dump to Atmosphere Valve.

Verify closed CV-3-6275C, 3C S/G Blowdown isolation valve.

Verify closed MOV-3-1425, 3C S/G Liquid Sample valve.

STEP 1

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When Implemented, 3-EOP-E-2 Faulted Steam Generator Isolation (3-EOP-ECA-3.1 steps follow)

TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Stabilize RCS Hot Leg Temperature:

Check faulted S/G Level - LESS THAN 9% Wide Range

[27% Narrow Range].

Stabilize RCS Hot Leg temperatures using Steam Dump by adjusting intact S/G Steam Dump To Atmosphere valves Auto setpoint to match current S/G pressures.

STEP 2 BOP:

Checks CST Level - GREATER THAN 12%.

STEP 3 BOP:

Check Secondary Radiation:

Direct Nuclear Chemistry to take periodic activity samples of all S/Gs.

Direct Radiation Protection to take radiation readings on Main Steam lines.

Check DCS DAM1 and other unisolated or secondary monitors readings - NORMAL.

Secondary radiation - NORMAL (no)

STEP 4 3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

US:

Directs 3-EOP-ECA-3.1 response.

Reviews Foldout Page Adverse Containment Setpoints SI Re-Initiation Criteria Secondary Integrity Criteria Cold Leg Recirculation Switchover Criteria CST Makeup Water Criteria Loss Of Offsite Power Or SI On Other Unit.

If RHR flow is less than 1100 gpm, then the RHR Pumps shall be shut down within 44 minutes of the initial start signal.

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3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED RCO:

Verify SI - RESET STEP 1 RCO:

Reset Containment Isolation Phase A & Phase B.

STEP 2 RCO:

Verify Instrument Air To Containment:

Verify CV-3-2803, Instrument Air Containment Isolation -

OPEN Verify Instrument Air pressure, as indicated on PI-3-1444 -

GREATER THAN 95 PSIG STEP 3 BOP:

Verify All 4KV Buses - ENERGIZED BY OFFSITE POWER A 4KV Bus B 4KV Bus C 4KV Bus D 4KV Bus STEP 4 BOOTH OPERATOR When TSC is called for minimum PRZ Water Level for heater coverage.

Report - TSC recommends PRZ Heaters be re-energized at 30% level RCO:

De-energize PRZ Heaters Positon PRZ Switches in OFF Consult TSC Reset PRZ Backup Heaters STEP 5 RCO:

Checks Containment Spray Pumps not running.

STEP 6 BOP:

Check Rupture S/G Level:

Check 3C S/G Level Narrow Range level - GREATER THAN 7%[27%]

Verify feed flow stopped to 3C S/G.

STEP 7

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3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED NOTE May become an additional critical task if the RHR Pump operating time at shutoff head on minimum flow recirculation is NOT limited to 44 minutes or less.

RCO:

Check If RHR Pumps should Be Stopped:

Suction aligned to RWST Check RCS pressure - GREATER THAN 275 PSIG[575 PSIG]

Check RHR flow - LESS THAN 1100 GPM Stop RHR Pumps and place in standby.

STEP 8 BOP:

Initiate Evaluation of Plant Status Using Attachment 8.

STEP 9 RCO:

Establish Charging Flow:

Charging Pumps - AT LEAST ONE RUNNING Check offsite power - AVAILABLE Start all available Charging Pumps.

Adjust speed controller as necessary to establish maximum Charging flow from the running Charging Pump(s).

Place RCS Makeup Control in STOP.

Adjust HCV-3-121, Charging Flow To Regen Heat Exchanger, to maintain proper Seal Injection flow.

Verify Charging Pump Suction auto transfers to RWST.

STEP 10 BOP:

Checks If S/Gs Are NOT Faulted NO S/G Pressure decreasing in an uncontrolled manner NO S/G completely depressurized Continues with 3-EOP-E-2 if NOT previously isolated STEP 11 BOP:

Check Intact S/G Level:

Any Narrow Range Level - GREATER THAN 7% [27%].

Maintain 3A and 3B S/G narrow range level between 21%

[27%] and 50%.

Narrow Level - LESS THAN 50%.

STEP 12

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3-EOP-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED RCO:

Initiate RCS Cooldown To Cold Shutdown:

Maintain cooldown rate in RCS Cold Legs - LESS THAN 100°F Check RHR in service (no)

Check Block valves - AT LEAST ONE OPEN STEP 13 RCO:

Check If Subcooled Recovery Is Appropriate:

RWST greater than 200k gallons S/G NR Level less than 94%.

STEP 14 RCO:

Check RCS Subcooling Based On Core Exit TCs - Greater than 19°F STEP 15 RCO:

Check HHSI Pump Status - Aligned to RWST STEP 16 NOTE The upper head region may void during RCS depressurization if RCPS are NOT running.

RCO:

Depressurize RCS to refill PRZ with one PORV.

Secure depressurization when PRZ level is greater than 26%.

STEP 17 The scenario is terminated after RCS depressurization which may be done in 3-EOP-ECA-3.1 or 3-EOP-E-3 or at the discretion of the Lead Evaluator.

- - END OF SCENARIO ***

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EVENT 8 - Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

3-EOP-E-0 Attachment 3 -Prompt Action Verifications TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Check Load Centers Associated With Energized 4 KV Buses -

ENERGIZED:

- 3A LC

- 3B LC

- 3C LC

- 3D LC

- 3H LC STEP 1 BOP:

Verify Feedwater Isolation:

a. Place Main Feedwater Pump switches in STOP

b. Feedwater Control Valves - CLOSED:

- FCV-3-478 - unable to close, failed open

- FCV-3-488

- FCV-3-498

c. Feedwater Bypass Valves - CLOSED:

- FCV-3-479

- FCV-3-489

- FCV-3-499

d. Feedwater Bypass Isolation Valves - CLOSED:

- POV-3-477

- POV-3-487

- POV-3-497

e. Feedwater Isolation MOVs - CLOSED:

- MOV-3-1407

- MOV-3-1408

- MOV-3-1409 - initially open unless closed immediately following trip - Manually closes to remove the ability to feed the 3C SG

f. Verify Standby Feedwater Pumps - OFF STEP 2

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EVENT 8 - Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

3-EOP-E-0 Attachment 3 -Prompt Action Verifications TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Check If Main Steam Lines Should Be Isolated: (no)

a. Check Main Steamline Isolation and Bypass valves -

ANY OPEN

b. Check if either Main Steam Isolation Signal has actuated:

High Steam Flow with either Low S/G Pressure 614 psig OR Low TAVE 543°F OR

Hi-Hi Containment Pressure 20 psig

c. Verify Main Steam Isolation and Bypass valves - CLOSED STEP 3 BOP:

Verify Containment Isolation Phase A Valve White Lights On VPB

- ALL BRIGHT Actuates SI and Phase A Containment Isolation, if NOT actuated.

STEP 4 BOP:

Verify Pump Operation:

a. At least two High-Head SI Pumps - RUNNING

b. Both RHR Pumps - RUNNING STEP 5 BOP:

Verify Proper CCW System Operation:

a. CCW Heat Exchangers - THREE IN SERVICE

b. CCW Pumps - ONLY TWO RUNNING

c. CCW Headers - TIED TOGETHER

d. MOV-3-626, RCP Thermal Barrier CCW Outlet - OPEN STEP 6 BOP:

Verify Proper ICW System Operation:

a. Verify ICW Pumps - AT LEAST TWO RUNNING

b. Verify ICW To TPCW Heat Exchanger - ISOLATED:

- POV-3-4882 - CLOSED

- POV-3-4883 - CLOSED

c. Check ICW Headers - TIED TOGETHER STEP 7

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EVENT 8 - Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

3-EOP-E-0 Attachment 3 -Prompt Action Verifications TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Verify Containment Cooling:

a. Check Emergency Containment Coolers - ONLY TWO RUNNING STEP 8 BOP:

Verify Containment Ventilation Isolation:

a. Unit 3 Containment Purge Exhaust And Supply Fans - OFF STEP 9 BOP:

Verify Containment Spray NOT Required: (yes)

a. Containment pressure - HAS REMAINED LESS THAN 20 PSIG:

- PR-3-6306A

- PR-3-6306B STEP 10 BOP:

Verify SI - RESET STEP 11 BOP:

Verify SI Valve Amber Lights On VPB - ALL BRIGHT STEP 12 BOP:

Verify SI Flow:

a. RCS pressure - LESS THAN 1625 PSIG[1950 PSIG]

b. High-Head SI Pump flow indicator - CHECK FOR FLOW

c. RCS pressure - LESS THAN 275 PSIG[575 PSIG]

d. RHR Pump flow indicator - CHECK FOR FLOW STEP 13

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EVENT 8 - Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

3-EOP-E-0 Attachment 3 -Prompt Action Verifications TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOOTH OPERATOR When directed by the crew, trigger LOA -

ALIGN U-4 HHSI SUCTION TO U-3 RWST.

BOP:

Realign SI System:

a. Check Procedure Entry Status - E-0 ENTERED FROM 3-ONOP-047.1, LOSS OF CHARGING FLOW IN MODES 1 THROUGH 4 (no)

e. Verify Unit 3 High-Head SI Pumps - TWO RUNNING (no)

f. Operate Unit 4 High-Head SI Pumps to ensure two HHSI Pumps running

g. Direct Unit 4 Reactor Operator to align Unit 4 High-Head SI Pump suction to Unit 3 RWST using Attachment 1.

STEP 14 BOP:

Verify Containment Isolation Phase A - RESET STEP 15 BOP:

Reestablish RCP Cooling:

a. Check RCPs -AT LEAST ONE RUNNING

b. Open CCW To Normal Containment Cooler Valves:

MOV-3-1417

MOV-3-1418

c. Reset and start Normal Containment Coolers STEP 16 BOP:

Verify Control Room Ventilation Isolation:

a. Emergency Air Supply Fan SF-1B - ON

b. Emergency Air Supply Fan SF-1A - OFF

c. Control Room Ventilation dampers - ALIGNED FOR RECIRC

d. TS-0002, TSC Emergency Vent Auto Initiate Key Switch - IN ENABLE STEP 17 BOOTH OPERATOR When requested by crew, trigger LOA - ALIGN PAHMS 3-NOP-094.

BOP:

Place Hydrogen Monitors In Service Using 3-NOP-094, CONTAINMENT POST ACCIDENT MONITORING SYSTEM STEP 18 BOP:

Verify All Four EDGs - RUNNING STEP 19

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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EVENT 8 - Feedwater Valves, MOV-3-1409 and FCV-3-498, fail to isolate on a Safety Injection.

3-EOP-E-0 Attachment 3 -Prompt Action Verifications TIME EVALUATOR ACTIVITIES & NOTES POSITION: APPLICANTS ACTIONS/BEHAVIORS*

Observable actions grayed.

BOP:

Verify Power To Emergency 4 KV Buses:

a. Check 3A, 3B AND 3D 4 KV Buses - ALL ENERGIZED STEP 20 BOP:

b. Check 3A AND 3B 4 KV Buses - ALL ENERGIZED FROM OFFSITE POWER BOP:

Notify Unit Supervisor Of The Following:

- Attachment 3 is complete

- Any safeguards equipment that is NOT In the required condition

- Status of Containment pressure continuous action STEP 21

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L-15-1 NRC SCENARIO 1 NRC Draft (Rev 11/21/14)

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Conduct a Post-Exam Examiner Debrief using section 3 of TR-AA-220-1004 Attachment 6, Dynamic Simulator Exam Briefings.

Discussion Points are intentionally NOT included in evaluated scenarios. However, space is available below to document follow-up questions when further information is required to determine an evaluation outcome.

FOLLOW-UP QUESTIONS QUESTION #1 ANSWER #1 QUESTION #2 ANSWER #2

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SIMULATOR POST-SCENARIO RESTORATION:

1. Restore per Simulator Operator Checklist.

2. Once exams are complete, restore from SEI-19, Simulator Exam Security.

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OPERATIONS SHIFT TURNOVER REPORT UNIT 3 RISK: GREEN (ACCEPTABLE)

PROTECTED TRAIN: B UNIT 4 RISK: GREEN (ACCEPTABLE)

PROTECTED TRAIN: B ONCOMING CREW ASSIGNMENTS Shift Mgr:

Inside SNPO:

Field Supv.:

Outside SNPO:

Admin RCO:

ANPO:

Unit 3 Unit 4 Unit Supv.:

Unit Supv.:

RCO:

NPO:

PLANT STATUS Unit 3 Unit 4 Mode:

1 Mode:

1 Power:

100%

Power:

100%

MWe:

857 MWe:

842 Gross Leakrate:

0.01 gpm Gross Leakrate:

0.03 gpm RCS Boron Conc:

784 RCS Boron Conc:

642 Operational Concerns:

3B HHSI Pump has a wiped motor bearing, and its return to service is expected in two days. B AFW Pump is having turbine repairs, and the expected return to service is late next shift. 3A1 CWP preventive maintenance is done with the clearance released. The 3A1 CWP is ready for start for a PMT.

U3 Anticipated LCO Actions:

None U4 Anticipated LCO Actions:

None Results of Offgoing Focus Area:

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UNIT 3 STATUS REACTOR OPERATOR UNIT RISK: GREEN (ACCEPTABLE) PROTECTED TRAIN: B Mode:

1 RCS Leakrate Accumulator Ref Levels Power:

100%

Gross:

0.01GPM A 6656 MWe 857 Unidentified 0.00 GPM B

6608 Tavg:

580°F Charging Pps:

0.01 GPM C 6646 RCS Pressure:

2245 RCS Boron Conc:

784 Abnormal Annunciators:

Annunciator:

H3/2 - SAFETY INJ PUMP 3B TRIP Comp Actions:

3A, 4A, and 4B HHSI Pumps are guarded.

Annunciator:

Comp Actions:

Annunciator:

Comp Actions:

Annunciator:

Comp Actions:

Annunciator:

Comp Actions:

Annunciator:

Comp Actions:

Current Tech Spec Action Statements: (Does Not Include For Tracking Only Items T.S.A.S / Component:

3.5.2.c - 30 day action / 3B HHSI Pump Reason:

Bearing Replacement Entry Date:

Yesterday T.S.A.S / Component:

3.7.1.2.c - 30 day action / B AFW Pump Reason:

Turbine Repairs Entry Date:

Yesterday T.S.A.S / Component:

Reason:

Entry Date:

T.S.A.S / Component:

Reason:

Entry Date:

T.S.A.S / Component:

Reason:

Entry Date:

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UNIT 3 STATUS REACTOR OPERATOR (CONT'D)

UNIT RISK: GREEN (ACCEPTABLE) PROTECTED TRAIN: B Changes to Risk Significant Equipment:

No recent changes from last shift.

OLRM: GREEN PROTECTED TRAIN: B Upcoming Reactivity Management Activities:

Maintain current power.

Upcoming Major POD Activities:

NONE Upcoming ECOs to Hang and /or Release:

Hang - None Release - None Evolutions or Compensatory Actions in Progress:

NONE General Information, Remarks, and Operator Work Around Status:

Weather forecast is overcast skies with scattered pockets of severe rain.

U3 supplying Aux Steam Air In-leakage = 0.0 SCFM

This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

L-15-1 NRC SCENARIO 2 NRC Draft (Rev 11/20/14)

SEG Page 2 of 43 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.

References Revision Number 3-ARP-097.CR various 0-ADM-115 8B 0-ADM-200 21 0-ADM-211 3

0-ADM-232 4

0-EPIP-20101 16 0-EPIP-20134 3

3-EOP-E-0 10 3-EOP-E-1 5

3-EOP-F-0 3

3-GOP-100 6

3-NOP-010 16 3-NOP-071 6

3-NOP-094 3

3-ONOP-008 0A 3-ONOP-028.1 1

3-ONOP-028.2 2

3-ONOP-067 6

3-ONOP-071.2 8

3-OSP-059.10 1A PTN Technical Specs 293

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L-15-1 NRC SCENARIO 2 NRC Draft (Rev 11/20/14)

SEG Page 3 of 43 This information is controlled by PTNs 2015 LOIT (L-15-1) NRC EXAMINATION SECURITY AGREEMENT.