ML15079A121

Initial Exam 2015-301 Final Simulator Scenarios

ML15079A121
Person / Time
Site:	Browns Ferry
Issue date:	03/20/2015
From:	NRC/RGN-II
To:	Tennessee Valley Authority
Shared Package
ML15079A140	List: ML15079A074 ML15079A082 ML15079A093 ML15079A098 ML15079A104 ML15079A106 ML15079A110 ML15079A121 ML15079A125 ML15079A128 ML15079A133
References
50-259/15-301, 50-260/15-301, 50-296/15-301
Download: ML15079A121 (221)

Text

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario Number: NRC-2 Op-Test Number: 1501 Examiners: Operators: SRO:

ATC:

BOP:

Initial Conditions: Reactor Power is 100%. The Steam Vault Exhaust Booster Fan is tagged out for lubrication PMs. Suppression Pool Cooling is in service due to a HPCI flow rate test on the previous shift.

Turnover: Secure Suppression Pool Cooling. MIG signed on and will be performing 2-SR-3.3.6.1.5(4NA) Core and Containment Cooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration.

Event Malfunction Event Event Description Number Number Type*

N-BOP Secure from Suppression Pool Cooling using 1 N/A N-SRO 2-01-74 MIG reports that 2-RLY-071-13A-K12 did not I-SRO 2 N/A energize when 2-PDT-71-1A was pressurized and TS-SRO that they have stopped at step 7.4[7]B.

3 TH12B C-SRO Recirc Pump 2B vibration high R-ATC 4 TH10B C-BOP Recirc Pump 2B seal failure/2-AOl-68-1A TS-SRO Override 2B C-BOP The 2A Steam Packing Exhauster will trip and the 2B 5

SPE Auto C-SRO Steam Packing Exhauster will not auto start.

CRD Flow Control Valve FCV develops an air leak on 6 IOR I-ATC the control air system.

TH22 7 RH01A&C M-All LOCNScram with inability to spray the Drywell/C4 RH06B SRO directs cool down or rapid depressurization of C-BOP 8 TC02 the RPV using Turbine bypass valves however they C-SRO fail closed and ED will be required.

Core Spray Loop II injection valve will fail to open on 9 CS02B C-BOP initiation signal but can be manually opened.

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

Appendix D Scenario Outline Form ES-D-1 Events

1. The BOP operator will secure from Suppression Pool Cooling using 2-01-74. When Suppression Pool Cooling is secured the scenario may continue.

2. The Instrument Mechanic Foreman will call the SRO and report that during performance of 2-SR-3.3.6.1.5(4A/A), Core and Containment Cooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration, 2-RLY-071-13A-K12 did not energize when 2-PDT-71-1A was pressurized and that they have stopped at step 7.4[7]B. The SRO will evaluate Tech Spec 3.3.6.1 and table 3.3.6.1-1 to determine that function 4a is the affected function and that the Tech Spec requires placing the channel in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Once the Tech Spec call is completed the scenario may continue.

3. 2-XA-55-4B window 20, RECIRC Pump Motor B Vibration High alarms.

The BOP operator will dispatch an AUO to 2-LPNL-925-0712 and he/she will report that 2-Xl-68-71 D and E are in alarm reading 12.0 mils and rising slowly. The SRO will direct lowering 2B Recirc speed to attempt to clear the alarm. The ATC operator will lower 2B Recirc pump speed. The AUO will report that 2-Xl-68-71 D and E lowered to -10 mils and will reset the alarm locally. When the lead examiner is ready the scenario may continue.

4. 2-XA-55-4B window 20, RECIRC Pump Motor B Vibration High alarms again and 2-XA-55-4B, window 25 Recirc Pump B no. 1Seal Leakage ABN, alarms. The number 2 seal pressure will rise to approximately Reactor Pressure. The SRO will direct tripping the 2B Recirc Pump and entering 2-AOl-68-1A. The ATC operator will lower Reactor Power IAW the RCP and 2-AOl-68-1A. The BOP operator will carry out the subsequent actions of the AOI. The SRO will address Tech Spec 3.4.1.

When conditions have stabilized and the lead examiner is ready the scenario may continue.

5. The running steam packing exhauster will trip and the standby exhauster will fail to auto start. The BOP operator will place the standby steam packing exhauster in service IAW 2-0l-47C section 6.3 and adjusts its operation to obtain 10-12 inches of H20 vacuum. At that point the scenario may continue.

6. The in service 2A CRD Flow Control Valve develops a control air leak causing the valve to slowly drift closed, causing the CRD Charging Header Pressure to go high. The ATC will enter the ARP for High Charging Pressure and attempt to control flow and pressure. The FCV will not respond to any attempt to control and after the AUO is sent to investigate, he reports an air leak, and the ATC uses 01-85 to swap Flow Control

Appendix D Scenario Outline Form ES-D-1 Valves. When the 2B CRD FCV is in service and the lead examiner is ready the scenario may continue.

7. A leak in the Drywell will develop causing Drywell Temperature and Pressure to rise. The SRO will set a trigger value for a Reactor Scram and when that value is reached a manual Scram will be inserted or the Reactor will Scram at 2.45 psig Drywell Pressure. All control rods will be inserted on the scram. The SRO will direct entry into 2-AOl-100-1. The SRO will direct Suppression Chamber spray per EOl-2 Appendix 17C.

The BOP operator will attempt to spray the Suppression Chamber however the select logic will fail on both loops of RHR.

The SRO/BOP operator will determine that neither the Suppression Chamber nor the Drywell can be sprayed. The SRO may attempt to cool down or anticipate that an ED will be required and attempt to rapidly depressurize the Reactor using the bypass valves however the bypass valves will fail closed. An ED will be required based on Drywell Temperature. As the Reactor depressurizes the action required area of curve 8 RPV Saturation Temp will be entered and Reactor Water Level indication will be lost. The SRO will direct entry into C-4 and the crew will inject using available systems until one of the conditions in C-4, Note 7 is met.

8. When the SRO directs a cool down or rapid depression of the RPV using the main turbine bypass valves, the operator will determine and report that the bypass valves have failed closed. This will lead to an ED being required.

9. With an accident signal present the Core Spray loop II injection valve will fail to automatically open, the BOP operator will manually open the injection valve.

The Scenario ends when the crew has performed an emergency depressurization and flooded the RPV to the Main Steam Lines.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks: 2

1. Emergency Depressurize before 300°F.

1. Safety Significance Precludes failure of Primary Containment

2. Cues Procedural compliance High Suppression Chamber or Drywell pressure

3. Measured by Observation-SRO updates or briefs the crew that ED is required based on not being able to restore and maintain drywell temperature below 280°F AND the operator opens 6 ADS/MSRVs

4. Feedback MSRV open indications RPV Pressure lowering

2. evel instruments flash, level unknown, inject into the available sources until one of the conditions in C-4, Note 7 is met.

1. Safety Significance Prevent fuel damage by establishing adequate core cooling

2. Cues Procedural compliance Loss of all RPV level indications

3. Measured by Observation-Indications that the Main Steam Lines are flooded are listed in C-4 Note 7

4. Feedback MSRV tail pipe temperature MSRV acoustic monitor RPV Pressure trend

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _1_ Page 1 of 3 Event

Description:

Secure from Suppression Pool Cooling using 2-01-74 Time Position Applicant's Actions or Behavior Directs BOP to secure Suppression Pool Cooling in accordance SRO with 2-01-74.

2-01-74 8.6 Shutdown of Loop 1(11) Suppression Pool Cooling NOTE

1) All operations are performed at Panel 2-9-3 unless otherwise noted.

2) RHR flow should be monitored while in operation with multiple flow paths (e.g., LPCI and Suppression Pool Cooling together, etc.). During any evolution, total system flow as indicated on RHR SYSTEM 1(11) FLOW, 2-Fl-74-50(64), should remain between 7,000 to 10,000 gpm for 1 pump operation or between 10,000 and 20,000 gpm for 2-pump operation.

BOP

[1] VERIFY Suppression Pool Cooling in operation.

REFER TO Section 8.5.

[2] REVIEW the precautions and limitations in Section 3.0.

[3] NOTIFY Radiation Protection of Suppression Pool Cooling loop removed from service. RECORD name and time of Radiation Protection representative notified in NOMS narrative log.

Verifies Suppression Pooling is in service, reviews P&L's in Section 3, and notifies RP that SPC is being removed from service.

DRIVER As RP, acknowledge that SPC is being removed from service.

Unit 2 Page 1of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No.__£__ Event No.: _1_ Page 2 of 3 Event

Description:

Secure from Suppression Pool Cooling using 2-01-74 Time Position Applicant's Actions or Behavior CAUTIONS

1) To prevent draining an RHR Loop, at least one of the RHR System test valves must be closed before stopping RHR Pumps in the associated loop.

2) To prevent excessive vibration, RHR pumps should not be allowed to operate for more than 3 minutes at minimum flow.

3) When closing throttle valve RHR SYS 1(11) SUPPR POOL BOP CLGffEST VLV, 2-FCV-74-59 and 2-FCV-74-73 from the control room, the handswitch should be held in the close position for approximately 6 seconds after the red light extinguishes. Failure to completely close these valves could provide a leak path to the suppression pool from the RHR discharge piping.

[4] IF both RHR Pumps in Loop 1(11) are in operation AND one pump is to be removed from service due to reduced heat load, THEN:

[4.1] THROTTLE RHR SYS 1(11) SUPPR POOL CLGffEST VLV, 2-FCV-7 4-59(73), to obtain a flow of between 7,000 to 10,000 gpm and Blue light illuminated as indicated on RHR SYS 1(11) FLOW, 2-Fl-74-50(64).

[4.2] STOP RHR PUMP 2A(2B) or 2C(2D) using 2-HS 5A(28A) or 16A(39A).

[4.3] CLOSE associated RHR HX 2A(2B) or 2C(2D) RHRSW OUTLET VALVE, 2-FCV-23-34(46) or 40(52).

[4.4] IF RHRSW for the Heat Exchanger removed from service is not required to support other unit operations, THEN STOP RH RSW pump for the Heat Exchanger removed from service.

Unit 2 Page 2 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _1_ Page 3 of 3 Event

Description:

Secure from Suppression Pool Cooling using 2-01-74 Time Position Applicant's Actions or Behavior

[5] CLOSE RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, 2-FCV-74-59(73).

[6] WHEN RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, 2-FCV-74-59(73) is CLOSED, THEN STOP RHR PUMPS 2A(2B) or 2C(2D) using 2-HS 5A(28A) and/or 16A(39A).

[7] CLOSE RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV, 2-FCV-74-57(71 ).

[8] CLOSE RHR HX(s) 2A(2B) and 2C(2D) RHRSW OUTLET VLV(s), 2-FCV-23-34(46) and 40(52).

BOP

[9] IF RHRSW for RHR Heat Exchanger(s) A(B) and C(D) is not required to support other unit operations, THEN STOP RHRSW Pump(s) for the Heat Exchanger(s) removed from service.

[1 O] CHECK RHR System discharge header pressure is greater than TRM 3.5.4 limit as indicated on 2-Pl-74-51(65),

RHR SYS 1(11) DISCH PRESS.

[11] IF the Drywell DP Compressor was removed from service in Section 8.5, THEN REFER TO 2-01-64 to return to service

[12] WHEN 2-Tl-74-136A and B, A/C and B/D RHR PUMP ROOM TEMP indicators at Panel 2-9-3 indicate less than 95°F, THEN RETURN EECW to its normal operating configuration, if desired. REFER TO 0-01-67.

BOP NOTE RHR Loop 1(11) is now in a Standby Readiness Condition Secures Suppression Pool Cooling IAW 2-01-74 Reports that Suppression PoolinQ is shutdown NRC End of Event #1 Unit 2 Page 3 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: -2_ Page 1 of 2 Event

Description:

MIG reports that 2-RLY-071-13A-K12 did not energize when 2-PDT-71-1 A was pressurized Time Position Applicant's Actions or Behavior When the NRG Chief Examiner is ready for the second event, as the l&C Foreman, call the SRO and report the following:

During performance of 2-SR-3.3.6.1.5(4A/A), Core and DRIVER ContainmentCooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration, 2-RLY-071-13A-K12 did not energize when 2-PDT-71-1A was pressurized and that they have stopped at step 7.4[7]8.

NRC Relay is on print 2-45E626, sheet 2 Refers to Technical Specifications Primary Containment Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6.1-1.

Unit 2 Page 4 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _2_ Page 2 of 2 Event

Description:

MIG reports that 2-RLY-071-13A-K12 did not energize when 2-PDT-71-1 A was pressurized Time Position Applicant's Actions or Behavior Table 3.3.6.1-1

4. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line Flow - 1,2,3 F High CONDITION REQACTION COMP TIME A. One or more required A.1 Place a channel 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for SRO channels inoperable. in trip Functions other than 2.a,2.b,5.h, 6.b, and 6.c Determines that a channel must be place in a tripped condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Conducts a crew brief on the test failure and Technical Specifications requirements due to the failure.

NRG End of Event #2 Unit 2 Page 5 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _3_ Page 1 of 2 Event

Description:

Recirc Pump 28 vibration high Time Position Applicant's Actions or Behavior DRIVER When the NRC Chief Examiner is ready for Event #3, insert F3 (imf th12b) for 'B' Reactor Recirc Pump high vibration.

Reports RECIRC PUMP MOTOR B VIBRATION HIGH (2-XA 48, Window 20) in alarm.

2-ARP-9-48, Window 20 A. CHECK the following on RECIRC PMP MTR 28 WINDING AND BRG TEMP recorder, 2-TR-68-71 on Panel 2-9-21 are below:

• Pump motor bearing temperatures (<190°F),

• Pump motor winding temperatures (<255°F),

• Pump Seal Cavity temperatures (<180°F),

• Pump motor closed cooling water temperature (<140°F),

ATC B. CHECK for rise in Drywell equip sump pump out rate due to seal leakage.

C. DISPATCH personnel to Panel 2-LPNL-925-0712 (Vibration Mon System) on El 565' (S-R10) and REPORT the Vibration Data for Pump 28 to the Unit Operator and any other alarm indications.

The person shall advise the Unit Operator of any changes in the vibration values and Acknowledge Alarms as necessary D. IF alarm seals in, THEN ADJUST pump speed slightly to try to reset the alarm.

Dispatches personnel to Panel 2-LPNL-925-0712 to report vibration on 28 recirc pump.

Unit 2 Page 6 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _3_ Page 2 of 2 Event

Description:

Recirc Pump 28 vibration high Time Position Applicant's Actions or Behavior As RB AUO, acknowledge going to Vibration Panel on 565 and DRIVER reporting recirc pump 'B' vibration Checks Recirc Pump 'B' temperatures on Panel 2-9-21 and BOP reports all temperatures within allowable values.

After 5 minutes, call ATC operator and report 68-71 D-G are DRIVER trending up slowly.68-71D and 68-71E are indicating 12 mils.

ATC Updates crew on vibration report on 'B' recirc pump Directs ATC to lower 'B' recirc pump speed to try to get vibration SRO alarm to reset.

ATC Lowers 'B' recirc speed at least 5 rpm.

NRG Operator may using 'Lower Slow' (1 rpm) or 'Lower Med' (5rpm)

When 'B' recirc pump speed has been lowered approximately 5 rpm, report that vibration is lowering but the alarm is still sealed in.

DRIVER When speed has been lowered approximately 10 rpm, insert dmf th12b to delete the vibration condition and report the vibration is approximately 1O mils and lowering very slowly.

Briefs crew on ARP operator actions if the vibration condition SRO would occur again and was unable to be reset.

SRO may not conduct a brief on a possible recurrence of the alarm NRG End of Event #3 Unit 2 Page 7 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _L Event No.: -1_ Page 1 of 5 Event

Description:

Recirc Pump 28 seal failure/2-AOl-68-1A Time Position Applicant's Actions or Behavior When the NRG Chief Examiner is ready for Event 4, insert F3 (imf DRIVER th12b) to cause recirc pump 'B' vibration again.

Reports RECIRC PUMP MOTOR B VIBRATION HIGH (2-XA 48, Window 20) in alarm.

ATC Dispatches personnel to Panel 2-LPNL-925-0712 to report vibration on 28 recirc pump.

Reported recirc pump vibration will continue to rise.

NRG Failure of the inner recirc pump seal will occur when vibration is 15 mils.

After 5 minutes, call ATC operator and report 68-71 D - Gare rising. 69-D and 68-71E are indicating 12 mils.

DRIVER One minute later, call ATC operator and report 68-71D- Gare rising.68-71F and 68-71Gare indicating 15 mils.

Insert F4 (imf th 1Ob 70) to fail the 'B' recirc pump inner seal SRO Directs ATC to trip 'B' Recirc Pump and enter 2-AOl-68-1 SRO may choose to enter 2-01-68, Section 7.2 ((Stopping a NRC Recirc Pump (Mode 1) & Single Loop Operation).

Unit 2 Page 8 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -2_ Event No.: -1._ Page 2 of 5 Event

Description:

Recirc Pump 28 seal failure/2-AOl-68-1A Time Position Applicant's Actions or Behavior 2-01-68 7.2 Stopping a Recirc Pump (Mode 1) & Single Loop Operation CAUTIONS

1) Prior to stopping a recirc pump, all attempts should be made to evaluate where the plant power to flow conditions will end up when a recirc pump is removed from service. If practical, the control rod line should be below 95.2% before stopping a recirc pump.

2) Per Technical Specifications, the reactor CAN BE operated indefinitely with one recirc loop out of service, provided the requirements of T.S. 3.4.1 are implemented within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of entering single loop operations.

3) In Single Loop Operation (SLO), reactor power is limited to less than or equal to 50%, core flow is limited to less than or equal to 50% and Active Recirc Drive flow is limited to 46.6kGPM.

ATC NOTE When depressing the switches which control the recirc drives, these switches must be firmly depressed to ensure all the contacts are made-up.

[1] IF stopping of the recirc pump is immediately required, THEN (Otherwise Continue at Step 7 .2[2])

[1.2] IF shut down of Recirc Drive 28 is desired, THEN (Otherwise N/A):

[1.2.1] FIRMLY DEPRESS RECIRC DRIVE 28 SID, 2-HS-96-20.

[1.2.2] CHECK recirc drive decelerates to 345 RPM.

[1.2.3] CHECK recirc drive shuts down.

[1.2.4] CHECK DRIVE RUNNING, 2-IL-96-40, is extinguished.

[1.3] REFER TO 2-AOl-68-1A or 1B for further required actions.

Trips 'B' Recirc Pump and refers to 2-AOl-68-1A Unit 2 Page 9 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.:__+/-___ Page 3 of 5 Event

Description:

Recirc Pump 28 seal failure/2-AOl-68-1A Time Position Applicant's Actions or Behavior 2-AOl-68-1A 4.2 Subsequent Actions (continued)

NOTE Power To Flow Map is maintained in O-Tl-248"Station Reactor Engineer" and on ICS.

[2] IF a single Recirc Pump tripped, THEN ATC CLOSE tripped Recirc Pump discharge valve.

[3] IF Region I or II of the Power to Flow Map is entered, THEN (Otherwise N/A)

IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. Refer to O-Tl-464, Reactivity Control Plan Development and Implementation.

[9] [NER/C] WHEN conditions allow, THEN MAINTAIN operating jet pump loop flow greater than 41x10 6 lbm/hr (2-Fl-68-46 or 2-Fl-68-48). [GE SIL 517]

When 'B' Recirc Pump is tripped insert dmf th12b to delete recirc DRIVER pump 'B' vibration ATC Closes 'B' Recirc Pump discharge valve Inserts control rods to lower reactor power below the 74% loadline ATC/ AND greater than 45% core flow IAW RCP.

BOP When conditions allow and at SRO direction, raises operating jet pump loop flow > 41 x 106 Reports RECIRC PUMP 8 NO 1 SEAL LEAKAGE ABN (2-XA 48, Window 25) in alarm.

2-ARP-9-48 NOTE Annunciator Window will not reset until all Alarms are ACK on the recorder 2-XR-68-2/5 (RECIRC PUMPS DISCH FLOW & TEMP).

A. DETERMINE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 2-9-4 or ICS.

Unit 2 Page 10of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _L Page 4 of 5 Event

Description:

Recirc Pump 28 seal failure/2-AOl-68-1A Time Position Applicant's Actions or Behavior

• Plugging of No. 2 RO - No. 2 seal pressure approaches no. 1 seal pressure.

• Failure of No. 1 seal - No. 2 seal pressure is greater than 50% of ATC the pressure of No. 1.

Reports that No. 1 seal pressure and No. 2 seal pressure are approximately equal 2-AOl-68-1A 4.2 Subsequent Actions (continued)

[9] NOTIFY Reactor Engineer to PERFORM the following:

• REFER TO Tech Specs 3.4.1

• 2-SR-3.4.1 (SLO), Reactor Recirculation System Single Loop Operation

• O-Tl-248, Core Flow Determination in Single Loop Operation

[1 O] [NER/C] WHEN the Recirc Pump discharge valve has been SRO closed for at least five minutes (to prevent reverse rotation of the pump) [GE SIL-517], THEN (N/A if Recirc Pump was isolated in Step 4.2[8])

OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibrium.

Calls RE to inform them of the Recirc Pump trip Directs ATC to open 'B' Recirc Pump discharge valve after 5 minutes ATC Opens 'B' Recirc Pump discharge valve after 5 minutes Unit 2 Page 11 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _L Page 5 of 5 Event

Description:

Recirc Pump 28 seal failure/2-AOl-68-1A

'::: _: _ Applicant's Actions or Behavior Reviews Technical Specification 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

SRO single loop operation limits specified in the COLR;

c. LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation; CONDITIONS REQ. ACTIONS COMP A. Requirements of A.1 Satisfy the requirements 24 hr LCO not met of the LCO Briefs crew on the Technical Specification requirements for only one recirc pump in operation.

DRIVER As RE, acknowledge call that 'B' recirc pump was tripped due to high vibration and TS requirements for single loop operations NRC End of Event #4 Unit 2 Page 12 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _5_ Page 1 of 2 Event

Description:

The 2A Steam Packing Exhauster will trip and the 28 Steam Packing Exhauster will not auto start.

Position Applicant's Actions or Behavior When the NRC Chief Examiner is satisfied with the Technical DRIVER Specification call and ready for Event #5, insert F5 to trip 2A Steam Packing Exhauster (bat NRC/1501-2-5)

If one SPE breaker trips, indication will be lost on BOTH SPEs.

NRC This is the normal logic scheme.

Reports STEAM PACKING EXHAUSTER VACUUM LOW (2-XA-55-?A, Window 12) in alarm.

Reports SPE Fan 2A has tripped and SPE Fan 28 has failed to start.

2-ARP-9-7A, Window 12 B. IF standby blower fails to start, THEN START standby OR VERIFY normal in service. REFER TO 2-01-4 ?C.

C. IF blower is running, PERFORM the following:

1. THROTTLE in-service STEAM PACKING EXHR 2B(2A)

BOP DISCHARGE VLV, 2-HS-66-35A (2-HS-66-34A), UNTIL SPE Vacuum, as indicated on STEAM PACKING EXH VACUUM, 2-Pl-66-54, is between 10" and 12" Vacuum, or as appropriate for plant conditions. REFER TO 2-0l-47C.

2. VERIFY SJAE/OG CNDR CNDS FLOW, 2-Fl-2-42, between 2 x 106 lbm/hr and 3 x 106 lbm/hr to prevent tripping SPE due to inadequate cooling. REFER TO 2-01-2.

Starts 2B SPE Fan and throttles discharge valve until SPE Vacuum is between 10" and 12" Vacuum. Verifies proper SJAE/OG CNDR condensate flow.

Reports 2B SPE Fan in service Unit 2 Page 13of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _5_ Page 2of2 Event

Description:

The 2A Steam Packing Exhauster will trip and the 28 Steam Packing Exhauster will not auto start.

Time Position Applicant's Actions or Behavior Calls Work Control to initiate a WO on the 2A SPE Fan OR directs SRO BOP to call.

DRIVER As Work Control, acknowledge initiating a WO on the 2A SPE fan.

NRC End of Event #5 Unit 2 Page 14of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _6_ Page 1of5 Event

Description:

CRD in service Flow Control Valve failure Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event #4, insert SF4 DRIVER (bat NRC/1501-2-4) to fail the 2A CRD Flow Control Valve.

Reports (2-XA-55-5A, Window 10) in alarm and CRD parameters abnormal.

Automatic Action: None Operator Action:

A. CHECK CRD ACCUM CHARGING WATER PRESSURE HI on 1-Pl-85-13A, Panel 1-9-5. It is reading about 1500 psig.

ATC B. CHECK 1-FCV-085-0011 A or B in service. A is in service C. IF in service controller has failed, THEN REFER TO 2-01-85.

Will refer to 2-01-85 D. IF pressure is still greater than 1510 psig after verifying proper controller operation, THEN THROTTLE CRD PUMP DISCH THROTTLING VLV, 1-THV-085-0527, to maintain between 1475 and 1500 psig.

SRO Directs entry into 2-AOl-85-3 2-AOl-85-3 4.2 Subsequent Actions

[3] IF operating CRD Pump has NOT tripped, THEN PERFORM the following:(REFER TO 2-01-85) (Otherwise N/A)

ATC [3.1] VERIFY FULL OPEN CRD PUMPS 2A & 18 UNIT 2 SUCTION, 2-FCV-85-65.

[3.4] IF system flow is high or low, THEN SHIFT CRD Flow Control Valves.

Dispatches the AUO to CRD Flow Control station to investigate.

DRIVER Report back to the ATC that there is an air leak on the 'A' FCV.

Unit 2 Page 15 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _6_ Page 2 of 5 Event

Description:

CRD in service Flow Control Valve failure Time Position Applicant's Actions or Behavior SRO Directs that CRD Flow Control Valves be shifted 2-01-85 6.3 Shifting CRD Flow Control Valves

[1] VERIFY Control Rod Drive Hydraulic System in operation.

Refer to Section 5.1.

[2] REVIEW all Precautions and Limitations in Section 3.6.

NOTE

1) Erratic operation of CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, may be observed during refueling/shutdown operations ATC when larger ~Ps exists due to low reactor pressure and CRD pressure.

2) As CRD Flow Control Valves are shifting, CRD System Flow Controller should be adjusted, as needed, to maintain a constant flow.

[3] PERFORM the following for Flow Control Valve being brought into service from Reactor Bldg El 565' Dispatches AUO to perform Step 6.3[3]

[3.1] OPEN FCV-85-11 B INLET SOV, 2-SHV-085-0561.

[3.2] OPEN FCV-85-11 B OUTLET SOV, 2-SHV-085-0562

[3.3] VERIFY OPEN PCV BYPASS SOV TO FCV-85-11 B, 2 318

[3.4] CHECK OPEN PCV 85-11 SOV, 2-85-247.

DRIVER

[3.5] CHECK OPEN HOR ISOL TO FCV-85-11A & B, 2-85-313.

[3.6] CHECK FCV-85-11A THREE WAY ISOL valve handle in Horizontal position for 2-85-251.

[3. 7] CHECK FCV-85-11 B, THREE WAY ISOL valve handle in Horizontal position for 2-85-252; Report to ATC that Step 6.3[3] is complete Unit 2 Page 16 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _6_ Page 3 of 5 Event

Description:

CRD in service Flow Control Valve failure Time Position Applicant's Actions or Behavior

[4] VERIFY CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, in BAL position on outer control selector wheel.

[4.1] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, by turning Manual Control Pot inside Control Selector Wheel UNTIL red deviation pointer is in Green Band.

ATC

[4.2] TURN CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, Control Selector from BAL position to MAN position.

[5] REDUCE CRD SYSTEM FLOW using 2-FIC-85-11, to approximately 40 gpm with Manual Control Pot on 2-FIC-85-11.

Completes Steps 4 and 5. Contacts AUO to perform 6.3[6]

[6] PLACE CRD SYSTEM FLOW CV SELECTOR SW 2-XS 11, on 2-LPNL-925-0018B, to select Flow Control Valve being DRIVER brought into service, in VALVE A(VALVE 8).

Contact ATC operator and report that Step 6~3[6] is complete

[7] CHECK selected in-service valve opening and out-of-service valve closing.

ATC Observes 'A' FCV closing and 'B' FCV opening Contacts AUO to perform Step 6.3[8]

[8] PERFORM the following for Flow Control Valve being removed from service:

[8.1] CLOSE FCV-85-11A(B) INLET SOV, 2-SHV-085-0563(0561 ).

DRIVER

[8.2] CLOS.E FCV-85-11A(B) OUTLET SOV, 2-SHV-085-0564(0562).

ContactATC operator and report Step 6.3[8] is complete Unit 2 Page 17of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _6_ Page 4 of 5 Event

Description:

CRD in service Flow Control Valve failure Time Position Applicant's Actions or Behavior

[9] ADJUST CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, to establish between 40 gpm and 65 gpm.

[1 O] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, ATC by TURNING Flow Demand Thumb Wheel until Red Deviation Pointer is in Green band, AND PLACE in AUTO OR BALANCE.

Contacts the AUO to perform Step 6.3[11]

[12] VERIFY CRD DR WATER HOR FLOW, 2-Fl-85-15A, is approximately 0 gpm.

[13] ESTABLISH the following by alternately adjusting tape setpoint of CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, AND throttled position of CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A:

• CRD CLG WTR HOR DP, 2-PDl-85-18A between 10 psid and 20 psid. o ATC

• CRD DRIVE WTR HOR DP, 2-PDl-85-17A between 250 psid and 270 psid. o

• CRD SYSTEM FLOW, 2-FIC-85-11 between 40 gpm and 65 gpm.

NOTE PUMP DISCH THROTTLING valve, 2-85-527, has been set to supply 1500 psig charging water pressure and Unit Supervisor authorization is required prior to changing valve position.

Unit 2 Page 18 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _6_ Page 5 of 5 Event

Description:

CRD in service Flow Control Valve failure Time Position Applicant's Actions or Behavior

[14] IF CRD ACCUM CHG WTR HOR PRESSURE, 2-Pl-85-13A, is less than 1475 psig, OR greater than 1500 psig, THEN ATC THROTTLE PUMP DISCH THROTTLING, 2-THV-085-0527, to maintain pressure within normal operating range of between 1475 psig and 1500 psig, as indicated on 2-Pl-8-13A. NA NRC End of Event #6 Unit 2 Page 19 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _]_ Page 1 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior DRIVER When NRC Chief Examiner is ready for Event 7, insert F7 (imf th22 100 20:00) bottom head leak.

Reports Drywell Pressure rising slowly Continues to monitor containment parameters Reports PRI CONTAINMENT N2 PRESS HIGH (2-XA-55-38, Window 10) in alarm Reports Drywell Pressure 1.5 psig and rising slowly 2-ARP-9-38 A. CHECK containment pressure using multiple indications:

B. CHECK containment temperature.

BOP C. REFER TO 2-01-64, Venting the drywell with standby gas treatment fan.

Verifies drywell pressure rising using other indications Reports containment temperature rising slowly Reports DRYWELL NORM OPERATING PRESS HIGH (2-XA-55 38, Window 19) in alarm Reports Drywell Pressure 1.6 psig and rising slowly 2-ARP-9-38 G. IF Drywell pressure is high, THEN REFER TO 2-AOl-64-1.

Unit 2 Page 20of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 2of21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior SRO Directs BOP to vent the drywell IAW 2-AOl-64-1 2-AOl-64-1 4.2 Subsequent Actions

[2] IF Drywell Pressure is High, THEN PERFORM the following: (Otherwise N/A)

BOP [2.4] ALIGN and START additional Drywell coolers and fans as necessary. REFER TO 2-01-64.

CAUTION Stack release rates exceeding 1.4 X 107 µci/sec, or a Sl-4.8. B.1.a.1 release fraction above one will result in ODCM release limits being exceeded

[2.5] VENT Drywell as follows:

[2.5.1] CLOSE SUPPR CHBR INBD ISOLATION VLV 2-FCV-64-34 (Panel 2-9-3).

[2.5.2] VERIFY OPEN, DRYWELL INBD ISOLATION VLV, 2-FCV-64-31 (Panel 2-9-3).

[2.5.3] VERIFY 2-FIC-84-20 is in AUTO and SET at 100 scfm (Panel 2-9-55).

[2.5.4] VERIFY RUNNING a Standby Gas Treatment Fan STGTS TRAIN C(A)(B) (Panel 2-9-25).

[2.5.5] IF required, THEN BOP REQUEST Unit 1 Operator to START Standby Gas Treatment Fans A or B. (Otherwise N/A)

[2.5.7] PLACE 2-FCV-84-20 CONTROL DW/SUPPR CHBR VENT, 2-HS-64-35, in OPEN (Panel 2 3).

Calls Unit 1 operator and requests that a standby gas train be started.

Vents the drywell IAW 2-AOl-64-1 Unit 2 Page 21of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No.__£__ Event No.: _7_ Page 3 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior When NRC Chief Examiner is satisfied with AOI actions, insert DRIVER SF7 (bat NRC/1501-2-7) steam leak in the drywell and a failure of RHR select switches.

Reports drywell pressure and temperature continuing to rise Reports DRYWELL PRESS APPROACHING SCRAM (2-XA 3B, Window 30) in alarm BOP 2-ARP-9-38 A.. CHECK containment pressure and temperature using multiple indications.

Continues to monitor and report containment parameters Briefs crew on current plant conditions.

Directs that a reactor scram be initiated at a point prior to the SRO automatic scram (line in the sand)

Calls line management and Load Dispatcher to inform them of upcoming scram DRIVER Acknowledge call concerning upcoming scram ATC Reports that drywell pressure is at the value directed by SRO Unit 2 Page 22of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _J_ Page 4 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior Initiates a manual reactor scram Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS

[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS-99-5A/S3B, on Panel 2-9-5.

[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A)

[3] Refuel Mode One Rod Permissive Light check:

[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1 I in REFUEL.

[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-Xl-85-46.

ATC

[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-Xl-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In.

(Otherwise N/A)

[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN.

[5] REPORT the following status to the US:

• Reactor Scram

• Mode Switch is in Shutdown

• "All rods in" or "rods out"

• Reactor Water Level and trend (recovering or lowering)

• Reactor pressure and trend

• MSIV position (Open or Closed)

• Power level Completes 'hard card' actions and makes scram report Unit 2 Page 23of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 5 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior Enters EOl-1 on RPV Water Level below +2" Verify RX scram - CHECKED RC/Q Monitor and Control reactor power - CHECKED The reactor is subcritical AND NO boron has been injected THEN EXIT RC/Q and ENTER AOl-100-1, Reactor Scram - CHECKED Directs ATC to enter 2-AOl-100-1 RC/L CAUTION: Ambient temp may affect RPV water level indication and trend - CHECKED MONITOR and CONTROL RPV water level - CHECKED SRO VERIFY each as required

• PCIS isolations (Groups 1, 2, and 3) - CHECKED It has NOT been determined that the reactor will remain subcritical without boron under all conditions - SUBCRITICAL RPV water level CANNOT be determined - CAN PC water level CANNOT be maintained below 105 ft - CAN OR Suppr chmbr press CANNOT be maintained below 55 psig - CAN CAUTION

1. 2 Pump NPSH and Vortex Limits

1. 3 Elevated suppr chmbr press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F - CHECKED Unit 2 Page 24 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __£__ Event No.: _7_ Page 6 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior RESTORE and MAINTAIN RPV water level between +2 in and

+51 in with ANY of the following CNDS and FW 5A CRD 58 RCIC with CST suction if available 5C HPCI with CST suction if available 50 CNDS 6A cs 60,6E LPCI 68, 6C Directs ATC to restore and maintain reactor level between +2 in and +51 in using Condensate and Feedwater in accordance with Appendix 5A RC/P OW press is above 2.4 psig - NO SRO Emergency RPV depressurization is anticipated - NO AND The reactor will remain subcritical without boron under all conditions Emergency RPV depressurization is or has been required - NO RPV water level CANNOT be determined - CAN Is ANY MSRV cycling - NO Steam cooling is required - NO Suppr pl temp and level CANNOT be maintained in a safe area of Curve 3 at the existing RPV press - CAN Suppr pl level CANNOT be maintained in the safe area - CAN of Curve 4 Unit 2 Page 25 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _7_ Page 7 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior OW control air becomes unavailable - AVAIL Boron injection is required - NO AND The main condenser is available AND SRO There has been no indication of a steam line break STABILIZE RPV press below 1073 psig with the main turbine bypass valves (APPX 8B)

Directs BOP to control reactor pressure 800 psig to 1000 psig in accordance with Appendix 8B Verifies that Main Turbine Bypass valves are controlling reactor BOP pressure 800 psig to 1000 psig.

Verifies and reports successful Gr. 2, 3, 6, and 8 PCIS Isolations 2-EOI APPENDIX-SA

1. IF ............. It is desired to use a reactor feed pump that is in operation, THEN ..... .CONTINUE at step 12 to control the operating pump

12. SLOWLY ADJUST RFPT speed UNTIL feedwater flow to the RPV is indicated, using ANY of the following methods on Panel 2-9-5:

• Individual 2-HS-46-8A(9A)(1 QA), RFPT 2A(2B)(2C) SPEED ATC CONT RAISE/LOWER switch in MANUAL GOVERNOR, OR

• Individual 2-SIC-46-8(9)(10), RFPT 2A(2B)(2C) SPEED CONTROL in MANUAL, OR

• 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL with individual 2-SIC

13. ADJUST RFPT speed as necessary to control injection using the methods of step 12.

Unit 2 Page 26 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 8 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 11 Time Position Applicant's Actions or Behavior

14. WHEN ....... RPV level is approximately equal to desired level AND automatic level control is desired, THEN ........... PLACE 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 2-SIC 8(9)(10), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO Restores reactor water level to +2 in to +51 in IAW App. 5A 2-AOl-100-1 ATC 2.0 SUBSEQUENT ACTIONS:

[2] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.

[3] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 2-9-5.

[4] MONITOR and CONTROL Reactor Water Level between

+2" and +51", or as directed by US, using RFP/RFPT.

[5] RETURN to body of procedure at step 4.2[5] AND CONTINUE with actions as required.

Performs AOl-100-1 subsequent actions 2-AOl-100-1 4.2 Subsequent Actions (continued)

VERIFY TRIPPED the Main Turbine as follows:

[9.1] DEPRESS the TRIP pushbutton, 2-HS-47-67D on Panel 2-9-7. [NER/C] [INPO SOER 81-015]

[9.2] PERFORM the following as required to VERIFY OPEN GENERATOR PCB 224:

[9.2.1] CHECK green light illuminated and red light not illuminated above handswitch GENERATOR BOP PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A.

[9.3] IMMEDIATELY PLACE VOLTAGE REGULATOR START/STOP SEL, 2-HS-57-24, to STOP and release.

[9.4] CHECK the following at 2-HS-57-24:

• GREEN light illuminated

• RED light extinguished Unit 2 Page 27of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: J_ Page 9 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior

[10] MONITOR Main Turbine Vibration on TURBINE GENERATOR VIBRATION, 2-XR-47-15, during coast down.

[11] ADJUST TURBINE OIL TEMPERATURE CONT, 2-TIC-24-75, setpoint to 85°F.

BOP

[12] WHEN turbine speed is less than 900 RPM, THEN START the following:

• TURBINE BEARING LIFT OIL PUMPS

• MOTOR SUCTION PUMP

• AC TURNING GEAR OIL PUMP ATC/BOP Updates crew on drywall pressure continuing to rise and reaching 2.45 psig - EOl-2 entry condition BOP Reports start and injection of HPCI into the RPV.

Verifies HPCI not needed for RPV level control and directs BOP to SRO trip and lockout HPCI.

Trips HPCI by holding the trip pushbutton depressed until turbine BOP speed in zero then taking HPCI Aux Oil Pump sw. to PTL.

Enters EOl-2 and re-enters EOl-1 on High Drywall Pressure.

EOl-2 SAMG entry is required and the TSC SAM team has assumed command and control - NO SAMG entry is required - NO CAUTION

1. 4 PC press vs pump NPSH - CHECKED SRO OW sprays have been initiated - NO Suppr chmbr sprays have been initiated - NO PC water level CANNOT be restored and maintained below 105 ft OR Suppr chmbr press CANNOT be restored and maintained below 55 psig - CAN CAUTION

1. 2 Pump NPSH and Vortex Limits - CHECKED Unit 2 Page 28of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 10 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior PC/P MONITOR and CONTROL PC press below 2.4 psig using the Vent system (AOl-64-1) - CHECKED WHEN PC press CANNOT be maintained below 2.4 psig - CHECKED BEFORE suppr chmbr press rises to 12 psig SRO CONTINUE - CHECKED CAUTION

1. 2 Pump NPSH and Vortex Limits INITIATE suppr chmbr sprays using only pumps NOT required to assure adequate core cooling by continuous inj (APPX 17C)

Directs BOP to initiate Suppression Pool Spray IAW App 17C 2-EOI APPENDIX-17C

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF ..... Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN ... BYPASS LPCI injection valve open interlock as necessary:

• PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BOP BYPASS SEL in BYPASS.

• PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

5. INITIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF ..... EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, Unit 2 Page 29of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page11of21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time I Position Applicant's Actions or Behavior THEN ... PLACE keylock switch 2-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 2-XS-74-121(129), RHR SYS 1(11)

CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF ..... 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN ... VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11)

BOP OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spray.

When aligning RHR for suppression pool spray, discovers that both the SELECT amber light and the 2/3 Core Height light will not illuminate when attempting the line up to spray.

Reports failure to SRO Directs using the other loop of RHR to spray the torus SRO Calls WC to investigate the RHR select logic When aligning the other loop of RHR for suppression pool spray, discovers that the SELECT amber light will not remain illuminated when the "SELECT" switch is taken to "SELECT" BOP Reports that Suppression Pool Spray cannot be initiated in either loop of RHR DRIVER As WC, acknowledge initiating WO to investigate and repair problem with RHR SELECT logic.

Reports that drywell pressure and temperature are continuing to CREW rise.

Unit 2 Page 30 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __£__ Event No.: _7_ Page 12 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior DW/T CAUTION

1. 1 Ambient temp may affect RPV water level indication and trend - CHECKED MONITOR and CONTROL DW temp below 160°F using available DW cooling - CHECKED WHEN DW temp CANNOT be maintained below 160°F - CHECKED OPERATE all available DW cooling - CHECKED BEFORE OW temp rises to 200° EOl-1, RPV Control at Step RC SCRAM BEFORE OW temp rises to 280°F CONTINUE SRO Is suppression pool level below 19 ft. - YES Is DW temp within the safe area of Curve 5 - YES SHUT DOWN Recirc pumps and DW blowers Directs ATC to Shutdown 'A' Reactor Recirc pump Directs BOP to shutdown DW blowers CAUTION

1. 2 Pump NPSH and Vortex Limits - CHECKED INITIATE DW sprays using only pumps NOT required to assure adequate core cooling by continuous injection (APPX17B) -

CANNOT ATC Reports 'A' recirc pump secured BOP Reports DW blowers secured ATC/BOP Reports drywell temperature and pressure continuing to rise.

Unit 2 Page 31of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.:_]_ Page 13of21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior IF the SRO decides to Rapidly Depressurize the RPV, this would NRC initiate Event No. 8 located on page 42 of 45 EOl-2, DW/T WHEN OW temp CANNOT be restored and maintained below 280°F EMERGENCY RPV DEPRESSURIZATION IS REQUIRED(EOl-1, RC/P-4; C1-1 C1-20; CS-12, CS-14)

I Updates crew that Emergency Depressurization is required EOl-1, RC/L IF Emergency RPV depressurization is or has been required THEN EXIT RC/P and ENTER C2, Emergency RPV Depressurization SRO SAMG entry is required and the TSC SAM team has assumed command and control - NO RPV water level CANNOT be determined - CAN Containment water level CANNOT be maintained below 44 ft-CAN OW control air becomes unavailable - AVAIL Will the reactor remain subcritical without boron under all conditions - YES Is OW press above 2.4 psig - YES PREVENT inj from ONLY those CS and LPCI pumps NOT required to assure adequate core cooling (Appx 4)

Direct ATC/BOP to terminate and prevent feedwater, condensate, CS and LPCI.

2-EOI APPENDIX-4

6. PREVENT injection from CONDENSATE and FEEDWATER by performing the following:

ATC

a. IF............. immediate injection termination from a reactor feedwater pump is required, THEN ........... PERFORM step 6.d for the desired pump.

Unit 2 Page 32 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: 7 Page 14 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4

' T *1me Position Applicant's Actions or Behavior

b. LOWER RFPT 2A(2B)(2C) speed to minimum setting (approximately 600 rpm) using ANY of the following methods on Panel 2-9-5:

• Using 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL AND individual 2-SIC-46-8(9)(10), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO, OR

• Using individual 2-SIC-46-8(9)(10), RFPT 2A(2B)(2CSPEED CONTROL in MANUAL, OR

• Using individual 2-HS-46-8A(9A)(1 QA), RFPT 2A(2B)(2C)

SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR.

ATC c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:

• 2-FCV-3-19, RFP 2A DISCHARGE VALVE

• 2-FCV-3-12, RFP 2B DISCHARGE VALVE

• 2-FCV-3-5, RFP 2C DISCHARGE VALVE

• 2-LCV-3-53, RFW START-UP LEVEL CONTROL

d. TRIP RFPTs as necessary to prevent injection by DEPRESSING the following push-buttons:

• 2-HS-3-125A, RFPT 2A TRIP

• 2-HS-3-151A, RFPT 2B TRIP

• 2-HS-3-176A, RFPT 2C TRIP.

Terminates and prevents condensate and feedwater IAW App 4 Reports when complete NOTE Following receipt of a CORE SPRAY automatic initiation signal, it is NOT necessary to wait until a pump starts before performing step 3.

BOP

3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.

Unit 2 Page 33of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 15 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior

4. PREVENT injection from LPCI SYSTEM I by performing the following:

NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.

a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 3-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE.

5. PREVENT injection from LPCI SYSTEM II by performing the BOP following:

NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 3-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

Terminates and prevents CS/LPCI IAW App. 4 Reports when complete Unit 2 Page 34of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _J__ Page 16 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior C-2 Is suppression pool level above 5.5 ft - YES OPEN all ADS Valves SRO (ok to exceed 100°F/hr cooldown rate)

Directs BOP to open all ADS valves Opens all ADS valves, verifies open using alternate indications BOP and reports 6 ADS valves open CRITICAL TASK Emergency Depressurize before 300°F.

1. Safety Significance Precludes failure of Primary Containment

2. Cues Procedural compliance High Drywell temperature

3. Measured by Observation-SRO updates or briefs the crew that ED is required based on Drywell temperature AND the operator opens 6 ADS/MSRVs

3. Feedback MSRV open indications RPV Pressure lowering This Critical Task is not met if the Crew does not ED by 300°F.

Unit 2 Page 35 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _J__ Page 17 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior EOl-1, RC/L RESTORE and MAINTAIN RPV water level between +2 in. and

+51 inches with ANY of the following:

RCIC with CST suction if available 5C HPCI with CST suction if available 50 SRO CS 60, 6E LPCI 6B, 6C Directs BOP to restore and maintain RPV water level between +2 in. and +51 in. using RCIC/HPCI, Appendices 5C/5D. Supplement with CS/LPCI, Appendices 6B, 6C, 60, and 6E.

Insert Shift F2 (cs02b) to prevent auto opening of CS II inboard DRIVER injection valve and Shift F3 (imf th34f) for flashing all reference legs.

When reactor pressure lowers below 450 psig, the operators will NRC verify that LPCl/CS injection valves open. This will initiate Event No. 9 located on page 43 of 45.

Coordinate restoring RPV water level to +2 in. to +51 in. using RCIC/HPCl/CS/LPCI ATC/BOP Trends condenser hotwell level and trips condensate pumps before pumps experience cavitation.

ATC/BOP Report isolation of RCIC and HPCI on low reactor pressure Briefs crew on current plant conditions including the potential to SRO flash RPV level reference legs.

Monitors RPV water level using all instrumentation Reports indication of 'notching' on RPV water level ATC/BOP instrumentation Reports RPV water level is 'unknown' Unit 2 Page 36 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 18 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior EOl-1, RC/L IF THEN RPV water level CANNOT be EXIT RC/L and determined ENTER C4, RPV Flooding SRO C-4 CAUTION

1. 1 Ambient temp may affect RPV water level indication and trend IF RPV water level can be determined - CANNOT AND It has NOT been determined that the reactor will remain subcritical without boron under all conditions - SUBCRITICAL RPV water level can be determined - CANNOT AND The reactor will remain subcritical without boron under all conditions SRO PC water level CANNOT be maintained below 105 ft - CAN OR Suppr chmbr press CANNOT be maintained below 55 psig- CAN OW control air becomes unavailable-AVAILABLE IF THEN The reactor will remain subcritical without boron under all conditions B>

Is suppr pl level above 5.5 ft. - YES OPEN all ADS vlvs - OPEN SRO (ok to exceed 100°F/hr cooldown rate Can 6 ADS vlvs be opened - YES IF THEN It has been determined that the RPV has been flooded to the main steam D>

lines Unit 2 Page 37 of 44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 19 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior Can any MSRV be opened - YES ISOLATE ANY of the following NOT needed for RPV injection Injection

• Main steam lines

• Main steam line drains SRO

• HPCI

• RCIC Directs BOP to isolate/verify isolated Main Steam lines, Main Steam line drains, HPCI and RCIC Verifies main steam lines, main steam line drains, HPCI, and BOP RCIC are isolated.

Reports isolated CAUTION

1. 2 Pump NPSH and Vortex Limits

1. 3 Elevated suppr chmbr press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F SRO FLOOD the RPV to the elevation of the main steam lines with the following:

CRD 5B CNDS 6A LPCI 6B, 6C SRO cs 6D, 6E Directs flooding the RPV to the steam lines using available sources of injection Inject to the RPV using all available sources of injection ATC/BOP Report that RPV pressure is rising Report that MSRV have reopened C-4 Can ANY MSRV be opened - YES SRO NOTE?

Unit 2 Page 38of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 20 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior Indication that the RPV is flooded to the main steam lines may include ANY of the following:

• Tail pipe temperatures of open SRVs decrease to subcooled values.

• Rising RPV pressure as noncondensibles are compressed.

SRO

• Actuation of SRV tail pipe acoustic monitors.

• MSRVs re-open and stay open at RPV pressures below 50 psig above suppression chamber pressure due to the head of water above the MSRVs. (MSRVs may open and close sluggishly if the discharge flow is subcooled.)

Monitors for indications that the RPV is flooded Determines that the RPV is flooded C-4 IF It CANNOT be determined that the RPV Is flooded to the Main Steam Lines - CAN SRO ISOLATE the following: - ISOLATED

• Main steam lines

• Main steam line drains

• HPCI

• RCIC CONTROL RPV inj as low as practicable to maintain the steam lines flooded Directs controlling RPV injection as low as practicable to maintain the steam lines flooded ATC/BOP Control RPV injection as directed.

Unit 2 Page 39of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _7_ Page 21 of 21 Event

Description:

LOCA/Scram with inability to spray the Drywell/C4 Time Position Applicant's Actions or Behavior After all RPV level instruments flash (level unknown), inject into the RPV with vailable sources until one of the conamons in C-4, ote 7 is met.

1. Safety Significance Prevent fuel damage by establishing adequate core cooling '

2. Cues Procedural compliance Loss of all RPV level indications

3. Measured by Observation-Indications that the Main Steam Lines are flooded are listed in C-4 Note 7

4. Feedback MSRV tail pipe temperature MSRV acoustic monitor RPV Pressure trend This Critical Task is not met if the Crew fails to continue to raise level and maintain pressure on the vessel.

NRC End of Event #7 and scenario Place simulator in FREEZE upon direction of the NRC Chief DRIVER Examiner Unit 2 Page 40of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _8_ Page 1 of 2 Event

Description:

Rapid depressurization of the RPV using Turbine bypass valves Time Position Applicant's Actions or Behavior When SRO makes the decision to rapidly depressurize the RPV, DRIVER insert SHIFT F1 (imf tc02 0) to fail the main turbine bypass valves closed EOl-2, DW/T WHEN OW temp CANNOT be restored and maintained below 280°FDW EMERGENCY RPV DEPRESSURIZATION IS REQUIRED(EOl-1, RC/P-4; C1-1, C1-20; C5-12, C5-14)

Determines that OW temperature will exceed 280°F and won't be able to restore and maintain below 280°F EOl-1, RC/P SRO IF Emergency RPV depressurization is anticipated AND The reactor will remain subcritical without boron under all conditions THEN RAPIDLY DEPRESSURIZE the RPV with the main turbine bypass vlvs (ok to exceed 100°F/hr cooldown Directs BOP to rapidly depressurize the RPV with the main turbine bypass valves.

Unit 2 Page 41of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ___£__ Event No.: _8_ Page 2 of 2 Event

Description:

Rapid depressurization of the RPV using Turbine bypass valves Time Position Applicant's Actions or Behavior 2-01-47 8.0 INFREQUENT OPERATIONS 8.1 Bypass Valve Operation NOTES BOP 1) Section 8.1.1 uses the RAISE, 2-HS-47-130B and LOWER, 2-HS-4 7-130A for the Bypass Jack to open and close the Bypass Valves.

2) When using the Jack to operate the Bypass valves, the valves will stay open at the desired position until closed by the LOWER, 2-HS-47-130A pushbutton. EHC Auto cooldown will not close the Bypass Valves, if the BPV DEMAND is greater than 0%

8.1.1 Using Bypass Valve Demand

[1] To open Bypass valves DEPRESS the BPV demand RAISE, 2-HS-47-130B pushbutton to slowly open the Bypass Valves until the desired number of Bypass valves are open.

BOP Attempts to open the Main Turbine Bypass valves to rapidly depressurize the RPV.

Reports that the Main Turbine Bypass valve will not open.

Calls WC to initiate troubleshooting on the bypass valves NRG End of Event #8 Unit 2 Page 42of44

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _2_ Event No.: _9_ Page 1 of 1 Event

Description:

Core Spray Loop II injection valve will fail to open on initiation signal but can be manually opened Time Position Applicant's Actions or Behavior When reactor pressure lowers below 450 psig, checks that LPCI and CS injection valves automatically open.

BOP Reports that CS II inboard injection valve (75-53) failed to automatically open.

Opens CS II inbd injection valve (75-53)

BOP Reports that CS II inbd injection valve is open NRC End of Event #9 Unit 2 Page 43of44

Appendix D Required Operator Actions Form ES-D-2 SIMULATOR SETUP IC 28 Exam IC 95 Batch File or pref/NRC/1501 NRC2 Pref File Malfunctions Description Event Delay Severity Ramp Initial Final

1. value value TH12B RR Pump B vibration 3 N/A N/A NIA False True TH10B RR Pump Seal 1 Fail 4 NIA N/A N/A False True TH22 LOCA inside OW 7 N/A 100 20:00 False True Containment Spray RH06A Active N/A N/A N/A True True Loqic failure A Loop Containment Spray RH06B Active N/A N/A N/A True True Logic failure B Loop Main Turb. Bypass TC02 8 N/A N/A N/A False True Vlvs fail to open Core Spray Loop 2 CS028 Injection vlvs fail to Active N/A N/A N/A True True auto open.

Remotes Description Event Delay Severity Ramp Initial Final

1. value value Overrides Description Event Delay Severity Ramp Initial Final

1. value value Override 28 SPE Auto 5 N/A N/A N/A False True Override CRD FCV to 6 N/A N/A N/A False True simulate air leak.

Batch I Pref File(s): NRC/1501 NRC2 bat NRC/1501-2-4 CRD FCV 11 Air leak bat NRC/1501-2-5 SPE Trip and other not auto start bat NRC/1501-2-7 LOCA/Scram with inability to spray the Drywell/C4 Unit2 Page 44 of 44

SHIFT MANAGER TURNOVER EQUIPMENT OOS/LCOS

• Steam Vault Exhaust Booster Fan is tagged out for lubrication PMs.

ANTICIPATED OPERATIONS/MAINTENANCE FOR ONCOMING SHIFT

• Secure Suppression Pool Cooling IAW 2-01-74

• Support MIG performance of 2-SR-3.3.6.1.5(4A/A) Core and Containment Cooling Systems RCIC Turbine Steam Line High Flow Instrument Channel A Calibration.

NO HEATERS ARE TO BE PLUGGED INTO OUTLETS IN THE CONTROL ROOM UNTIL DETERMINED IF WE CAN USE SPP 7.3 RISK REVIEW DOCUMENTATION ON P DRIVE FOLDER WCC-RISK/BFN EXPECTATION FOR 1500 MEETING IS FOR THE SM TO COVER CURRENT STATUS ON ALL CONTROL ROOM DEFICIENCIES AND ANNUNCIATOR WOs THAT SHOULD BE WORKING. THIS EXPECTATION STARTED 3/26/09 DO NOT DISABLE ANY ANNUNCIATORS WITHOUT GOING THRU MANAGEMENT REVIEW.

IF ALTERNATE HEAT BALANCES ALARM COMES IN ON ANY UNIT'S ICS, REDUCE POWER BY 10 MWT (3448 IF AT RATED POWER) AND CALL REACTOR ENGINEERING IN TO PLANT TO EVALUATE.

STATUS WORK WEEK - DIV I !8], DIV II 0 OUTSIDE AIR TEMP 45°F COMMON UNIT 1

• 100%

UNIT2

• 100%

• Suppression Pool Cooling in service due to a HPCI Flow Rate test last shift UNIT3

• 100%

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U2 SIMULATOR Urgent Load Reduction/Recirculation Pump Trip Contingency Plans Operations Plan

IUnit BFN 2

Reactivity Control Plan Attachment 7 (Page 1of5)

Reactivity Control Plan Form BFN Unit:~ Valid Date(s): (l)""~ Reactivity Maneuver Plan#: U2 SIMULATOR Are Multiple Activations Allowed: Yes {If yes, US may make additional copies)

Prepared by: ~~ev- /W{~ Reviewed by:Q,t:...5~ I* t:.v~

Refi'ctor Engineer Date Qualified Reactor E~gineer Date Approved by: J(..s rY~.r jCv'...-r*..-Vi-Concurrence: WCL- A~~ L c.1c<~

R.b. na r Date "1.i'V'CC/Risk/Us SRO Date Approved by: 0. ~ I cv~r~ Authorized by: S", rYJ~ I w.r-r~

Ops ~R Supt. Date Shift Manager Date RCP Activated: I .!/.!dLs Date RcP Terminated:

Unit Supervisor

/ _ __

Date Title of Evolution: Ur ent Load Reduction/Recirculation Pum Plan Purpose/Overview of Evolution: Maneuver Reactor to maintain acceptable operating conditions during and following unexpected plant conditions. These Urgent Load Reduction steps do not support a

. shutdown using the Improved BPWS Control Rod Insertion Process.

Maneuver Steps 1

Recirculation Pump Trip:

1. IF both pumps trip, SCRAM the reactor per 2-AOl-68-1A.

For a single recirculation pump trip, IMMEDIATELY insert rod groups to lower rod line below 74%

using Shove Sheets.

2. Change speed of the Operating recirculation pump, as needed, until Core Flow is 46-49% to exit POWER/FLOW region II and Operating Recirculation pump drive flow is below 46,610 gpm.

Urgent Load Reduction:

1A Lower power by lowering core flow until either the desired operating power level is reached or core flow is -60% of rated core flow. (If a recirculation pump trip is imminent, only lower using the pump to be tripped).

2A. Insert control rods per Shove Sheets to lower power until either the desired operating power level is reached or until Load Line is between 55% and 66% .

. 3A. Lower recirculation pump speed to -28% (-480 rpm).

• 4A. Insert control rods per Shove Sheets until desired ower level is reached or SCRAM the reactor.

IUnit2 BFN Reactivity Control Plan Attachment 7 (Page 2 of 5)

Reactivity Control Plan Form Reactivity Maneuver Plan #U2 SIMULA TOR Operating Experience ANO General Issues:

Operating Experience Perry Nuclear Plant scrammed on OPRM signals -9 minutes after receiving a dual recirculation pump shift to slow speed, before control rods were inserted.

CONCLUSIONS/RECOMMENDATIONS: In the event of a Recirculation pump trip or both pumps running back to minimum speed, the control rod insertion should be accomplished as soon as practical due to potential OPRM trip.

General Issues This should be considered a guideline, not covering all possible situations. Contact On-Call Reactor Engineer as soon as possible to provide assistance.

Note that in the event of a Recirculation Pump trip the Load Line may go over 113.6% and MFLCPR may exceed 1.0 but the insertion of the recommended control rod groups will return Load Line and MFLCPR below the required limits.

If a Powerplex case runs in the transient condition, the case is considered invalid. An automatic case may be triggered in the event of a recirculation pump trip and that case should be considered invalid and another case requested as soon as reactor power stabilizes.

ff shutdown is required while operating in single roop with the OOS loop isolated, Scram unit prior to going to minimum flow on operating recirculation pump to avoid flow stratification concerns. Recommend scramming between 40 to 45% RTP.

Cautions/Error Likely Situations/Special Monitoring Requirements/Contingencies:

During SLO, reactor power and flow conditions will be near to Region 2. Operation in SLO requires the following conditions to be met:

• Reactor power ,:::50% of rated power

• Core Flow _:::50% of rated core flow

• Operating pump drive flow .::;46.610 GPM Operating pump speed ,:::100% of rated speed

• POWERPLEX using a SLO thermal limit set

• MFLCPR <0.94 The RE must be contacted to change the PowerPlex calculation using the correct Thermal Limit set while in SLO. Th RE must also change the Thermal Limit set if the plant is operating with reduced feedwater heating.

In the event of change in Feedwater heating, then evaluate Final FW Temperature versus Core Thermal Power graph to determine if change required to Fuel Thermal Limits. Graph may be found on ICS as FWTCTP. in Appendix S of O-Tl-248 or in this Attachment.

In the event of change in steam path to turbine (such as a MSIV or turbine valve malfunction). then evaluate Dome Pressure versus Fraction Rate Thermal Power to determine if operation is within limits. Graph may be found on ICS as RXPCTP, in Appendix S of O-Tl-248 or in this Attachment.

IBFN Unit2 Reactivity Control Plan Attachment 7 (Page 3 of 5)

Reactivity Control Plan Form Reactivity Maneuver Plan #U2 SIMULATOR The expectation criteria shown below should be used to provide guidance on when it will be required to contact the On-Call Reactor Engineer. Document discussion and conclusions on Attachment 8 in the Comments/Notes section.

Parameter Expectation (from Prediction)

MFLCPR major change+/- 0.07, minor change** ::i::0.03 MAPRAT major change +/-0.15, minor change** +/- 0.05 MFLPD I MFDLRX major change+/- 0.15, minor change**+/- 0.05 Load Line +/- 5% of predicted Power Level +/- 5% Rated Thermal Power Core Flow +/- 5% of predicted

• Major power change - when the entire RCP **Minor power change -when the entire RCP involves a CTP change> 15% involves a CTP change 0 - 15%

IUnit2 BFN Reactivity Control Plan Attachment 7 (Page 4 of 5)

Reactivity Control Plan Form Reactivity Maneuver Plan #U2 SIMULATOR

--1 t---- l 11.--__., i 1

_..-i .- I I

i 3 I

I I

l_.....-

I I

!-- - -- - - - ; A

.~

JI_..,--

~

- l_,

.~

..JV"_... I 1--- ~~

I c:::.::

.. I,,_,./

I

~ ~

___.., i.--- I

)---- L---- ----- I I

~ l v------- Normal FW Heating - Minimum Normal FW Heaiing - Nominal

.....- Last Stage FW Heaters Out - Minimum_

- - Last Stage FW Heaters Out - Target

___,, ,..._Start-up \Vilh Admin Thermal limits

__, 5 25

__:)

- ~

50 55 t50 I

I Core Them1al POWl!r (%RTP) 7'J ao j

25 I

lCQ

• IF >50% RTP AND below line #4, (Last Stage FW Heaters Out - Minimu.m), OR <50% RTP AND more than 5°F below line #5. THEN enter Tech Spec thermal limit LCOs.

• IF <50% RTP and between lines #4 AND #5, (Startup without FW Heaters), THEN ensure the following:

o The CMSS limit set used includes FHOOS1 (and TBVOOS, if needed) .

o Redline Load Line does not exceed 100%.

c:o Power remains less than 50%.

• IF FVV temperature is below line #2. (Normal FW Heating - Minimum). THEN implement a Feedwater Heater OOS limit set (FHOOS) appropriate for current core exposure.

• IF <50% RTP and up to 5°F below line #5. (Startup without FW Heaters), THEN ensure one of the following:

o The CMSS limit set used includes FHOOS2 (and TBVOOS. if needed).

Rodline Load Line does not exceed 95%.

0 Power remains less than 50%.

\ BFN Reactivity Control Plan Unit 2 Attachment 7 (Page 5 of 5)

Reactivity Control Plan Form Reactivity Maneuver Plan #U2 SIMULATOR When operating at or above line #2 (Normal FW Heating - Minimum) of the Feedwater Temperature Versus Core Thermal Power graph, Reactor Dome Pressure is to be maintained between the Nominal Upper Pressure Limit and the Nominal Lower Pressure Limit.

When operating below line #2 (Normal FW Heating - Minimum) of the Feedwater Temperature Versus Core Thermal Power graph, then Reactor Dome Pressure is to be maintained between the Nominal Upper Pressure Limit and the FW Heater OOS Lower Pressure Limit.

~:: -*-_-_ -~.~l . ----~1--=-

Reactor Dome Pressure Allowable Values Versus Core Thermal Power

---~- *-**- -*-*

Cl 1020 i

-r -

iii I I

Q.

I !I

~ I I

1000 *;- . _____I - ----*-*--._. --**---!-***** - ... -~-~----1 a.

I

<IJ E: l 0

980 . --*-~***--**---4-- --

...0 Cl I

1 unl I I

! i'

<IJ a:: 960 I ------ ----*---~ .. **~* ..... ---*- ---! --**- *------~

r Nominal Upper Pressure Umii

-+-Nominal Lower Pressure Limit 940 .--*----

~FW Heater OOS Lower Pressure Limit 920 *- . --*------ - - * - - - - --:-----*---*---...A--*- - - **~*---.

20 30 40 50 60 70 80 90 100 Core Power, % of rated

IUnit2 BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 1 0f 2 Reacf1v1tv

. M aneuver Pl an # U2 SIMULATOR Description of Stem Recirculation Pump Trip Response IF both pumps trip. SCRAM the reactor per 2-0l-68-1A.

For a single recirculation pump trip, IMMEDIATELY insert_<?ontrol rods to lower Lgad Line below 74% using Emergency Shove Sheet. (Stop after any rod.) ---

Conditions:'f[o be recorded at the completion of step.) Recorded: A4 /~

(by RO) (Dfu1Y} '-0/ r

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

QRE presence required in the Control Room? Yes _ _ No _lL (check)

Predicted Actual Predicted Actual I (may be (may be ranges}

ranqes)

• Core Power 40 -50% ~!' i;--'fc. *MFLCPR 0.68 -0.88 .. '7 '1?/

• Core Flow 45-53.5% '-/&.&, bfr:. *MAPRAT 0.34 - 0.54 * 'I~ 3'

• Load Line 66 - 74% '8-S.. '1 -l *MFDLRX 0.56-0.76

• iC'~ (

Critical Parameters: (To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section_

Description including frequency, method of monitoring, and contingency High Low actions:

Load Line: Monitor completing control rod insertions using ICS NHB or 74% 55%

PWRFLW (Power/Flow Map) displays. If HIGH, then insert additional rods. (NOTE: May take 3-4 minutes after finishing rod insertions for heat balance to show correct power level.) If LOW, then do not insert any more control rods.

Reactor Pressure: Verify relationship between reactor pressure and See ICS "RXPCTP" display reactor power remains within allowable region during power reduction by or attached chart.

using ICS display RXPCTP. lf HIGH or LOW, then stop power reduction and adjust reactor oressure before resuming power reduction.

Comments I Notes:

Core Flow indication may be incorrect if the active loop jet pump flow (2-Fl-68-48 or 2-Fl-68-46) is less than 41 Mlbm/hr.

Step Complete and Reviewed by: I Unit Supervisor J Date

IBFN Unit 2 Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 2 0f 2 React*1v1:v "t Maneuver Pl an # U2 SIMULATOR Description of Steo: ReCirculation Pump Trip Response Change speed of the Operating recirculation pump until:

• Core Flow is 46 - 49% to exit/remain out of POWER/FLOW Map Region II

• Operating Recirculation Pump Drive Flow is,::: 46,610 GPM

• Operating Recirculation Pump Speed is:; 100%

Conditions: (To be recorded at the completion of step.) Recorded: j (by RO) (Date)

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

QRE presence required in the Control Room? Yes _ _ No _x_ (check)

Predicted Actual Predicted Actual (may be (may be ranges) ranqes)

• Core Power 37 - 49% *MFLCPR 0.67 - 0.87

• Core Flow 46-49% *MAP RAT 0.32 - 0.52

• Load Line 65 - 74% *MFDLRX 0.55 - 0.75 Critical Parameters: (To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section.

Description including frequency, method of monitoring, and contingency High Low actions:

Core Flow: Monitor while changing core flow using ICS NHB or PWRFLW 50% 45%

(Power/Flow Map) displays. If LOW. then raise core flow. If HIGH, then lower core flow.

Reactor Pressure: Verify relationship between reactor pressure and See ICS "RXPCTP" display .

reactor power remains within allowable region during power reduction by or attached chart.

using ICS display RXPCTP. If HIGH or LOW, then stop power reduction and adjust reactor pressure before resumini:i power reduction.

Load Line: Monitor while changing core flow using ICS NHB or PWRFLW 74% NIA displays. If HIGH, then first verify proper core flow used in calculation.

Insert control rods per Shove Sheets to reduce <74%. {NOTE: May take 3-4 minutes after finishing rod inserts for heat balance to show correct power level.)

Comments I Notes:

Operating Recirculation Pump drive flow is not to exceed 46,610 GPM and its speed is not to exceed 100% While in SLO . .

Core Flow must be greater than 45% to exit Region II at the expected load line, as well as prevent loop delta temperature concerns.

Step Complete and Reviewed by: I Unit Supervisor I Date

I BFN Unit 2 Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 1A 0f 4A Reacf1v1tv

. Maneuver Pl an # U2 SIMULATOR DescriQtion of SteQ: Urgent Load Reduction:

Lower power by lowering core flow until either the desired operating power level is reached or core flow is -60% of rated core flow.

(If a recirculation pump trip is imminent, only lower using the pump to be tripped).

Conditions: (To be recorded at the completion of step.) Recorded: I (by RO} (Date)

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

QRE presence required in the Control Room? Yes _ _ No _x_ (check)

Predicted Actual Predicted Actual (may be (may be ranges) ranges)

• Core Power 65-99% *MF LC PR 0.80 - 0.95

• Core Flow 55 -101% *MAP RAT 0.49 - 0.80

• Load Line 90 - 109% *MFDLRX 0.60- 0.85 Critical Parameters: (To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section.

Description including frequency, method of monitoring, and contingency High Low actions:

Core Flow: Monitor while changing core flow using ICS NHS or PWRFLW NIA 55%

(Power/Flow Map) displays. If LOW, then raise core flow Reactor Pressure: Verify relationship between reactor pressure and See ICS "RXPCTP" display reactor power remains within allowable region during power reduction by or attached chart.

using ICS display RXPCTP. If High or Low, then stop power reduction and adjust reactor pressure before resuminq oower reduction.

Comments I Notes:

If desired, stop power reduction at any point before core flow reaches 60%. However, if continuing to next step, verify core flow is between 55% - 62%.

If a recirculation pump trip is a concern, insert control rods before going below 90% core flow unless flow is only reduced on the pump expected to trip.

2-AOl-6-1A, 18, or 1 C specifv the oower level required for loss of feedwater heaters.

Step Complete and Reviewed by: I Unit Supervisor I Date

~nit2 I BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 2A of 4A Reactivity Maneuver Plan # U2 SIMULATOR Description of Step: Urgent Load Reduction Insert control rods per Shove Sheet until either the desired operating power level is reached or until Load Line is between 52% and 66%.

Conditions: (To be recorded at the completion of step.) Recorded: I (by RO) (Date)

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

QRE presence required in the Control Room? Yes __ No~ (check)

Predicted Actual Predicted Actual (may be (may be ranges) ranges)

• Core Power 39- 55% *MFLCPR 0.61 - 0.81

• Core Flow 55- 65% *MAPRAT 0.32 - 0.52

• Load Line 55- 66% *MFDLRX 0.54- 0.74 i

Critical Parameters: (To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section.

Description including frequency, method of monitoring, and contingency High Low actions:

Load Line: Monitor while inserting control rods using ICS NHS or 66.7% 55%

PWRFLW (Power/Flow Map) displays. If LOW, then stop inserting control rods. If HIGH, then continue inserting control rods.

Reactor Pressure: Verify relationship between reactor pressure and See ICS "RXPCTP" display reactor power remains within allowable region during power reduction or attached chart.

using ICS display RXPCTP. If High or Low, then stop power reduction and adiust reactor pressure before resuming power reduction.

Comments I Notes:

If desired, stop power reduction after any control rod (not group) is fully inserted. However. before continuing to next Step. Load Line must be between 55% and 66%.

Note that Load Line may increase -1.5% as flow is reduced in the next Step.

Note OPS can remove heaters during this step after power is below 45%. RE will change TL set if heaters removed from service. Before continuing to next Step, Load Line must be between 55 and 66%.

Step Complete and Reviewed by: I Unit Supervisor I Date

l~nit2 BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 3A of 4A Reactivitv Maneuver Plan# U2 SIMI JI A TOR Description of Step: Urgent Load Reduction Lower recirculation pump speed to -28% (-480 rpm, -38% core flow).

(Control rods may be adjusted during this step.)

Conditions: (To be recorded at the completion of step.) Recorded: I (by RO) (Date}

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

ORE presence required in the Control Room? Yes _ _ No _L (check)

Predicted Actual Predicted Actual (may be {may be ranges) rani:ies)

"Core Power 28-40% *MFLCPR 0.65 - 0.85 "Core Flow 36 - 43% *MAP RAT 0.24 - 0.44

• Load Line 55 - 66% "MFDLRX 0.46 - 0.66 Critical Parameters: {To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section.

Description including frequency, method of monitoring, and contingency High Low actions:

Load Line: Displayed on Heat Balance or Power/Flow Map 66.7% N/A If HIGH, then insert control rods.

Reactor Pressure: Verify relationship between reactor pressure and See ICS "RXPCTP" display reactor power remains within allowable region during power reduction or attached chart.

using ICS display RXPCTP. If High or Low, then stop power reduction and adjust reactor pressure before resuming power reduction.

Comments I Notes:

I Step Complete and Reviewed by: I Unit Supervisor J Date

I BFN Unit 2 Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions s TEP 4A of 4A Reactivity Maneuver Plan# U2 SIMULATOR Description of Step: Urgent Load Reduction Insert control rods per Shove Sheet until desired power level is reached or SCRAM the reactor.

Conditions: (To be recorded at the completion of step.) Recorded: I (by RO) {Date)

• Contact the On-Call RE for parameters which do not meet the requirements in the expectation criteria as found in the Special Monitoring Requirements Section on Attachment 7.

QRE presence required in the Control Room? Yes _ _ No _x_ (check)

Predicted Actual Predicted Actual (may be (may be ranges) ranQes)

• Core Power 14 - 24% *MFLCPR 0.38 - 0.58

• Core Flow 36-42% *MAPRAT 0.06 - 0.26

• Load Line 25-45% *MFDLRX 0.17 - 0.37 Critical Parameters: (To be recorded DURING step in the OPs log.) If critical parameters are outside of the specified criteria, then discuss with the RE and record conclusions in the Comments/Notes section.

Description including frequency, method of monitoring, and contingency High Low actions:

Load Line: Displayed on Heat Balance or Power/Flow Map 66.7% N/A If HIGH, then insert control rods Reactor Pressure: Verify relationship between reactor pressure and See ICS ' 1RXPCTP" display reactor power remains within allowable region during power reduction or attached chart.

using ICS display RXPCTP. If High or Low, then stop power reduction and adiust reactor pressure before resuminQ power reduction.

Comments I Notes:

Load Line/Power should continue to drop because of the xenon transient.

Bypass the RWM if power drop is to continue below 30%.

The unit should be scrammed or lined up to the correct RWM sequence before reaching 10% RTP.

Step Complete and Reviewed by: I Unit Supervisor I Date

1 BFN Reactivity Control Plan Unit2 Attachment 9 (Page 1 of 1)

Power to Flow Map Recirculation Pump Trip (Steps 1 - 2)

Reactivity Control Plan# U2 SIMULATOR 110.--,---.-:,..----r-.,...--r----,l~-,--.,.-~~...,.--.,...~.,.--.....,---:-~...,.--.,.--r---i 100 I -~- .. ---- ----- -.. j- ~-l-O+--+--+--IL---+--+--+~-+--1---+--+--+--+---lf---l--~-l--+~-+-~r---l 20 25 30 35 40 45 50 55 60 65 70 75 BO 85 90 95 100 105 110 115 PERCENT RATED CORE FLOW

- - - - - - - - - - - - - - - - - - - - - - - - - - *--*- --------------- .. ..I

1 BFN Reactivity Control Plan Unit2 Attachment 9 (Page 1of1)

Power to Flow Map Urgent Load Reduction (Steps 1A - 4A)

Reactivity Control Plan # U2 SIMULA TOR 110..--...--.r.,--~~-.-----,~-.---,,--____,___,,...--.,..-~~-----,-..----......~-.-~~-,----,~~

1 I

100 +----+--~

--*-T-90 I ---;-~...--_,_-t--1-::..-"----,--=<0""1""':;._t---t--1--+t--;---i I

_80 --+--+---+-

cJr70 w

~60

~ --- 1 j

~50+--+-++----"'"-

ci:

~40 1-

@30~---+--~~-+--.-~---+-

l-i:520 .l.~+--t---+----t------,~-+--,----!-~-~-1

0. I '

l 10 +----+-~+---# -- -*I -**- --*---t----"------i-------

O-t--t-~-t---l'~-t----i~-t---1t---t-~r--;-~-1---;-~+---t-~-r---r~-t---1i---t 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 PERCENT RATED CORE FLOW

- - - - - _ _ _ _ _ __ _ _ J

I~nit2 BFN Reactivity Control Plan Attachment 10 (Page 1of1)

Recirculation Flew Maneuver Instructions

.. controI Plan # U2 SIMULATOR React1v1ty RCP Flow Time Target Delta Target Completed (RO)

Step# Step# Power Flow

(%RTP or +/-(MWe) (MLb/Hr)

MWe)

Comments I Notes:

Reviewed by: I Unit Supervisor I Date

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.5(A)

Unit2 Rev.0024 Page 16 of 363 Attachment 1

,,..,,'\

(Page 1of2)

Control Rod Movement and Control Rod Drive Problem Data Sheets Date CUIA.fl'f Part 1: CONTROL ROD MOVEMENT DATA SHEET RWM' ROD Rod Movement Completed GP NUMBER FROM TO Signoffs 3

UO <AC} 2 Peer Check NIA 30-39 08 00 NIA 38-31 08 00 NIA 30-23 08 00 NIA 22-31 08 00 NIA 14-39 18 00 NIA 46-39 18 00 NIA 46-23 18 00 NIA 14-23 18 00 NIA 22-47 18 00 NIA 38-47 18 00 NIA 38-15 18 00 NIA 22-15 18 00 REMARKS 4 : CONTINUOUS INSERT Page 1 NOTES:

RWM Group may be marked "NIA" when NOT applicable (i.e., when above the LPSP).

2 For all rod moves to the "full our position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

3 Documentation of Peer Check by a second qualified member of the plant staff (i.e .* RE. STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thennal power s10%.

4 Record the rod number and any problems encountered on Attachment 1, page 2 of 2. as applicable.

5 Peer Check by RE OR SRO (SRO to be utilized if a second RE is unavailable). The SRO should be checking the FROM and TO control rod positions as a minimum. The RE OR SRO should be checking the positions identified for agreement with the Predictor Cases. Anytime the SRO feels the Peer Check is beyond his/her knowledge level, then call-in a second RE to perform the required Peer Check.

5 Issued By: I Reviewed By  : I Reactor Engineer Date RE OR SRO Date Authorization to perfonn the Control Rod manipulations identified on this sheet tklt S'~'" I tJW'f'eltt Unit Supervisor Date

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.S(A)

Unit2 Rev.0024 Page 16 of 363 Attachment 1 (Page 1of2)

Control Rod Movement and Control Rod Drive Problem Data Sheets Datecvr<t.~

Part 1: CONTROL ROD MOVEMENT DATA SHEET RWM' ROD Rod Movement Completed GP NUMBER FROM TO Signoffs UO (AC) 2 Peer Check 3 NIA 14-31 48 00 NIA 30-47 48 00 NIA 46-31 48 00 NIA 30-15 48 00 NIA 22-39 48 00 NIA 38-39 48 00 NIA 38-23 48 00 NIA 22-23 48 00 NIA 30-31 48 00 NIA 06-31 48 00 N/A 30-55 48 00 NIA 54-31 48 00 NIA 30-07 48 00 REMARKS4 : CONTINUOUS INSERT Page2 NOTES:

RWM Group may be marked "NIA" when NOT applicable {i.e., when above the LPSP).

2 For all rod moves to the "full ouf' position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

3 Documentation of Peer Check by a second qualified member of the plant staff (i.e .. RE, STA OR UO} is required ONLY when the RWM is inoperable OR bypassed with core thennal power s10%.

4 Record the rod number and any problems encountered on Attachment 1, page 2 of 2, as applicable.

5 Peer Check by RE OR SRO [SRO to be utilized if a second RE is unavailable]. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE OR SRO should be checking the positions identified for agreement with the Predictor Cases. Anytime the SRO feels the Peer Check is beyond his/her knowledge level, then call-in a second RE to perfonn the required Peer Check.

5 Issued By: I Reviewed By  : I Reactor Engineer Date REOR SRO Date Authorization to perform the Control Rod manipulations identified on this sheet

(;{/(it cf~#'" I e<<1W1rt Unit Supervisor Date

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.S(A)

Unit 2 Rev.0024 Paae 16 of 363 Attachment 1 (Page 1of2)

Control Rod Movement and Control Rod Drive Problem Data Sheets Part 1: CONTROL ROD MOVEMENT DATA SHEET RWM' ROD Rod Movement Completed GP NUMBER FROM TO Signoffs UO (AC) 2 Peer Check 3 NIA 14-47 48 00 NIA 46-47 48 00 NIA 46-15 48 00 NIA 14-15 48 00 NIA 22-55 48 00 NIA 38-55 48 00 NIA 38-07 48 00 NIA 22-07 48 00 REMARKS4 : CONTINUOUS INSERT Page 3 NOTES:

RWM Group may be marked "NIA" when NOT applicable (i.e .* when above the LPSP).

2 For all rod moves to the "full our' position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

3 Documentation of Peer Check by a second qualified member of the plant staff (i.e .* RE, STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal power :S10%.

4 Record the rod number and any problems encountered on Attachment 1, page 2 of 2, as applicable.

5 Peer Check by RE OR SRO [SRO to be utilized if a second RE is unavailable). The SRO should be checking the FROM and TO control rod positions as a minimum. The RE OR SRO should be checking the positions identified for agreement with the Predictor Cases. Anytime the SRO feels the Peer Check is beyond his/her knowledge level, then call-in a second RE to perform the required Peer Check.

Issued By: I Reviewed By5: I Reactor Engineer Date RE OR SRO Date Authorization to perform the Control Rod manipulations identified on this sheet tf,,;t .f~n,. I ellJ'Wl(t Unit Supervisor Date

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario No.: NRC - 3 Op-Test No.: 1501 Examiners: Operators: SRO:

ATC:

BOP:

Initial Conditions: Reactor Power is:::: 3%. Unit1 and Unit 3 are at 100% power. IRM D failed upscale last shift and is bypassed.

Turnover: Secure purging the Drywell and Suppression Chamber in accordance with 2-01-64 Section 8.1. Continue plant startup IAW 2-GOl-100-1A section 5.4, mode change from Mode 2 to Mode 1.

Event Malfunction Event Event Description Number Number Type*

N-BOP 1 N/A Secure form Drywell /Torus Purge.

N-SRO R-ATC 2 N/A Power increase with Control Rods to 8% IAW GOI R-SRO C-ATC RD07R0239 3 TS- Control Rod Drift in RD06R0239 SRO C-BOP 4 OG04A Loss of SJAE 'A' I Swap to STBY SJAE 'B' C-SRO I-ATC 5 NM05 IRM 'B' Failure Upscale/Half Scram TS-SRO C-BOP 6 DG03D Loss of 4KV Shutdown Board D, D DIG fails to AUTO tie TS-SRO FW14C C-ATC 7 Trip of RFP 2C/ recover with already warm RFP 2B C-SRO PC 14 (e20 0) 100 Loss of Torus Water level /SCRAM (ATWS)and ED on 8 300 75 M-ALL Torus water level 9 FW30 C-ATC Failure of RFP 2B governor/pump needs tripped.

AD01D C-BOP 10 ADS SRV Failures AD01E C-SRO

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

Appendix D Scenario Outline Form ES-D-1 Events

1. BOP secures Drywell Purge in accordance with 2-01-64 Section 8.1. After the OW and Torus Purge has been secured, the scenario may continue.

2. ATC will withdraw control rods in order to raise power to 8% for a mode change from 2 to 1. Once the Chief Examiner has seen an appropriate amount of power rise he may choose to continue with the next event which will halt control rod withdrawal.

3. Control Rod 02-39 will begin to drift in to approximately position 10, and the ATC will respond IAW 2-AOl-85-5 and bypass the RWM and insert Control Rod 02-39 to position 00. The SRO will declare Control Rod 02-39 inoperable and refer to Tech Spec 3.1.3 condition C. The SRO will also refer to Tech Spec 3.1.6 condition A for BPWS and 3.3.2.1 condition C for RWM. Once the rod has been inserted and the Tech Spec call has been made the scenario may continue.

4. Loss of SJAE A, BOP operator swaps to B SJAE IAW 2-AOl-47-3, Loss of Condenser Vacuum or IAW 2-01-66 or the hardcard. After the standby SJAE has been placed in service and Main Condenser vacuum has recovered, the scenario may continue.

5. The ATC will respond to a failure of IRM 'B' upscale and notice and report that a% scram failed to come in on the upscale condition. The crew will be unable to bypass IRM 'B' since IRM 'D' is currently bypassed. The SRO will refer to Tech Spec 3.3.1.1 and enter condition 'A'. The required action is to place a channel in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. After the SRO has completed the Tech Spec call the scenario may continue.

6. D 4KV Shutdown Board will lose power and the D Diesel Generator will fail to automatically tie to the Shutdown Board. The BOP will manually tie the Diesel to the board. SRO will refer to Tech Specs and determine TS 3.8.1 condition A, B, and G, and TS 3.8.7.A. The ATC will be resetting RPS and PCIS. After the BOP Operator ties the Diesel to the Buss and the SRO has completed the Tech Spec call the scenario may continue.

7. The ATC will respond to a trip of the 2C RFP IAW 2-AOl-3-1 by raising the speed of the warm RFP 2B to feed the RPV. Once the ATC has entered AOl-3-1 and raised the speed on the standby Reactor Feedpump to maintain RPV Water Level the scenario may continue.

8. At the cue of the Chief Examiner initiate the next event. An unisolable leak will develop on the suppression chamber. The US will direct entry into EOl-3 on secondary containment area flood alarms and EOl-2 on suppression pool water

Appendix D Scenario Outline Form ES-D-1 level. Prior to 12. 75 ft, in the Suppression Pool, the US will direct HPCI to be secured and locked out. Prior to 11.5 ft in the Suppression Pool the US will transition to EOl-1 and direct a SCRAM. An ATWS will exist on the SCRAM. The crew will work through EOl-1 and C-5 to insert control rods, maintain reactor water level, and reactor pressure. The US will transition to C-2 to emergency depressurize before Suppression Pool water level lowers to 11.5 feet.

9. The US will direct terminating and preventing IAW EOI Appendix 4, and the 2B RFP governor will fail as is. The ATC/BOP will Trip the 2B RFP.

10. The BOP will report that two of the ADS SRV's failed for Emergency Depressurization. Two additional non ADS SRV's will be opened at the direction of the SRO.

The Scenario ends when Emergency Depressurization and Reactor Water Level is restored and maintained within the assigned band or upon request of Lead Examiner.

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 3

1. When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.

1. Safety Significance:

Prevent failure of Primary Containment from pressurization of the Suppression Chamber

2. Cues:

Procedural compliance Suppression Pool Level indication

3. Measured by:

Observation - HPCI Auxiliary Pump placed in Pull to Lock

4. Feedback:

HPCI does not Auto initiate No RPM indication on HPCI

2. When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of Containment

2. Cues:

Procedural compliance Suppression Pool Level Trend

3. Measured by:

Observation - US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet.

Observation - RO opens at least 6 SRV's during performance of Emergency Depressurization actions.

Appendix D. Scenario Outline Form ES-D-1

4. Feedback:

RPV pressure trend SRV status indications Suppression Pool temperature trend

3. With RPV pressure <MSCP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.

1. Safety Significance:

Maintaining adequate core cooling and preclude possibility of large power excursions.

2. Cues:

Procedural compliance RPV pressure indication

3. Measured by:

Observation - Injection not commenced until less than MSCP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF

4. Feedback:

RPV level trend RPV pressure trend Injection system flow rate into RPV

SHIFT MANAGER TURNOVER EQUIPMENT OOS/LCOS

• IRM 'D' ANTICIPATED OPERATIONS/MAINTENANCE FOR ONCOMING SHIFT

• Secure purging Drywell and Suppression Chamber IAW 2-01-64

• Continue plant startup IAW 2-GOl-100-1A Section 5.4 NO HEATERS ARE TO BE PLUGGED INTO OUTLETS IN THE CONTROL ROOM UNTIL DETERMINED IF WE CAN USE SPP 7.3 RISK REVIEW DOCUMENTATION ON P DRIVE FOLDER WCC-RISK/BFN EXPECTATION FOR 1500 MEETING IS FOR THE SM TO COVER CURRENT STATUS ON ALL CONTROL ROOM DEFICIENCIES AND ANNUNCIATORWOs THAT SHOULD BE WORKING. THIS EXPECTATION STARTED 3/26/09 DO NOT DISABLE ANY ANNUNCIATORS WITHOUT GOING THRU MANAGEMENT REVIEW.

IF ALTERNATE HEAT BALANCES ALARM COMES IN ON ANY UNIT'S ICS, REDUCE POWER BY 10 MWT (3448 IF AT RATED POWER) AND CALL REACTOR ENGINEERING IN TO PLANT TO EVALUATE.

STATUS WORK WEEK - DIV I 18], DIV II 0 OUTSIDE AIR TEMP 45°F COMMON UNIT 1

• 100%

UNIT2

• 3%

UNIT3

• 100%

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _1_ Page 1 of 4 Event

Description:

Secure purging the Drywell and Suppression Chamber with Primary Containment Purge Filter Fan IAW 2-01-64, Sec. 8.1 Applicant's Actions or Behavior 2-01-64 8.1 Purging the Drywell and Suppression Chamber with Primary Containment Purge Filter Fan.

[24] WHEN both of the following instruments have stabilized AND Chemistry samples indicate Drywell and Suppression Chamber oxygen concentration is greater than or equal to 19.5%, THEN STOP PRI CTMT PURGE FILTER FAN using 2-HS-64-131 (Rx Bldg El 621 '):

• H2/02 CONCENTRATION, 2-XR-76-110 (Panel 2-9-54)

• H2/02 ANALYZER, 2-MON-76-110 (Panel 2-9-55)

[25] CLOSE the following valves (Panel 2-9-3):

BOP

• DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-29, using 2-HS-64-29

• DRYWELL VENT OUTBD ISOLATION VLV, 2-FCV-64-30, using 2-HS-64-30

• SUPPR CHBR VENT INBD ISOL VALVE, 2-FCV-64-32, using 2-HS-64-32

• SUPPR CHBR VENT OUTBD ISOLATION VLV, 2-FCV 33, using 2-HS-64-33

• DW/SUPPR CHBR AIR PURGE ISOL VLV, 2-FCV-64-17, using 2-HS-64-17

• DRYWELL ATM SUPPLY INBD ISOLATION VLV, 2-FCV-64-18, using 2-HS-64-18

• SUPPR CHBR ATM SPLY INBD ISOLATION VLV 2-FCV-64-19 using 2-HS-64-19 Page 1 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: Page 2 of 2 Event

Description:

Secure purging the Drywell and Suppression Chamber with Primary Containment Purge Filter Fan IAW 2-01-64, Sec. 8.1 Time Position Applicant's Actions or Behavior

[26] RECORD time purging was stopped in the Narrative log.

[27] PLACE the following key lock switches in the NORMAL position (Panel 2-9-3):

• PC PURGE DIV I RUN MODE BYPASS, 2-HS-64-24

• PC PURGE DIV II RUN MODE BYPASS, 2-HS-64-25

[28] VERIFY CTMT PURGE FILTER HEATER ON, 2-HS-64-133 (RB, R12-S LINE, Primary Containment Purge Unit, El 621).

NRG End of Event #1 Page 2 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _£___ Page 1 of 2 Event

Description:

Power increase with Control Rods to 8% IAW GOI Time Position Applicant's Actions or Behavior Conducts crew brief to focus on the continued reactor startup SRO Directs ATC to continue startup IAW GOl-100-1A 2-GOl-100-1 A 5.4 Withdrawal of Control Rods while in Mode 2 (continued)

(67] CONTINUE to withdraw control rods to raise Reactor power to approximately 8% per 2-01-85 and 2-SR-3.1.3.5(A).

2-01-85 6.6.3 Control Rod Notch Withdrawal

[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

[2] OBSERVE the following for selected control rod:

• CRD ROD SELECT pushbutton is brightly ATC ILLUMINATED.

• White light on the Full Core Display ILLUMINATED

• Rod Out Permit light ILLUMINATED.

[3] VERIFY ROD WORTH MINIMIZER operable and LATCHED into correct ROD GROUP when Rod Worth Minimizer is enforcing.

[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE

[5] OBSERVE control rod settles into desired position AND ROD SETTLE light extinguishes.

[6] IF control rod is notch withdrawn to rod notch Position 48, THEN Page 3 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _2_ Page 2 of 2 Event

Description:

Power increase with Control Rods to 8% IAW GOI Time Position Applicant's Actions or Behavior PERFORM control rod coupling integrity check as follows:

[6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.

[6.2] CHECK control rod coupled by observing the following:

• Four rod display digital readout AND full core display digital readout AND background light remain illuminated.

ATC *CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window 14) does not alarm.

[6.3] CHECK control rod settles into Position 48 and ROD SETTLE light extinguishes.

[6.4] IF control rod coupling integrity check fails, THEN Refer to 2-AOl-85-2.

Withdraws control rods IAW 2-01-85 and 2-SR- 3.1.3.5(A).

NRC End of Event #2 Page 4 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 1of6 Event

Description:

Control Rod Drift In Time Position Applicant's Actions or Behavior When the NRG Chief Examiner is satisfied with the reactivity Driver change and is ready for Event No. 3, insert F3 (bat NRC/1501-3-3) to drift control rod 2243 in.

Reports CONTROL ROD DRIFT (2-XA-55-5A, Window 28) in alarm.

2-ARP-9-5A A.DETERMINE which rod is drifting from Full Core Display C. IF rod drifting in, THEN REFER TO 2-AOl-85-5 and 2-AOl-85-7.

2-AOl-85-5.

4.0 OPERATOR ACTIONS 4.1 Immediate Actions

[1] IF multiple rods are drifting into core, THEN MANUALLY SCRAM Reactor. Refer to 2-AOl-100-1.

4.2 Subsequent Actions NOTE If Reactor Power is less than 24%, the Rod Worth Minimizer (RWM) system may impose rod blocks ATC

[1] IF Reactor Power is less than less than 24% RTP, THEN MANUALLY BYPASS the Rod Worth Minimizer on Panel 2-9-5 as follows: (Otherwise NIA)

[1.1] PLACE RWM SWITCH PANEL, 2-XS-85-9025, in BYPASS position.

[1.2] CHECK the Manual Bypass light is illuminated.

[1.3] NOTIFY the Unit Supervisor that Rod Worth Minimizer has been manually Bypassed. (Reference Tech Spec Sections 3.3.2.1 and 3.1.6)

[1.4] When time permits, PERFORM the remaining actions to Manually Bypass the Rod Worth Minimizer per 2 85.

[2] IF a Control Rod is moving from its intended position without operator actions, THEN INSERT the Control Rod to position 00 using CONTINUOUS IN. (Otherwise N/A)

Page 5 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 2 of 6 Event

Description:

Control Rod Drift In Time Position Applicant's Actions or Behavior Identifies and reports control rod 22-43 drifting into the core ATC Monitors the full core display for other control rod drifts When control rod 22-43 stops moving (stuck), insert Shift F3 (dmf Driver rd07r2243) to delete the drift then insert Shift F4 (dmf rd06r2243) to allow the rod to be inserted 2-AOl-85-5 Rod Drift In

[1] IF Reactor Power is less than less than 24% RTP, THEN MANUALLY BYPASS the Rod Worth Minimizer on Panel 2-9-5 as follows:

[1.1] PLACE RWM SWITCH PANEL, 2-XS-85-9025, in BYPASS position.

[1.2] CHECK the Manual Bypass light is illuminated.

[1.3] NOTIFY the Unit Supervisor that Rod Worth Minimizer has been manually Bypassed. (Reference Tech Spec Sections 3.3.2.1 and 3.1.6)

[2] IF a Control Rod is moving from its intended position without operator actions, THEN INSERT the Control Rod to position 00 using CONTINUOUS IN.

[3] IF a Control Rod Block occurs during rod insertion due to Rod Worth Minimizer, THEN BYPASS the RWM per step 4.2[1] above.

[4] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

[5] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 2-AOl-100-1.

Page 6 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 3 of 6 Event

Description:

Control Rod Drift In Tim::-

• Position Applicant's Actions or Behavior

[6] CHECK Thermal Limits on ICS (RUNMON).

[7] ADJUST control rod pattern as directed by Reactor Engineer and CHECK Thermal Limits on ICS (RUNMON).

[9] VERIFY scram pilot air header aligned to scram inlet and outlet valves.

[1 O] CHECK CRD SCRAM OUTLET, 2-FCV-085-398, for leakage as indicated by the following:

• Scram riser for affected HCU has higher than normal ATC temperature.

• CRD SCRAM OUTLET, 2-FCV-085-398, producing flow noise.

NOTE The CRD accumulator is required to be considered inoperable per Technical Specifications 3.1.5 when the charging water is isolated.

Page 7of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 4 of 6 Event

Description:

Control Rod Drift In Time Position Applicant's Actions or Behavior

[11] CHECK CRD SCRAM INLET VALVE, 2-FCV-085-39A, for leakage as follows:

[11.1] CHECK insert riser for affected HCU for higher than normal temperature.

[11.2] CLOSE CHARGING WATER SOV, 2-SHV-085-588, AND OBSERVE CRD ACCUMULATOR NITROGEN SIDE PRESS, 2-Pl-85-34, for a lowering trend.

[4] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.

Dispatches AUO to investigate the rod drift at the HCU As RE, acknowledge control rod 22-43 drifting in and subsequently being inserted fully into the core. Will evaluate core thermal limits and preconditioning limits.

Driver As AUO, acknowledge going to HCU for control 22-43 to investigate the rod drift Page 8 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 5 of 6 Event

Description:

Control Rod Drift In Time Position Applicant~s Actions or Behavior Evaluates Technical Specifications 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE C. One or more control rods C.1 Fully insert 3 hr inoperable for reasons inoperable control other than Condition A or B rod AND C.2 Disarm the 4 hr associated CRD NRC - T.S. 3.1.6 only 3.1.6 Rod Pattern Control applies until the rod is LCO 3.1.6 OPERABLE control rods shall comply with the declared inoperable. requirements of the banked position withdrawal sequence If-----.,.----~ (BPWS).

SRO A. One or more OPERABLE A.2 Declare associated 8 hr control rods not in control rod inoperable compliance with BPWS 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.

C. Rod worth minimizer C.1 Verify~ 12 rods Immediately (RWM) inoperable during withdrawn reactor startup. AND C.2.2 Verify movement of control rods is in compliance with banked position withdrawal sequence by a second licensed operator or other qualified member of the technical staff Page 9 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _3_ Page 6of6 Event

Description:

Control Rod Drift In Position Applicant's Actions or Behavior Updates crew on Technical Specification requirements for the inoperable control rod, RWM, and BPWS.

As Rx Bldg AUO, report that the scram outlet valve has an air Driver leak.

NRC End of Event #3 Page 10 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _A_ Page 1 of 2 Event

Description:

Loss of SJAE 'A' I Swap to STBY SJAE 'B' Time Position Applicant's Actions or Behavior When NRC Chief Examiner is ready for Event No. 4, insert F4 (imf Driver og04a) to isolate 'A' SJAE.

Reports OG HOLDUP LINE INLET FLOW LOW (2-XA-55-53, Window 4) in alarm.

BOP 2-ARP-9-53 D. VERIFY that SJAE auto isolation has NOT occurred Reports that SJAE 'A' has isolated Directs BOP to swap to SJAE 'B' using the hard card SRO Directs ATC to monitor main condenser vacuum 2-01-66 Appendix C Standby SJAE System Lineup Hard Card 2.0 OPERATOR ACTION FOR SJAE 28 NOTES Radiation Protection should be notified prior to placing a SJAE in BOP service. If time does not permit this due to plant conditions then notification should be made when possible.

2-HS-001-0375, SJAE TRAIN PERMISSIVE (located on 2-LPNL-925-0105, U2 TB, el 586') should normally be in the position of the standby SJAE. If problems are encountered while placing a SJAE in service and time permits, operate this switch as required during the performance of this section.

Page 11 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _A_ Page 2of2 Event

Description:

Loss of SJAE 'A' I Swap to STBY SJAE 'B' Time Position Applicant's Actions or Behavior

[1] VERIFY RESET Off-Gas isolation using 2-HS-90-155, OG OUTLET/DRAIN ISOLATION VLVS.

[2] VERIFY OPEN the following valves:

• 2-HS-66-15, SJAE 28 INLET VALVE.

• 2-HS-1-156A, STEAM TO SJAE 28.

[3] VERIFY in AUTO/OPEN 2-HS-66-18, SJAE 28 OG OUTLET VALVE.

[4] PLACE 2-HS-1-152, SJAE 28 PRESS CONTROLLER, in CLOSE and then in OPEN.

[5] VERIFY OPEN the following valves (red light illuminated):

• 2-PCV-1-153/167, STEAM TO SJAE 28 STAGES 1,2, BOP AND3.

• 2-FCV-1-152, SJAE 28 INTMD CONDENSER DRAIN.

[6] MONITOR hotwell pressure as indicated on CONDENSATE recorder, 2-XR-002-0026 (Point 3), on Panel 2-9-6.

[7] FOR the SJAE not being placed in service, VERIFY CLOSED the following valves:

• 2-HS-66-14, SJAE 2A OG OUTLET VALVE.

• 2-HS-1-150, SJAE 2A PRESSURE CONTROLLER.

• 2-HS-1-155A STEAM TO SJAE 2A Reports 'B' SJAE in service Calls RP to report 'B' SJAE placed in service Reports that main condenser vacuum had started to lower but has ATC returned to normal Driver As RP, acknowledge that 'B' SJAE has been placed in service.

NRC End of Event #4 Page 12 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _5_ Page 1 of 2 Event

Description:

IRM 'B' Failure of Upscale/Half Scram 1ime Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 5, insert F5 Driver (imf nm05b 100) to fail IRM 'B' upscale.

Reports IRM B HI-HI/I NOP condition with no reactor half scram.

ATC Immediately stops reactivity changes.

Reports that "IRM CH B, D, F, H Hl-Hl/INOP" (2-XA-55-5A, Window 34) is in alarm and REACTOR CHANNEL B AUTO SCRAM (2-XA-55-58, Window 2) failed to come in on a valid IRM Upscale condition.

Reviews ARPs 2-ARP-9-5A A. STOP any reactivity changes.

B. VERIFY alarm by multiple indications.

ATC C. RANGE initiating channel or BYPASS initiating channel to reset half-scram. REFER TO 2-0l-92A.

H. NOTIFY Reactor Engineer 2-ARP-9-5B A. VERIFY channel B relays dropped out by checking scram solenoid and backup scram valve lights extinguished.

Checks indications and reports IRM 'B' has failed upscale Verifies scram solenoid and backup scram valve lights did not extinguish.

Page 13of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _5_ Page 2of2 Event

Description:

IRM 'B' Failure Upscale/Half Scram 11me Position Applicant's Actions or Behavior Refers to Technical Specifications 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

Table 3.3.1.1-1 Reactor Protection System Instrumentation SRO 1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G Determines that:

Condition A applies Required Action A.1 Place the channel in trip in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR A.2 Place the associated trip system in trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Completion time 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Directs ATC and BOP to insert a half scram IAW 01-99 SRO (place the IRM mode switch out of operate)

The SRO will have IM's investigate and repair the IRM especially since he has 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to place a channel in trip.

The next event may be started before the crew has the NRG opportunity to placing the 1h scram in on B RPS.

End of Event #5 Page 14of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _§__ Page 1 of 4 Event

Description:

Loss of 4KV Shutdown Board D, D DIG fails to AUTO tie e Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 6, insert F6 to cause the loss of 4kV SD Bd D (ed09d) and a failure of 'D' DG Driver to automatically tie to the board (dg03d). After diesel has started, insert Shift F6 (dmf ed09d) to delete to overcurrent trip of the board.

ATC Reports a reactor half scram and loss of RPS 'B' Recognizes and reports that 'D' 4kV Shutdown Board is de-energized and 'D' Diesel Generator failed to automatically tie to the board.

Verifies no lockout conditions on the board and closes 'D' Diesel Generator output breaker BOP Reports 'D' 4kV Shutdown Board re-energized Call WC to investigate the loss of 'D' 4kV Shutdown Board and the failure of 'D' to automatically tie to the board.

Calls to have RPS 'B' restored As WC, acknowledge initiating investigation of the loss of 'D' 4kV Shutdown Board and the failure of 'D' Diesel Generator to Driver automatically tie to the board.

As outside US, acknowledge restoring power to RPS 'B' Refers to Technical Specifications 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; Page 15 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: __§__ Page 2of4 Event

Description:

Loss of 4KV Shutdown Board D, D DIG fails to AUTO tie Time Position Applicant's Actions or Behavior

b. Unit 1 and 2 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; SRO A. One required offsite A. 1 Verify power availability 1 hr circuit inoperable. from the remaining AND OPERABLE offsite Once transmission network 18 hr B. One required Unit 1 B.1 Verify power availability 1 hr and 2 DG inoperable. from the offsite AND G. One required offsite G.1 Declare the affected Immediately circuit inoperable. 4.16 kV shutdown AND board inoperable One Unit 1 and 2 DG Inoperable 3.8 ELECTRICAL POWER SYSTEMS SRO 3.8.7 Distribution Systems - Operating LCO 3.8.7 The following AC and DC electrical power distribution subsystems shall be OPERABLE:

a. Unit 1 and 2 4.16 kV Shutdown Boards; A. One Unit 1 and 2 A.1 Restore the Unit 1 and 2 5 day 4.16 kV Shutdown Board 4.16kV Shutdown Bd inoperable. to OPERABLE status Briefs crew on Technical Specification requirements based on the loss of 4kV Shutdown Board 'D' Insert Shift F7 (mrf rp02 reset) to reset RPS 'B' circuit protectors.

Driver As outside US, call and report that RPS 'B' tlas been restored SRO Directs ATC to reset RPS and PCIS IAW 01-99 Page 16of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: __§___ Page 3 of 4 Event

Description:

Loss of 4KV Shutdown Board D, D DIG fails to AUTO tie 1ime Position Applicant's Actions or Behavior 2-01-99 8.5 Restoration to Normal Following RPS Bus Power Loss or Transfer NOTES

1) This section provides instructions for resetting the various ATC system isolations and reopening affected valves to allow those systems to be restored to normal operation in accordance with their respective operating instructions.

2) The following steps are performed at Panel 2-9-5 unless otherwise noted.

3) When RPS Bus power is lost to some scram discharge volume level switches, their RTD heater is de-energized. Following the restoration of power, a time delay, dependent on how long the level switch was de-energized, prevents resetting the half scram signal. This may take up to 37 seconds after RPS power is restored. Precaution 3.0 0 can be referred to for more information on these level switches

[1] OBTAIN Unit Supervisor/SRO's permission to restore to normal.

[2] MOMENTARILY PLACE SCRAM RESET, 2-HS-99-5A-S5, as follows:

[2.1] RESET FIRST position. (Group 2/3)

[2.2] RESET SECOND position. (Group 1/4)

[2.3] NORMAL position.

[3] CHECK the following conditions:

A. All eight SCRAM SOLENOID GROUP A/B LOGIC RESET lights ILLUMINATED. D B. The following four lights ILLUMINATED:

• SYSTEM A BACKUP SCRAM VALVE, 2-IL-99-5A/AB.

• SYSTEM B BACKUP SCRAM VALVE, 2-IL-99-5A/CD.

C. Scram Discharge Volume vent and drain valves indicate OPEN.

D. Points SOE033 and SOE035 on ICS computer or on the First Out Printer reads "NOTTRIP" for RPS "A".

Page 17of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _§___ Page 4 of 4 Event

Description:

Loss of 4KV Shutdown Board D, D DIG fails to AUTO tie Time Position Applicant's Actions or Behavior E. Points SOE034 and SOE036 on ICS computer or on the First Out Printer reads "NOTTRIP" for RPS "B".

[4] At Panel 2-9-4, RESET PCIS trip logic as follows:

[4.1] MOMENTARILY PLACE PCIS DIV I RESET, 2-HS-64-16A-S32, to left and right RESET positions.

[4.2] CHECK the following red lights ILLUMINATED:

• MSIV GROUP A 1.

• MSIV GROUP 81.

ATC

[4.3] MOMENTARILY PLACE PCIS DIV II RESET, 2-HS-64-16A-S33, to left and right RESET positions.

[4.4] CHECK the following red lights ILLUMINATED:

• MSIV GROUP A2.

• MSIV GROUP 82.

Resets RPS and PCIS IAW 01-99 Reports RPS and PCIS have been reset 2-AOl-70-1 Loss of Reactor Building Closed Cooling Water 3.0 AUTOMATIC ACTIONS A. RBCCW SECTIONALIZING VALVE, 2-FCV-70-48, closes automatically on RBCCW Pump discharge header pressure at or below 57 psig.

BOP 4.1 Immediate Actions

[1] IF RBCCW Pump(s) has tripped, THEN Perform the following:

• SECURE RWCU Pumps.

• VERIFY RBCCW SECTIONALIZING VLV, 2-FCV-70-48 CLOSED.

Page 18of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 2-AOl-99-1 Loss of Power to One RPS Bus Automatic Actions A. RPS trip logic A(B) half-scram occurs.

B. PCIS Group 1 half-trip logic de-energizes.

D. PCIS Group 3 isolation, RWCU:

E. PCIS Group 6 isolation, Pri Cont Vent and Purge and RB Vent F. Group 8 isolation, TIP.

G. Control Room Emergency Ventilation System start.

H. Standby Gas Treatment System starts.

4.1 Immediate Action

[1] STOP all testing with potential RPS half-scrams or PCIS logic isolation signals.

4.2 Subsequent Actions

[1] VERIFY automatic actions occur.

[2] VERIFY Steam Tunnel Booster FAN operating by observation of MAIN STEAM LINE TUNNEL TEMPERATURE,2-TIS-1-60A.

[4] IF desired to place RPS Bus B on Alternate Supply, THEN PERFORM Step 4.2[9].

[6] IF desired to Startup and Load RPS MG Set 2B, THEN PERFORM Step 4.2[11]. (Otherwise N/A) 0:>

[7] IF Alternate RPS Supply AND RPS MG Set unavailable, THEN PERFORM the following

[7 .1] ATTEMPT to determine the cause of loss of RPS Circuit Protector by using the indicating lights inside the RPS cabinets.(Otherwise N/A)

[7.2] CONTACT Electrical Maintenance to correct the cause. e:>

[9] RESTORE Alternate Power to RPS Bus B as follows:

[9.1] VERIFY memory lights inside RPS Circuit Protector cabinets are RESET.

RPS CIRCUIT PROT RESET PB CAB 2C1, 2-HS-099-0002C1 RPS CIRCUIT PROT RESET PB CAB 2C2, 2-HS-099-0002C2 Page 19 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2

[9.2] VERIFY Circuit Protectors 2C1 and 2C2 are RESET.

RPS CIRCUIT PROTECTOR 2C2, 2-HS-099-0002C1/1 RPS CIRCUIT PROTECTOR 2C2, 2-HS-099-0002C2/1

[9.3] VERIFY ALTERNATE SOURCE AVAILABLE amber light ILLUMINATED.

[9.4] In Battery Board Room No. 2 PLACE RPS BUS B NORMAL/ALTERNATE TRANSFER SWITCH, 2-XS-099-0002B, to ALT.

[9.5] NOTIFY Unit 2 Operator that RPS 2B is on Alternate power supply.

[9.6] PROCEED to Step 4.2[12].

NRG End of Event #6 Page 20of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _7_ Page 1 of 1 Event

Description:

Trip of RFP 2C/ recover with already warm RFP 28 Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 7, insert F7 Driver (imf fw14c) to trip reactor feed pump 2C Reports RFPT TRIPPED (2-XA-55-6C, Window 29) in alarm.

ATC Monitors indications and reports RPT 'C' tripped.

Monitors reactor water level SRO Directs entering AOl-3-1 to control reactor water level 2-AOl-3-1 4.2 Subsequent Actions 5.0 LOW REACTOR WATER LEVEL OR LOSS OF FEEDWATER

[1] IF Feedwater Control System has failed, THEN PERFORM the following:

[1.1] PLACE individual RFPT Speed Control Raise/Lower switches in Manual Governor (depressed position with amber light illuminated).

[1.2] ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain Reactor Water Level.

Verifies '28' RFPT speed control switch in manual and raises RFPT speed to control and maintain reactor water level.

Places '28' RFPT in automatic control Reports '28' RFPT in service SRO Calls WC to investigate the trip of '2C' RFPT Driver As WC, acknowledge WO to investigate the trip of '2C' RFPT NRC End of Event #7 Page 21 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 1 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 8, insert F8 Driver (bat NRC/1501-3-8) to insert a torus leak and ATWS Reports multiple room flood alarms in secondary containment and ATC/BOP SUPPR CHAMBER WATER LEVEL ABNORMAL (2-XA-55-3B, Window 15) in alarm 2-ARP-9-38 A. CHECK Suppression Pool level using multiple indications.

B. IF level is low, THEN DISPATCH personnel to check for leaks.

E. REFER TO Tech Spec 3.6.2.2.

F. IF level is above -1" or below -6.25" AND NOT in Mode 4 or Mode5 THEN (otherwise N/A)

ENTER 2-EOl-2 Flowchart.

Reports suppression pool water level lowering BOP Calls AUO to check for leaks in reactor building pump rooms and torus area Enters EOl-3 on secondary containment flood alarms EOl-3 IF SAMG entry is required and the TSC SAM team has assumed command and control - NO Rx Zone ventilation exh radiation lvl is above 72 mR/hr - NO Refuel Zone ventilation exh radiation lvl is above 72 mR/hr - NO Rx Zone ventilation is isolated AND SRO Rx Zone ventilation exh radiation lvl is below 72 mR/hr - NO Refuel Zone ventilation is isolated AND Refuel Zone ventilation exh radiation lvl is below 72 mR/hr - NO CAUTION

1. 7 Spent fuel pool temp and level Page 22 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 2of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior SC/L WHEN any floor drain sump water level is above 66 in. - YES WHEN any area water level is above 2 in. - YES RESTORE and MAINTAIN floor drain sump water lvls Below 66 in. using all available sump pumps - CHECKED RESTORE and MAINTAIN area water lvls below 2 in. using all available sump pumps - CHECKED SRO WHEN any floor drain sump water lvl cannot be restored and maintained below 66 in - YES WHEN any area water lvl cannot be restored and maintained below 2 in. - YES ISOLATE all systems that are discharging into the area EXCEPT systems required:

For damage control OR To be operated by EOls - CHECKED After being dispatched to check water levels, wait 5 minutes and report that water level is approximately 4 inches and rising in the Driver southwest quad. Water is flowing in from the torus area. Source of the leak in unknown Will emergency depressurization reduce discharge into Secondary Cntmt- NO WHEN water lvls in 2 or more areas are above 20 in.

Enters EOl-2 on Supp Pl Lvl below -6.25 in SRO SAMG entry is required and the TSC SAM team has assumed command and control - NO SAMG entry is required - NO CAUTION

1. 4 PC press vs pump NPSH - CHECKED Page 23 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 3of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level 1 ime Position Applicant's Actions or Behavior OW sprays have been initiated - NO Suppr chmbr sprays have been initiated - NO PC water level CANNOT be restored and maintained below 105 ft - CAN OR Suppr chmbr press CANNOT be restored and maintained below 55 psig - CAN SP/L Primary Containment Flooding is required - NO AND SAMG entry is NOT required MONITOR and CONTROL suppr pl lvl between -1 in. and -6 in.

SRO (APPX 18)

Directs BOP to control suppression pool water level IAW App. 18 Can suppr pl lvl be maintained above-6 in. - NO IF THEN Suppr pl lvl CANNOT be SECURE HPCI irrespective of Maintained above 12.75 ft whether adequate core cooling Is assured Sets a trigger value, on suppression pool level, for securing HPCI Briefs crew on current plant conditions including securing HPCI at the trigger value for SPL.

EOI APPENDIX-18 SUPPRESSION POOL WATER INVENTORY REMOVAL AND MAKEUP BOP NOTE: All panel operations performed at Control Room Panel 2 3 unless otherwise stated.

1. IF ..... Suppression Pool Water makeup is required, THEN ... CONTINUE in this procedure at Step 5.

Page 24of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 4of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior

5. IF ..... Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN ... CONTINUE in this procedure at Step 9 using RHR Loop I OR Step 10 using RHR Loop II.

6. IF ......... Directed by SRO to add water to suppression pool, THEN ... MAKEUP water to Suppression Pool as follows:

BOP

a. VERIFY OPEN 2-FCV-73-40, HPCI CST SUCTION VALVE.

b. OPEN 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE.

Adds water to the suppression pool IAW App. 18 Monitors Suppression Pool water level and reports that level is continuing to lower.

SRO Directs securing HPCI BOP Secures HPCI by placing HPCI Aux Oil Pump in PTL.

When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.

1. Safety Significance:

Prevent failure of Primary Containment from pressurization of the Suppression Chamber

2. Cues:

Procedural compliance Suppression Pool Level indication

3. Measured by:

Observation - HPCI Auxiliary Pump placed in Pull to Lock

4. Feedback:

HPCI does not Auto initiate No RPM indication on HPCI This Critical Task is not met if the Crew fails to place the HPCI AOP in PTL at 12.75 ft.

Page 25of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 5of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior MAINTAIN suppr pl lvl above 11.5 ft (APPX 18)

SRO BEFORE Suppr pl lvl drops to 11.5 ft CONTINUE As Suppression Pool level lowers, the SRO will take the conservative action to start the shutdown with 2-AOl-100-1.

SRO Directs the ATC to enter AOl-100-1 and insert control rods in accordance with 2-01-85; 2-01-85 Control Rod Drive System Section 6.7 Control Rod Insertion.

[4] PERFORM the following to insert the control rod as appropriate.

Control Rod Continuous Insertion per Section 6.7.3.

6.7.3 Continuous Insertion of Control Rod

[1] VERIFY Section 6.7.1 has been performed.

[2] SELECT desired control rod by depressing appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.

ATC

[3] OBSERVE the following for selected control rod:

D CRD ROD SELECT pushbutton is brightly ILLUMINATED.

D White light on Full Core Display ILLUMINATED

[4] PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, to RODIN.

[5] WHEN control rod notch reaches even rod notch position prior to desired final control rod notch position, THEN RELEASE CRD CONTROL SWITCH, 2-HS-85-48.

[6] OBSERVE the control rod settles into desired position AND ROD SETTLE light extinguishes.

Page 26 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 6 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level e Position Applicant's Actions or Behavior Before Torus Level reaches 11.5 feet the SRO will enter EOl-1 SRO and direct the ATC to initiate a manual reactor scram.

Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS

[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS-99-5A/S3B, on Panel 2-9-5.

[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START &

HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A)

[3] Refuel Mode One Rod Permissive Light check:

ATC [3. 1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S 1, in REFUEL.

[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-Xl-85-46.

[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-Xl-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In.

(Otherwise N/A)

[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN.

Page 27of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 7 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior

[5] REPORT the following status to the US:

• Reactor Scram

• Mode Switch is in Shutdown

• "All rods in" or "rods out"

• Reactor Water Level and trend (recovering or lowering)

• Reactor pressure and trend

• MSIV position (Open or Closed)

• Power level Completes immediate operator actions and makes scram report including 'rods out' and 'power level <5%.

ATC 2.0 SUBSEQUENT ACTIONS:

[1] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following (otherwise N/A):

[1.1] INITIATE ARI by Arming and Depressing BOTH of the following:

• ARI Manual Initiate, 2-HS-68-119A

• ARI Manual Initiate, 2-HS-68-1198

[1.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 2-9-4.

[1.3] REPORT "ATWS Actions Complete" and power level.

Completes subsequent actions Enters EOl-1, RPV Control at Step RC-1 SAMG entry is required and the TSC SAM team has assumed SRO command and control - NO VERIFY RX scram Directs ATC to initiate a manual reactor scram Page 28 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 8 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time - _::**on Applicant's Actions or Behavior EOl-1 RC/Q MONITOR and CONTROL reactor power - CHECKED The reactor will remain subcritical without boron under all conditions - NO SRO The reactor is subcritical - NO AND NO boron has been injected VERIFY reactor mode switch in SHUTDOWN - CHECKED INITIATE ARI - CHECKED Will tripping Recirc pumps cause trip of Main Turbine, RFP, HPCI, or RCIC-NO Is reactor power above 5% OR unknown - NO BEFORE suppr pl temp rises to 110°F CONTINUE - CHECKED RESET ARI DEFEAT ARI logic trips if necessary (APPX 2)- CHECKED INSERT control rods using ANY of the following methods:

Scram valves opened 1. RESET scram But SDV is full DEFEAT RPS logic trips SRO if necessary 1F 2.DRAIN SDV

3. RECHARGE accumulators

4. INITIATE reactor scram Manual control rod 1. DRIVE control rods Insertion methods BYPASS RWM if 1D necessary Directs ATC to perform App. 1F and 2 and to drive control rods with App. 1D Directs personnel to perform App. 2 and the outside portions of ATC App. 1F Page 29 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 8 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Acknowledge performing App. 2 and the outside portions of App.

1F. Wait 3 minutes then insert Shift F8 (bat app02) to defeat ARI Driver logic. Wait 5 minutes then insert Shift F9 (bat app01f) to defeat RPS logic EOl-1 RC/L CAUTION

1. 1 Ambient temp may affect RPV water lvl SRO indication and trend - CHECKED MONITOR and CONTROL RPV water lvl - CHECKED VERIFY each as required:

• PCIS isolations (Groups 1, 2 and 3)

• ECCS

• RCIC It has NOT been determined EXIT RC/L and that the reactor will remain ENTER C5, Level/Power Control subcritical without boron under all conditions C-5 CAUTION SRO

1. 1 Ambient temp may affect RPV water lvl indication and trend RPV water lvl CANNOT be determined - CAN The reactor will remain subcritical without boron under all conditions - NO PC water lvl CANNOT be maintained below 105 ft- CAN OR Suppr chmbr press CANNOT be maintained below - CAN 55 psig OW Control Air becomes unavailable -AVAIL Page 30 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 8 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior INHIBIT ADS Is ANY main steam line open - YES BYPASS the following isolation interlocks:

• MSIV low low low RPV water lvl (APPX 8A)

• RB ventilation low RPV water lvl (APPX 8E)

Directs BOP to inhibit ADS Calls for App 8A and 8E to be performed Acknowledge performing App 8A and App8E.

Driver Wait 5 minutes then insert Shift F10 (bat app08a) and Shift F11 (bat app08e). Call US and report App 8A and 8E are complete.

Inhibits ADS BOP Reports ADS inhibited Page 31of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 9 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Suppr pl temp is above 110°F - BELOW AND Reactor power is above 5% OR unknown - BELOW AND An MSRV is open or cycling OR OW press is above 2.4 psig - NO AND RPV water lvl is above -162 in. - YES Is reactor power above 5% OR unknown - NO Emergency RPV depressurization is required - NO Reactor power is above 5% - BELOW SRO OR unknown AND RPV water lvl is above -50 in.

CAUTION

1. 5 Rapid RPV inj may cause core damage

1. 2 Pump NPSH and Vortex limits

1. 3 Elevated suppr chmbr press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F MAINTAIN RPV water lvl between -180 in. and +51 in. with the following inj sources CNDS and FW 5A 1210 psig Directs ATC to maintain RPV water level +2 in. to +51 in. using App. 5A Condensate and Feedwater Page 32 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 10of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time  ?::::ition Applicant's Actions or Behavior Maintains RPV water level +2 in to +51 in IAW App 5A When it is reported that App 2 and the outside portions of App 1F are complete, continues with App 1F ATC EOI Appendix-1F

[2] WHEN RPS Logic has been defeated, THEN RESET Reactor Scram.

[3] VERIFY OPEN Scram Discharge Volume vent and drain valves EOI Appendix-1 D

[1] VERIFY at least one CRD pump in service.

[3] VERIFY REACTOR MODE SWITCH in SHUTDOWN.

[4] BYPASS Rod Worth Minimizer.

[5] REFER TO Attachment 2 and INSERT control rods in the area of highest power as follows:

ATC [5.1] SELECT control rod.

[5.2] PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.

[5.3] REPEAT Steps 1.0[5.1] and 1.0[5.2] for each control rod to be inserted Completes control room portions of App 1F and inserts control rods IAW App 1D Reports driving control rods Page 33 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 10of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior RC/P DW press is above 2.4 psig - NO Emergency RPV depressurization is anticipated - YES AND The reactor will remain subcritical without boron under all conditions - NO Emergency RPV depressurization is or has been required - NO SRO RPV water lvl CANNOT be determined - CAN Is ANY MSRV cycling - NO Steam cooling is required - NO Suppr pl temp and lvl CANNOT be maintained in a safe area of Curve 3 at the existing RPV press - CAN Page 34of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 11 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Suppr pl lvl CANNOT be maintained in the safe area - CAN of Curve 4 OW control air becomes unavailable - AVAIL Boron inj is required - NO AND The main condenser is available SRO AND There has been no indication of a steam line break STABILIZE RPV press below 1073 psig with the main turbine bypass vlvs (APPX 8B)

Directs BOP to control reactor pressure 800 psig to 1000 psig in accordance with Appendix 8B Verifies that main turbine bypass valves are controlling reactor pressure 800 psig to 1000 psig BOP Reports that suppression pool water level is continuing to lower Page 35of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 12of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Sets a trigger valve and briefs the crew on the requirement to emergency depressurize the RPV before reaching 11.5 ft. in the torus.

EOl-2 SP/L WHEN suppr pl lvl CANNOT be maintained above 11.5 ft.

EMERGENCY RPV DEPRESSURIZATION IS REQUIRED(EOl-1, SRO RC/P-4; C1-1, C1-20; C5-12, C5-14)

EOl-1 RC/P Emergency RPV depressurization EXIT RC/P and

~ is or has been required ENTER C2, Emergency Depressurization Enters C2 for Emergency Depressurization SAMG entry is required and the TSC SAM team has assumed command and control - NO RPV water lvl CANNOT be determined - CAN Containment water lvl CANNOT be maintained below 44 ft- CAN OW control air becomes unavailable - AVAIL Will the reactor remain subcritical without SRO Boron under all conditions - NO WHEN all inj into the RPV is stopped and prevented EXCEPT from RCIC, CRD, and SLC per C5, Level/Power Control, Step C5-22 Page 36 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 12of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior C-5 Emergency RPV depressurization is Required STOP and PREVENT ALL inj into RPV EXCEPT from RCIC, CRD and SLC (APPX 4)

Directs ATC to terminate and prevent condensate and feedwater using APP 4 Directs BOP to terminate CS and LPCI using App 4 When App 4 is directed, insert F9 (imf fw30b 55.7) to fail RFP 'B' Driver Woodward governor.

When ATC is using App 4 to terminate and prevent condensate NRC and feedwater, he will recognize the failure of 'B' RFP speed to lower. This will start Event No. 9.

Page 37of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 13of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior EOI APPENDIX-4 NOTE Following receipt of a CORE SPRAY automatic initiation signal, it is NOT necessary to wait until a pump starts before performing step 3.

3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.

4. PREVENT injection from LPCI SYSTEM I by performing the following:

NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.

a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig, BOP 1) PLACE 2-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE.

5. PREVENT injection from LPCI SYSTEM II by performing the following:

NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.

a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.

OR

b. BEFORE RPV pressure drops below 450 psig,

1) PLACE 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

AND

2) VERIFY CLOSED 2-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.

Page 38 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 14of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Insert F10 (bat NRC/1501-3-10) to prevent two ADS valves from Driver opening Terminates and prevents injection on 9-3 IAW App 4.

BOP Reports terminate and prevent complete

6. PREVENT injection from CONDENSATE and FEEDWATER by performing the following:

c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:

• 2-FCV-3-19, RFP 2A DISCHARGE VALVE ATC

• 2-FCV-3-12, RFP 2B DISCHARGE VALVE

• 2-FCV-3-5, RFP 2C DISCHARGE VALVE

• 2-LCV-3-53, RFW START-UP LEVEL CONTROL Terminates and prevents injection on 9-6 IAW App 4.

Reports terminate and prevent complete C-2 Is suppr pl lvl above 5.5 ft. - YES SRO OPEN all ADS vlvs Directs BOP to open all ADS valves When SRO directs opening all ADS valves, Event No. 1 O would NRC start C-2 Are at least 4 MSRVs open - YES SRO The reactor is NOT subcritical - >A Page 39 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 15of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required, RO initiates Emergency Depressurization as directed by US.

1. Safety Significance:

Precludes failure of Containment

2. Cues:

Procedural compliance Suppression Pool Level Trend

3. Measured by:

Observation - US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet.

Observation - RO opens at least 6 SRV's during performance of Emergency Depressurization actions.

4. Feedback:

RPV pressure trend SRV status indications Suppression Pool temperature trend This Critical Task is not met if Emergency Depressurization is not initiated until Suppression Pool Level gets to 11.5 feet.

Page 40of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 16 of 19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior C-5 Can at least two MSRVs be opened per C2, Emergency RPV Depressurization - YES WHEN RPV press is below MSCP (Table 1A)

SRO Table 1A Minimum Steam Cooling Press 6 or more 190 psig Briefs crew that injection to the RPV will begin when RPV pressure is less than 190 psig CAUTION

1. 5 Rapid RPV inj may cause core damage

1. 2 Pump NPSH and Vortex limits

1. 3 Elevated suppr chmbr press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F SRO START and SLOWLY RAISE RPV inj with the following inj sources to restore and maintain RPV water lvl above -180 in.

CNDS and FW 5A 1210 psig Directs ATC to start and slowly raise injection to the RPV, using condensate IAW App 5A, when reactor pressure lowers to less than 190 psig.

Page 41of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 17of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior During an A TWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection from ECCS and Feedwater until reactor pressure is below the MSCP as directed by US.

1. Safety Significance:

Prevention of fuel damage due to uncontrolled feeding.

2. Cues:

Procedural compliance.

3. Measured by:

Observation - No EGGS injection prior to being less than the MSCP.

Observation - Feedwater terminated and prevented until less than the MSCP.

4. Feedback:

Reactor power trend, power spikes, reactor short period alarms.

Injection system flow rates into RPV.

Page 42 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 18of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior When reactor pressure lowers to less than 190 psig, slowly starts injection to the RPV IAW App 5A ATC Reports injection to the RPV Determines that RPV water level is above -180 in. and rising Can RPV water lvl be restored and maintained above -180 in. -YES Was RPV water lvl lowered for Step C5 NO >D SRO Maintain RPV water lvl between -180 in and +51 in. with the following inj sources:

CNDS and FW 5A Directs ATC to restore RPV water level to +2 in to +51 in. using condensate App 5A Page 43 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _8_ Page 19of19 Event

Description:

Loss of Torus Water level /SCRAM (ATWS)and ED on Torus water level Time Position Applicant's Actions or Behavior Continues to restore RPV water level to +2 in to +51 in using condensate Appendix-1 F 1.0 INSTRUCTIONS (continued)

[4] DRAIN SDV UNTIL the following annunciators clear:

• WEST CRD DISCH VOL WfR LVL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 1). D

• EAST CRD DISCH VOL WfR LVL HIGH HALF SCRAM (Panel 2-9-4, 2-XA-55-4A, Window 29).

ATC NOTES

1) If EOI Appendix 2 has been executed, ARI initiation or reset will NOT be possible or necessary in Step 1.0[6].

2) If reactor pressure is greater than 600 psig, SRO may direct performance of step 1.0[6] prior to accumulators being fully recharged.

[6] WHEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.

Monitors for indication that the Scram Discharge Volumes are drained.

Reports when SDV are drained Reports 'initiating a manual reactor scram' ATC Initiates a manual reactor scram Reports rod movement Resets Reactor Scram and verifies Scram Discharge Volume vent and drain valves open Scans full core display and reports all control rods inserted.

NRG End of Event #8 and scenario Place simulator in FREEZE upon direction of the NRG Chief Driver Examiner Page 44 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _9_ Page 1 of 1 Event

Description:

Failure of RFP 28 governor/pump needs tripped.

Time Position Applicant's Actions or Behavior EOI APPENDIX-4

6. PREVENT injection from CONDENSATE and FEEDWATER by performing the following:

b. LOWER RFPT 2A(28)(2C) speed to minimum setting (approximately 600 rpm) using ANY of the following methods on Panel 2-9-5:

• Using 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL AND individual 2-SIC-46-8(9)(10), RFPT 2A(28)(2C) SPEED CONTROL in AUTO, OR

• Using individual 2-SIC-46-8(9)(10), RFPT 2A(28)(2C)

SPEED CONTROL in MANUAL, OR

• Using individual 2-HS-46-8A(9A)(1 QA), RFPT ATC 2A(28)(2C)SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR.

Attempts to lower RFPT '8' speed using App 4 Reports that speed cannot be lowered in manual or automatic

a. IF ................. Immediate injection termination from a reactor feedwater pump is required, THEN ........... PERFORM step 6.d for the desired pump.

d. TRIP RFPTs as necessary to prevent injection by DEPRESSING the following push-buttons:

• 2-HS-3-151A, RFPT 28 TRIP Trips 28 RFPT Reports 28 RFPT tripped NRG End of Event #9 Page 45of47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _3_ Event No.: _1Q_ Page 1 of 1 Event

Description:

ADS SRV Failures Time Position Applicant's Actions or Behavior Opens six ADS valves Observes acoustic monitor and determines that only four ADS BOP valves have opened Reports that only four ADS valves have opened C-2 Can 6 ADS vlvs be opened - NO OPEN additional MSRVs as necessary to establish 6 MSRVs SRO open (ok to exceed 100°F/hr cooldown rate)

Directs BOP to open additional MSRVs as necessary to get six MSRVs open Opens two additional MSRVs BOP Reports that two additional MSRVs have been opened and that six total MSRVs are now open NRG End of Event #10 Page 46 of 47 Unit 2

Appendix D Required Operator Actions Form ES-D-2 Scenario Setup IC 9 Exam IC 96 Batch File or Pref File I1501NRC3

?~;~t~titl:i j;11111ral' ¥~~~~1;

• • ~iilij~ * *vallJcf RD07R2243 Rod Drift In 22-43 NIA NIA NIA NIA NIA RD06R2243 Rod Stuck 22-43 00:35 NIA NIA NIA NIA OG04A SJAEA Tri NIA NIA NIA NIA NIA NM05B NIA 100 NIA NIA 100 D Diesel Fails to Auto DG03D Active NIA NIA NIA NIA NIA Tie to SID Bus FW14C 7 NIA NIA NIA NIA NIA PC14 Torus Water Leak 8 NIA 100 5:00 NIA NIA AD01D ADS Valve Failure 10 NIA NIA NIA NIA NIA AD01E ADS Valve Failure 10 NIA NIA NIA NIA NIA r*>s**'~ver1

• *+.:~

.:7;.: 'R*.*..*.*a'*.*.*.*.*.m*.*.* .**...P** . * .***.s.*.* **.*1n**,*1**,**a,."1t F. .*1*n**a**1 zi:,~el'lt~: ?QJ!laY: *:~~v~ri~ ~"Raft'.Jp, .. lfli~~'< ~J11~1

'****:w> ** .<. *ya,llla *:Value.

SID Bus D Norm ZDIOHS2110D22A 6 0 NIA NIA CLOSE TRIP Feeder Bkr Tri SID Bus D Norm ZLOOHS211 OD22A Feeder Bkr White 6 0 NIA NIA OFF ON Li ht Batch I Pref File(s):

Page 47 of 47 Unit 2

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.5(A)

Unit2 Rev.0024 Paae 126 of 363 Attachment 5 (Page 25 of 39)

A2 Startup Sequence Control Rod Movement Data Sheet RWM ROD FROM TO Rod Movement Completed GP NUMBER Signoffs UO {AC' 1 Peer Check 2 36 02-35 00 12 o~ <l(

00 12 fX 36 36 26-59 34-59 00 12 1A7 !5fj R

If~

36 58-35 00 12 r~kL 36 58-27 00 12 ty 36 34-03 00 12 1r 36 26-03 00 12 f v f.:.:::

36 02-27 00 12 (,, v rl IL 37 06-47 00 12 DV 37 14-55 00 12 <<-fi v 37 46-55 00 12 37 54-47 00 12 -/; v <(

37 54-15 00 12 l.1 v 37 46-07 00 12 6 cl (_

37 14-07 00 12 /)f"'

37 06-15 00 12 IY NOTES:

1 For all rod moves to the "full out" position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

2 Documentation of Peer Check by a second qualified member of the plant staff (i.e., RE, STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal powers10%.

3 Record the rod number and any problems encountered, as applicable.

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.S(A)

Unit2 Rev.0024 Page 127 of 363 Attachment 5 (Page 26 of 39)

A2 Startup Sequence Control Rod Movement Data Sheet RWM ROD FROM TO Rod Movement Completed GP NUMBER Signoffs UO (AC) 1 Peer Check 2 38 02-35 I 12 48 38 26-59 12 48 38 34-59 12 48 38 58-35 12 48 38 58-27 12 48 38 34-03 12 48 38 26-03 12 48 38 02-27 12 48 39 06-47 12 48 39 14-55 12 48 39 46-55 12 48 39 54-47 12 48 39 54-15 12 48 39 46-07 12 48 39 14-07 12 48 39 06-15 12 48 REMARKS 3 :

NOTES:

1 For all rod moves to the "full out" position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

2 Documentation of Peer Check by a second qualified member of the plant staff (i.e .. RE, STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal powerS10%.

3 Record the rod number and any problems encountered, as applicable.

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.5(A)

Unit 2 Rev.0024 Paae 128 of 363 Attachment 5 (Page 27 of 39)

A2 Startup Sequence Control Rod Movement Data Sheet RWM ROD FROM TO Rod Movement Completed GP NUMBER Signoffs UO (AC) 1 Peer Check 2 40 02-43 00 12 40 18-59 00 12 40 42-59 00 12 40 58-43 00 12 40 58-19 00 12 40 42-03 00 12 40 18-03 00 12 40 02-19 00 12 41 10-51 00 12 41 50-51 00 12 41 50-11 00 12 41 10-11 00 12 NOTES:

1 For all rod moves to the "full out" position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

2 Documentation of Peer Check by a second qualified member of the plant staff (i.e., RE, STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal powers10%.

3 Record the rod number and any problems encountered, as applicable.

BFN Control Rod Coupling Integrity Check 2-SR-3.1.3.S(A)

Unit2 Rev.0024 Page 129 of 363 Attachment 5 (Page 28 of 39)

A2 Startup Sequence Control Rod Movement Data Sheet RWM ROD FROM TO Rod Movement Completed GP NUMBER Signoffs UO (AC) 1 Peer Check 2 42 02-43 12 48 42 18-59 12 48 42 42-59 12 48 42 58-43 12 48 42 58-19 12 48 42 42-03 12 48 42 18-03 12 48 42 02-19 12 48 43 10-51 12 48 43 50-51 12 48 43 50-11 12 48 43 10-11 12 48 NOTES:

1 For all rod moves to the "full out" position (notch position 48), this signoff verifies coupling integrity was checked in accordance with 2-01-85.

2 Documentation of Peer Check by a second qualified member of the plant staff (i.e., RE, STA OR UO) is required ONLY when the RWM is inoperable OR bypassed with core thermal powers10%.

3 Record the rod number and any problems encountered, as applicable.

Appendix D Scenario Outline Form ES-D-1 Facility: BFN Scenario No.: NRC-4 Op-Test No.: 1501 Examiners: Operators: SRO:

ATC:

BOP:

Initial Conditions: Reactor Power is 90%. EECW Pump A3 is out of service for maintenance.

Turnover: Complete Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9); start at step 7.3.

Raise reactor power to 100% after 3-SR-3.3.1.1.8(9).

Event Malfunction Event Event Description No. Number Type*

Conducts Turbine Control Valve Fast Closure, or Turbine 1 N/A N-BOP Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Failure of RPS B to trip.

2 EG13A C-BOP Bus Duct Cooling Fan 3A trip C-BOP Loss of 480V S/D Board 3B and failure of RBCCW 3 ED10B C-ATC sectionalizing valve to auto-close.

TS-SRO C-BOP 4 MC04 R-ATC Loss of Condenser Vacuum and power reduction.

C-SRO 5 RC03 C-BOP RCIC Suction Pressure Trip 6 ED01 M-ALL Loss of Offsite Power 7 DG01A C-BOP DG 3EA Fails to Auto start 8 TH21 M-All LOCA 9 HP04 C-BOP HPCI Steam Supply Valve fails to auto open.

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

Appendix D Scenario Outline Form ES-D-1 Critical Tasks 2

1. RPV Level maintained above TAF ((-)162 inches)

1. Safety Significance:

Maintaining adequate core cooling

2. Cues:

RPV level indication

3. Measured by:

Reactor level indication above (-)162 inches

4. Feedback:

RPV level trend HPCI injection valve open indication

2. When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and before Drywell temperature rises to 280°F.

1. Safety Significance:

Precludes failure of containment

2. Cues:

Procedural compliance High Drywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation - US directs Drywell Sprays IAW with EOI Appendix 178 AND Observation - RO initiates Drywell Sprays

4. Feedback:

Drywell and Suppression Pressure lowering RHR flow to containment

Appendix D Scenario Outline Form ES-D-1 Events

1. BOP will conduct Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9). When 3-SR-3.3.1.1.8(9) has been secured and the Tech Spec LCO declared. The Crew will call to have the RPS Logic Fuses pulled to insert a half scram. At the direction of the lead examiner the scenario may continue.

2. Bus Duct Cooling Fan 3A will trip and annunciator 3-XA-55-7A window 31, GEN BUS DUCT FAN FAILURE, will be received. The BOP operator will place the 3B Bus Duct Cooling Fan in service. When the 3B Bus Duct Cooling Fan is in service the scenario may continue.

3. The crew will respond to a loss of 480V Shutdown Board 3B. This will cause a loss of RPS B, loss of 480V RMOV BO 3B, 3C and 3E. The crew will need to restore power to the 480V RMOV Boards, reset RPS, reset PCIS and restore systems. The SRO will refer to Technical Specification 3.8.7 and determine conditions A, B, C, and Dare required. Loss of the Shutdown Board will also result in entry into T.S 3.5.1 for Division 2 RHR LPCI because the power supply to the valves has been placed on its alternate source. PAM instrumentation LCO 3.3.3.1 and ODCM 1 I 2.1.1 for Radioactive Liquid Monitors. When power has been restored to the RMOV boards, the crew will respond to a trip of RBCCW Pump 3B IAW 3-AOl-70-1, and notice the RBCCW sectionalizing valve failing to auto close and manually closing the sectionalizing valve. When the Tech Spec calls are completed and the lead examiner is ready the scenario may continue.

4. Condenser Vacuum will begin to degrade causing the Crew to enter 3-AOl-47-3 and direct reducing reactor power in an attempt to maintain condenser vacuum. Condenser Vacuum will continue to degrade. The SRO will set a trigger value to trip the main turbine and scram the reactor before an automatic turbine trip occurs at approximately 24.3 inched Hg.

5. If the Crew attempts to place RCIC in service, it will trip on low suction pressure. It will remain unavailable from that point forward.

6. After the Reactor Scram on vacuum, a Loss of Offsite Power will occur. The crew will respond to the Reactor Scram IAW 3-AOl-100-1 and O-AOl-57-1A.

7. During the LOOP DG 3EA will fail to automatically start and will have to be manually started and after it starts auto tie to the buss.

6. Sometime after the LOOP a LOCA will develop requiring the crew to use systems to maintain Reactor Level and Containment parameters.

7. The HPCI Steam Supply Valve, 3-FCV-73-16, will fail to OPEN on an automatic HPCI initiation signal, but can be initiated manually.

The scenario ends when Drywell Sprays have been initiated and Reactor Level is maintained above TAF ((-)162 inches) or upon request of Lead Examiner.

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _1_ Page 1 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior 7.3 Turbine Control Valve, 3-FCV-1-75 (CV #1) Test NOTE Indicator light 3-IL-099-5A-DS15B on Panel 9-17, Bay 1, will extinguish when jumper is installed in Step 7.3[1].

[1] PERFORM the following in Panel 3-9-17 (Bay 1), to simulate Turbine Control Valve, 3-FCV-1-80 (CV #2), is CLOSED:

[1.1] REMOVE the back cover from RPS CH B 1 CONTROL BOP VALVE 3A FAST CLOSURE, 3-RLY-099-05AK08F (H3),

if required (Otherwise N/A)

[1.2] INSTALL jumper across contacts 3 to 4 on relay 3-RLY-099-05AK08F.

Calls operator in Aux Instrument Room to install jumper on relay 99-05AK08F As operator in AIR, acknowledge jumpering contacts 3 to 4 on Driver relay 3-RLY-099-05AK08F. Insert Trigger 1 (irf rp15b jumpered)

[3] CHECK the following indications:

• RPT SYS A TURB CONTROL VLV, 3-IL-099-5A-DS15A, on Panel 9-15, Bay 1, is illuminated.

• RPT SYS A TURB CONTROL VLV, 3-IL-099-5A-DS15B, on Panel 9-17, Bay 1, is extinguished.

[4] ENSURE Reactor power less than 95%.

[5] RECORD valve position in terms of percent below.

• CV-1, 3-Zl-1-75  %

BOP

• CV-2, 3-Zl-1-80  %

• CV-3, 3-Zl-1-85  %

• CV-4, 3-Zl-1-89  %

Calls operator in Aux Instrument Room to verify indications on panels 9-15 and 9-17 Verifies reactor power less than 95%

Records control valve position Page 1 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No.~ Event No.: _1_ Page 2 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior As operator in AIR, report that 3-IL-099-5A-DS15A is illuminated Driver and 3-IL-099-5A-DS15 is extinguished.

CAUTION If power is not simulated, 3-HS-47-158 should be held in until reactor water level, reactor pressure, and neutron flux signals have stabilized to prevent flux spiking, which would cause a reactor scram.

[6] NOTIFY the UO that performance of the following step will result in a channel A Half Scram.

[7] DEPRESS and HOLD CV-1 TEST push-button, 3-HS-47-158, on Panel 3-9-7 until Step 7.3[12].

Updates crew that the next step will cause a channel A half scram Depresses and hold CV-1 TEST pushbutton BOP

[8] CHECK the following on Panel 3-9-5:

• 3-IL-99-5A/AB, SYSTEM A BACKUP SCRAM VALVE, light on Panel 3-9-5 is extinguished.

• 3-IL-99-5A/CD, SYSTEM B BACKUP SCRAM VALVE, light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP A LOGIC RESET 1 red light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP A LOGIC RESET 2 red light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP A LOGIC RESET 3 red light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP A LOGIC RESET 4 red light on Panel 3-9-5 is extinguished.

Page 2 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _1_ Page 3 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior

[9] CHECK the following Annunciators are in alarm:

• Annunciator REACTOR CHANNEL A AUTO SCRAM (3-XA-55-5B, window 1) is in alarm.

• Annunciator TURB CONTROL VLV FAST CLOSURE HALF SCRAM (3-XA-55-4A, window 15) is in alarm.

BOP

• Annunciator RPT SYS A TRIP 3-XA-68-72 (3-XA-55-4A window 11) is in alarm.

Checks appropriate backup scram valve lights are extinguished and all four solenoid group A logic lights are extinguished Checks appropriate alarms are in alarm

[1 O] CHECK RPT SYS A TURB CONTROL VLV light indicator 3-IL-099-5A-DS15A, on Panel 9-15, Bay 1, is extinguished

[11] CHECK the following on ICS:

• ICS point SOE042 displays TRIP condition.

BOP

• ICS point SOE035 displays TRIP condition.

• ICS point DIG081 displays TRIP condition Calls operator in aux instrument room to verify indication on panel 9-15 Verifies appropriate ICS points display TRIP condition As operator in AIR, report that 3-IL-099-5A-DS15A, on Panel 9-15, Driver Bay 1, is extinguished.

Page 3 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __£_ Event No.: _1_ Page 4 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior

[12] WHEN Rx Water Level, Rx Pressure, and Neutron Flux signals have stabilized, THEN RELEASE CV-1 TEST push-button, 3-HS-47-158.

[13] RESET RPS HALF SCRAM SIGNAL on 3-9-5.

[14] RESET the following Panel annunciators:

• RESET affected annunciators 3-HS-55-4 on Panel 3-9-4.

• RESET affected annunciators 3-HS-55-5 on Panel 3-9-5.

Releases TEST push-button Resets half scram signal and panel annunciators BOP

[15] CHECK the following on Panel 3-9-5:

• All four SYSTEM A and SYSTEM B BACKUP SCRAM VALVE lights on Panel 3-9-5 are illuminated.

• All eight SCRAM SOLENOID GROUP A & B LOGIC RESET lights on Panel 3-9-5 are illuminated.

[16] CHECK RESET the following Annunciators:

• REACTOR CHANNEL A AUTO SCRAM (3-XA-55-5B, window 1).

• TURB CONTROL VLV FAST CLOSURE HALF SCRAM (3-XA-55-4A, window 15).

• RPT SYS A TRIP 3-XA-68-72 (3-XA-55-4A window 11).

Checks backup scram valve lights illuminated, scram solenoid lights illuminated and annunciators reset

[17] CHECK the following on Panel 3-9-15:

BOP *Relay RPS CH A1 CONTROL VALVE 3A FAST CLOSURE, 3-RLY-099-05AK08E in Panel 9-15, Bay 1, is energized.

• RPT SYS A TURB CONTROL VLV light indicator, 3-IL-099-5A-DS 15A, on Panel 9-15, Bay 1, is illuminated.

Page 4of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _1_ Page 5 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior

[18] CHECK the following on ICS:

• ICS point SOE042 displays NOT TRIP.

• ICS point SOE035 displays NOT TRIP.

• ICS point DIG081displays NOT TRIP.

Calls operator in aux instrument room to verify 3-RLY-099-05AK08E in Panel 9-15, Bay 1, is energized and 3-IL-099-5A-DS15A, on Panel 9-15, Bay 1, is illuminated.

Verifies ICS computer points are indicating NOT TRIP As operator in aux instrument room report 3-RLY-099-05AK08E in Driver Panel 9-15, Bay 1, is energized and 3-IL-099-5A-DS15A, on Panel 9-15, Bay 1, is illuminated.

[20] PERFORM the following for Relay 3-RLY-099-05AK08F (Panel 3-9-17, Bay 1, H3):

[20.1] REMOVE jumper from contacts 3 to 4 on relay 3-RLY-099-05AK08F.

[20.2] INSTALL cover on back of relay if removed in Step 7.3[1] (Otherwise N/A).

BOP

[21] CHECK RPT SYS A TURB CONTROL VLV light indicator, 3-IL-099-5A-DS15B, on Panel 9-17, Bay 1, is illuminated. (N/A this step if turbine first stage pressure is less than 30%.)

Calls operator in aux instrument room to remove jumper and verify 3-IL-099-5A-DS15B, on Panel 9-17, Bay 1, is illuminated As operator in aux instrument room, insert mrf rp15b norm and Driver report jumper removed and 3-IL-099-5A-DS15B, on Panel 9-17, Bay 1, is illuminated Page 5of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _L Event No.: _1_ Page 6 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior NOTE Failure to satisfactorily complete any step designated by (AC) requires immediate notification of the Unit Supervisor. The Cognizant Engineer should be notified in a timely manner.

7.4 Turbine Control Valve, 3-FCV-1-80 (CV #2) Test

[1] PERFORM the following in Panel 3-9-15 (Bay 1), to simulate Turbine Control Valve, 3-FCV-1-75 (CV #1), is CLOSED NOTE BOP Indicator light 3-IL-099-5A-DS15A on Panel 9-15, Bay 1, will extinguish when jumpers are installed in Step 7.4[21].

[1.1] REMOVE the back cover from RPS CH B1 CONTROL VALVE 3A FAST CLOSURE, 3-RLY-099-05AK08E (H3),

if required (Otherwise N/A)

[1.2] INSTALL jumper across contacts 3 to 4 on relay 3-RLY-099-05AK08E.

Calls operator in Aux Instrument Room to install jumper on relay 99-05AK08E As operator in AIR, acknowledge jumpering contacts 3 to 4 on Driver relay 3-RLY-099-05AK08E. Insert Trigger 11 (mrf rp15a jumpered)

[3] CHECK the following indications:

• RPT SYS A TURB CONTROL VLV, 3-IL-099-5A-DS15A, on Panel 9-15, Bay 1, is extinguished.

• RPT SYS A TURB CONTROL VLV, 3-IL-099-5A-DS15B, on Panel 9-17, Bay 1, is illuminated.

[4] ENSURE Reactor power less than 95%.

BOP

[5] RECORD valve position in terms of percent below.

• CV-1, 3-Zl-1-75  %

• CV-2, 3-Zl-1-80  %

• CV-3, 3-Zl-1-85  %

• CV-4, 3-Zl-1-89  %

Page 6 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: _1_ Page 7of9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior Calls operator in Aux Instrument Room to verify indications on panels 9-15 and 9-17 BOP Verifies reactor power less than 95%

Records control valve position As operator in AIR, report that 3-IL-099-5A-DS1 SA is extinguished Driver and 3-IL-099-5A-DS15 is illuminated CAUTION If power is not simulated, 3-HS-47-159 should be held in until reactor water level, reactor pressure, and neutron flux signals have stabilized to prevent flux spiking, which would cause a reactor scram.

[6] NOTIFY the UO that performance of the following step will result in a channel A Half Scram.

[7] DEPRESS and HOLD CV-1 TEST push-button, 3-HS-47-159, on Panel 3-9-7 until Step 7.4[12].

Updates crew that the next step will cause a channel B half scram Depresses and hold CV-2 TEST pushbutton

[8] CHECK the following on Panel 3-9-5:

BOP

• 3-IL-99-SA/AB, SYSTEM A BACKUP SCRAM VALVE, light on Panel 3-9-5 is extinguished.

• 3-IL-99-SA/CD, SYSTEM B BACKUP SCRAM VALVE, light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP B LOGIC RESET 1 red light on Panel 3-9-5 is extinguished

• SCRAM SOLENOID GROUP B LOGIC RESET 2 red light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP B LOGIC RESET 3 red light on Panel 3-9-5 is extinguished.

• SCRAM SOLENOID GROUP B LOGIC RESET 4 red light on Panel 3-9-5 is extinguished.

Notes that B scram solenoid logic lights are illuminated.

Informs SRO that a Channel B half scram did not occur.

Page 7of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _1__ Event No.: _1_ Page 8 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior Directs that the surveillance procedure be stopped SRO Directs that jumper be removed to place system back to initial conditions As operator in aux instrument room, insert mrf rp15a norm and Driver report that jumper across contacts 3 to 4 on relay 3-RLY-099-05AK08E has been removed As operator in AIR, report that Relay 5A-K88 and 5A-K8F did Driver not open (they are still shut)

Refers to Technical Specifications 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

9. Turbine Control Valve Fast >30% RTP 2 E Closure, Trip Oil Pressure Low SRO One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable.

OR A.2 ------NOTE----------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Place associated trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system in trip.

Updates crew on Technical Specification requirements Page 8 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _1_ Page 9 of 9 Event

Description:

Conducts Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9).

Time Position Applicant's Actions or Behavior Contacts the Work Control Center to pull fuse FU 1-1-SOCA (5A-SRO F88) in accordance with 3-01-99, Illustration #3.

NRC End of Event #1 Page 9 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _2_ Page 1 of 1 Event

Description:

Bus Duct Cooling Fan 3A trip Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 2, insert Driver Trigger 2 (malfunction to cause 'A' Bus Duct Cooling Fan to trip).

Reports GEN BUS DUCT FAN FAILURE (3-XA-55-7A, Window

31) in alarm ARP-9-7A A. VERIFY Main Bus Cooling Fans, 3-HS-262-1A or 3-HS-262-2A, indicates running on Panel 3-9-8 AND START GEN BUS DUCT HX FAN A(B) using 3-HS-2-6 1A(2A), on panel 3-9-8 to start the standby fan.

Reports 'A' Bus Duct Cooling Fan tripped Starts 'B' Bus Duct Cooling Fan Reports 'B' Bus Duct Cooling Fan in service Dispatches personnel to check 'A' Bus Duct Cooling Fan breaker and fan motor Acknowledge checking breaker and fan motor for 'A' Bus Duct Driver Cooling Fan.

NRC End of Event #2 Page 10 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _1_ Event No.: _3_ Page 1 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

When the NRC Chief Examiner is ready for Event #3, insert Driver Trigger 3 (malfunction to cause of loss of 480V S/D Board 3B).

Responds to numerous alarms CREW Diagnoses a loss of 480V Shutdown Board 3B and 480V RMOV Boards 3B and 3C and 3E.

Reports that Reactor Power, Reactor Pressure and Reactor Level are stable following the board loss.

ATC Reports a loss of RPS 'B' and Ch. 'B' half scram.

Responds to the following alarms; 8B-30, 8C-24, and 8C-31 480V SHUTDOWN BO 3B UV OR XFR (3-XA-55-8B, Window 30)

A. Overcurrent - trips and locks out normal Bkr and prevents manual transfer to alternate.

B. IF 480V Shutdown Bd 3B is lost, THEN BOP Manually TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 3-9-8.

D. IF manual transfer is NOT accomplished, THEN REFER to Tech Spec 3.8.7 and 3.8.8 Determines and reports that 38 480V SID Bd. cannot be transferred to alternate Page 11 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. --1_ Event No.: _3_ Page 2of14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Refers US to TS 3.8.7 and 3.8.8 Dispatches personnel to investigate the loss of 3B 480V SID Bd 480V REACTOR MOV BO 3B OR 3E UV (3-XA-55-8C, Window 24)

A. CHECK light indications for loss of any 480V equipment.

D. IF undervoltage or transfer has occurred, THEN

1. REFER TO TS Section 3.8.7

2. RESET possible half-scram. REFER TO 3-01-99.

BOP E. REFER TO O-Ol-57B to re-energize or transfer board.

BOP 480V REACTOR MOV BO 3C UV OR XFR (3-XA-55-8C, Window 31)

A. VERIFY automatic action:

B. CHECK light indications for loss of 480V equipment.

Checks that Diesel Aux Board 3EB auto-transfers to 480 SID Bd A.

E. REFER TO O-Ol-57B to re-energize or transfer the board.

Directs entry into AOl-99-1, Loss of Power to One RPS Bus AOl-99-1 4.0 OPERATOR ACTIONS 4.1 Immediate Action

[1] STOP all testing with potential RPS half-scrams or PCIS logic isolation signals.

Page 12of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _3_ Page 3of14 Event

Description:

Loss of 480V S/D Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior 4.2 Subsequent Actions NOTES

1) If power cannot be restored promptly to a de-energized RPS Bus, plant operation may continue until repairs are made BOP provided all plant operational limits are met

2) Loss of RPS will isolate 3-RM-90-256, Drywell Air Monitor, and TS LCO 3.4.5 Condition B should be entered.

[1] VERIFY automatic actions occur.

[2] VERIFY Steam Tunnel Booster FAN operating by observation of MAIN STEAM LINE TUNNEL TEMPERATURE, 3-TIS 60A.

[4] IF desired to place RPS Bus B on Alternate Supply, THEN SRO PERFORM Step 4.2[9]. (Otherwise N/A)

Dispatches personnel to transfer RMOV Bds 3B, 3C, and 3E to the alternate power source, then transfer RPS 'B' to alternate.

After being dispatched to restore power, wait 5 minutes then insert (bat) to restore power to RMOV Bds 38/3C/3E. Then insert (bat) to restore power to RPS 'B' and reset ATU Gross Driver Failures Call US and report power restored to RPS 'B' and RMOV Bds 3B, 3C, and 3E. ATU GROSS FAILURES reset in AIR Updates the crew on power restoration SRO Directs ATC to reset RPS logic half scram and PCIS Page 13of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _3_ Page 4of14 Event

Description:

Loss of 480V S/D Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior AOl-99-1 4.2 Subsequent Actions

[12] RESET the RPS trip logic half scram at Panel 3-9-5 as ATC follows:

NOTE The eight CONTROL ROD TEST SCRAM SOLENOID GROUP A and B LIGHTS SHOULD ILLUMINATE

[12.1] MOMENTARILY PLACE SCRAM RESET, 3-HS-99-5A-S5, as follows:

[12.2] RESET FIRST position. (Group 2/3)

[12.3] RESET SECOND position. (Group 1/4)

[12.4] NORMAL position.

[13] VERIFY the following:

ATC

[13.1] All eight SCRAM SOLENOID GROUP A/B LOGIC RESET lights ILLUMINATED.

[13.2] The following four lights ILLUMINATED:

[13.2.1] SYSTEM A BACKUP SCRAM VALVE, 3-IL-99-5A/AB. D

[13.2.2] SYSTEM B BACKUP SCRAM VALVE

[13.3] Scram Discharge Volume vent and drain valves indicate OPEN.

Page 14of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _1_ Event No.: _3_ Page 5of14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[14] RESET PCIS trip logic at Panel 3-9-4 as follows:

[14.1] MOMENTARILY PLACE PCIS DIV I RESET, 3-HS 16A-S32, to left and right RESET positions.

[14.2] CHECK the following red lights ILLUMINATED:

[14.2.1] MSIV GROUP A1, 3-IL-64-A1 BOP [14.2.2] MSIV GROUP B1, 3-IL-64-B1

[14.3] MOMENTARILY PLACE PCIS DIV II RESET, 3-HS-64-16A-S33, to left and right RESET positions.

[14.4] CHECK the following red lights ILLUMINATED:

[14.4.1] MSIV GROUP A2, 3-IL-64-A2

[14.4.2] MSIV GROUP B2, 3-IL-64-B2

[16] RESET the secondary containment isolation logic at Panel 3-9-25, as follows:

[16.1] PLACE the REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11A, to OFF.

[16.2] PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, to OFF.

[16.3] VERIFY only one SBGT trains operating.

[17] START the Refuel Zone supply and exhaust fans, at Panel 3-9-25, as follows:

[17.1] PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, in SLOW A (SLOW B) position. (!)

[18] START the Reactor Building supply and exhaust fans, at Panel 3-9-25, as follows:

[18.1] PLACE the REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11A, to the SLOW A(B) position.

Page 15of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __£_ Event No.: _3_ Page 5 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[19] SECURE remaining SBGT train per 0-01-65 section 7.1. e:i

[20] IF fans transfer to fast speed is required after five minutes, THEN: PLACE the REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11A, to the FAST A(B) position.

[21] IF fans transfer to fast speed is required after five minutes, THEN: PLACE the REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, to the FAST A(FAST B) position.

[22] REFER to 3-01-99 to restore remaining affected systems to operation following RPS Bus Power loss and RETURN to Step 4.2[23]

Page 16 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: _3_ Page 5 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Directs BOP to restore remaining affected systems IAW 01-99 Refers to Technical Specifications 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution Systems - Operating LCO 3.8. 7 The following AC and DC electrical power distribution subsystems shall be OPERABLE:

b. Unit 3 480 V Shutdown Boards;

c. Unit 3 480 V RMOV Boards 2A, 3B, 2D, and 3E; B. One Unit 3 480 V B.1 Restore Board to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Shutdown Board OPERABLE inoperable. status SRO OR 480 V RMOV Board 3B inoperable C. Unit 3 480 V RMOV C.1 Declare the affected Immediately Board 3E inoperable RHR subsystem Inoperable D. One Unit 3 DG Auxiliary D.1 Restore Unit 3 DG 5 days Board inoperable. Auxiliary Board to OPERABLE status. AND 12 days from discovery of failure to meet LCO Page 17 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _3_ Page 6 of 14 Event

Description:

Loss of 480V S/D Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior 01-99 8.5 Restoration to Normal Following RPS Bus Power Loss or Transfer NOTE Steps 8.5[5] through 8.5[22] can be performed in any order.

BOP

[5] VERIFY the green lights are ILLUMINATED on all 5 of the QLVPS located at Panel 9-14.

[6] RESTORE Reactor and Refuel Zone Ventilation to normal operation. REFER TO 3-AOl-64-2D, Group 6 Ventilation System Isolation.

AOl-64-20 4.2 Subsequent Actions

[1 O] WHEN initiating signal has been corrected AND necessary repairs have been made, THEN

[10.1] VERIFY PCIS RESET:

• RESET PCIS DIV I RESET, 3-HS-64-16A-S32.

• RESET PCIS DIV II RESET, 3-HS-64-16A-S33.

[10.2] RESET Reactor/Refuel isolation logic as required:

• PLACE REFUELING ZONE FANS AND DMPRS, 3-HS-64-3A in OFF.

• PLACE REACTOR ZONE FANS AND DMPRS, 3-HS-64-11A in OFF.

[10.3] START Reactor/Refuel zone ventilation as required:

• PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11A, in SLOW A (SLOW B).

• PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, in SLOW A (SLOW B).

Page 18 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _3_ Page 7of14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[10.4] VERIFY for the fans started the dampers open and fans start as indicated by illuminated red lights above the following switches:

• The two green lights A(B) above REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, extinguish and the two red lights A(B) illuminate.

• The two green lights A(B) above REACTOR ZONE FANS AND DAMPERS Switch 3-HS-64-11A, extinguish and the two red lights A(B) illuminate.

• REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13.

• REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14.

• REACTOR ZONE EXH INBD ISOL DMPR, 3-HS-64-42.

• REACTOR ZONE EXH OUTBD ISOL DMPR, 3-HS-64-43.

• REACTOR ZONE FANS AND DAMPERS, 3-HS-64-11A.

• REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5.

BOP

• REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6.

• REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9.

• REFUEL ZONE EXH INBD ISOL DMPR, 3-HS 10.

[11] NOTIFY RADCON of the isolation restoration AND REQUEST surveys BOP Restores ventilation and contacts radcon for surveys Page 19of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _3_ Page 8 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior As Radcon, acknowledge that ventilation has been restored and Driver surveys need to be performed.

01-99 8.5 Restoration to Normal Following RPS Bus Power Loss or Transfer BOP

[7] RESTORE Standby Gas Treatment System to standby readiness. REFER TO 0-01-65, Section 7.0.

Calls Unit 1 to have Standby Gas Treatment restored to standby readiness.

As Unit 1 operator, acknowledge securing SBGT. Insert (bat) to Driver secure SBGT trains.

[8] At Panel 3-9-3, PLACE PSC head tank pumps in service as follows:

• PLACE PSC PUMP SUCTION INBD ISOL VALVE, 3-HS-75-57A, in AUTO After OPEN. D

• PLACE PSC PUMP SUCTION OUTBD ISOL VALVE, 3-HS-75-58A, in AUTO After OPEN. D BOP *PLACE PSC HEAD TANK PUMPS 2A & 3B, 3-HS 75A & 3-HS-75-76A in AUTO.

NOTE 3-FCV-64-139 and 3-FCV-64-140 opens and closes automatically when the Drywell DP Compressor starts and stops.

Page 20of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. 4 Event No.: _3_ Page 9 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[1 O] At Panel 3-9-3, RESTORE Drywell DP Compressor to automatic operation as follows:

[10.1] DEPRESS DRYWELL DP COMP SUCTION VLV RESET pushbutton, 3-HS-64-139A.

[10.2] DEPRESS DRYWELL DP COMP DISCH VLV RESET pushbutton, 3-HS-64-140A.

[10.3] VERIFY OPEN DRYWELL INBD ISOLATION VLV using 3-HS-64-31.

[10.4] VERIFY OPEN SUPPR CHBR INBD ISOLATION VLV using 3-HS-64-34.

[11] At Panel 3-9-4, RESTORE Drywell Floor and Equipment Drain Systems to normal operation as follows:

[11.1] NOTIFY Radwaste Operator that Drywell Equipment and Floor Drain Sump isolation valves are being reopened.

[11.2] PLACE OW EQPT DRAIN INBD ISOL VALVE, 3-HS-77-15A, in AUTO After OPEN.

[11.3] PLACE OW EQPT DRAIN OUTBD ISOL VALVE, 3-HS-77-15B, in AUTO After OPEN.

[11.4] PLACE OW FLOOR DRAIN INBD ISOL VALVE, 3-HS-77-2A, in AUTO After OPEN.

[11.5] PLACE OW FLOOR DRAIN OUTBD ISOL VALVE, 3-HS-77-3B, in AUTO After OPEN.

BOP

[12] IF OW Radiation Monitor CAM, 3-RM-90-256 was secured due to a preplanned transfer, THEN (otherwise N/A)

Page 21 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _3_ Page 10of14 Event

Description:

Loss of 480V S/D Board 38 and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[13] IF OW Radiation Monitor CAM, 3-RM-90-256, isolated due to loss of RPS, THEN (otherwise N/A)

MOMENTARILY DEPRESS the following RESET pushbuttons on Panel 3-9-2 and verify the associated valves open.

• UPPER INBD SUPPLY ISOL VALVE RESET, 3-HS 254A-A (opens FCV-90-254A).

• LOWER INBD SUPPLY ISOL VALVE RESET, 3-HS 2548-A (opens FCV-90-2548).

• OUTBD RETURN ISOL VALVE RESET, 3-HS-90-257A-A (opens FCV-90-257A).

• OUTBD SUPPLY ISOL VALVE RESET, 3-HS-90-255A (opens FCV-90-255).

• INBD RETURN ISOL VALVE RESET, 3-HS-90-2578-A (opens FCV-90-2578).

[14] At Panel 3-9-54, PLACE H2/02 Analyzer in service per 3-01-76.

[15] At Panel 3-9-55, VERIFY PATH A VENT FLOW CONT, 3-FIC-84-20, in AUTO with setpoint at 100 scfm.

Page 22of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _3_ Page 11 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior

[ 19] At Panels 3-9-10 and 3-9-11 , RESTORE Radiation Monitoring System as follows:

[19.1] DEPRESS applicable RESET pushbuttons.

[19.2] RESTORE Radiation Monitoring System to normal.

REFER TO 3-01-90, Section 5.0.

[21] At Panel 3-9-14, VERIFY APRM and RBM Memory lights RESET (If current plant conditions allow).

[22] At Panel 3-9-13, DEPRESS TIP ISOLATION RESET pushbutton.

BOP

[14] RESET PCIS trip logic at Panel 3-9-4 as follows:

[14.1] MOMENTARILY PLACE PCIS DIV I RESET, 3-HS-64-16A-S32, to left and right RESET positions.

[14.2] VERIFY the following red lights ILLUMINATED:

[14.2.1] MSIV GROUP A1.

[14.2.2] MSIV GROUP B1.

[14.3] MOMENTARILY PLACE PCIS DIV II RESET, 3-HS-64-16A-S33, to left and right RESET positions.

This completes restoration from the loss of RPS 'B' with the NRC exception of RWCU.

Page 23of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __1__ Event No.: _3_ Page 12 of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Loss of l&C Bus B Enters 3-AOl-57-5B 3.0 AUTOMATIC ACTIONS A. Panel 3-9-2 Cabinet 3 transfers to Alt. Pwr. Supply.

B. RFP A and B min flow vlvs open causing FW flow to lower.

C. Loss of 1/2 pwr sources to RFPT 3B Gov and Final Driver D. Loss of RFPT/RFP Vib Mon Equip for all three RFPT/RFPs.

E. SJAE 3B isolates if in service.

F. Reactor/Refuel zone ventilation isolates.

BOP G. Drywell Leak Detection System isolates.

H. +/-24V DC neutron monitoring battery chargers B1-3 & B2-3 trip.

I. LP FW Htrs ext bypass vlvs open causing MWe to lower.

J. FPO vlvs 78-64 & 67, Close. 78-65.

K. RWCU filler demin 3B isolates.

L. Loss of RM-90-259B (Control Bay Vent Rad Mon.).

M. Loss of pwr to RR monitoring relay 3-RLY-068-MMR3/B, and digital frequency relay 3-RLY-068-DFR3/B. These relays will fail to initiate a trip to the 2 out of 3 logic trip system.

Reports that l&C Bus has a loss of power and that the .:t 24V DC Neutron Monitor Battery chargers tripped.

Acknowledges the report of the l&C Bus loss.

SRO Evaluates the operability of the SRMs due to a loss of the 24V DC chargers.

Page 24of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _..1_ Event No.: _3_ Page 13 of 14 Event

Description:

Loss of 480V SID Board 38 and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Reports RBCCW PUMP DISCH. HOR PRESS LOW (3-XA-55-4C, Window 12) in alarm.

Recognizes an entry into 3-AOl-70-1 Loss of Reactor Building Closed Cooling Water.

3.0 AUTOMATIC ACTIONS RBCCW SECTIONALIZING VLV, 3-FCV-70-48, closes automatically on RBCCW Pump discharge header pressure o 57 psig.

ARP-9-4C ATC A. VERIFY 3-FCV-70-48 CLOSING/CLOSED.

B. VERIFY RBCCW pumps A and B in service.

Recognizes and reports that sectionalizing valve 70-48 is not closing and attempts to close 70-48. If power has not been restored the valve will not move. Once the power has been transferred the ATC will have to give the valve a closed signal from the control switch.

Reports 70-48 closed Contacts Work Control for a work order to investigate the failure of the 70-48 valve Page 25of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1._ Event No.: _3_ Page 14of 14 Event

Description:

Loss of 480V SID Board 3B and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Contacts Unit 1 to have the spare RBCCW Pump aligned and started to Unit 3, REFER TO 3-01-70.

3-AOl-70-1

[5] IF RBCCW flow was restored to two pump operation, THEN BOP PERFORM the following (Otherwise N/A):

[5.1] REOPEN RBCCW SECTIONALIZING VLV, 3-HS-70-48A

[5.2] RESTORE the RWCU system to operation. (REFER TO 3-01-69).

List of LCOs entered on a loss of SID Bus 3B.

1. 3.8.7 Power Distribution

a. B, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />

b. C, Immediately declare RHR inop 3.5.1

c. D, 5 day

2. 3.5.1 A.1 RHR Loop II; 7 day (RMOV Board Eon Alt.)

3. ODCM 1/2.1.1 Radioactive Liquid Effluent Table 1.1 D.

SRO (3-RM-132 Raw Cooling Water, RBCCW, RHRSW Disch Rad monitors have lost power)

4. 3.3.3.1 PAM Instrumentation A.1 Drywell Temperature; 30 Day These LCOs were in until the RMOV Boards are transferred to alternate. Then they were exited.

5. 3.4.5 Leakage Detection - places the unit in TS 3.0.3 Page 26of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _3_ Page 14of 14 Event

Description:

Loss of 480V S/D Board 38 and the RBCCW Sectionalizing valve does not close at the low pressure signal.

Time Position Applicant's Actions or Behavior Acknowledge investigating the failure of the 70-48 valve to automatically close.

Driver Also the request to have the spare RBCCW Pump placed in service to Unit 3. Insert Remote function SW02 NRC End of Event #3 Page 27of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: ~ Page 1 of 11 Event

Description:

Loss of Condenser Vacuum Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 4, insert Driver malfunction for a loss of condenser vacuum.

Reports OG HOLDUP LINE INLET FLOW HIGH (3-XA-55-53, Window 14) in alarm.

ARP-9-53 BOP A. VERIFY holdup line flow on:

2. 3-FR-66-20, Off-gas flow to holdup volume, Panel 9-8.

Checks off-gas flow and reports flow is high ATC Reports degrading condenser vacuum Directs entering AOl-47-3 AOl-47-3 NOTE Turbine trip is expected around 24.3 inches Hg as indicated on 3-XR-002-0026 due to differences between instrument taps for turbine trip and indicated vacuum.

4.2 Subsequent Actions SRO (1] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOl(s).

CAUTION

[NRC/C] Operations outside of the allowable regions shown on the Recirculation System Operating Map could result in thermal-hydraulic power oscillations and subsequent fuel damage. REFER TO 3-GOl-100-12A for required actions and monitoring to be performed during a power reduction. [NCO 940245001]

Page 28of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. 4 Event No.: _A_ Page 2of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior

[2] IF unable to maintain hotwell pressure below -25 inches Hg as indicated on 3-XR-002-0026, with Reactor power less than 30%, THEN TRIP the main turbine.

[3] IF condenser vacuum is lost, THEN OPEN the HOTWELL SAMPLE TO FL DR, 3-DRV-043-1019 (557'@ T-10 C-Line) and CON DEMIN SAMPLE TO FL DR, 3-DRV-043-1020 (557'@ T-6 G-Line), to establish flow through the sample lines.

[4] REDUCE reactor power in an attempt to maintain condenser vacuum.

[5] VERIFY automatic actions SRO Sets a trigger value, on main condenser vacuum, to initiate a manual reactor scram and main turbine trip.

Directs ATC to initiate any of the RR Power Runback in an attempt to maintain main condenser vacuum.

Initiates any of the Reactor Recirc Power Runbacks and reports reactor power lowering.

ATC Monitors main condenser vacuum and reports that vacuum is continuing to degrade.

Page 29of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _!___ Event No.: _!___ Page 3 of 11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior AOl-47-3 4.2 Subsequent Actions

[6] CHECK CCW pumps for proper operation.

[7] VERIFY CLOSED CONDENSER VAC BREAKERS 1A and 1B, 3-HS-66-1A, Panel 9-8.

[8] CHECK OFF-GAS FLOW TO 6-HOUR HOLDUP VOLUME, 3-FR-66-20, Pnl 9-8, between 20 and 180 scfm.

[9] IF VERIFY OPEN, 3-FCV-66-28, OFF-GAS SYSTEM ISOLATION VALVE

[10] IF SJAE 3A is in service, THEN VERIFY the following:

• 3-PCV-001-0151 and 3-PCV-001-0166 OPEN using BOP STEAM TO SJAE 3A STAGES 1,2, AND 3, 3-Zl-1-151/166 on Panel 3-9-7.

• SJAE 3A INTMD CONDENSER DRAIN, 3-Zl-1-150, on Panel 3-9-7, indicates OPEN.

• 3-FCV-066-0011 OPEN using SJAE 3A INLET VALVE, 3-HS-66-11 on Panel 3-9-8.

• Main Steam supply pressure at SJAE 3A STAGE I & II STEAM PRESS, 3-Pl-001-0150, on 3-LPNL-925-0105, is being maintained between 190 and 225 psig. (TB EL 586' T12-C)

[11] IF SJAE 3B is in service, THEN VERIFY the following NIA

[12] IF a failure of the in-service SJAE is indicated, THEN PLACE the standby SJAE in service as follows: NIA Verifies proper operation of CCW and in-service SJAE Page 30of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: -1_ Page 4of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior SRO Directs a manual reactor scram and main turbine trip Initiates a manual reactor scram AOl-100-1 Reactor Scram OATC Hard Card 1.0 IMMEDIATE ACTIONS ATC

[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5AfS3A and 3-HS-99-5AfS3B, on Panel 3-9-5.

[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A)

[3] Refuel Mode One Rod Permissive Light check:

[3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL.

[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-Xl-85-46.

[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-Xl-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A)

Page 31 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __.1_ Event No.: __.1_ Page 5 of 11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior

[4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in SHUTDOWN.

[5] REPORT the following status to the US:

• Reactor Scram

• Mode Switch is in Shutdown

• "All rods in" or "rods out"

• Reactor Water Level and trend (recovering or lowering)

• Reactor pressure and trend

• MSIV position (Open or Closed)

• Power level Completes immediate operator actions and makes scram report Continues to monitor main condenser vacuum Page 32of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -4_. Event No.: -4_. Page 6of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior Reactor Scram BOP Unit Operator Hard Card 1.0 SUBSEQUENT ACTIONS: PANELS 3-9-7 & 3-9-8 NOTES

1) To OPEN PCB 224 with the control room handswitch, ONE of the following is required:

3-XA-55-8A window 7, GEN REVERSE PWR FIRST RELAY OPERATION 3-EA-57-136, WITH GENERATOR PCB 224 BOP CNTR W/REV BYPASS, 3-HS-242-224A, placed in TRIP.

OR GENERATOR PCB 224 CNTR W/REV BYPASS, 3-HS-242-224A, placed in BYPASS.

2) The following steps are not required to be performed in order, but only as required to maintain stable conditions.

[1] Ats 50 MWe, or as directed by the Unit Supervisor, VERIFY TRIPPED the Main Turbine as follows:

[1.1] DEPRESS the TRIP pushbutton, 3-HS-47-670 on Panel 3-9-7.

[1.2] VERIFY OPEN GENERATOR PCB 224.

BOP [1.3] VERIFY TRIPPED VOLTAGE REGULATOR

[2] ANNOUNCE Reactor SCRAM over PA system.

Trips the main turbine and announces the reactor scram over the PA system.

Reports the main turbine tripped.

Page 33of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: ~ Page 7of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior 2.0 SUBSEQUENT ACTIONS: PANELS 3-9-3 NOTE The following steps are not required to be performed in order, but only as required to maintain stable conditions

[1] MONITOR and CONTROL RPV pressure to keep below 1073 psig and stable.

[2] IF any PCIS isolation signal is received, THEN VERIFY PCIS isolations using any of the following:

(Otherwise N/A)

• Containment Isolation Status System on Panel 3-9-4 Verifies that main turbine bypass valves are controlling reactor pressure Checks CISS panel and reports 2,3,6, and 8 PCIS isolations Enters EOl-1 on RPV Water Level below +2" Verify RX scram RC/Q Monitor and Control reactor power SRO The reactor is subcritical AND NO boron has been injected THEN EXIT RC/Q and ENTER AOl-100-1, Reactor Scram Page 34of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. 4 Event No.: _L Page 8of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior Directs ATC to enter 3-AOl-100-1 RC/L CAUTION: Ambient temp may affect RPV water level indication and trend - CHECKED MONITOR and CONTROL RPV water level - CHECKED VERIFY each as required

• PCIS isolations (Groups 1, 2, and 3) - CHECKED It has NOT been determined that the reactor will remain subcritical without boron under all conditions - SUBCRITICAL RPV water level CANNOT be determined - CAN PC water level CANNOT be maintained below 105 ft - CAN OR Suppression chamber press CANNOT be maintained below 55 SRO psig - CAN CAUTION

1. 2 Pump NPSH and Vortex Limits

1. 3 Elevated suppression chamber press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F - CHECKED RESTORE and MAINTAIN RPV water level between +2 in and

+51 in with ANY of the following CNDS and FW 5A CRD 58 RCIC with CST suction if available 5C HPCI with CST suction if available 50 CNDS 6A cs 6D,6E LPCI 68, 6C Page 35of68 Unit 3

Appendix D Required Operator Actions Form ES-0-2 Op Test No.: 15-01 Scenario No. _1__ Event No.: _1__ Page 9of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior Directs ATC to restore and maintain reactor level between +2 in and +51 in using Condensate and Feedwater in accordance with Appendix 5A RC/P OW press is above 2.4 psig - NO Emergency RPV depressurization is anticipated - NO AND The reactor will remain subcritical without boron under all conditions Emergency RPV depressurization is or has been required - NO RPV water level CANNOT be determined - CAN Is ANY MSRV cycling - NO SRO Steam cooling is required - NO Suppr pl temp and level CANNOT be maintained in a safe area of Curve 3 at the existing RPV press - CAN Suppression pool level CANNOT be maintained in the safe area -

CAN of Curve 4 OW control air becomes unavailable - AVAIL Boron injection is required - NO AND The main condenser is available AND There has been no indication of a steam line break Page 36 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 STABILIZE RPV press below 1073 psig with the main turbine bypass vlvs (APPX 8B)

Op Test No.: 15-01 Scenario No. ~ Event No.: ~ Page 10of11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior Directs BOP to control reactor pressure 800 psig to 1000 psig in SRO accordance with Appendix 8B Recognizes and reports that main turbine bypass valves have BOP closed due to the loss of vacuum Recognizes and reports Reactor Feedwater Pumps tripped due to ATC loss of vacuum Directs BOP to restore and maintain RPV water level between +2 in. and +51 in. with RCIC using App 5C EOl-1 RC/P SRO AUGMENT RPV press control as necessary with ANY of the following MSRVs ONLY when suppr level is above 5.5 ft 11A Directs BOP to maintain RPV pressure 800 psig to 1000 psig using MSRVs, App. 11A Page 37of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: ~ Page 11 of 11 Event

Description:

Loss of Condenser Vacuum Time Position Applicant's Actions or Behavior Appendix-11A 1.0 INSTRUCTIONS

((3] OPEN MSRVs using the following sequence to control RPV pressure as directed by SRO:

1 3-PCV-1-179 MSL A RELIEF VALVE 2 3-PCV-1-180 MSL D RELIEF VALVE 3 3-PCV-1-4 MSL A RELIEF VALVE 4 3-PCV-1-31 MSL C RELIEF VALVE 5 3-PCV-1-23 MSL B RELIEF VALVE 6 3-PCV-1-42 MSL D RELIEF VALVE 7 3-PCV-1-30 MSL C RELIEF VALVE 8 3-PCV-1-19 MSL B RELIEF VALVE 9 3-PCV-1-5 MSL A RELIEF VALVE 10 3-PCV-1-41 MSL D RELIEF VALVE 11 3-PCV-1-22 MSL B RELIEF VALVE 12 3-PCV-1-18 MSL B RELIEF VALVE 13 3-PCV-1-34 MSL C RELIEF VALVE BOP Controls RPV pressure 800 psig to 1000 psig IAW App 11A Page 38of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: _5_ Page 1 of 2 Event

Description:

RCIC Isolation Time Position Applicant's Actions or Behavior EOI APPENDIX-5C CAUTION

• Operating RCIC turbine below 2100 rpm may result in unstable system operation and equipment damage.

BOP

• High Suppression Chamber pressure may trip RCIC.

• Operating RCIC Turbine with suction temperatures above 140°F may result in equipment damage.

3. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRIP/THROT VLV.

4. VERIFY 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 620 gpm.

5. OPEN the following valves:

• 3-FCV-71-39, RCIC PUMP INJECTION VALVE BOP

• 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE

• 3-FCV-71-25, RCIC LUBE OIL COOLING WTR VLV.

6. PLACE 3-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.

7. OPEN 3-FCV-71-8, RCIC TURBINE STEAM SUPPLY VLV, to start RCIC Turbine.

Page 39 of 69 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No.~ Event No.: _5_ Page 2 of 2 Event

Description:

RCIC Isolation Time Position Applicant's Actions or Behavior

8. CHECK proper RCIC operation by observing the following:

a. RCIC Turbine speed accelerates above 2100 rpm.

b. RCIC flow to RPV stabilizes and is controlled automatically at 620 gpm.

c. 3-FCV-71-40, RCIC TESTABLE CHECK VLV, opens by observing 3-Zl-71-40A, DISC POSITION, red light illuminated.

d. 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE, closes as flow rises above 120 gpm.

10. ADJUST 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller as necessary to control injection.

Insert Trigger 5 to cause the Inboard Steam Supply Valve to Driver close.

BOP or NRC End of Event #5 Page 40 of 69 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _L Event No.: _§._ Page 1 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 6, insert Driver Trigger 6 (loss of offsite power with a failure of 'A' Diesel Generator to automatically start)

CREW Recognize and report that a loss of offsite power has occurred.

SRO Directs entry into AOl-57-1A AOl-57-1A 4.0 OPERATOR ACTIONS 4.1 Immediate Actions NOTE Performing this instruction, in conjunction with an earthquake, may BOP require resetting the individual Diesel Generator's 86G Lockout Relay and the Field Breaker(both locally at the Diesel Generator electrical cabinet).

[1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators When the BOP operator checks on the status of the Diesel NRC Generators, it would initiate Event No. 8, Failure of the 'A' Diesel Generator to automatically start.

BOP Dispatches personnel to monitor the Diesel Generators Driver As AUO, acknowledge monitoring .diesel generators Page 41of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.:_§__ Page 2 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior

[2] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators.

[3] IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes) (Otherwise N/A).

BOP

[4] PERFORM the following to ensure at least one train of Diesel Generator Room Fans are energized:

• VERIFY 480V DSL Aux Board A or B energized.

NOTE The following subsequent actions may be performed out of order, depending on plant conditions.

Performs actions IAW AOl-57-1A. Recognizes Diesel Generator

'A' failed to start. Starts 'A' DIG and verifies tied to 4kV S/D Bd.

4.2 Subsequent Actions NOTES

1) SBO Unit does attachment 12 only; the other two units perform subsequent actions of AOI.

2) If a Unit is in a Station Blackout condition, performance of this instruction will also require implementation of 1(2)(3)-AOl-30B-BOP 1, Reactor Building Ventilation Failure, on the Unit in Station Blackout.

3) The following is the preferred pump combinations of Unit 1 and Unit 3 RHR pumps used in SOC: 1B/20, 1A/3C, 1A/2D, 1B/3C.

By using these pumps it ensures that a loss of a Diesel during a station blackout would not result in a loss of Shutdown cooling for both Unit 1 and Unit 3.

Page 42of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _6_ Page 3 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior

[1] IF ANY EOI entry condition is met, THEN REFER TO the appropriate EOl(s). (Otherwise N/A)

[2] IF any Unit is under a "Station Blackout" THEN ONLY PERFORM Attachment 12 for that Unit: - NA

[3] VERIFY automatic actions and PERFORM any that failed to occur.

[4] REFER TO 1(2)(3)-AOl-78-1, FPC System Failure for a complete Loss of AC POWER, as necessary. NOT NEC NOTES

1) EECW supply valves to the Control Air Compressors and RBCCW are air operated. If initial air pressure is low, air compressors may trip on high temperature, until cooling water flow is established.

2) At US discretion, the O-FCV-67-53 valve can be placed in the open position with hand switch. The valve will automatically come open once EECW pressure is above setpoint. REFER TO 01-67 for valve operation.

3) The North header supply to Unit 1 RBCCW, the North header supply to Unit 3 RBCCW and the South header supply to Unit 3 RBCCW are normally isolated with a manual valve; therefore no flow will occur when either 1-FCV-67-50, 3-FCV-67-50 or 3-FCV-67-51 opens.

BOP [5] WHEN EECW header pressure is restored above the reset pressure setpoint (psig) for the valves listed below, THEN Common Unit 1 Unit 2 Unit 3 O-FCV-67-53 106 FCV-67-50 90 91 92 FCV-67-51 107 109 113 Page 43of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No.~ Event No.: _Q__ Page 4 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel 1-LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. REFER TO the EECW to the RCW Crossties for Control Air & RBCCW section of 0-01-67.

[6] START Control Air Compressors A, D and Gas required and MONITOR system pressure. REFER TO O-AOl-32-1.

[6.1] IF an air compressor trips on high temperature, THEN (Otherwise N/A)

NOTIFY Unit Supervisor for instructions.

[8] PLACE RPS MG Sets A and 8 in service. REFER TO 1(2,3)-01-99.

[9] START the Diesel Driven Fire Pump. REFER TO 0-01-26.

Calls for Control Air Compressors and EECW to be reset and for RPS to be reenergized.

As outside personnel, acknowledge restoring control air, resetting EECW and restoring RPS. Wait 5 minutes, then insert (bat)

Driver Call control room and report that control air has been restored, EECW has been reset and RPS has been restored.

Reports that MSIVs are closed ATC/BOP IF bulk suppression pool temperature reaches 95°F, notifies SRO of EOl-2 entry condition.

When suppression pool temperature reaches 95°F, enters EOl-2 SRO Enters EOl-2 on Supp Pl Temp above 95°F Page 44of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _1_ Event No.: _Q_ Page 5 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior SAMG entry is required and the TSC SAM team has assumed command and control - NO SAMG entry is required - NO CAUTION

1. 4 PC press vs pump NPSH - CHECKED OW sprays have been initiated - NO Suppression chamber sprays have been initiated - NO PC water level CANNOT be restored and maintained below 105 ft - CAN OR Suppression chamber press CANNOT be restored and maintained below 55 psig - CAN SRO SPIT CAUTION

1. 2 Pump NPSH and Vortex Limits MONITOR and CONTROL suppr pl temp below 95°F using available suppr pl cooling (APPX 17A) - CHECKED WHEN suppr pl temp CANNOT be maintained below 95°F OPERATE all available suppr pl cooling using only RHR pumps NOT required to assure adequate core cooling by continuous injection (APPX 17A)

Directs BOP to put all available suppression pool cooling in service IAW App 17A EOI APPENDIX-17A BOP NOTE: Placing a BYPASS SEL switch in BYPASS in step 1 below prevents automatic opening of the affected RHR loop's outboard injection valve. This makes LPCI mode of that RHR loop inoperable.

Page 45of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. --1_ Event No.: _Q_ Page 6 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior

1. IF .............. Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN ....... BYPASS LPCI injection valve open interlock AS NECESSARY:

• PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

• PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:

BOP

• 3-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

• 3-FCV-23-46, RHR HX 3B RHRSW OUTLET VLV

• 3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV

• 3-FCV-23-52, RHR HX 20 RHRSW OUTLET VLV

d. IF ............. Directed by SRO, THEN ....... PLACE 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERIDE

e. IF ............. LPCI Initiation signal exists, THEN ....... MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.

f. IF ............. 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN ....... VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.

g. OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV

h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.

Page 46of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: __§__ Page 7 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior CAUTION RHR system flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage.

i. THROTTLE OPEN 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLGfTEST VLV, to maintain EITHER of the following as indicated on 3-Fl-74-50(64), RHR SYS 1(11) FLOW:

• Between 7000 and 10000 gpm for one-pump operation.

OR BOP *At or below 13000 gpm for two-pump operation. _ _

j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.

k. MONITOR RHR Pump NPSH using Attachment 1.

I. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.

Places RHR in Suppression Pool Cooling IAW App 17A Reports Suppression Pool Cooling in service Contacts Chemistry to reports RHRSW aligned to in service RHR Heat Exchangers.

As Chemistry, acknowledge that RHRSW has been aligned to Driver RHR Heat Exchangers.

Refers to Technical Specifications for SPT 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature SRO LCO 3.6.2.1 Suppression pool average temperature shall be:

a. <95°F when any OPERABLE intermediate range monitor (IRM) channel is> 70/125 divisions offull scale on Range 7 and no testing that adds heat to the suppression pool is being performed; Page 47of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1__ Event No.: _6_ Page 8 of 8 Event

Description:

Loss of Offsite Power Time Position Applicant's Actions or Behavior A Suppression pool Verify suppression Once/hr average temperature pool temperature

~ 95°F but S110°F. S110°F AND AND Any OPERABLE IRM Restore suppression 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> channel > 70/125 pool average SRO divisions of full scale on temperature to Range 7 S95°F AND Not performing testing That adds heat to the suppression pool NRC End of Event #6 Page 48of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ....L Event No.: J_ Page 1 of 1 Event

Description:

DG A Fails to Auto start Time Position Applicant's Actions or Behavior Recognizes and reports that 'A' Diesel Generator has failed to start Starts 'A' Diesel Generator BOP Verifies the 'A' Diesel Generator ties to 4kV Shutdown Board 'A' Reports that 'A' Diesel Generator has been started and that the diesel generators are tied to all 4kV Shutdown Boards.

NRC End of Event #7 Page 49of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. 4 Event No.: _8_ Page 1 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior When the NRC Chief Examiner is ready for Event No. 8, insert Driver Trigger 8 (LOCA and a failure of the HPCI steam supply valve to open automatically).

A TC Reports reactor water level lowering BOP Reports drywell pressure rising Reenters EOl-1 on reactor water level RESTORE and MAINTAIN RPV water level between +2 in. and

+51 in. with ANY of the following:

CRD 5B SRO RCIC with CST suction if available 5C HPCI with CST suction if available 5D Directs BOP to start HPCI IAW App 5D Directs ATC to maximize CRD IAW App 5B When BOP operator starts HPCI OR if HPCI gets an initiation NRC signal due to rising drywell pressure or low reactor water level refer to Event 9 page 62.

Page 50 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 2 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior EOI Appendix-SD 1.0 INSTRUCTIONS

[4] VERIFY at least one SGTS train in operation.

CAUTIONS

1) Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.

2) Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.

[1] VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller is in one of the following configurations, as desired:

• in AUTO and set for 5300 gpm for rapid injection

• in AUTO and set for approximately 2500 gpm for slower injection

• in MANUAL with output at approximately 50% for slower injection.

NOTE HPCI Auxiliary Oil Pump will NOT start UNTIL 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, starts to open.

[2] IF high reactor water level trip logic is actuated, THEN

[2.1] DEPRESS HPCI TURBINE TRIP RX LEVEL HIGH RESET pushbutton.

[2.2] CHECK HPCI TURBINE TRIP LEVEL HIGH amber light has extinguished.

[3] PLACE HPCI AUXILIARY OIL PUMP handswitch in START.

[4] PLACE HPCI STEAM PACKING EXHAUSTER handswitch in START.

[5] OPEN the following valves:

• 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE

• 3-FCV-73-44, HPCI PUMP INJECTION VALVE.

Page 51 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _A_ Event No.: _8_ Page 3of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior

[6] OPEN 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, to start HPCI Turbine.

[7] CHECK proper HPCI operation by observing the following:

A. HPCI Turbine speed accelerates.

B. 3-FCV-73-45, HPCI TESTABLE CHECK VLV, opens by observing 3-Zl-73-45A, DISC POSITION, red light illuminated.

BOP C. HPCI flow to RPV stabilizes and is controlled automatically at the setpoint. (N/A if controller in manual).

D. 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds approximately 1200 gpm.

[8] ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.

[9] VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly.

[10] WHEN HPCI Auxiliary Oil Pump stops, THEN BOP PLACE HPCI AUXILIARY OIL PUMP handswitch in AUTO.

Injects to the reactor vessel with HPCI IAW App 50 EOI Appendix-58 1.0 INSTRUCTIONS

[2] IF BOTH of the following exists,

• CRD is NOT required for rod insertion, AND

• Maximum injection flow is required, THEN LINE UP ALL available CRD pumps to the RPV as follows:

[2.2] IF CRD Pump 3B is available, THEN VERIFY RUNNING CRD Pump 3B.

Page 52of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: _8_ Page 4of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior CAUTION Failure to maintain CRD system pressure above 1450 psig on 3-Pl-85-13A could result in pump runout.

[2.3] OPEN the following valves to increase CRD flow to the RPV:

• 3-PCV-85-23, CRD DRIVE WATER PRESS CONTROL VLV ATC

• 3-PCV-85-27, CRD CLG WATER PRES CONTROL VLV

• 3-FCV-85-50, CRD EXH RTN LINE SHUTOFF VALVE.

[2.4] ADJUST 3-FIC-85-11, CRD SYSTEM FLOW CONTROL to control injection WHILE maintaining 3-Pl-85-13A, CRD ACCUM CHG WTR HOR PRESS, above 1450 psig, if possible.

Maximizes CRD flow to the RPV IAW App 5B ATC Reports that RPV water level is continuing to lower EOl-1, RC/L RPV water level can be restored and INHIBIT ADS maintained above (-)162 inches.

SRO AND The ADS timer has initiated Directs BOP to Inhibit ADS BOP Inhibits ADS. Verifies inhibited and reports ADS inhibited.

Page 53of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __L Event No.: _8_ Page 5 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior

1. 3 Elevated suppression chamber press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F IF RPV water level CANNOT be restored and maintained between

+2 in. and +51 in.

SRO THEN RESTORE and MAINTAIN RPV water level above (-)162 inches. AUGMENT RPV water level control as necessary with ANY of the following:

SLC (boron tank) 7B Directs ATC to initiate SLC from the boron tank IAW App 7B EOI Appendix-78 1.0 INSTRUCTIONS

[2] IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step 1.0[1 O] to inject SLC Boron Tank to RPV ATC [1 O] UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A/3B, control switch in START-A or START-B (Panel 9-5).

[11] CHECK SLC injection by observing the following:

• Selected pump starts, as indicated by red light illuminated above pump control switch.

• Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished. D

• SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (3-XA-55-5B, Window 20).

Page 54of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __+/-___ Event No.: _8_ Page 6of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior

• 3-Pl-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.

• System flow, as indicated by 3-IL-63-11, SLC FLOW, red light illuminated, ATC

• SLC INJECTION FLOW TO REACTOR Annunciator in alarm (3-XA-55-58, Window 14).

Injects SLC to the RPV IAW App 78 EOl-1, RC/L Can RPV water level be restored NO C 1, Alternate SRO and maintained above -162 in. Level Control Enter C 1, Alternate Level Control BOP Reports drywell temperature and pressure continuing to rise Re-enters EOl-2 on Drywell Temperature and Pressure PC/H H2 and 02 monitoring system is inoperable - OPERABLE Offsite radioactivity release rate reaches ODCM limits - NO OR H2 is NO longer detected in PC (2.4% on control room indicators) - DETECTED Verify H202 analyzer in service (APP 19)

When H2 is detected in PC (2.4% on control room indicators continue - STOPS SP/L Primary Containment Flooding is required - NO AND SAMG entry is NOT required Page 55of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 7 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior MONITOR and CONTROL suppression pool level between -1 in.

and -6 in. (APPX 18) - CHECKED SRO Can suppression pool level be maintained above -6 in. - YES Can suppression pool level be maintained below -1 in. - YES PC/P MONITOR and CONTROL PC press below 2.4 psig using the Vent system (AOl-64-1) - CHECKED WHEN PC press CANNOT be maintained below 2.4 psig - CONTINUES SRO BEFORE suppression chamber press rises to 12 psig CONTINUE - CONTINUES CAUTION

1. 2 PUMP NPSH and Vortex Limits INITIATE suppression chamber sprays using only pumps NOT required to assure adequate core cooling by continuous injection (APPX 17C)

Directs BOP to initiate Suppression Chamber Sprays using APP 17C EOI APPENDIX-17C

1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.

2. IF ..... Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN ... BYPASS LPCI injection valve open interlock as necessary:

Page 56of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _1_ Event No.: _8_ Page 8 of 17 Event

Description:

LOCA me Position Applicant's Actions or Behavior

• PLACE 3-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.

• PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.

3. IF ..... Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN ... CONTINUE in this procedure At Step 7 using RHR Loop I OR At Step.8 using RHR Loop II.

4. IF ..... Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN ... CONTINUE in this procedure at Step 9.

BOP

5. INITIATE Suppression Chamber Sprays as follows:

a. VERIFY at least one RHRSW pump supplying each EECW header.

b. IF ..... EITHER of the following exists:

• LPCI Initiation signal is NOT present, OR

• Directed by SRO, THEN ... PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.

c. MOMENTARILY PLACE 3-XS-74-121(129), RHR SYS 1(11)

CTMT SPRAY/CLG VLV SELECT, switch in SELECT.

d. IF ..... 3-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN ... VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11)

OUTBD INJECT VALVE.

e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spra .

Page 57of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. 4 Event No.: _8_ Page 9 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior

f. VERIFY OPEN 3-FCV-74-57(71), RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV.

g. OPEN 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.

h. IF ..... RHR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN ... CONTINUE in this procedure at Step 5.k.

k. MONITOR RHR Pump NPSH using Attachment 2.

I. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).

BOP m.THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:

• 3-FCV-23-34, RHR HX 2A RHRSW OUTLET VLV

• 3-FCV-23-46, RHR HX 38 RHRSW OUTLET VLV

• 3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV

• 3-FCV-23-52, RHR HX 20 RHRSW OUTLET VLV.

n. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers Initiates and reports Suppression Chamber sprays in service IAW App 17C Calls Chemistry to report that RHRSW is aligned to in-service RHR Heat Exchangers As Chemistry, acknowledge that RHRSW is aligned to RHR heat Driver exchangers Page 58of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. __1___ Event No.: _8_ Page 10of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior DW/T CAUTION

1. 1 Ambient temp may affect RPV water level indication and trend MONITOR and CONTROL OW temp below 160°F using available OW cooling OW - CHECKED WHEN OW temp CANNOT be maintained - CONTINUES SRO below 160°F OPERATE all available OW cooling OW - IN SERVICE BEFORE OW temp rises to 200°F - CONTINUES Verifies reactor scrammed BOP Reports Suppression Chamber pressure is greater than 12 psig PC/P WHEN suppression chamber press exceeds 12 psig -

CONTINUES Is suppression pool level below 19 ft. - YES SRO Is OW temp within the safe area of Curve 5 Verifies in the safe area - YES SHUTDOWN Recirc pumps and OW blowers Verifies recirc pumps shutdown and directs BOP to shutdown OW blowers BOP Secures all OW blowers and reports all OW blowers shutdown Page 59of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 11 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior

1. 2 Pump NPSH and Vortex Limits INITIATE OW sprays using only pumps NOT required to assure adequate core cooling by continuous injection (APPX 17B SRO Directs BOP to initiate OW Sprays IAW App 17B EOI APPENDIX-178

1. BEFORE drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.

6. INITIATE Orywell Sprays as follows:

e. VERIFY OPERATING the desired System 1(11) RHR pump(s) for Orywell Spray.

f. OPEN the following valves:

BOP

• 3-FCV-74-60(74), RHR SYS 1(11) OW SPRAY OUTBO VLV

• 3-FCV-74-61(75), RHR SYS 1(11) OW SPRAY INBO VLV.

g. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.

h. IF ..... Additional Orywell Spray flow is necessary, THEN ... PLACE the second System 1(11) RHR Pump in service.

Initiates and reports OW Sprays in service IAW App 17B Reports OW pressure and temperature lowering Page 60of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. -1_ Event No.: _8_ Page 12of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior Critical Task When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit (DSIL) curve and before Drywell temperature rises to 280°F.

1.Safety Significance:

Precludes failure of containment 2.Cues:

Procedural compliance High Orywell Pressure and Suppression Chamber Pressure

3. Measured by:

Observation - US directs Orywell Sprays IAW with EOI Appendix 178 Observation - RO initiates Orywell Sprays 4.Feedback:

Orywell and Suppression Pressure lowering RHR flow to containment This Critical Task is not met if OW Pressure reaches 26 psig or OW Temperature reaches 280°F before Sprays are initiated.

Briefs crew on current plant status.

C/1 CAUTION

1. 1 Ambient temp may affect RPV water level indication and trend Emergency RPV depressurization is required - NO It has NOT been determined that the reactor will remain subcritical without boron under all conditions - HAS RPV water level CANNOT be determined - CAN PC water level CANNOT be maintained below 105 ft - CAN Page 61 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. _L Event No.: _8_ Page 13of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior OR Suppression chamber press CANNOT be maintained - CAN below 55 psig Verifies ADS inhibited CAUTION SRO

1. 2 Pump NPSH and Vortex Limits

1. 3 Elevated suppression chamber press may trip RCIC

1. 6 HPCI or RCIC suction temp above 140°F RESTORE and MAINTAIN RPV level above -162" using ANY of the following LPCI system I (pumps A or C) 6B LPCI system II (pumps B or D) 6C CS system I (pumps A or C) 6D SRO CS system II (pumps B or D) 6E Directs BOP to align CS and LPCI for injection to the RPV IAW Appendices 6B, 6C, 6D and 6E Verifies that LPCI and CS are aligned with pumps running.

Reports LPCI and CS aligned BOP Reports all four diesels have started Calls for AUO to monitor diesel generators As AUO, acknowledge monitoring diesel generators for proper Driver operation.

Page 62 of 68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. --1_ Event No.: _8_ Page 14 of 17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior C-1 Can 2 or more CNDS, LPCI or CS injection subsystems be lined up - YES WHEN RPV water level drops to -162 in.

Direct crew to report when RPV water level lowers to -162 in.

Is ANY CNDS, LPCI or CS injection subsystem lined up for injection with at least one pump running - YES SRO Is ANY RPV injection source lined up with at least one pump running - YES BEFORE RPV water level drops to -180 in. CONTINUE- CONTINUES EMERGENCY RPV DEPRESSURIZATION IS REQUIRED Updates crew that emergency depressurization is required EOl-1, RC/L IF Emergency RPV depressurization is or has been required THEN EXIT RC/P and ENTER C2, Emergency RPV Depressurization Enters C2, Emergency RPV Depressurization SAMG entry is required and the TSC SAM team has assumed command and control - NO RPV water level CANNOT be determined - CAN Containment water level CANNOT be maintained below 44 ft- CAN Page 63of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 15of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior DW control air becomes unavailable - AVAIL Will the reactor remain subcritical without boron under all conditions - YES SRO Is DW press above 2.4 psig - NO Is suppression pool level above 5.5 ft - YES OPEN all ADS vlvs (ok to exceed 100°F/hr cooldown rate)

Directs BOP to open all ADS valves Opens all ADS valves, verifies open using alternate indications BOP and reports 6 ADS valves open C-1 INJECT into the RPV with ANY available sources Directs injection of water to the RPV using all available sources C-2 Can 6 ADS vlvs be opened - YES IF Less than 4 MSRVs are open - NO SRO AND RPV press is 70 psi or more above suppression chamber press The reactor is NOT subcritical - SUBCRITICAL WHEN the reactor will remain subcritical without boron under all conditions - SUBCRITICAL WHEN shutdown cooling RPV press interlock clears - STOP Page 64of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 16of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior When RPV pressure lowers below 450 psig, verifies and reports ATC/BOP that LPCI and CS injection valves have opened.

Report water injection into the RPV and reactor water level risinq C-1 IF RPV water level can be restored and maintained above -162 in.

THEN EXIT this procedure and ENTER EOl-1, RPV Control, at Step RC/L-1 SRO Exits C-1, re-enters EOl-1 RC/L Directs coordinating injection of LPCI and CS to RESTORE and MAINTAIN RPV water level between +2 in. and +51 in. using LPCI and CS, Appendices 6B, 6C, 60, and 6E

1. RPV Level maintained above TAF (-162 inches) 1.Safety Significance:

Maintaining adequate core cooling 2.Cues:

RPV level indication 3.Measured by:

Reactor level indication above -162 inches 4.Feedback:

RPV level trend HPCl/RCIC injection valve open indication This Critical Task is not met if RPV Level remains below-162 inches.

ATC/BOP Coordinate restoring RPV level to +2 in. to +51 in. using App. 6B, 6C, 60, and 6E Page 65of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. ~ Event No.: _8_ Page 17of17 Event

Description:

LOCA Time Position Applicant's Actions or Behavior NRC Once RPV level has been restored, the scenario can end.

When NRC Chief Examiner directs, place the simulator in Driver FREEZE Page 66of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 15-01 Scenario No. --1._ Event No.: _9_ Page 1 of 1 Event

Description:

HPCI Steam Supply Valve fails to auto open Time Position Applicant's Actions or Behavior HPCI may receive an initiation signal from either reactor low water NRC level -45 in. or drywell high pressure 2.45 psig.

Recognize and report HPCI failure to initiate (initiation signal present and not injecting to the reactor vessel)

Recognizes that the HPCI Steam Supply valve (73-16) is not open Opens HPCI Steam Supply valve and verifies HPCI injection to BOP the Reactor Vessel.

Reports that HPCI is injecting to the RPV Calls Work Control to request a work order on the steam supply valve for HPCI As Work Control, acknowledge investigating failure of 73-16, HPCI Driver Steam Supply valve.

NRG End of Event #9 Page 67of68 Unit 3

Appendix D Required Operator Actions Form ES-D-2 SIMULATOR SETUP IC 28 Exam IC 188 Batch File or I G:\1501NRC Pref File

.* Pt1~1t1Jflgtinti§.~ ';,;z'1oitia1;;;; ',~ii)a1*.****

~.;i~~l'll~~l t0~i1~~'.*

Bus Duct Cooling EG13A 2 N/A N/A N/A N/A N/A Fan Tri ED10B SID Board Loss 3 N/A N/A N/A N/A N/A Sectionalizing Valve S&M 3 N/A N/A N/A N/A N/A fails to close.

Loss of Condenser MC04 4 N/A 50 N/A N/A N/A Vacuum RC03 RCIC Sue Press Tri 5 N/A N/A N/A N/A N/A ED01 LOOP 6 N/A N/A N/A N/A N/A DG A fails to Auto DG01A Active N/A N/A N/A N/A N/A Start TH21 LOCA 8 N/A 5 15:00 N/A N/A HPCI Steam Supply HP04 Valve fails to auto Active N/A N/A N/A N/A N/A o en Batch I Pref File{s): 1501 NRG Page 68of68 Unit 3

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U3 NRC Exam Urgent Load Reduction/Recirc Pump Trip Contingency Plans Operations Plan

BFN Reactivity Control Plan Page 2of15 Unit 3 Attachment 7 (Page 1 of 4)

Reactivity Control Plan Form BFN Unit: 3 Valid Date(s): 10/28/14 - 05/2/15 Reactivity Maneuver Plan#: U3 NRC Exam Are Multiple Activations Allowed: Yes (If yes, US may make additional copies)

Prepared by: ----'~=~==-:;...;::;~~*

.;.;.;;..;___.;../_.:.ToN.,="-'+ Reviewed by: ~RE.~ I 1o4, Reactor Engineer Date Qualified Reactor Engineer Date Approved by: R.E. ~ / ToN., Concurrence: JJ,,.. W. SRO I 1o4, RE Supervisor Date WCC/Risk/US SRO Date Approved by: O.P. H~ I 1o4, Authorized by: H,,. S43 H~ I 1o4, Ops Manager or Supt. Date Shift Manager Date RCP Activated: ~~~~~~~~'--~-

I RCP Terminated: ~~~~~~~~~~-

I Unit Supervisor Date Unit Supervisor Date

_!itle of Evolution: Urgent Load Reduction/Recirc Pump Trip Contingency Plan Purpose/Overview of Evolution: Maneuver Reactor to maintain acceptable operating conditions during and following unexpected plant conditions.

Maneuver Steps Recirc Pump Trip:

1. IF both pumps trip, SCRAM the reactor per 3-AOl-68-1A For a single recirculation pump trip, IMMEDIATELY insert rod groups to lower rod line below 95%.

2. Change core flow with the Operating recirculation pump to exit POWER/FLOW region 2 and prevent loop delta temperature concerns (>45%).

Urgent Load Reduction:

1A. Lower power by reducing core flow until either the desired operating power level is reached or core flow is 74% to 75% of rated core flow. (If a recirculation pump trip is imminent, only lower using the pump to be tripped).

2A. Insert control rods by groups per Emergency Shove Sheet to obtain the desired power level.

BFN Reactivity Control Plan Page 3of15 Unit 3 Attachment 7 (Page 2 of 4)

Reactivity Control Plan Form Reactivity Maneuver Plan #: U3 NRC Exam Maneuver Steps (continued)

Urgent Load Reduction (continued):

3A. Lower core flow to 46% core flow. (If a recirculation pump trip is imminent, only lower using the pump to be tripped.)

4A. Insert control rods Emergency Shove Sheet until load line is between 55% and 66.7%.

SA. Reduce recirculation pump speed to -28% (-480 rpm).

SA. SCRAM the reactor or continue control rod insertion until SCRAM.

BFN Reactivity Control Plan Page 4of15 Unit 3 Attachment 7 (Page 3 of 4)

Reactivity Control Plan Form Reactivi Maneuver Plan #: U3 NRC Exam Operating Experience AND General Issues:

Operating Experience Perry Nuclear Plant scrammed on OPRM signals -9 minutes after receiving a dual recirc pump shift to slow speed, before control rods were inserted.

CONCLUSIONS/RECOMMENDATIONS: In the event of a Recirc pump trip or both pumps running back to minimum speed, the control rod insertion should be accomplished as soon as practical due to potential OPRM trip.

General Issues This should be considered a guideline, not covering all possible situations. Contact On-Call Reactor Engineer as soon as possible to provide assistance.

Note that in the event of a Recirc Pump trip the loadline may go over 113.6% and MFLCPR may exceed 1.0 but the insertion of the recommended control rod groups will return loadline and MFLCPR below the required limits.

If a Powerplex case runs in the transient condition, the case is considered invalid. An automatic case may be triggered in the event of a recirc pump trip and that case should be considered invalid and another case requested as soon as reactor power stabilizes.

If shutdown is required while operating in single loop with the OOS loop isolated, Scram unit prior to going to minimum flow on operating recirc pump to avoid flow stratification concerns.

Recommend scram min between 40 to 45% RTP.

BFN Reactivity Control Plan Page 5of15 Unit 3 Attachment 7 (Page 4 of 4)

Reactivity Control Plan Form Reactivity Maneuver Plan#: U3 NRC Exam Cautions/Error Likely Situations/Special Monitoring Requirements/Contingencies:

During SLO, reactor power and flow conditions will be near to Region 2.

RE must calculate and substitute core flow value when reducing flow in the active jet pump loop below 41 Mlbm/hr per 3-SR-3.4.1 (SLO).

Prerequisites for starting an idle Recirc Pump, (3-01-68):

• Operating recirc pump flow is below 46,600 gpm.

• Operating recirc pump speed is less than 860 RPM speed.

• Reactor operating conditions are outside of Region 1,2 and 3.

• 10% margin between reactor power and APRM rod block setpoint.

The RE must be contacted to change the PowerPlex calculation using the correct Thermal Limit set while in SLO. (#15 SLO NSS BOC to NEOC RPTOOS). The RE must also change the Thermal Limit set if the plant is operating with reduced feedwater heating. (#16 - TLO NSS ICF BOC to NEOC RPTOOS, FHOOS)

BFN Reactivity Control Plan Page 6of15 Unit 3 Attachment 8 Reactivity Maneuver Instructions STEP 1of2 Reactivity Maneuver Plan # U3 NRC Exam Description of Step: Recirc Pump Trip Response IF both pumps trip, SCRAM the reactor per 3-0l-68-1A.

For a single recirculation pump trip, IMMEDIATELY insert control rod groups to lower Load Line below 95% using Emergency Shove Sheet.

(Case reflects inserting 2 groups)

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranges)

MW Electric N/A MFLCPR 0.85 - 0.95 MW Thermal 1800 - 1900 MAPRAT 0.50 - 0.65 Core Flow 45 - 51.5% MFDLRX 0.60 - 0.75 Load line 79- 87 Core Power 50 - 55% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Load Line: Displayed on Heat Balance or Power/Flow Map 95% N/A If HIGH, then insert rods.

Comments I Notes:

Core Flow indication may be incorrect if the active loop jet pump flow (3-Fl-68-48 or 3-Fl 46) is less than 41 Mlbm/hr.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 7of15 Unit 3 Attachment 8 Reactivity Maneuver Instructions STEP2of2 Reactivity Maneuver Plan # U3 NRC Exam Description of Step: Recirc Pump Trip Response Change speed of the Operating recirculation pump until:

• core flow is 46 - 50% to exit/remain out of POWER/FLOW Map Region II

• operating recirculation pump drive flow is below 46,600 gpm Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (may be ranqes) (may be ranqes)

MW Electric N/A MFLCPR 0.70 - 0.85 MW Thermal 1750-1850 MAPRAT 0.50 - 0.65 Core Flow 48-52% MFDLRX 0.60 - 0.75 Load line 79- 86 Core Power 51 - 56% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low continQency actions Core Flow: Displayed on Heat Balance or Power/Flow Map N/A 45%

If LOW, then raise core flow Comments I Notes:

Recirculation Pump flow must be kept below 46,600 gpm while in SLO.

Core flow must be greater than 45% to exit Region II at the expected load line, as well as prevent loop delta temperature concerns.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 8of15 Unit3 Attachment 8 Reactivity Maneuver Instructions STEP 1Aof6A Reactivity Maneuver Plan# U3 NRC Exam Description of Step: Urgent Load Reduction:

Lower power by reducing core flow to -74% or until the desired operating power level is reached. (If a recirculation pump trip is imminent, only lower using the pump to be tripped).

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranges)

MW Electric N/A MFLCPR 0.90 - 0.97 MW Thermal 2450 - 2700 MAPRAT 0.50 - 0.60 Core Flow ~60% MFDLRX 0.65 - 0.75 Load line 100 - 108 Core Power 70 - 78% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Core Flow: Displayed on Heat Balance or Power/Flow Map N/A 55%

If LOW, then stop lowering core flow.

Load Line: Displayed on Heat Balance or Power/Flow Map 113.6% N/A If HIGH, then stop flow reduction and allow load line calculation to stabilize to confirm. If still HIGH, then proceed to next step to insert rods.

MFLCPR: Displayed on CSUM if HIGH, then stop flow 0.985 NIA reduction and proceed to next step to insert rods.

Comments I Notes:

If desired, stop power reduction at any point between 100% and 70%. However, if continuing to next step, verify core flow is between 56% - 62%.

If a recirc pump trip is a concern, insert control rods before going below 90% core flow unless flow is only reduced on the pump expected to trip.

3-AOl-6-1A, 1B, or 1C specify the power level required for loss of feedwater heaters.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 9of15 Unit3 Attachment 8 Reactivity Maneuver Instructions STEP 2A of 6A Reactivity Maneuver Plan # U3 NRC Exam Description of Step: Urgent Load Reduction:

Insert control rods by groups per Emergency Shove Sheet till either the desired power level is reached or Load line is below 90%. Stop after completion of any rod group. (Case shows 2 rod groups inserted)

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranges)

MW Electric N/A MFLCPR 0.80 - 0.90 MW Thermal 1950 - 2150 MAPRAT 0.40 - 0.50 Core Flow 60% MFDLRX 0.50 - 0.60 Load line 80 - 95%

Core Power 50 - 70% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Load Line: Displayed on Heat Balance or Power/Flow Map N/A 55%

If LOW, then stop inserting control rods.

Comments I Notes:

3-AOl-6-1A, 1B or 1C specify the power level required for loss of feedwater heaters.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 10of15 Unit3 Attachment 8 Reactivity Maneuver Instructions STEP 3Aof6A Reactivity Maneuver Plan# U3 NRC Exam Description of Step: Urgent Load Reduction:

Lower core flow to 46% core flow. (If a recirculation pump trip is imminent, only lower using the pump to be tripped. Stop flow reduction when the pump to be tripped is at -28% speed and then trip the pump.)

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (mav be ranaes) (mav be ranaes)

MW Electric N/A MFLCPR 0.80 - 0.90 MW Thermal 1590 MAPRAT 0.40 - 0.50 Core Flow 46% MFDLRX 0.50 - 0.60 Load line 75%

Core Power 46% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low continQencv actions Core Flow: Displayed on Heat Balance or Power/Flow Map NIA 45%

If LOW, then stop lowerinQ core flow.

Load Line: Displayed on Heat Balance or Power/Flow Map 95% NIA If HIGH, then insert control rods.

Comments I Notes:

3-AOl-6-1A, 1B, or 1C specify the power level required for loss of feedwater heaters.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 11 of 15 Unit3 I Attachment 8 Reactivity Maneuver Instructions STEP 4Aof6A Reactivity Maneuver Plan # U3 NRC Exam Description of Step: Urgent Load Reduction:

Insert control rods per Emergency Shove Sheet until Load Line is between 55%

and 66%. ( Case shows total of 4 groups inserted )

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

ORE presence required in the Control Room? Yes No x (check)

Predicted Actual Predicted Actual (may be ranQes) (may be ranQes)

MW Electric N/A MFLCPR 0.70 - 0.75 MW Thermal 1150 - 1315 MAPRAT 0.25 - 0.35 Core Flow 46.0% MFDLRX 0.35 - 0.45 Load line 57 -64%

Core Power 35-40% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Load Line: Displayed on Heat Balance or Power/Flow Map N/A 55%

If LOW, then stop inserting control rods.

Comments I Notes:

If desired, stop power reduction after any control rod (not group) is fully inserted. Note that Load Line may increase -1.5% as flow is reduced in the next Step.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 12of15 Unit3 Attachment 8 Reactivity Maneuver Instructions STEP 5Aof6A Reactivity Maneuver Plan# U3 NRC Exam Description of Step: Urgent Load Reduction:

Lower recirculation pump speed to -28% (-480 rpm or -38% core flow).

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No x (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranges)

MW Electric N/A MFLCPR 0.70 - 0.75 MW Thermal 1045 - 1280 MAPRAT 0.25 - 0.30 Core Flow 38% MFDLRX 0.35 - 0.45 Load line 55-66%

Core Power 33 - 37% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the  !"'" _ -** . .t.* *--

THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Load Line: Displayed on Heat Balance or Power/Flow Map 66% N/A If HIGH, then insert control rods.

Comments I Notes:

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 13of15 Unit3 Attachment 8 Reactivity Maneuver Instructions STEP SA of SA Reactivity Maneuver Plan # U3 NRC Exam Description of Step: Urgent Load Reduction:

Insert control rods per Emergency Shove Sheet until desired power level is reached or SCRAM the reactor.

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No x (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranaes)

MW Electric N/A MFLCPR 0.60 - 0.70 MW Thermal 700 - 875 MAPRAT 0.20 - 0.30 Core Flow 38% MFDLRX 0.35 - 0.50 Load line 40 -45%

Core Power 21 - 30% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Load Line: Displayed on Heat Balance or Power/Flow Map 66% N/A If HIGH, then insert control rods.

Comments I Notes:

Load line/Power should continue to drop because of the xenon transient.

Bypass the RWM if power drop is to continue below 30%.

The unit should be scrammed or lined up to the A2 RWM sequence before reaching 10%

RTP.

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Reactivity Control Plan Page 14of151 Unit3 Attachment 9 (Page 1 of 1)

Power to Flow Map Recirculation Pump Trip (Steps 1 - 2)

Reactivity Control Plan # U3 NRC Exam 110 100 70

....0.

...u 20

" " " ... " . "5

%RATED CORE FLOW 70 715 80 . 1015 110 1115 120

BFN Reactivity Control Plan Page 15of15 Unit 3 Attachment 9 (Page 1 of 1)

Power to Flow Map Urgent Load Reduction (Steps 1A - SA)

Reactivity Control Plan # U3 NRC Exam R R E E 100 -~--------T-------1-------- G G

I I

.. ----------+--- --- --*--- 0 0 ----1-------------~---- ---------L----~"'------ -b~~~.,,.,.-::-----+----------*---+--11------\----+-----

N N 80 20 10

"' " ,. ... ~ ~ ~ e ~ n ~ 15 100 105 110 116

%RATED CORE FLOW

I BFN Reactivity Control Plan Page 16of15 1 Unit3 I Attachment 10 (Page 1 of 1)

Recirc Flow Maneuver Instructions Reactivity Control Plan # U3 NRC Exam RCP Flow Time Target Delta Target Completed (RO)

Step# Step# Power Flow

(%RTP or +/-..(MWe) (Mlb/Hr)

MWe)

Comments I Notes:

Reviewed by: I Unit Supervisor I Date

SHIFT MANAGER TURNOVER EQUIPMENT OOS/LCOS

• A3 EECW Pump is out of service for maintenance.

ANTICIPATED OPERATIONS/MAINTENANCE FOR ONCOMING SHIFT

• Continue the in progress Turbine Control Valve Fast Closure, or Turbine Trip and RPT Initiate Logic testing IAW 3-SR-3.3.1.1.8(9), starting at step 7 .3.

See copy attached.

• Following the Surveillance on the TCVs, return the plant to 100% power.

NO HEATERS ARE TO BE PLUGGED INTO OUTLETS IN THE CONTROL ROOM UNTIL DETERMINED IF WE CAN USE SPP 7.3 RISK REVIEW DOCUMENTATION ON P DRIVE FOLDER WCC-RISK/BFN EXPECTATION FOR 1500 MEETING IS FOR THE SM TO COVER CURRENT STATUS ON ALL CONTROL ROOM DEFICIENCIES AND ANNUNCIATORWOs THAT SHOULD BE WORKING. THIS EXPECTATION STARTED 3/26/09 DO NOT DISABLE ANY ANNUNCIATORS WITHOUT GOING THRU MANAGEMENT REVIEW.

IF ALTERNATE HEAT BALANCES ALARM COMES IN ON ANY UNIT'S ICS, REDUCE POWER BY 10 MWT (3448 IF AT RATED POWER) AND CALL REACTOR ENGINEERING IN TO PLANT TO EVALUATE.

STATUS WORK WEEK - DIV I jg], DIV II 0 OUTSIDE AIR TEMP 45°F COMMON UNIT1

• 100%

UNIT2

• 100%

UNIT3

• 90% power.

BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3.1.3.5(A)

UNIT3 REV 0026 ATTACHMENT 7 (Page 1 of2)

Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS UO(AC) 1 2nd(AC) I Peer Check2 NIA 14-31 16 00 NIA 30-47 16 00 NIA 46-31 16 00 NIA 30-15 16 00 NIA 22-31 48 00 NIA 30-39 48 00 NIA 38-31 48 00 NIA 30-23 48 00 NIA 14-39 48 00 NIA 46-39 48 00 NIA 46-23 48 00 NIA 14-23 48 00 NIA 22-47 48 00 NIA 38-47 48 00 NIA 38-15 48 00 NIA 22-15 48 00 REMARKS 4 : SHOVE SHEETS - Continuously Insert NOTES:

(1) RWM Group may be marked "N/A" ifnot applicable (i.e., when above the LPSP).

(2) For all rod moves to position "full out" (notch position 48), this signoffverifies coupling integrity was checked in accordance with 3-0I-85.

(3) Documentation of Peer Check by a second qualified member of the plant staff (i.e., RE, STA, OR UO) is required ONLY when the RWM is Inoperable OR bypassed with core thermal power <10%

(4) Record the rod number and any problems encountered on attachment 1, page 2 of2, as applicable.

(5) Peer check by RE or SRO [SRO to be utilized ifa second RE is unavailable]. The SRO should be checking FROM and TO control rod positions as a minimum. The RE OR SRO should be checking the positions identified for agreement with the predictor cases.

Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.

Issued by: ~VJ~ I Today Reviewed by5: H~ K.td I Today_

Reactor Engineer Date RE OR SRO Date Authorization to perform the control rod manipulations identified on this sheet

~~ I Today Unit Supervisor Date

BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3. l .3.5(A)

UNIT3 REV 0026 ATTACHMENT 7 (Page 2 of2)

Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS UO(AC) 1 2nd(AC) I Peer Check2 NIA 06-31 48 00 NIA 30-55 48 00 NIA 54-31 48 00 NIA 30-07 48 00 NIA 06-39 48 00 NIA 54-39 48 00 NIA 54-23 48 00 NIA 06-23 48 00 NIA 22-55 48 00 NIA 38-55 48 00 NIA 38-07 48 00 NIA 22-07 48 00 NIA 06-47 48 00 NIA 54-47 48 00 NIA 54-15 48 00 NIA 06-15 48 00 REMARKS 4 : SHOVE SHEETS-Continuously Insert NOTES:

(1) RWM Group may be marked "NIA" if not applicable (i.e., when above the LPSP).

(2) For all rod moves to position "full out" (notch position 48), this signoffverifies coupling integrity was checked in accordance with 3-01-85.

(3) Documentation of Peer Check by a second qualified member of the plant staff(i.e., RE, STA, OR UO) is required ONLY when the RWM is Inoperable OR bypassed with core thermal power <10%

(4) Record the rod number and any problems encountered on attachment 1, page 2 of2, as applicable.

(5) Peer check by RE or SRO [SRO to be utilized ifa second RE is unavailable]. The SRO should be checking FROM and TO control rod positions as a minimum. The RE OR SRO should be checking the positions identified for agreement with the predictor cases.

Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.

Issued by:_gai VJ~_! Today Reviewed by5: H~ ~ I Today Reactor Engineer Date RE OR SRO Date Authorization to perform the control rod manipulations identified on this sheet

~R~ I Today Unit Supervisor Date

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U3 NRC 4 Unit3 Turbine Valve Testing

IBFN Reactivity Control Plan Attachment 7 (Page 1of2)

Reactivity Control Plan Form BFN Unit: 3 Valid Date(s): 1/11/15 - 2/9/15 Reactivity Control Plan#: U3 NRC 4 Are Multiple Activations Allowed: --'-N"'""o__ (If yes, US may make additional copies)

Prepared by: ----'-~""""'~~-"-~--+---*

____ /--'-1*'-'-'-+ Reviewed by: Q.~ ~~I 1ffl Reactor Engineer Date Qualified Reactor Engineer Date Approved by: R.E ~ I 1ffl Concurrence: JJ.,,. W. S.R.O I 1ffl RE Supervisor Date WCC/Risk/US SRO Date Approved by: O.P. H~ I 1ffl Authorized by: H,,. Stq;t H~ I 1ffl Ops Manager or Supt. Date Shift Manager Date RCP Activated: _ _ _ _ _ _ _----'/'---- RCP Terminated: ~-------___;,_--~ I Unit Supervisor Date Unit Supervisor Date Title of Evolution: Unit 3 Turbine Control Valve Testing Purpose/Overview of Evolution: Reduce Reactor Power to 90% with Reactor Recirc flow, for valve testing and return Rector Recirc flow to raise power back to 100%.

Maneuver Steps

1. Reduce Reactor Power using Reactor Recirc flow reduction. Initial power level 100%, final power level 90%.

2. At the completion of Turbine Valve Testing, raise Reactor Power back to 100%

using Reactor Recirc flow, with no additional restrictions on rate of return.

IBFN Reactivity Control Plan Attachment 7 (Page 2 of 2)

Reactivity Control Plan Form Operating Experience and General Issues: U3 NRC 4 Previously known control rod issues:

None Cautions/Error Likely Situations/Special Monitoring Requirements/Contingencies:

IBFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 1of1 Reactivity Maneuver Plan # U3 NRC 4 Description of Step:

Conditions : To be recorded at the Completion of Step Recorded: I (by RO) (Date)

QRE presence required in the Control Room? Yes No X (check)

Predicted Actual Predicted Actual (may be ranges) (may be ranges)

MW Electric 1152 MFLCPR 0.830- 0.840 MW Thermal 3455 MAPRAT 0.649 - 0.655 Core Flow 87.1 mlbm/hr MFDLRX 0.785- 0.810 Load line 110.2 Core Power 100% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.

Description including frequency, method of monitoring, AND High Low contingency actions Comments I Notes:

Step Complete AND Reviewed by: I Unit Supervisor I Date

BFN Control Rod Coupling Integrity Check 3-SR-3.1.3.S(A)

Unit2 Rev.0026 ATTACHMENT 9 POWER TO FLOW MAP U3 Startup Scenario 8 110 100 90 80

~

0::

w 70

=0 D..

....I 60 0::

w

...:x:c 50

...<w 40 c

D..

> 30 20 10 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 PERCENT RATED CORE FLOW