ML072920097
| ML072920097 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/22/2007 |
| From: | Florida Power & Light Energy Seabrook |
| To: | NRC Region 1 |
| Caruso J | |
| Shared Package | |
| ML062050109 | List: |
| References | |
| 50-443/07-301 | |
| Download: ML072920097 (22) | |
Text
i - ,
Appendix D, Rev. 9 Scenario Outline Form ES-D-I Facility: Seabrook Station Scenario No.: 2 (6) OpTestNo.: __
Examiners: Operators:
Initial Conditions: Middle of Life. 75% power after downpower at 200/dhr.
Turnover: Continue power increase @ 1Oo&hr. SUFP Is aligned to Bus 4 for breaker testing of Bus 5 SUFP breaker.
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Event Type*
R-SRO Event Description Power Increase, Reactivity Change N-ATC N-BOP R-ATC R-BOP I-SRO Controlling channel of PZR level fails low.
I-ATC C-ATC C-SRO Charging pump 2A overcurrent trip.
C-ATC C-SRO 'B' Main Feed Pump shaft shear.
C-BOP M-SRO Failure of reactor to trip.
M-ATC R-SRO R-ATC R-BOP c-BOP Turbine Driven Emergency Feedwater Pump trip.
C-SRO C-BOP Main Turbine fails to trip on demand 0 mfED038 M-SRO Loss of Offsite Power FR-H.1.
M-ATC M-BOP
FPL E Seabrook Station SIMULATOR EXAMINATION 2005 - 2007 LOlT DEMONSTRATIVE EXAMINATION Scenario 6 Rev. 02 This material is developed for FPL Energy Seabrook training programs by the Training Group.
Text materials and figures contained in this document are developed for purposes of instructior and should not be used in connection with either plant maintenance or plant operation. This material may not be reproduced without the authorization of the Training Manager.
PREPARED BY: DATE:
INSTRUCTOR REVIEWED BY: DATE:
SME (OPTIONAL)
APPROVED BY: DATE:
TWINING SUPERVISOR
SCENARIO The simulator is reset to IC #31, Middle of Core Life, 75% Power. The Reactor will not trip automatically or manually and the main turbine will fail to trip automatically.
After the reactor is locally tripped, a loss of offsite power occurs and the B Emergency Diesel Generator breaker does not close. No means of emergency boration will exist at this point. The motor driven EFW pump is not available because it is powered from Bus 6 (B EDG). The turbine driven EFW pump trips on overspeed and will be restored in FR-H.1.
The crew completes FR-S.1 and goes to the procedure and step in effect (Normally E-0). The crew should go straight to FR-H.l based upon a Red Path on Heat Sink.
At step 1 of FR-H.l, the crew verifies heat sink required At step 2 the crew will note that no CCP (charging pump) is available and go directly to step 10 (bleed and feed).
The crew will perform steps 10 through 18 to establish bleed and feed colling of the RCS and at step 20 begins to look for sources to feed the steam generators.
The crew will establish the Turbine Driven Emergency Feedwater Pump as a feed source and return to step 21 to continue with the procedure.
At steps 24 and 25 the crew will close the PORVs to terminate bleed and feed cooling and transition to E-I, Loss of Reactor or Secondary Coolant.
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SIMULATOR SETUP 1 Reset the simulator to IC #31,Middle of Core Life, 75% Power 2 Reactor will not trip automatically or manually.
Insert the following Reactor Protection system malfunctions:
0 mfRPSOOl , Automatic Reactor Trip Failure Train A 0 mfRPS002, Automatic Reactor Trip Failure Train B 0 mfRPS027, Reactor trip Switches Fail to Actuate (Train A) mfRPS028, Reactor trip Switches Fail to Actuate (Train B) 3 EDG-1B breaker will not close on demand. Insert Malfunctions, Electrical Distribution Component bkEDEGDG, Select Fail Open 4 Align SUFP to Bus 4 for SUFP Breaker testing on Bus 5.
Place a Danger Tag on SUFP Breaker Control Switch.
5 On Sim Diagram FW3, FW-P-37B, Component Malfunction 0 bkFWP37B SELECT: FAIL OPEN 0 INSERT 6 Main turbine will fail to automatically trip. Insert the following RPS malfunction:
0 mfRPS003, Automatic Turbine Trip Failure Page 3 of 21
SHIFT TURNOVER The plant is at 75% power. The current boron concentration is 1171 ppm.
The SUFP is aligned to Bus 4 for Breaker Testing of the BUS 5 SUFP Breaker.
Axial Flux Difference is on target at - 1. I 2 %.
Continue the power increase at 10% per hour.
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DEMONSTRATIVE EXAMINATION Rev. 01 I SCENARIO OUTLINE EVENT INSTRUCTION COMMENTS Event f Shift Turnover Shift turnover information as stated. Operators should review alarms and indications.
Crew continues power increase to 100% at 10% per hour.
At the Chief Examiners discretion continue to the next event.
Unit Supervisor: A brief reactivity review will take place discussing the temperature limits for Tave to be used in the power increase. The temperature band will normally be +2"F. Control rods will be used for AFD and temperature control.
Turbine Operations: The BOP operator will increase the turbine Load Pot in 3 to 5 increments, called "Flats".
This opens the turbine control valves to increase turbinelreactor power.
BOP operator should verify the change with control valve position change, temperature change and power change.
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DEMONSTRATIVE EXAMINATION Rev. 01 EVENT INSTRUCTION COMMENTS Event 1 Reactor Power change: The crew will use control rods initially to increase temperature during the turbine load increases. Using ODI-56, a dilution value will be determined to change the boron concentration and increase power/temperature. If control rods are used in manual the operator will verify rod speed, place the Rod Motion Selector switch to the OUT direction and withdraw the rods a maximum of three steps and monitor temperature and power.
Using the laminated sheets for the Figure 2: Dilution Check List, the RO will set up the controllers for the required dilution volume and rate.
The high level steps are:
Verify the pumps in AUTO Verify the makeup valves are in AUTO Activate Non-RCS Make Up Function on plant computer Place Blender Mode Start Switch to STOP Place the Mode Selector Switch to Dilute or Alt. Dilute Set the quantity on CIS-FIQ-111 controller Set the Mode Start Switch to START Verify the pumps and valves respond Verify Plant Response.
Restore System Deactivate Non-RCS Makeup on plant computer.
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DEMONSTRATIVE EXAMINATION Rev. 01 EVENT INSTRUCTION COMMENTS EVENT 2 0 Insert Malfunction: Reactor Coolant Alarms:
VAS D4461, PZR LVL Low 8, HTR Interlock Actuated The Component Malfunctions controlling VAS F4323, PZR Control Heaters Tripped channel of VAS F4324, PZR Group C Backup Heaters Tripped Pressurizer [7 ltRCLT459, FAILS LOW VAS F4325, PZR Group D Backup Heaters Tripped level will fail VAS 88186, PZR STMlCHG Line AT Approach Limit low. INSERT Conditions:
-RC-LCV-459 and CS-V-145 will go closed and isolate letdown flow.
At 5% pressurizer level deviation alarm D4436, PZR LVL Deviation High &
BU HTRS On will actuate and the Pressurizer Backup Heaters will energize.
The crew responds using OS1201.07, PZR Level Instrument Failure.
Step I)The RO should identify the controlling level channel has failed low.
Step 2) The step calls for manually controlling pressurizer level at program.
0 The RO will have to control charging to supply just the RCP seals.
0 The operator should take manual control of CS-FK-121, Charging Flow Control Valve, and adjust CS-HCV-182, Seal Injection flow control to accomplish this task.
The RO should select alternate channels for Control/Backup. The operator should select the 461/460 combination.
The RO should select an alternate channel (460 or 461) for the pressurizer level recorder.
Step 3) The RO should reset the control group pressurizer heaters Step 4) The RO should verify that RC-LCV-459 has closed Page 7 of 21
DEMONSTRATIVE EXAMINATION Rev. 01 Step 5) a. The RO should verify or establish greater than 50 gpm charging flow by adjusting CS-FCV-121 andlor CS-HCV-182 accordingly.
- b. The RO should verify that pressurizer level is greater than 17%.
Step 6) The RO should reestablish letdown flow per the following process:
Verify PCCW is aligned to the Letdown Heat Exchanger and that TCV-130 is in auto.
Close Letdown Flow Control Valves CS-HCV-189/190.
Open Letdown Line Isolation Valves. Operator should reopen RC-LCV-459 and CS-V-145.
Establish letdown flow with CS-HCV-189 or 190. The operator should utilize the backpressure control valve CS-PK-131 to maintain adequate backpressure. The operator may initially utilize CS-PK-131 in manual and then return the valve to automatic when letdown flow is established.
Alarms:
During letdown flow restoration the following alarms may occur:
67399, RCS Unidentified Leak Rate High 88266, RCS Unidentified Leak Rate Warning The leak rate monitor is accurate during steady state conditions. The variances introduced during this transient will cause these alarms.
NOTE: It takes a period of time to restore pressurizer level back to its program value. The operator should establish a letdown flowrate that is greater than the charging flowrate to return level to program. When level finally returns to setpoint the crew should utilize procedure OS1002.08,
'Pressurizer Level Control System Operation' as additional guidance for returning level control to automatic.
Step 7) When pressurizer level returns to setpoint the RO should verify proper controller setpoint and return RC-LK-459 and/or CS-FK-121 to automatic.
Step 8) The RO should verify that there are no redundant Pressurizer Level Hi bistables lit on UL-6.
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DEMONSTRATIVE EXAMINATION Rev. 01 Step 9)
- a. The Unit Supervisor should identify the following Tech. Spec. items as applicable:
3.3.1, Reactor Trip System Instrumentation, Table 3.3-1, Item 11, Action 6 requires the channel bistable to be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 3.3.3.6, Accident Monitoring Instrumentation, Table 3.3-10, Itern 5, Action A requires that, with one less than the total channels, the affected channel must be restored to operable within 30 days.
3.3.3.5, Remote Shutdown System, Table 3.3-9, Itern 5. the crew may discuss that the remote safe shutdown instrument comes off of the same instrument tap as level channel 459. The crew may discuss the possibility of investigating whether the RSS channel is affected.
- b. If desired to test or troubleshoot, Unit supervisor may direct I&C to place bistables to "bypass" using BTI for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Allow the crew to complete the abnormal procedure or at Lead Examiners discretion prior to entering the next malfunction.
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DEMONSTRATIVE EXAMINATION Rev. 01 EVENT Event 3 U
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INSTRUCTION Insert Malfunction: Chemical and COMMENTS Alarms:
Volume Control System, mfCSO16, UA-52,C-3, RCP Seal Injection Flow Lo An over- 'CS-P-2A OC Trip'. D4652, CNTRFGL CHG Pump A BKR Trip & UO current tric, D7853, CHG PMP DISCH Combined HDR Low Low of Charging D4676 through D4679, RCP Seal Injection Flow Low alarms Pump, CS-P-2A will F6958, CVCS Train A h o p occur.
The crew responds using OS1202.02, 'Charging System Failure'.
Step 1) RO should state that charging flow is unavailable.
RO should reduce letdown flow andlor isolate letdown flow by closing CS-V-145. This step is prudent to prevent flashing conditions in the regenerative heat exchanger.
Step 2) The crew should be monitoring the affects on RCP seals due to the loss of seal injection. The crew should use Main Plant Computer color graphics Reactor Coolant Pump displays and/or computer analog points. Should be monitoring seal water inlet temperature <230"F.
Instructor CUE: After charging pump trip the Step 3) The RO should state that there are no charging pumps running.
crew may dispatch an NSO to check the pump The RO will start the 'B' charging pump utilizing the following sequence:
breaker. If so, then after a time delay have the Place the charging flow controller CS-FK-121 in manual and reduce the NSO report back that the breaker has tripped output signal to 0%.
on over-current. The RO should start the '8' centrifugal charging pump from the control board. The operator may utilize computer color graphics and the analog ammeter to observe pump conditions.
The RO should slowly establish charging flow and establish desired seal injection flow by operating the charging flow controller, CS-FK-121 in conjunction with the seal injection flow control valve, CS-HCV-182.
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DEMONSTRATIVE EXAMINATION Rev. 01 Step 4)
- a. The RO should verify proper seal injection filter AP. The in service filter should be reading approx. 5 psid.
- b. The operator should verify seal injection flow greater than 6 gpm or adjust flow accordingly utilizing CS-HCV-182.
Step 5) The crew should verify that the charging system is intact. The operators should look at RDMS displays, building sump levels, and dispatch personnel to walk down the PAB.
The RCS is intact, the crew should go to Step 9.
Step 9) The RO should verify that VCT level is greater than 30%
Step I O ) The crew should check if normal letdown can be established The RO should verify that charging flow is greater than 50 gpm and pressurizer level is greater than 17%.
w The RO should re-establish letdown flow per the following process:
Verify PCCW is aligned to the Letdown Heat Exchanger.
Close Letdown Flow Control Valves CS-HCV-189/190.
Open Letdown Line Isolation Valves. Operator should reopen RC-LCV-459 and CS-V-145.
Establish letdown flow with CS-HCV-189 or 190.
The operator should utilize the backpressure control valve CS-PK-131 to maintain adequate backpressure.
The operator may initially utilize CS-PK-131 in manual and then return the valve to automatic when letdown flow is established.
The crew should go to Step 12.
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DEMONSTRATIVE EXAMINATION Rev. 01 ,
Step 12) : It takes a period of time to restore pressurizer level back to it's program value.
The operator should establish a letdown flowrate that is greater than the charging flowrate to return level to program.
When level returns to setpoint the crew should utilize procedure OS1002.08, 'Pressurizer Level Control System Operation' as additional guidance for returning level control to automatic.
Step 13) The Unit Supervisor should identify the following Tech. Spec. as applicable:
3.5.2, ECCS Subsystems, Action A, requires restoration of the inoperable ECCS subsystem to operable status within 7 days.
Allow the crew to complete the abnormal procedure or at Lead Examiners discretion prior to entering the next malfunction.
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DEMONSTRATIVE EXAMINATION Rev. 01 :
EVENT INSTRUCTION COMMENTS 17 Malfunctions, Feedwater The B Main Feedwater Pump Shaft Shear will initiate the major event.
Event 4 Component. Feedwater flow will be drastically reduced. No automatic plant setback will The B Main occur because a pump trip signal is not generated. Steam Generator levels Feedwater Pump 0 Insert Malfunction: FWP32B decrease rapidly and a SG Lo-Lo level reactor trip demand occurs @! 20%
shaft will shear (no Shaft Shear. SG Narrow Range level.
pump trip signal).
Select: INSERT Conditions:
BOP may report low 'B' Feed Pump suction flow (FW-FI-4065-1) s BOP may report mismatch in 1 stage flow and governor valve position for Feed Pump ' B as compared to Feed Pump ' A .
Alarms:
D4700, MFP B Suction Header Flow Low D4704, SG D STM FlowIFeed Flow Mismatch D4706, SG A STM Flow/Feed Flow Mismatch D4708, SG B STM FlowlFeed Flow Mismatch D4710, SG C STM Flow/Feed Flow Mismatch Alarms @ 45% level decreasing:
UA35, A3, SG A Level Hi/Lo UA35, B3,SG B Level Hi/Lo UA35, C3, SG C Level Hi/Lo UA35, D3, SG D Level HilLo D4770, SG A LVL Ref Deviation D4771, SG B LVL Ref Deviation D4772, SG C LVL Ref Deviation D4773. SG D LVL Ref Deviation Alarms @ 35% level decreasing:
D4876, SG A Level Low D4877, SG B Level Low D4878, SG C Level Low D4879, SG D Level Low Page 13 of 21
DEMONSTRATIVE EXAMINATION Rev. 01 Alarms @ 20% level decreasing:
Primary VAS Red Alarms:
D4865, RX Trip SG A LVL Lo Lo D4866, RX Trip SG B LVL Lo Lo D4867, RX Trip SG C LVL Lo Lo D4868, RX Trip SG D LVL Lo Lo BOP VAS Alarms:
F4756, SG A Level Lo Lo F4840, SG B Level Lo Lo F4841, SG C Level Lo Lo F4850, SG D Level Lo Lo Crew Response:
NOTE: With the B MFP
- This event will lead to the necessity for a reactor trip. The crew may shaft sheared, its recirc address the VPRO procedures for the steam generator feed flow mismatch valve will open based alarms. For example; D4704, SG D STM FlowlFeed Flow Mismatch upon low flow through the includes the following actions:
pump. The BOP may
- 1) Check steam and feed flow indications on the main control board.
diagnose this as the recirc
- 2) Check for proper feed reg valve response, take manual control if valve failing open. necessary.
- 3) Check FW system parameters.
- 4) Check steam generator parameters.
- 5) Check for any possible steam leaks.
- The loss of a main feed pump will degrade feed flow to all 4 steam generators.
The crew should recognize the loss of steam generator levels and realize that a reactor trip is warranted, based upon approaching the Lo-Lo level trip setpoint of 20% Narrow Range level.
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DEMONSTRATIVE EXAMINATION Rev. 01
..
Event 5 On the reactor trip attempt:
The crew will respond by performing the immediate actions of E-0, The reactor does 0 Sim Diagram MS7, MS-V-129 REACTOR TRIP OR SAFETY INJECTION.
not trip and 0RF:Manual Adjust Step 1) The RO should recognize that the reactor did not trip.
Event 6 Step 1-RNO) The RO should attempt to manually trip the reactor using both LOSSof EFW-P- SELECT:O for value main control board reactor trip switches. The RO should inform the Unit 37A, Turbine Supervisor that the reactor did not trip.
Driven EFW Pump 0INSERT The crew should transition to FR-S.1, RESPONSE TO NUCLEAR Event 7 POWER GENERATIONIATWS.
Main Turbine fails CUE: -40 seconds after the Rx Trip The RO and BOP should perform the immediate actions of FR-S.1.
to automatically attempt Fail FW-P-32A via Shaft Step 1) The RO should recognize that the reactor did not trip and should be trip Shear: verifying that control rods are inserting in automatic or the RO should be Malfunctions, Feedwater manually inserting rods.
Component. Step 2) The BOP should recognize that the main turbine has failed to 0Insert Malfunction: FWP32A Shaft automatically trip and should manually trip the turbine by depressing the Shear. turbine trip pushbutton.
0Select: INSERT NOTE:
The motor driven EFW pump breaker will trip.
The turbine driven EFW pump trips on overspeed and will not be restored.
The crew should dispatch an operator to reset MS-V-129.
Step 3) The BOP should verify EFW status Step 4) The RO should initiate Emergency boration:
Verify a charging pump running.
Start a boric acid pump/ Open emergency borate valve.
Place charging flow control valve CS-FK-121 open to charge at max rate.
Open CCP suction valves from RWST, CS-LCV-112D and CS-LCV-112E.
Close CCP suction valves from VCT, CS-LCV-1128 and CS-LCV-112C Check pressurizer pressure less than 2385 psig Page 15 of 21
DEMONSTRATIVE EXAMINATION Rev. 01 Step 5) The RO should verify Containment Ventilation Isolation. This can be done by looking at UL status lights.
Step 6) The crew should have dispatched an NSO to locally trip the reactor.
Step 7) Once the reactor is tripped locally the RO should verify that the reactor is subcritical.
Return to procedure and step in effect. This would normally be E-0 however a Red path on the Heat Sink CSF should exist at this point and the crew should transition to FR-H.l, 'RESPONSE TO LOSS OF SECONDARY HEAT SINK'.
After the reactor After the reactor is manually tripped, a loss of offsite power occurs and the Instructor CUE:
is manually 'B' Emergency Diesel Generator breaker does not close.
tripped, a loss of Manually trip the reactor after FR-offsite power S.1, step 4 is complete, No means of emergency boration will exist at this point.
occurs and the -0R-
'B' Emergency The motor driven EFW pump is not available because it is powered from Diesel Generator Wait 2 minutes and perform the Bus 6 ('B' EDG).
breaker does not following to trip the reactor:
close.
[7 DMF mfRPSOOl , Automatic Restoration of the Turbine Driven EFW Pump will be the ultimate success Reactor Trip Failure Train A path in FR-H.1.
0DMF mfRPS002 , Automatic Reactor Trip Failure Train B 0DMF mfRPS027 , Reactor trip Switches Fail to Actuate (Train A) 0DMF mfRPS028 , Reactor trip Switches Fail to Actuate (Train B)
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DEMONSTRATIVE EXAMINATION Rev. 01 Event 8 Malfunctions, Electrical FR-H.l, RESPONSE TO A LOSS OF SECONDARY HEAT SINK:
After the reactor Distribution; is manually Step 1) The crew should verify that RCS pressure is greater than non-faulted tripped a LOP 0IMF mfED038 steam generator pressure and that RCS Hot Leg temperature is greater than occurs 350°F.
Step 2) The RO should verify that no centrifugal charging pump is available The crew should go to step 10 to establish RCS bleed and feed.
Step IO) The RO should verify all RCPs stopped Step 11) The RO should actuate SI using the manual SI switch.
Step 12) Verification of feed status. The RO should identify that there are no CCPs running and that there is one SI pump Running, SI-P-6A.
Step 13) The RO should perform the following to establish a bleed path:
- a. Verify power to RC-VI24
- b. Verify both PORV block valves open.
- c. Open both PORVs.
Step 14) The RO should verify an adequate bleed path exists Page 17 of 21
DEMONSTRATIVE EXAMINATION Rev. 01 Step 15) The RO or BOP should perform Attachment 'D' to verify proper SI equipment actuation.
Step 16) The RO should verify ECCS flow and maintain both PORV's open as a continuous action throughout the bleed and feed process.
Step 17) The RO should reset SI using the control board SI reset switches Step 18) The RO should reset Phase A and Phase B signals using the control board 'T' and 'P' signal reset switches.
Step 19) The BOP should check the status of instrument air.
Step 20) The RO should verify that no CBS pumps are running Step 21) The crew should continue attempts to establish a secondary heat sink.
The crew refers to Step 3 to attempt to establish flow from the Turbine Driven EFW Pump, FW-P-37A.
The crew will establish flow from the EFW pump and should continue with Step 22.
Restore the Turbine Driven EFW Restoration Of Pump, FW-P-37A, in FR-H.l: NOTE: Due to the RCS Bleed and Feed containment pressure may be EFW Flow. elevated to greater than the Adverse Containment value of 4 psig. If so, a 0Sim Diagram MS7 check of SG levels should be done before restoring feed flow as "Hot Dry" SG conditions may apply. Hot Dry SG is defined as a WR level 4 4 %
0On MS-V129, Select RF: Manual (30% Adverse Containment).
Adjust.
0Value = 1.o 0Ramp = 10 seconds Page 18 of 21
DEMONSTRATIVE EXAMINATION Rev. 01 If bleed and feed has been established AND RCS temperatures are increasing recovery of a dry SG should be initiated by selecting a single
-
intact SG and feeding at the maximum rate.
If bleed and feed has been established AND RCS temperatures are stable or decreasinq, recovery of a dry SG should be initiated by selecting a single intact SG and feeding less than or equal to 100 gpm. When WR level is greater than 14% (30% Adverse Containment), Feedwater flow can be adjusted as necessary to restore adequate level.
Step 22) The BOP should observe SG levels for verification of adequate heat sink. If levels are adequate the crew should go to Step 23. If levels are not adequate the crew must wait until level recovers.
Evaluator CUE: At step 22, Steam Generator levels may not have recovered enough to continue with the procedure. The exam may be stopped here. If steam generator level has recovered, continue to steps 25 and 26 to stop bleed and feed.
Step 23) The RO should verify that Core Exit Thermocouple and RCS Hot Leg temperatures are decreasing.
Step 24) The RO should verify that the Reactor Head Vent valves are closed.
Step 25) The RO should report that subcooling is not adequate for SI termination. The crew should go to Step 26.
Step 26) The RO should close one of the PORV's. The crew should return to Step 25.
Step 25) The RO should report that subcooling is not adequate for SI termination. The crew should go to Step 26.
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DEMONSTRATIVE EXAMINATION Rev. 01 Step 26) The RO should close the remaining open PORV's. The crew should return to Step 25.
Step 25) The RO should report that subcooling is not adequate for SI termination. The crew should go to Step 26.
Step 26) The RO should verify no bleed paths open The crew should transition to E-I, LOSS OF REACTOR OR SECONDARY COOLANT.
Terminate the examination when the crew is at the transition to E-I, or at Lead Examiners discretion.
CUE: When the simulator is placed in FREEZE the evaluator should administer SRO ADMIN JPM05, E-Plan Classification and Notification. The candidate should be directed to classify the event based upon the worst conditions experienced during the scenario.
Emergency Plan Classifications:
Initial: Site Area Emergency due to FR H RED.
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II
- h DEMONSTRATIVE EXAM # REV.00 CREW CRITICAL TASKS
- 1. INSERT negative reactivity into the core by at least one of the following methods in accordance with FR-S.1.
Automatic andlor manual insertion of the RCCAs.
Establish Emergency Boration flow to the RCS.
- 2. ISOLATE the main turbine from the SG's before proceeding to step 5 of FR-S.l on an ATWS initiated by a loss of feedwater.
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