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| {{#Wiki_filter:ppendix D Scenario Outline Form ES-D-1 Facility: Watts Bar October 2011 Scenario No. 5 Op Test No.: 2 Examiners: Operators: SRO RO BOP Initial Conditions: 100% power, RCS boron concentration 747 ppm. Control Bank D is at 220 steps. | | {{#Wiki_filter:}} |
| Turnover: Train A/Channel 1 Work Week. lA-A CCP is tagged for motor bearing replacement. LCO 3.5.2 and TR 3.1.4 were entered 8 hours ago. Protected equipment signs have been posted for lB-B CCP. An oil leak developed on the 1A MFP during the last shift. Leak rate is being monitored, and Operations Management is meeting to determine whether the pump will be removed from service.
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| Event MaIf. No. Event Type* Event Description No.
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| 1 rx05a 1-RO 1-LT-68-339, PZR LEVEL fails low.
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| TS-SRO 2 rxl 8 C-RO T-avg control signal fails high.
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| 3 n/a R-RO Operations Superintendent contacts the control room and states N-BOP/SRO that the 1 A MFP is to be taken out of service. Power is to be reduced at 3%/mm to 75% using AOI-39, Rapid Load Reduction.
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| 4 fwo5a C-BOP At 95% power, 1A Main Feedwater pump trips due to low oil TS-SRO pressure.
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| 5 fw94 C-BOP 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1-fw95 FCV-2-21 0, MFPT A CONDENSER CNDS INLET fail to close automatically.
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| 6 fw29c M-ALL 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing fw22c SG 3 level to drop. The automatic reactor trip function has failed, fw49a requiring a manual trip of the reactor. E-0, Reactor Trip or Safety rpOl b Injection, will be entered and a transition to ES-0.1, Reactor Trip ed06b Response, will be made. At the trip, the TDAFW pump becomes steam bound, the 1A MDAFW pump shaft shears. lB-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink, will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4. Bleed-and-feed must start immediately, since neither CCP is available.
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| (N)ormal, (R)eactivity, (I )nstrument, (C)omponent, (M)ajor Appendix D Watts Bar Examination October 2011 NUREG 1021 Revision 9
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| ppendix D Scenario Outline Form ES-D-1 Scenario 5 - Summary Initial Condition 100% power, RCS boron concentration 747 ppm. Control Bank D is at 220 steps.
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| Turnover Train A/Channel 1 Work Week. lA-A CCP is tagged for motor bearing replacement. LCO 3.5.2 and TR 3.1.4 were entered 8 hours ago. Protected equipment signs have been posted for I B-B CCP. An oil leak developed on the IA MFP during the last shift. Leak rate is being monitored, and Operations Management is meeting to determine whether the pump will be removed from service.
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| Event 1 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal. Requires a Tech Spec evaluation Event 2 T-avg control signal fails high. Requires the RO to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled rod movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
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| Event 3 Operations Superintendent contacts the control room and states that the 1A MFP is to be taken out of service due to the slow oil leak. Power is to be reduced at 3%/mm to 75% using AOl-39, Rapid Load Reduction.
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| Event 4 At 95% power, 1A Main Feedwater pump trips due to low oil pressure. Requires entry into AOl-i 6, Loss of Normal Feedwater, Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
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| Requires a Tech Spec evaluation of Axial Flux Difference. Requires a boration to return AFD to target band.
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| Event 5 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1-FCV-2-210, MFPT A CONDENSER CNDS INLET fail to close automatically. Requires the BOP to manually close valves.
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| Event 6 1 -LCV-3-90, SG 3-MEW REG VLV fails to 25% position causing SG 3 level to drop. Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection, will be entered, and a transition to ES-0.1, Reactor Trip Response, will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. I B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-Hi, Loss of Secondary Heat Sink, will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4. Bleed-and-feed must start immediately, since neither CCP is available.
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| Appendix D Watts Bar Examination October 2011 NUREG 1021 Revision 9
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| ppendix D Scenario Outline Form ES-D-1 Scenario 5 - Critical Task Summary Critical Task 1 Manually trip the reactor from the control room upon recognition of the failure of the automatic trip circuit.
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| Critical Task 2 Initiate RCS bleed and feed so that the RCS depressurizes sufficiently for safety injection pumps to inject into the RCS.
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| Appendix D Watts Bar Examination October 2011 NUREG 1021 Revision 9
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| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 5 Simulator Console Operators Instructions SIMULATOR SETUP INFORMATION ENSURE Examination Security has been established.
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| 2 RESET to Initial Condition 304 by performing the following actions:
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| : a. Select lCManager on the THUNDERBAR menu (right hand side of Instructor Console Screen).
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| : b. Locate IC# 304.
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| : c. Right click on IC# 304.
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| : d. Select Reset on the drop down menu.
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| : e. Right click on RESET.
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| : f. Enter the password for IC 304.
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| : g. Select Yes on the INITIAL CONDITION RESET pop-up window.
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| : h. Perform SWITCH CHECK.
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| 3a SELECT Director on the THUNDERBAR menu (right hand side of Instructor Console Screen).
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| 4 ENSURE the following information appears on the Director Screen:
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| Key Description Type Event Delay Inserted Ramp Initial Final Value cvola charging pump a trip M 00:00:00 00:00:00 00:00:00 Active Active 1W94 fail auto close 2-205 M 00:00:00 00:00:00 00:00:00 Active Active fw22c airbound tdafw pump M 00:00:00 00:00:00 00:00:00 Active Active lt95 fail auto close 2-210 M 00:00:00 00:00:00 00:00:00 Active Active rpolb automatic reactor trip signal failure (atws) M 00:00:00 00:00:00 00:00:00 Active Active rx05a pzr level transmitter fails to position chnl 1 68-339 M 1 00:00:00 00:00:30 0 59.4271 Page 1 of 5
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| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 5 Simulator Console Operators Instructions SIMULATOR SETUP INFORMATION Key Description Type Event Delay Inserted Ramp Initial Final Value
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| ] ed06b loss of 6.9 kv shutdown board bus I b-b M 19 00:00:00 00:00:00 Active InActive fw49a afti pump a sheared shaft M 19 00:00:00 00:00:00 Active InActive rxl8 t-avg control signal failure M 2 00:00:20 00:00:00 100 56.4271 fw05a turbine driven feed pump a trip M 4 00:00:00 00:00:00 Active InActive fw29c main fw reg vlv fcv-3-90 fail position M 6 00:00:00 00:00:00 25 0
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| . rprl8 blocktrainaautosi R 21 00:00:00 00:00:00 Block Normal rprl9 blocktrainbautosi R 21 00:00:00 00:00:00 Block Normal pi-46-20 02120 mfp a brg oil press 0 4 00:00:00 00:00:00 0 18.9983 5 Place simulator in RUN and acknowledge any alarms.
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| 6 ENSURE 1-HS-62-108A, CCP A-A (ECCS (ECCS) is in Stop, Pull-to-Lock position and a Hold Notice (Red) Tag is placed on the handswitch. Place pink PROTECTED EQUIPMENT tag on 1-HS-62-104A, CCP B-B (ECCS).
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| 7 -
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| ENSURE the Train A Week Channel I sign is placed on 1-M-30.
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| 8.. Place simulator in FREEZE.
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| 9 ENSURE Watts Bar Nuclear Plant Unit I Reactivity Briefing Book MOL (Middle Of Life) is provided to the crew as part of the Turnover Package, and that the MOL placards are on 1-M-6, below the Boric Acid and Primary Water Integrators.
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| : 10. WHEN prompted by the Chief Examiner, place the Simulator in RUN.
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| Page 2 of 5
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| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 5 Simulator Console Operators Instructions Exam Simulator Event Event DescriptionlRole Play No. No.
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| ROLE PLAY: When the applicants assume shift, the BOP may contact the TB AUO for the latest status of the oil leak on IA MFW pump. If/when contacted, state that the oil leak rate is stable, approximately 0.25 gpm. The oil is contained and Fire Operations personnel are at the IA MFW pump.
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| I I 1-LT-68-339, PZR LEVEL fails low.
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| ROLE PLAY: When contacted as Work Control, repeat back request to remove the failed channel from service.
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| ROLE PLAY: When contacted as Work Control, repeat back request for a troubleshooting and repair package for I-L T-68-339, PZR LEVEL.
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| 2 2 T-avg control signal fails high.
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| ROLE PLAY: When contacted as Work Contro4 repeat back request to prepare a troubleshooting and repair package for the Tavg auctioneering circuit.
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| 3 n/a Operations Superintendent contacts the control room and states that the IA MFP is to be taken out of service. Power is to be reduced at 3%/mm to 75% using AOI-39, Rapid Load Reduction.
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| ROLE PLA Y: When the SRO contacts Shift Manager, the Console Operator will repeat back the request to evaluate conditions using EPIP-I, Emergency Plan Classification Flowchart.
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| ROLE PLAY: When the SRO contacts Load Coordinator, the Console Operator will repeat back the information provided.
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| ROLE PLAY: When contacted as Chemistry, repeat back request perform power change samples to be performed.
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| Page3of5
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| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 5 Simulator Console Operators Instructions Exam Simulator vent Event DescriptionlRole Play No. No.
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| 4 4 At 95% power, 1A Main Feedwater pump trips due to low oil pressure.
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| ROLE PLAY: When contacted as the Turbine Building AUO, repeat back request to investigate the cause of the IA MFP trip. Report that the oil leak caused the trip and that Fire Ops is on the scene containing the oil. Report that there is no fire.
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| ROLE PLAY: When the BOP contacts the AUO to perform this step(LOCALLY MAINTAIN oil temp between 110 to 130°F on running Standby MFP using 1-THV-24-94&), the Console Operator will repeat back the request.
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| ROLE PLAY: When contacted as Chemistry, repeat back request perform power change samples to be performed.
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| ROLE PLAY: When the SRO contacts Work Control, the Console Operator will repeat back request to prepare a troubleshooting and repair package for the IA Main Feedwater Pump.
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| 5 n/a 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1-FCV-2-210, MFPT A CONDENSER CNDS IN LET fail to close automatically.
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| ROLE PLAY: None 6 n/a IA-A MD AFW pump shaft failure.
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| ROLE PLAY: When dispatched as an AUO, report that the IA-A MD AFW pump shaft appears to be broken.
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| nla TDAFW pump becomes steam bound.
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| ROLE PLAY: When dispatched as an AUO to the TD AFW pump, state that the discharge piping and pump casing are very hot. If requested to vent the pump, report back that the vent valve is stuck and that assistance from Maintenance has been requested.
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| n/a 1 B-B 6.9 KV Shutdown Board Trip.
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| ROLE PLAY: When dispatched as an AUO to the lB-B 6.9 KVShutdown Board, state that the board tripped on differential relay operation. State that the board is damaged severely, and that there is NO fire.
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| Page 4 of 5
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| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 5 Simulator Console Operators Instructions exam Simulator Event Event DescriptionlRole Play No. No.
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| n/a Block of AUTO SI. (From FR-H.lStep 11 .b, contained in attachment 3 of the scenario guide)
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| . ROLE PLAY: If/when notified as Work Control OR Instrument Maintenance to block auto SI using IMI-99.040, AUTO SI Block repeat back the request. Wait 10 minutes then insert remote functions rprl8, block train a auto si and rprl9, block train b auto si. Report back that IMI-99.040 is complete.
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| Page 5 of 5
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| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario # 5 Event # 1 Page 1 of 29 Event
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| == Description:==
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| 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal.
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| Requires a Tech Spec evaluation.
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| E Time Position Applicants Actions or Behavior Indications:
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| 92-A PZR LEVEL LO 92-C PZR LEVEL LO HTRS OFF & LTDN CLOSED Diagnoses and announces the failure of the I -LT-68-339. PZR RO LEVEL.
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| Takes the IMMEDIATE ACTION of AOl-20, Malfunction of RO Pressurizer Level Control System.
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| Enters and directs actions of AOI-20, Malfunction of Pressurizer SRO Level Control System.
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| The following actions are taken from AOl-20, Malfunction of AOI-20 Pressurizer Level Control System.
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| NOTE Step 1 is an IMMEDIATE ACTION step.
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| RO 1. CHECK charging in service.
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| : 2. CHECK pzr level program signal NORMAL (green pen):
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| * 1 -LR-68-339 RO RO determines that the GREEN indicating pen is indicating normal full power PZR level (60%).
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| : 3. ENSURE 1-XS-68-339E selected to OPERABLE channels for control and backup:
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| * LT-68-339 & 335 OR RO
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| * LT-68-335 & 320 OR
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| * LT-68-339 & 320 SRO directs the RO to position 1-XS-68-339E to select L T-68-335&320.
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| : 4. CHECK letdown in SERVICE
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| * 1-FCV-62-69 OPEN
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| * 1-FCV-62-70 OPEN
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| * 1-FCV-62-77 OPEN RO
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| * Letdown orifice OPEN
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| * Letdown flow NORMAL RO determines letdown was isolated by the failure of 1-LT 339 low.
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| 2011-10 Watts Bar NRC Examination
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| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario# 5 Event# 1 Page 2 of 29 Event
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| == Description:==
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| 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal.
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| Requires a Tech Spec evaluation.
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| Time Position Applicants Actions or Behavior
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| : 4. RESPONSE NOT OBTAINED:
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| ENSURE pzr heater banks D and C ON.
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| RO ESTABLISH charging and letdown:
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| * REFER TO Attachment 1 EXAMINER: The following actions are taken from AOl-20, Malfunction of Pressurizer Level Control System, Attachment 1, ALIGNMENT OF CHARGING AND LETDOWN.
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| NOTE This section can be used in conjunction with SOI-62.O1 for local control of 1-FCV-62-89, 1-PCV-62-81 or 1-FCV-62-93. Substitute local actions as appropriate for Steps 1.IA.1, 1.IA.2, 1.IA.6, 1.1D, 1.IG and lIE.
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| 1.1 Establish Charging and Letdown.
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| A. IF charging NOT established, THEN PERFORM the following:
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| RO Since charging is in service, the SRO continues to the next step.
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| B. ENSURE letdown isol valves OPEN:
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| : 1. 1-FCV-62-69, CVCS LETDOWN ISOLATION.
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| : 2. 1-FCV-62-70, CVCS LETDOWN ISOLATION RO RO rotates 1-FCV-62-70 CVCS LETDOWN ISOLATION to the right to the OPEN position, and holds the handswitch OPEN until the RED indicating light is LIT and the GREEN indicating light is DARK. V
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| : 3. 1-FCV-62-77, CVCS LP LETDOWN ISOLATION.
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| C. PLACE 1-HIC-62-78A, LETDOWN HX OUTLET TEMP 1-TCV V 192 CNTL, in MANUAL at 25% OPEN.
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| RO 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL is placed in MANUAL by lifting the toggle switch from the AUTO position, the toggle is pushed to the right to open the valve to 75% (as read on the controller.)
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| D. PLACE 1- HIC-62-81A, LETDOWN PRESS CONTROL, in MANUAL at 40-50% OPEN if using 75 gpm orifice (20-30% OPEN if using 45 gpm orifice)
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| RO 1-HIC-62-81A, LETDOWN PRESS CONTROL is placed in MANUAL by lifting the toggle switch from the AUTO position, the toggle is pushed to the right to open the valve to 50-60% (as read on the controller).
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| 2011-10 Watts Bar NRC Examination
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| Appendix D Required Operator Actions Form ES-D-2j Op Test No.: 2 Scenario # 5 Event # Page 3 of 29 Event
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| | |
| == Description:==
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| 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal.
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| Requires a Tech Spec evaluation.
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| Time Positionj Applicants Actions or Behavior E. ESTABLISH 75 gpm or greater charging flow while maintaining seal injection flow between 8 and 13 gpm for each RCP using 1-FCV-62-93 and 1 -FCV-62-89.
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| RO RO establishes 75 gpm charging flow on 1-FI-62-93A, and 8 to 13 gpm on 1-Fl-62-IA, RCP I SEAL SUP FLOW, 1-Fl-62-14A,,
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| RCP 2 SEAL SUP FLOW, 1-Fl-62-27A, RCP 3 SEAL SUP FLOW, and 1-Fl-62-40A, RCP 4 SEAL SUP FLOW.
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| F. OPEN letdown orifices as needed:
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| * 1-FCV-62-72 (45 gpm) 1 -FCV-62-73 (75 gpm).
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| * 1-FCV-62-74 (75 gpm).
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| RO
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| * 1-FCV-62-76 (5 gpm).
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| RO selects either 1-HS-62-73A. LETDOWN ORIFICE B 75 GPM -
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| ClV-qA, or 1-FCV-62-74, LETDOWN ORIFICE C 75 GPM CIV- - SA 9
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| and rotates the selected handswitch to the right to the OPEN position, and holds the handswitch OPEN until the RED indicating light is LIT and the GREEN indicating light is DARK.
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| G. ADJUST 1..HIC-.62-81A, LETDOWN PRESS CONTROL, for desired press, (320 psig at normal letdown temp), and PLACE in AUTO.
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| RO 1-HIC-62-81A, LETDOWN PRESS CONTROL toggle switch is moved to the right to close the valve and raise pressure to 320 psig. 1-HIC-62-81A is placed in AUTO by pushing the toggle switch down to the AUTO position.
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| H. PLACE 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL, in AUTO.
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| RO 1-HIC-62-78A, LETDOWN HX OUTLET TEMP TCV-70-192 CNTL, is placed in A UTO by pushing the toggle switch down to the AUTO position.
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| I. RETURN pzr level to program.
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| RO makes periodic adjustments to 1-HIC-62-93A, RO CHARGING FLOW PZR LEVEL CONTROL and 1-HIC 89A, CHG HDR RCP SEALS FLOW CONTROL to return PZR level to program leveL J. WHEN ready to return 1-FCV-62-93 to AUTO control, THEN RO PERFORM Section 1.2 1.2 Establish AUTO Control of 1-FCV-62-93 A. RETURN 1-FCV-62-93 to AUTO.
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| RO 1-HIC-62-93A, CHARGING FLOW PZR LEVEL CONTROL, is placed in AUTO once PZR level is on program if required.
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| 2011-10 Watts Bar NRC Examination
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| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 1 Page 4 of 29 Event
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| == Description:==
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| 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal.
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| Requires a Tech Spec evaluation.
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| Time 1 Position Applicants Actions or Behavior EXAMINER: After completing Attachment 1, ALIGNMENT OF CHARGING AND LETDOWN, the applicants will return to AOI-20, Subsection 3.2, at Step 5.
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| : 5. ENSURE pzr level returning to program:
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| RO
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| * CONTROL 1-HIC-62-93A in MAN as necessary
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| * MAINTAIN regen hx letdown temp < 380 F.
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| : 6. CHECK pzr heaters ENERGIZED:
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| : a. Control Heaters D red light LIT.
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| RO observes handswitch 1-HS-68-341F, CONTROL HEATERS D RED indicating light LIT.
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| RO b. Backup Heaters C red light LIT.
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| RO places handswitch 1-HS-68-341H, BACKUP HEATERS C in the OFF position, then may place the handswitch to the ON position for boron concentration control. Handswitch 1-HS 341H is then returned to P-AUTO position.
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| : 7. ENSURE operable channel selected for recording with 1-XS RO 339B [1-M-5j:
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| RO observes that 1-XS-68-339E is selected LT-68-335&320.
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| : 8. CHECK 1-HIC-62-93A in AUTO.
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| RO RO checks that 1-HIC-62-93A, CHARGING FLOW PZR LEVEL CONTROL, was placed in AUTO as a result of performing Attachment 1, ALIGNMENT OF CHARGING AND LETDOWN.
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| : 9. REFER TO the following Tech Specs:
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| * 3.3.1, Reactor Trip System (RTS) Instrumentation.
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| Function 9 Pressurizer Water Level High, Condition X, With one channel inoperable, place the failed channel in trip within 72 hours OR reduce THERMAL POWER to <P-7 within 78 hours.
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| 3.3.3, Post Accident Monitoring (PAM) Instrumentation.
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| RO Function 13, RCS Pressurizer Level, Condition A, With one or more Functions with one required channel inoperable, restore required channel to OPERABLE status within 30 days.
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| * 3.4.9 Pressurizer not applicable
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| * 3.5.2, Emergency Core Cooling Systems (ECCS) not applicable TR 3.1.4 Charging Pump, Operating not applicable
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| : 10. NOTIFY Work Control to remove any failed channel from service.
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| SRO When the SRO contacts Work Contro4 the Console Operator will repeat back request to remove the failed channel from service.
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| 2011-10 Watts Bar NRC Examination
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| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 1 Page 5 of 29 Event
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| | |
| == Description:==
| |
| 1-LT-68-339, PZR LEVEL fails low. Requires entry into AOl-20, Malfunction of Pressurizer Level Control System. Requires return of letdown flowpath to normal.
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| Requires a Tech Spec evaluation.
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| Time I Position Applicants Actions or Behavior 1 1. INITIATE repairs to failed instrument/circuitry.
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| SRO When the SRO contacts Work Control, the Console Operator will repeat back request for a troubleshooting and repair package for 1-LT-68-339, PZR LEVEL.
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| SRO 12. RETURN TO instruction in effect.
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| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
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| Crew Brief would typically be conducted for this event as time SRO allows prior to the next event.
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| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
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| SRO Operations Manacement Typically Shift Manager.
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| Maintenance Personnel Typically Work Control Center (WCC).
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| (Note: Maintenance notification may be delegated to the Shift Manager).
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| Cue Console Operator to insert Event 2, if not previously inserted.
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| 2011-10 Watts Bar NRC Examination
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| | |
| rAppendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 2 Page 6 of 29 Event
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| | |
| == Description:==
| |
| T-avg control signal fails low. Requires the RD to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Rod Bank Movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
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| Time Position Applicants Actions or Behavior Indications:
| |
| 94-A AUCT TAVG-TREF DEVIATION 94-B LOOP TAVG & AUCT TAVG DEVN Diagnoses and announces the failure of the Auctioneered Tavg RO signal.
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| Takes the IMMEDIATE ACTION of AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Bank Movement, RO by placing 1-RBSS ROD BANK SELECT in MAN and determining that rod motion has stopped.
| |
| Enters and directs actions of AOl-2, Malfunction of Reactor Control SRO System, Section 3.2, Uncontrolled Rod Bank Movement.
| |
| The following actions are taken from AOI-2, Malfunction of AOI-2 Reactor Control System, Section 3.2, Uncontrolled Rod Bank Movement.
| |
| NOTE Step 1 is an IMMEDIATE ACTION step.
| |
| : 1. STOP uncontrolled rod motion:
| |
| : a. PLACE control rods in MAN.
| |
| RO placed 1-RBSS ROD BANK SELECT in MAN during RO performance of IMMEDIATE ACTION.
| |
| : b. CHECK control rod movement STOPPED.
| |
| RO observed rod motion stopped when 1-RBSS ROD BANK SELECT was placed in MAN.
| |
| : 2. MAINTAIN T-ave on PROGRAM. (Reference Attachment 1)
| |
| * (p) USE control rods.
| |
| RO RO evaluates Tavg and may adjust control rod position in MAN.
| |
| OR (p) ADJUST turbine load.
| |
| EXAMINER: AOI-2, Malfunction of Reactor Control System, Attachment 1, Reactor Power VS T-avglT-ref Temperature and PZR Level, is included as Attachment I to this scenario.
| |
| : 3. CHECK loop T-ave channels NORMAL.
| |
| RD RO observes Tavg normal on 1-Tl-68-2E, LOOP I TAVO, I-TI 25E, LOOP 2 TAVG, I-TI-68-44E LOOP 3 TAVG, and I-TI-68-67E, LOOP 4 TAVG.
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| 2011-10 Watts Bar NRC Examination
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| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario# 5 Event# 2 Page 7 of 29 Event
| |
| | |
| == Description:==
| |
| T-avg control signal fails low. Requires the RO to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Rod Bank Movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
| |
| Time Position Applicants Actions or Behavior
| |
| : 4. CHECK Auct T-avg NORMAL on 1-TR-68-2B.
| |
| RD RO observes that the GREEN pen (Tavg) is off-scale high on 1-TR-68-2B, TREF & AUCT TAVG - F.
| |
| : 4. RESPONSE NOT OBTAINED:
| |
| CONTROL PZR level in MAN with 1-FCV-62-93 and 1-FCV 89. (Reference Attachment 1)
| |
| RD RO places 1-HIC-62-93A CHARGING FLOW PZR LEVEL CONTROL in MAN. Since the PZR level program is capped at 60%, no adjustment is expected.
| |
| : 5. CHECK NIS power range channels NORMAL.
| |
| RD RO observes 1-NI-41B, PR FLUX % POWER, 1-Nl-42B, PR FLUX
| |
| % POWER, 1-Nl-43B, PR FLUX % POWER, 1-NI-44B, PR FLUX %
| |
| POWER are indicating normal (approximately 100% power.)
| |
| : 6. CHECK the following:
| |
| * Turbine impulse pressure channel 1-Pl-1-73, NORMAL.
| |
| RO observes I -P1-1-73, % HP TURBINE POWER TR A, and after comparing the output to 1-Pl-1-72, % HP TURBINE POWER TR B RD determines that it is NORMAL.
| |
| * T-ref and Auct T-avg NORMAL on 1-TR-68-2B (Reference Attachment 1).
| |
| RO observes that the GREEN pen (Tavg) is off-scale high on 1-TR-68-2B, TREF & AUCT TAVG - F.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario# 5 Event# 2 Page 8 of 29 Event
| |
| | |
| == Description:==
| |
| T-avg control signal fails low. Requires the RO to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Rod Bank Movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
| |
| Time Position N Applicants Actions or Behavior
| |
| : 6. RESPONSE NOT OBTAINED:
| |
| PLACE steam dumps in pressure mode as follows:
| |
| : a. PLACE steam dumps to OFF.
| |
| BOP rotates 1-HS-1-103A, STEAM DUMP FSV A to the left to the OFF RESET position. BOP rotates 1-HS-1-103B, STEAM DUMP FSV B to the left to the OFF RESET position.
| |
| : b. PLACE mode selector HS to STEAM PRESS.
| |
| BOP rotates 1-HS-1-103D, STEAM DUMP MODE to the right to the STEAM PRESS position.
| |
| : c. ADJUST steam dump demand to zero.
| |
| BOP BOP observes 1-Xl-1-33, STEAM DUMP DEMAND dropping to zero.
| |
| : d. PLACE steam dumps to ON.
| |
| BOP rotates 1-HS-1-103A, STEAM DUMP FSV A to the right to the ON position. BOP rotates 1-HS-1-103B, STEAM DUMP FSV B to the right to the ON position.
| |
| : e. ENSURE controller set at 84% (1092 psig).
| |
| BOP observes that 1-PIC-1-33, STM DUMP PRESS CONTROL dial is set to 84%.
| |
| : f. WHEN conditions allow, THEN REFER TO SQl-I .02 and PLACE steam dumps in TAVG Mode.
| |
| SRO maintains this step open until repairs are made.
| |
| : 7. MONITOR core power distribution parameters:
| |
| * Power range channels
| |
| * A Flux Indicators SRO *T-ave
| |
| * Loop AT
| |
| * Incore TCs
| |
| * Feed flow/Steam flow
| |
| : 8. INITIATE repairs to failed equipment.
| |
| SRO When the SRO contacts Work Contro4 the Console Operator will repeat back request to prepare a troubleshooting and repair package for the Tavg auctioneering circuit.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 2 Page 9 of 29 Event
| |
| | |
| == Description:==
| |
| T-avg control signal fails low. Requires the RO to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOI-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Rod Bank Movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
| |
| Time Position Applicants Actions or Behavior EXAMINER: Since the auctioneering circuit effects control functions only, Tech Specs are not affected.
| |
| : 9. REFER TO Tech Specs:
| |
| * 3.1.1, Shutdown Margin Not applicable
| |
| * 3.1 .5, Rod Group Alignment Limits Not applicable
| |
| * 3.1.6, Shutdown Bank Insertion Limits Not applicable
| |
| * 3.1.7, Control Bank Insertion Limits Not applicable
| |
| * 3.2.1, Heat Flux Hot Channel Factor Not applicable
| |
| * 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor Not SRO -
| |
| applicable
| |
| * 3.2.4, Quadrant Power Tilt Ratio Not applicable
| |
| * 3.2.3, Axial Flux Difference Not applicable
| |
| * 3.3.1, Reactor Trip System (RTS) Instrumentation Not -
| |
| applicable
| |
| * 3.3.2, Engineered Safety Features Actuation System (ESFAS)
| |
| Instrumentation Not applicable
| |
| : 10. NOTIFY Chemistry of any reactor power changes greater than SRO 15% in one hour.
| |
| 1 1. WHEN ready to restore repaired T and loop T-avg channels, THEN:
| |
| SRO SRO determines that the conditions of the step cannot be made and continues to the next step.
| |
| CAUTION Allowing at least 5 minutes between any rod control input change (i.e., T-ave, T-ref, or NIS) and placing rods in AUTO, will help prevent undesired control rod movement.
| |
| : 12. WHEN auto rod control desired, THEN:
| |
| SRO SRO determines that the conditions of the step cannot be made and continues to the next step.
| |
| : 13. WHEN conditions allow auto PZR level control, THEN SRO SRO determines that the conditions of the step cannot be made and continues to the next step.
| |
| SRO 14. RETURN TO Instruction in effect.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario# 5 Event# 2 Page 10 of 29 Event
| |
| | |
| == Description:==
| |
| T-avg control signal fails low. Requires the RD to take IMMEDIATE OPERATOR ACTION to place rod control to MANUAL to stop rod motion. Requires entry into AOl-2, Malfunction of Reactor Control System, Section 3.2, Uncontrolled Rod Rod Bank Movement. Requires manual control of 1-FCV-62-93, CHARGING FLOW PZR LEVEL CONTROL.
| |
| Time I Position Applicants Actions or Behavior Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| Cue Console Operator to contact the crew as the Operations Superintendent and direct a plant shutdown using AOl-39, Rapid Load Reduction to 75% at 3%Imin.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| [Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 3 Page 11 of 29 Event
| |
| | |
| == Description:==
| |
| Operations Superintendent contacts the control room and states that the IA MFP is to be taken out of service. Power is to be reduced at 3%/mm to 75% using AOl-39, Rapid Load Reduction, Section 3.3, Reactor power is greater than 50%.
| |
| Time Position Applicants Actions or Behavior The following actions are taken from AOl-39, Rapid Load AOI-39 Reduction, Section 3.3, Reactor power is greater than 50%.
| |
| CAUTION Over boration may result in excessive rod withdrawal, T-avg lower than desired, and AFD oscillations.
| |
| NOTE
| |
| * Rod Control should remain in automatic for T-avg Control
| |
| * Reactivity Briefing Sheet, Thumb Rules (page 3), lists boration flows and volumes for different reduction rates.
| |
| * Effect of boration will lag behind turbine load reduction and can be compensated for by temporarily increang boric acid flow rate above recommended rate.
| |
| : 1. INITIATE a manual boration:
| |
| : a. DETERMINE recommended boration flow rate and volume from Reactivity Briefing Sheet:
| |
| RO determines the recommended boration flow rate to be 40 gpm (for 3%/Mm or greater) and the volume to add to be 341 gallons of boron.
| |
| : b. INITIATE normal boration:
| |
| : 1) ADJUST BA flow controller, I -FC-62-1 39, to desired flow rate.
| |
| RO adjusts 1-FC-62-139 to 100% which corresponds to 40 gpm.
| |
| : 2) ADJUST BA batch counter 1 -FQ-62-1 39 to required quantity.
| |
| RO adjusts 1-FQ-62-139 BA BATCH COUNTER as follows:
| |
| : 1. Depresses and holds the black pushbutton.
| |
| RO 2. While holding the pushbutton, the applicant raises the red translucent cover.
| |
| : 3. While still holding the pushbutton, the applicant enters 000341 in the display.
| |
| : 4. While still holding the pushbutton, the applicant lowers the red translucent cover, and then releases the pushbutton.
| |
| : 3) PLACE mode selector 1-HS-62-140B to BOR.
| |
| RO rotates 1-HS-62-140B, VCT MAKEUP MODE from AUTO to the right to BOR position.
| |
| : 4) (p) PLACE VCT makeup control 1-HS-62-140A, to START.
| |
| RO rotates 1-HS-62-140A, VCT MAKEUP CONTROL to the right to the START position.
| |
| : 5) VERIFY desired boric acid flow indicated on 1-Fl-62-139.
| |
| RO observes approximately 40 gpm flow on 1-FI-62-139, BA TO BLENDER FLOW.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 3 Page 12 of 29 Event
| |
| | |
| == Description:==
| |
| Operations Superintendent contacts the control room and states that the IA MFP is to be taken out of service. Power is to be reduced at 3%/mm to 75% using AOl-39, Rapid Load Reduction, Section 3.3, Reactor power is greater than 50%.
| |
| L Time Position Applicants Actions or Behavior CAUTION
| |
| * Condenser Backpressure limits are on page 5.
| |
| * TURBINE MANUAL Operation requires continuous operator monitoring and control.
| |
| * LOSS OF CONDENSER VACUUM may be made worse if steam dumps are actuated. AOl 11 requires T-ave and T-ref be maintained within 3°F.
| |
| NOTE If the initiating condition is corrected, the power reduction may be terminated
| |
| : 2. ESTABLISH a turbine load reduction, to less than or equal to 76%
| |
| power, at a rate greater than or equal to 2%/rn in, and less than 5%/mm:
| |
| : a. PLACE turbine in IMP IN.
| |
| BOP depresses the IMP IN pushbutton and observes the transfer from IMP OUT to IMP IN control complete.
| |
| : b. SET a desired load in the SETTER with the REFERENCE CONTROL.
| |
| BOP depresses the reference control V (down) button to BOP reduce the setter display to less than 75% load as directed by the SRO.
| |
| : c. SET the LOAD RATE at greater than or equal to 2%/rn in, and less than 5%/mm.
| |
| BOP selects the load reduction rate directed by the SRO using the LOAD RATE % PER MIN thumbwheel selector. It is expected that the load reduction rate of 3%/minute, specified by the Operations Superintendent will be used.
| |
| : d. (p) DEPRESS GO pushbutton.
| |
| BOP depresses the REFERENCE CONTROL GO button.
| |
| NOTE AFD green target band can be monitored using ICS Turn On code DOGHOUSE.
| |
| : 3. MONITOR rod position:
| |
| RO
| |
| * Rods above Lo-Lo insertion limit
| |
| * AFD within Target Band
| |
| : 4. REFER TO EPIP-1, Emergency Plan Classification Flowchart SRO When the SRO contacts Shift Manager, the Console Operator will repeat back the request to evaluate conditions using EPIP 1, Emergency Plan Classification Flowchart.
| |
| : 5. NOTIFY the Load Coordinator of the required load reduction and expected ramp rate.
| |
| SRO When the SRO contacts Load Coordinator, the Console Operator will repeat back the information provided.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario # 5 Event # 3 Page 13 of 29 Event
| |
| | |
| == Description:==
| |
| Operations Superintendent contacts the control room and states that the 1A MFP is to be taken out of service. Power is to be reduced at 3%/mm to 75% using AOl-39, Rapid Load Reduction, Section 3.3, Reactor power is greater than 50%.
| |
| Time Position Applicants Actions or Behavior
| |
| : 6. MONITOR T-avg and T-ref:
| |
| RO
| |
| * T-ave trending to T-ref.
| |
| * Mismatch less than 5°F.
| |
| : 7. CHECK rate of power reduction is rapid enough for existing plant SRO conditions.
| |
| BOP 8. NOTIFY Cnds Demin AUO of impending pmp shutdowns.
| |
| : 9. WHEN rated thermal power change exceeds 15% in one hour, THEN NOTIFY Chemistry to initiate 1-SI-68-28.
| |
| SRO When the SRO contacts Chemistry, the Console Operator will repeat back the request.
| |
| EXAMINER: When load has been reduced to 95%, the IA MFW pump will trip (Event 4).
| |
| When the IA MFW pump trips, the MFWT condenser isolation valves fail to close automatically (Event 5). The SRO will implement AOl-I 6, Loss of Normal Feedwater, Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67%
| |
| Turbine Load).
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 4 and 5 Page 14 of 29 Event
| |
| | |
| == Description:==
| |
| : 4) At 95% power, IA Main Feedwater pump trips due to low oil pressure. Requires entry into Aol-i 6, Loss of Normal Feedwater, Sub Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| Requires a Tech Spec evaluation of Axial Flux Difference. Requires a boration to return AFD to target band. 5) 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and i-FCV-2-210, MFPT A CONDENSER CNDS INLET fail to close automatically. Requires the BOP to manually close valves.
| |
| TimeD Position N Applicants Actions or Behavior Indications:
| |
| 50-A MFPT 1A ABNORMAL 51-A MFPT 1ATRIPPED BOP Diagnoses and announces the trip of the 1A Main Feedwater pump.
| |
| May determines MFWP turbine condenser valves did not BOP automatically close and closes them manually Enters and directs actions of AOl-i 6, Loss of Normal Feedwater, SRO Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| The following actions are taken from AOl-I 6, Loss of Normal AOI-16 Feedwater, Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| : 1. (p) IF loss of S/G level is imminent, THEN TRIP reactor, and **
| |
| GO TO E-0, Reactor Trip or Safety Injection.
| |
| BOP BOP determines from SG level trends that loss of SG level is NOT imminent.
| |
| : 2. ENSURE failed MFP TRIPPED.
| |
| BOP BOP determines that 1-HS-46-9A, MFPTA TRIP RESET, RED -
| |
| indicating light is DARK and GREEN indicating light is LIT.
| |
| EXAMINER: If contacted as the Turbine Building AUO, console operator will repeat back request to investigate the cause of the IA MFP trip. Report that the oil leak caused the trip and that Fire Ops is on the scene containing the oil.
| |
| Report that there is no fire.
| |
| : 3. CHECK turbine load less than or equal to 1000 MWe (85%).
| |
| BOP BOP observes 1-EI-57-16A, MEGAWATTS or the MEGAWATT meter on 1-XX-47-2000 EHC DISPLAY less than 1000 MWe.
| |
| : 4. PLACE tripped MFP recirc valve controller in MANUAL, and CLOSE recirc valve.
| |
| BOP BOP places l-FIC-3-70, MFWP A RECIRC CONTROL, in MANUAL and presses and holds the close pushbutton until the red arrow indicates 100% (as read on the controller).
| |
| : 5. CHECK turbine load less than 800 MWe (67%).
| |
| BOP BOP observes 1-El-57-16A, MEGAWATTS or the MEGAWATT meter on 1-XX-47-2000 EHC DISPLAY greater than 800 MWe.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 4 and 5 Page 15 of 29 Event
| |
| | |
| == Description:==
| |
| : 4) At 95% power, IA Main Feedwater pump trips due to low oil pressure. Requires entry into AOl-16, Loss of Normal Feedwater, Sub Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| Requires a Tech Spec evaluation of Axial Flux Difference. Requires a boration to return AFD to target band. 5) 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1.FCV-2-210, MFPT A CONDENSER CNDS INLET fail to close automatically. Requires the BOP to manually close valves.
| |
| Time Position Applicants Actions or Behavior
| |
| : 5. RESPONSE NOT OBTAINED:
| |
| ENSURE Standby MFWP running. (p) IF Standby MFWP NOT BOP available, THEN REDUCE turbine load to less than 800 MWe with valve position limiter.
| |
| BOP verifies the Standby MFWP is running.
| |
| : 6. ENSURE MFWP speed rising to control S/G &P and levels on program.
| |
| BOP BOP verifies MFPT B SPEED 1-Sl-46-20B is rising or stable, and
| |
| #1 HTR INLET PRESS 1-Pl-3-1 is indicating greater than MAIN STEAM PRESS 1-PI-1-33.
| |
| CAUTION Continued load reductions below 800 MWe should be done using normal turbine controls at less than or equal to 5% mm.
| |
| NOTE Load will not change until VALVE POS LIMIT light is cleared.
| |
| : 7. ENSURE adequate feed flow for existing conditions:
| |
| * Feed flow greater than or equal to steam flow.
| |
| BOP
| |
| * SIG levels returning to program.
| |
| BOP verifies feed flow greater than steam flow for all steam generators, and all steam generator levels are rising or stable.
| |
| RO 8. MONITOR T-avg trending to within 3°F of T-ref.
| |
| : 9. MAINTAIN AFD within limits (p) INITIATE boration as required REFER TO ATTACHMENT 1, Manual Boration.
| |
| SRO Based on the power reduction, LCO 3.2.3, Axial Flux Difference (AFD) Condition A, must be entered. With AFD not within limits, reduce THERMAL POWER to < 50% RTP within 30 minutes.
| |
| EXAMINER: AOI-16, Loss of Normal Feedwater, Attachment 1, Manual Boration, is included as Attachment 2 to this scenario.
| |
| EXAMINER: The IA MFWP turbine condenser valves fail to automatically close when the IA MFW pump trips. The BOP must close valves I-FCV-2-205 and I-FCV-2-2I0 manually.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario # 5 Event # 4 and 5 Page 16 of 29 Event
| |
| | |
| == Description:==
| |
| : 4) At 95% power, IA Main Feedwater pump trips due to low oil pressure. Requires entry into AOl-i 6, Loss of Normal Feedwater, Sub Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| Requires a Tech Spec evaluation of Axial Flux Difference. Requires a boration to return AFD to target band. 5) 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1-FCV-2-210, MFPT A CONDENSER CNDS INLET fail to close automatically. Requires the BOP to manually close valves.
| |
| Time Position Applicants Actions or Behavior
| |
| : 10. ENSURE tripped MFWP turbine condenser valves CLOSED:
| |
| * Pump A, 1-FCV-2-205 and 210 OR
| |
| * Pump B, 1-FCV-2-211 and 216 BOP BOP determines MFWP turbine condenser valves did not automatically close and closes them manually by rotating MFPT A CONDENSER CNDS INLET 1-HS-2-210A and MFPT A CONDENSER CNDS OUTLET 1-HS-2-205A switches to the left and verifying the GREEN indicating lights are LIT and the RED indicating lights are DARK.
| |
| BOP 11. MONITOR reg valves controlling S/C levels on program.
| |
| : 12. LOCALLY MAINTAIN oil temp between 110 to 130°F on running Standby MFP using 1-THV-24-948.
| |
| BOP When the BOP contacts the AUO to perform this step, the Console Operator will repeat back the request.
| |
| EXAMINER: Since the steam dump controls are in STEAM PRESSURE mode at this time, the SRO may elect to:
| |
| 1.) Not reset C-7 interlock OR 2.) Direct the BOP to place the steam dump controls in OFF prior to resetting C-7.
| |
| : 13. ENSURE reset of 0-7:
| |
| : a. CHECK C-7 LOSS OF LOAD STM DUMP INTERLOCK annunciator LIT. [66-E]
| |
| BOP will verIfy C-7 LOSS OF LOAD STM DUMP INTERLOCK (66-E) is LIT.
| |
| BOP b. ENSURE steam dump valves have zero demand.
| |
| BOP will verify STEAM DUMP DEMAND 1-XI-1-33 indicates zero.
| |
| : c. RESET loss-of-load interlock with steam dump mode switch.
| |
| BOP will rotate STEAM DUMP MODE switch 1-HS-1-103D to the left and verify C-7 LOSS OF LOAD STM DUMP INTERLOCK (66-E) is DARK.
| |
| : 14. ENSURE Condensate System Pumps in service as necessary:
| |
| * REFER TO GO-4, Normal Power Operation.
| |
| BOP BOP verifies all condensate pumps are running and MFW PMPS SUCT PRESS adequate.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D2 I OpTestNo.: 2 Scenario # 5 Event# 4 and 5 Page 17 of 29 Event
| |
| | |
| == Description:==
| |
| : 4) At 95% power, IA Main Feedwater pump trips due to low oil pressure. Requires entry into Aol-I 6, Loss of Normal Feedwater, Sub Section 3.4, TDMFWP Trip OR Loss of Flow GREATER than or equal to 800 MWe (67% Turbine Load).
| |
| Requires a Tech Spec evaluation of Axial Flux Difference. Requires a boration to return AFD to target band. 5) 1-FCV-2-205, MFPT A CONDENSER CNDS OUTLET and 1-FCV-2-210, MFPT A CONDENSER CNDS INLET fail to close automatically. Requires the BOP to manually close valves.
| |
| Time Position Applicants Actions or Behavior
| |
| : 15. IF reactor power dropped by greater than or equal to 15% in one hour, THEN NOTIFY Chemistry to initiate power change SRO sampling requirements.
| |
| When SRO contacts Chemistry to initiate sampling, console operator will acknowledge the request.
| |
| BOP 16. CHECK VALVE POS LIMIT LIT.
| |
| : 17. RETURN valve position limiter to normal:
| |
| : a. ENSURE TURBINE in IMP OUT.
| |
| SRO b. (p) REDUCE turbine load setpoint using REFERENCE CONTROL V (lower) AND GO button until VALVE POS LIMIT LIGHT not LIT.
| |
| : c. SET valve position limiter to 95%.
| |
| : 18. INITIATE repairs on failed pump.
| |
| SRO When the SRO contacts Work Control, the Console Operator will repeat back request to prepare a troubleshooting and repair package for the IA Main Feedwater Pump.
| |
| SRO 19. RETURN TO Instruction in effect.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO -
| |
| Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| Cue Console Operator to enter Event 6.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 18 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 Level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time Position Applicants Actions or Behavior Indications:
| |
| 62-C SG 3 STM-EW FLOW MISMATCH 63-F SG 3 LEVEL LO Diagnoses and announces the failure of 1 -LCV-3-90 SG 3 MEW BOP REG VLV closed.
| |
| May recommend to the SRO that a reactor trip be initiated based on BOP imminent loss of SG level.
| |
| May attempt to open 1-LCV-3-90 manually, in an attempt to recover BOP feedwater flow.
| |
| May direct the RO to manually trip the reactor based on loss of SG SRO level.
| |
| Enters and directs the actions of E-O, Reactor Trip or Safety Injection.
| |
| The following actions are taken from E-O, Reactor Trip or Safety Injection.
| |
| NOTE
| |
| * Steps 1 thru 4 are IMMEDIATE ACTION STEPS.
| |
| e Status Trees I SPDS should be monitored when transitioned to another instruction.
| |
| EXAMINER: The next step (E-O Step 1) assumes the reactor is manually tripped prior to entering E-O completing Critical Task 1. If Reactor has not been manually tripped, Critical Task I will be accomplished in E-O step I Response Not Obtained column. as indicated below.
| |
| 2011-10 Watts Bar NRC Examination
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| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 19 of 29 Event
| |
| | |
| == Description:==
| |
| 1 -LCV-3-90, SG 3-MEW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. I B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. ER-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 1. ENSURE reactor trip:
| |
| * Reactor trip and bypass breakers OPEN.
| |
| RO checks 1-52RTB, R)( TRIP BKR A GREEN indicating light LIT on panel 1-M-4.
| |
| RO checks 1-52RTB, RC TRIP BKR B GREEN indicating light LIT on panel 1-M-4.
| |
| RO checks 1-52BYA, BYPASS BKR A lights DARK.
| |
| RO checks 1-52BYB, BYPASS BKR B lights DARK.
| |
| * RPIs at bottom of scale.
| |
| RO RO normally observes 1-MON 85 5000/1 CERPI Monitor I and 1-MON-85 5000/2 CERPI MONITOR 2 for indication that all SHUTDOWN and CONTROL bank rods are inserted.
| |
| Note: CERPI Monitors are not working due to power loss.
| |
| * Neutron flux DROPPING.
| |
| RO observes neutron flux trending down on 1-NR-92-145, NEUTRON FLUX LEVEL RECORDER. May also observe levels decreasing on 1-Nl-92-135A, CH I NEUTRON MON % PWR, and 1-NI-92-136A, CH II NEUTRON MON % PWR.
| |
| Critical Task I Manually trip the reactor from the control room upon recognition of the failure of the automatic trip circuit.
| |
| Critical 1. RESPONSE NOT OBTAINED:
| |
| Task I Manually TRIP reactor.
| |
| RO rotates 1-RT-1, REACTOR TRIP to the right to the TRIP position. Alternatively, the RO may elect to rotate 1-RT-2, RO REACTOR TRIP to the right to the TRIP position.
| |
| IF reactor will NOT trip, THEN GO TO FR-Si, Nuclear Power Generation I ATWS.
| |
| : 2. ENSURE Turbine Trip:
| |
| * All turbine stop valves CLOSED.
| |
| RO RO observes that indicating lights on 1-XX-47-1000 EHC CONTROL for individual throttle and governor valves are GREEN.
| |
| 2011-10 Watts Bar NRC Examination
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| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 20 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MEW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-HI, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 3. CHECK 6.9 kV shutdown boards:
| |
| : a. At least one board energized from:
| |
| CSST (offsite),
| |
| OR DIG (blackout).
| |
| RO At the trip, the lB-B 6.9 KV Shutdown Board suffers a differential relay operation causing the board to be lost.
| |
| When dispatched as an AUO to the lB-B 6.9 KV Shutdown Board, console operator will state that the board tripped on differential relay operation. State that the board is damaged severely, and that there is NO fire.
| |
| : 4. CHECK SI actuated:
| |
| : a. Any SI annunciator LIT.
| |
| RO RO checks window 70-A, SI ACTUATED; 76-G SI MANUAL; 77-G IS PZR PRESS LO; SI CNTMT PRESS HI; and SI STM PRESS LO DARK.
| |
| : 4. RESPONSE NOT OBTAINED:
| |
| DETERMINE if SI required:
| |
| : a. IF ANY of the following exists:
| |
| * S/G press less than 675 psig, OR RO observes SG pressures greater than 675 psig (stable, at approximately 1092 psig.)
| |
| * RCS press less than 1870 psig, OR RO observes RCS pressure greater than 1870 psig (approximately RO 2235 psig.)
| |
| Cntmt press greater than 1 .5 psig RO observes containment pressure less than 1.5 psig (approximately 0 psig.)
| |
| THEN ACTUATE SI manually.
| |
| IF SI NOT required, THEN ** GO TO ES-0.1, Reactor Trip Response.
| |
| SRO determines that SIis NOT required and transitions to ES-0.1, Reactor Trip Response.
| |
| 2011-10 Watts Bar NRC Examination
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| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.: 2 Scenario# 5 Event# 6 Page 21 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. lB-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time N Position 0 Applicants Actions or Behavior The following actions are taken from ES-O.1, Reactor Trip ES-O.1 Response.
| |
| CAUTION Plant conditions, AFW pump start signals and flow requirements should be evaluated as time allows.
| |
| : 1. MONITOR SI actuation criteria:
| |
| RO
| |
| * IF SI actuation occurs during the performance of this Instruction, THEN ** GO TO E-O, Reactor Trip or Safety Injection.
| |
| : 2. CHECK Generator PCBs OPEN.
| |
| BOP observes that 1-HS-57-26, PCB 5044 MAIN GENERATOR 500 KV BUS red INDICATING LIGHT IS DARK and GREEN INDICATING LIGHT IS lit. BOP observes 1-HS-57-24, PCB 5088 MAIN GENERA TOR 500 KV BUS red INDICA TING LIGHT IS DARK and GREEN INDICA TING LIGHT IS lit.
| |
| : 3. MONITOR RCS temperature stable at or trending to 557°F using:
| |
| * RCS Loop T-avg with any RCP running.
| |
| OR
| |
| * RCS Loop T-cold with RCPs out-of-service.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 22 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the IA MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-HI, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time Position 1 Applicants Actions or Behavior
| |
| : 4. ENSURE AFW operation:
| |
| : a. AFW established:
| |
| * Both MD AFW pumps RUNNING.
| |
| BOP determines that IA-A MD AFW pump is running but not pumping forward. BOP dispatches an AUO to locally inspect the IA-A MD AFW pump.
| |
| When dispatched as an AUO, console operator will report that the IA-A MD AFW pump shaft appears to be broken.
| |
| lB-B MD AFW pump has no power due to the loss of lB-B 6.9 KV Shutdown board.
| |
| * TD AFW pump RUNNING.
| |
| BOP determines that the TD AFW pump is running but has no discharge flow. BOP dispatches an AUO to inspect the TD AFW pump locally.
| |
| When dispatched as an AUO to the TD AFW pump, console operator will state that the discharge piping and pump casing are very hot. If requested to vent the pump, console operator will report back that the vent valve is stuck and that assistance from Maintenance has been requested.
| |
| * LCVs in AUTO or controlled in MANUAL.
| |
| : 4. a. RESPONSE NOT OBTAINED:
| |
| RO ESTABLISH feed flow from AFW or MFW as necessary.
| |
| SRO may direct the BOP to attempt to restore MFW.
| |
| : b. Heat sink available:
| |
| RO
| |
| * Total feed flow greater than 410 gpm, OR
| |
| * At least one SIG NR level greater than 29%.
| |
| : 4. b. RESPONSE NOT OBTAINED:
| |
| RO IF heat sink can NOT be established, THEN GO TO FR-H.1, Loss Of Secondary Heat Sink.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 23 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MEW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the IA MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time Position N Applicants Actions or Behavior The following steps are taken from E-2, Faulted Steam FR-RI Generator Isolation.
| |
| CAUTION
| |
| * If total feed flow CAPABILITY of 410 gpm is available, this Instruction should NOT be performed.
| |
| * If an Intact SIG is available, feed flow should NOT be reestablished to any faulted SIG.
| |
| : 1. CHECK if secondary heat sink is required:
| |
| : a. RCS pressure greater than any Intact SIG pressure.
| |
| BOP b. RCS temperature greater than 375°F [360°F ADV].
| |
| RO determines that both of the conditions requiring a heat sink exist.
| |
| : 2. ENSURE at least one charging pump RUNNING.
| |
| BOP RO reports that NO charging pumps are running or available to be placed in service.
| |
| : 2. RESPONSE NOT OBTAINED:
| |
| BOP IF at least one charging pump NOT RUNNING, THEN STOP all RCPs AND GO TO Cautions prior to Step 18 to initiate RCS bleed and feed.
| |
| CAUTION
| |
| * Step 18 Through 20 must be performed quickly in order to establish RCS heat removal by RCS bleed and feed.
| |
| * Termination of bleed and feed is required prior to transitioning out of FR-H.1 when heat sink is restored.
| |
| Critical Task 2 Initiate RCS bleed and feed so that the RCS depressurizes sufficiently for safety injection pumps to inject into the RCS.
| |
| Critical 18. ACTUATE SI.
| |
| Task 2 RO rotates 1-HS-63-133B SIACTUATE TR A & B to the right to Begin RO the ACTUATEposition. RO observes window 70-A SI ACTUATED light is LIT. RO observes that window 76-G, SI MANUAL is LIT>
| |
| 2011-10 Wafts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I OpTestNo.: 2 Scenario # 5 Event# 6 Page 24 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-Hi, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time I Position Applicants Actions or Behavior Critical 19. ENSURE at least one of the following RCS feed paths:
| |
| Task 2
| |
| * At least one charging pump injecting thru BIT, OR RO
| |
| * At least one SI Pump running with its injection valves open.
| |
| RO determines that the IA-A SI pump is running and its discharge valve is open.
| |
| CAUTION
| |
| * When the reactor vessel head vent block valve is opened, the throttle valve will cycle open and closed.
| |
| * Slowly opening (5 seconds stroke time) the head vent valve will prevent water hammer and pipe damage.
| |
| Critical 20. ENSURE adequate RCS bleed path:
| |
| Task 2 a. ENSURE all pzr PORVs and pzr PORV block valves OPEN.
| |
| END
| |
| * RO rotates 1-HS-68-340AA, PZR PORV 340A to the right to the OPEN position. RO observes the Red indicating light LIT, GREEN indicating light DARK.
| |
| RO rotates 1-HS-68-334A, PZR PORV 334 to the right to the RO OPEN position. RO observes the Red indicating light LIT, GREEN indicating light DARK.
| |
| RO observes 1-HS-68-333A, BLOCK VLVFORPORV34OA RED indicating light LIT, GREEN indicating light DARK.
| |
| RO observes 1-HS-68-332A, BLOCK VLVFOR PORV 334 RED and GREEN indicating lights are DARK due to power loss. RO reports that the last position of 1-HS-68-332A was OPEN.
| |
| CAUTION WHEN feedwater source is AVAILABLE, THEN feed rate will be controlled by Steps 30 and 31.
| |
| NOTE The details of Steps 4 through 15 may be referred to as necessary to establish feed flow in the following step but procedure performance must continue to terminate RCS bleed and feed.
| |
| EXAMINER: The applicants are expected to refer to FR-H.1 Steps 4 through 15, to establish Main feedwater using the Standby Main Feedwater Pump. Steps 4 through 15 along with Appendix A, Establishing MFW Following Reactor Trip, are contained in Attachment 3 to this scenario.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 25 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.1, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 21. RESET SI, AND CHECK the following:
| |
| RO depresses 1-HS-63-134A, SI RESET TR A, and depresses 1-HS-63-134B SI RESET TR B.
| |
| * SI ACTUATED permissive DARK.
| |
| RO RO observes window 70-A SIACTUA TED is DARK on Panel 1-XX-55-4A.
| |
| * AUTO SI BLOCKED permissive LIT.
| |
| RO observes window 70-B AUTO SI BLOCKED is LIT on Panel 1-XX-55-4A.
| |
| : 22. RESET Containment Isolation Phase A and Phase B.
| |
| RO depresses 1-HS-30-63D, çk4 CNTMT ISOL RESET TR-A and 1-HS-30-63E, qfA CNTMT ISOL RESET TR-B and observes RED i/iA lights on 1-XX-55-6C MASTER ISOL STATUS PNL and 1-XX-55-6D are DARK.
| |
| RO RO depresses 1-HS-30-64D, i/B CNTMT ISOL RESET TR-A and 1-HS-30-64E, i/B CNTMT ISOL RESET TR-B.
| |
| If the i/B signal was present then the RO will observe RED B lights on 1-XX-55-6C MASTER ISOL STATUS PNL and 1-XX 6D DARK.
| |
| : 23. ENSURE cntmt air in service:
| |
| : a. Aux air press greater than 75 psig [M-1 5].
| |
| RO observes 1-Pl-32-104A AUXAIR A PRESS indicating approximately 95-100 psig.
| |
| RO observes 1-Pl-32-105A AUX AIR B PRESS indicating approximately 95-100 psig.
| |
| : b. Cntmt air supply valves OPEN [M-15]:
| |
| * 1 -FCV-32-80.
| |
| RO
| |
| * 1-FCV-32-102.
| |
| * 1-FCV-32-11O.
| |
| BOP observes the RED indicating lights are LIT and the GREEN indicating lights are DARK on 1-HS-32-80A, AUXAIR TO R)(
| |
| BLDG TR A ClV-i/sB/TO PSI D/S CLOSES, 1-HS-32-102A, AUXAIR TO R)( BLDG TR B ClV-qBI70 PSI D/S CLOSES, and 1-HS IIOA, NON ESS AUXAIR TO RXBLDG CIV-çtB/70 PSI DIS CLOSES.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 26 of 29 Event
| |
| | |
| == Description:==
| |
| l-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.i, Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the IA MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.i, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 24. PERFORM Steps I through 6 of E-O, REACTOR TRIP OR SAFETY INJECTION, while continuing with this Instruction.
| |
| RO SRO assigns BOP to perform Steps 1-6 of E-O, Reactor Trip or Safety Injection.
| |
| : 25. MAINTAIN RCS bleed and feed paths:
| |
| * MAINTAIN charging pump injection thru BIT.
| |
| RO Charging flow is NOT available.
| |
| * MAINTAIN SI pump flow.
| |
| * MAINTAIN both pzr PORVs and block valves OPEN.
| |
| CAUTION If containment pressure rises to greater than 2.8 psig, containment spray should be verified.
| |
| : 26. DETERMINE if cntmt spray should be stopped:
| |
| : a. Spray pumps running.
| |
| : b. MONITOR cntmt pressure less than 2.0 psig.
| |
| RO c. RESET containment spray signal.
| |
| : d. STOP cntmt spray pumps AND PLACE in A-AUTO.
| |
| : e. CLOSE cntmt spray discharge valves i-FCV-72-2 and I FCV-72-39.
| |
| : 27. WHEN RWST level is less than 34%, THEN ** GO TO ES-i .3, RO TRANSFER TO RHR CONTAINMENT SUMP.
| |
| : 28. ENSURE CCS alignment for RHR operation:
| |
| : a. RHR heat exchanger B outlet 1-FCV-70-i53 OPEN.
| |
| RO
| |
| : b. RHR heat exchanger A outlet i-FCV-70-i56 OPEN.
| |
| : c. SFP heat exchanger A supply 0-FCV-70-197 CLOSED.
| |
| NOTE The details of Steps 4 through 15 may be referred toas necessary to establish feed flow in the following step but procedure performance must continue to terminate RCS bleed and feed.
| |
| EXAMINER: The applicants are expected to refer to FR-H.1 Steps 4 through 15, to establish Main feedwater using the Standby Main Feedwater Pump. Steps 4 through 15 along with Appendix A, Establishing MEW Following Reactor Trip, are contained in Attachment 3 to this scenario.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ESD-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 27 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1 Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 29. EVALUATE the following to restore level in at least one S/G:
| |
| : a. AFW pumps.
| |
| b.MFW pumps.
| |
| RO
| |
| : c. Condensate pumps.
| |
| : d. ERCW valves to AFW suction. V
| |
| : e. HPFP spool piece (AC 1-7.06).
| |
| CAUTION Feedwater flow rates should be controlled to prevent excessive RCS cooldown.
| |
| NOTE If possible, a S/G should be selected to feed which has WR level greater than 15% [25% ADVI and RCS Loop WR hot leg temperature less than 550F.
| |
| : 30. ESTABLISH feedflow to one Selected S/G:
| |
| : a. Feed source AVAILABLE RNO leading to Step 33 will be required until Standby Main Feedwater Pump is in service
| |
| : b. Selected SIG WR level less than 15% [25% ADV]
| |
| RO When Standby Main Feedwater Pump is in service, RNO for this step is expected leading to Step 31.
| |
| : c. Selected S/G RCS Loop WR hot leg temperature greater -
| |
| than 550F
| |
| : d. Core exit TCs RISING
| |
| : e. ESTABLISH feedflow to selected S/G at maximum rate.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 28 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LcV-3-90, SG 3-MEW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-0, Reactor Trip or Safety Injection will be entered, and a transition to ES-0.1, Reactor Trip Response will be made. At the trip, the TDAEW pump becomes steam bound and the 1A MDAEW pump shaft shears. lB-B 6.9KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-0.1 Step 4.
| |
| Time Position Applicants Actions or Behavior
| |
| : 30. RESPONSE NOT OBTAINED:
| |
| : a. GO TO Step 33.
| |
| : b. ESTABLISH feedflow to selected SIG at a rate which causes SIG WR level to rise and RCS Loop WR hot leg temperature to drop.
| |
| **GO TO Step 31.
| |
| : c. ESTABLISH feedflow to selected SIG at a rate which causes RO SIG WR level to rise and RCS Loop WR hot leg temperature to drop.
| |
| GO TO Step 31.
| |
| : d. ESTABLISH feedflow to selected Sf0 at less than 100 GPM (40,000 PPH) UNTIL selected Sf0 WR level is greater than 15% [25% ADV], THEN ADJUST feedflow as necessary to obtain SIG NR level greater than 29% [39% ADV).
| |
| **GO TO Step 31.
| |
| : 31. CHECK Selected Sf0:
| |
| : a. Selected SIG RCS Loop WR hot leg temperature less than-RO 550°F RNO for this step is expected when Standby Main Feedwater Pump is in service.
| |
| 31 .a. RESPONSE NOT OBTAINED:
| |
| MAINTAIN feedflow to selected Sf0 at rate established in Step 30.
| |
| RO WHEN selected SIG RCS Loop Hot Leg temperature less than 550° F, THEN PERFORM Steps 31 b, c and d.
| |
| **GO TO Step 33.
| |
| : 33. CHECK all RCS bleed and feed termination criteria met:
| |
| * At least one SIG NR level greater than 29% [39% ADV].
| |
| RO
| |
| * Incore TIC dropping.
| |
| * T-hot dropping.
| |
| : 33. RESPONSE NOT OBTAINED:
| |
| CONTINUE RCS bleed and feed UNTIL all criteria met.
| |
| RO CONTINUE actions to restore secondary heat sink.
| |
| GO TO Note prior to Step 29.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.: 2 Scenario # 5 Event # 6 Page 29 of 29 Event
| |
| | |
| == Description:==
| |
| 1-LCV-3-90, SG 3-MFW REG VLV fails to 25% position causing SG 3 level to drop.
| |
| Requires a manual trip of the reactor. E-O, Reactor Trip or Safety Injection will be entered, and a transition to ES-O.1 Reactor Trip Response will be made. At the trip, the TDAFW pump becomes steam bound and the 1A MDAFW pump shaft shears. 1 B-B 6.9 KV Shutdown Board trips due to differential relay operation. This results in a loss of secondary heat sink. FR-H.1, Loss of Secondary Heat Sink will be entered either from the Status Tree evaluation OR at ES-O.1 Step 4.
| |
| Time IL Position Applicants Actions or Behavior EXAMINER: When the SRO enters Step 33 RNO with the Standby Main Feedwater Pump in service and states that a return to Step 29 is required, direct the Console Operator to place the simulator in STOP and inform the applicants that the scenario is terminated.
| |
| END OF SCENARIO 2011-10 Watts Bar NRC Examination
| |
| | |
| Scenario 5 Attachment I AOI2, MALFUNCTION OF REACTOR CONTROL SYSTEM Attachment I
| |
| | |
| WBN Malfunction of Reactor Control System AOl-2 Unit I Rev. 0038 Attachment I (Page 1 of 1)
| |
| Reactor Power VS T-avglT-ref Temperature and PZR Level (Tavg-Tref values rounded to one tenth of a degree)
| |
| RX TAVG- PZR RX TAVE- PZR POWER TREE LEVEL POWER TREF LEVEL L
| |
| 2% 557.6 °F 25.7 % 52% 572.2 °F 43.0 %
| |
| 4% 5582 °F 26.4 % 54% 572.8 °F 43.7 %
| |
| 6% 558.8 °F 27.1 % 56% 573.4 °F 44.4 %
| |
| 8% 559.3 °F 27.8 % 58% 573.9 °F 45.1 %
| |
| 10% 559.9 °F 28.5 % 60% 574.5 °F 45.8%
| |
| 12% 560.5 °F 29.2 % 62% 575.1 °F 46.5 %
| |
| 14% 561.1°F 29.8% 64% 575.7 °F 47.1 %
| |
| 16% 561.7°F 30.5% 66% 576.3 °F 47.8 %
| |
| 18% 562.3°F 31.2% 68% 576.9°F 48.5%
| |
| 20% 562.8°F 31.9% 70% 577.4 °F 49.2 %
| |
| 22% 563.4 °F 32.6 % 72% 578.0 °F 49.9 %
| |
| 24% 564.0 °F 33.3 % 74% 578.6 °F 50.6 %
| |
| 26% 564.6 °F 34.0 % 76% 579.2°F 51.3%
| |
| 28% 565.2 °F 34.7 % 78% 579.8 °F 52.0 %
| |
| 30% 565.8 °F 35.4 % 80% 580.4 °F 52.7 %
| |
| 32% 566.3°F 36.1 % 82% 580.9 °F 53.4 %
| |
| 34% 566.9 °F 36.8 % 84% 581.5°F 54.1%
| |
| 36% 567.5 °F 37.5 % 86% 582.1 °F 54.8 %
| |
| 38% 568.1 °F 38.1 % 88% 582.7 °F 55.4 %
| |
| 40% 568.7 °F 38.8 % 90% 583.3 °F 56.1 %
| |
| 42% 569.3 °F 39.5 % 92% 583.9 °F 56.8 %
| |
| 44% 569.8 °F 40.2 % 94% 584.4°F 57.5%
| |
| 46% 570.4 °F 40.9 % 96% 585.0 °F 58.2 %
| |
| 48% 571.0°F 41.6% 98% 585.6 °F 58.9 %
| |
| 50% 571.6°F 42.3% 100% 586.2 °F 59.6 %
| |
| Page 48 of 48
| |
| | |
| Scenario 5 Attachment 2 AOI16 Loss of Normal Feedwater Attachment I
| |
| | |
| WBN Loss of Normal Feedwater AOl-16 Unit I Rev. 0033 Attachment I (Page I of 2)
| |
| Manual Boration 1.0 MANUAL BORATION METHODS NOTE The required Boric Acid needed to compensate for a runback may be obtained from Reactivity Briefing Sheet A. INITIATE boration of RCS to restore AFD to normal using Section 1.2 OR 1.3.
| |
| 1.2 Normal Manual Boration A. INITIATE normal boration to change CB as necessary:
| |
| : 1. PLACE BA flow controller 1 -FC-62-1 39, to desired flow rate.
| |
| : 2. ADJUST BA batch counter 1-FQ-62-139 to ensure boration continues.
| |
| : 3. PLACE mode selector 1-HS-62-140B to BOR.
| |
| : 4. PLACE VCT makeup control 1-HS-62-140A to START.
| |
| : 5. VERIFY boric acid flow indicated on 1-FI-62-139.
| |
| : 6. WHEN adequate amount of boric acid injected, THEN CONTINUE with this instruction.
| |
| : 7. PLACE 1-H-62-140A, VCT Makeup Control to STOP.
| |
| : 8. VERIFY 1-FI-62-139 Boric Acid to Blender, indicating ZERO.
| |
| : 9. ENSURE 1-FC-62-142, PW to Blender on 35% (7Ogpm), and Manual-Auto toggle in AUTO.
| |
| : 10. ADJUST 1-FC-62-139, BAto Blender, to new RCS CB.
| |
| : 11. PLACE 1-HS-62-140B, VCT MAKEUP MODE, in AUTO.
| |
| : 12. TURN 1-HS-62-140A, VCT MAKEUP CONTROL, to START, and VERIFY red light lit.
| |
| Page 35 of 36
| |
| | |
| WBN Loss of Normal Feedwater AOI-16 Unit I Rev. 0033 Attachment I (Page 2 of 2) 1.3 Emergency Boration A. IF manual boration unavailable, THEN ESTABLISH required emergency boration flow:
| |
| : 1. PLACE both BA pumps in FAST speed.
| |
| : 2. ADJUST emergency borate valve 1-FCV-62-138 to obtain required flow.
| |
| : 3. CHECK emergency borate flow on 1-FI-62-137A.
| |
| : 4. WHEN adequate amount of boric acid injected, THEN CONTINUE with this instruction.
| |
| : 5. PLACE both BA pumps in SLOW speed.
| |
| : 6. CLOSE emergency borate valve 1-FCV-62-138.
| |
| : 7. VERIFY emergency borate flow 1-Fl-62-137A, indicating zero flow.
| |
| Page 36 of 36
| |
| | |
| Scenario 5 Attachment 3 FRH.1 LOSS OF SECONDARY HEAT SINK Steps 415 and Appendix A
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Uniti Rev.0018 Step Action/Expected Response Response Not Obtained
| |
| : 4. ENSURE S/G blowdown ISOLATED. Manually CLOSE valves.
| |
| : 5. MONITOR CST volume INITIATE CST refill USING SOI-59.01, greater than 200,000 gal. Demineralized Water System.
| |
| IF CST volume drops to less than 5000 gal, THEN MONITOR AFW pumps to ensure suction transfer.
| |
| NOTE If the use of condensate flow is anticipated, then a higher pzr level will better accommodate the level shrink from S/G cooldown and depressurization.
| |
| : 6. CONTROL pzr level between 29% and 63%
| |
| [47% and 58% ADV].
| |
| Page 5 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained
| |
| : 7. ESTABLISH MD AFW pump flow:
| |
| : a. CHECK MD AFW a. **
| |
| GO TO Step 8.
| |
| pump AVAILABLE.
| |
| : b. ENSURE both MD AFW b. START pumps from pumps RUNNING. the shutdown boards.
| |
| : c. ENSURE MD AFW LCVs OPEN. c. OPEN MD LCVs from the auxiliary control room, OR Locally OPEN MD LCVs and manual isolation valves USING SOI-3.02, Auxiliary Feedwater System.
| |
| : d. CHECK MD AFW pump flow d. ENSURE AFW valve alignment greater than 410 gpm. USING S0I-3.02, Auxiliary Feedwater System.
| |
| : e. CHECK NR level in at least e. MAINTAIN total feed flow to S/Gs one S/G greater than 29% greater than 410 gpm UNTIL NR
| |
| [39% ADV]. level in at least one S/G greater than 29% [39% ADV].
| |
| WHEN NR level in at least one S/G greater than 29% [39% ADV],
| |
| THEN RETURN TO instruction in effect.
| |
| GOTOStep8.
| |
| : f. RETURN TO Instruction in effect.
| |
| Page 6 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1
| |
| * Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained
| |
| : 8. ESTABLISH TD AFW pump flow:
| |
| : a. CHECKTDAFW pump a. **
| |
| GO TO Step 9.
| |
| AVAILABLE.
| |
| : b. ENSURE turbine steam b. OPEN steam supply valves supply valves OPEN: from reactor MOV boards.
| |
| * Either 1-FCV-1-15 or 1FCV116. IF trip and throttle valve
| |
| * 1-FCV-1-17 and 1-FCV-1-51 closed, THEN 1-FCV-1-18 CHECK the following:
| |
| * Trip and throttle valve.
| |
| * Valve latched to motor operator.
| |
| * Mechanical overspeed reset.
| |
| * Thermal overloads reset.
| |
| : c. ENSURE TD AFW LCVs OPEN. c. OPEN TD LCVs at auxiliary control panel [TD pump room, 692],
| |
| OR Locally OPEN TD LCVs and manual isolation valves:
| |
| * SIG 1 and 4 [south vlv room].
| |
| * SIG 2 and 3 [Aux Bldg 737].
| |
| : d. CHECK TD AFW d. Locally CONTROL TD AFW pump.
| |
| pump speed NORMAL.
| |
| : e. CHECK TD AFW pump flow e. ENSURE AFW valves aligned greater than 410 gpm. USING SOI-3.02, Auxiliary Feedwater System.
| |
| Step continued on next page Page7of37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-RI
| |
| * Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained
| |
| : 8. (continued)
| |
| : f. CHECK NR level in at least f. MAINTAIN total feed flow to S/Gs one S/G greater than 29% greater than 410 gpm UNTIL NR
| |
| [39% ADV]. level in at least one S/G greater than 29% [39% ADVJ.
| |
| WHEN NR level in at least one SIG greater than 29% [39% ADV],
| |
| THEN RETURN TO Instruction in effect.
| |
| GO TO Step 9.
| |
| : g. RETURN TO Instruction in effect.
| |
| : 9. STOP all four RCPs.
| |
| : 10. IF Secondary pumps will be used to feed SIGs, THEN REFER TO Appendix A (FR-H.1),
| |
| Establishing MEW following Reactor Trip, while continuing this Instruction.
| |
| Page 8 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.l Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained CAUTION
| |
| * If offsite power is lost after SI reset, manual action will be required to restart the SI pumps and RHR pumps due to loss of SI start signal.
| |
| If plant conditions degrade after automatic SI is blocked, manual actuation may be required.
| |
| NOTE After the low steamline pressure SI signal is blocked, main steamline isolation will occur if the high steam pressure rate setpoint is exceeded.
| |
| : 11. BLOCK SI signals:
| |
| : a. INITIATE RCS depressurization to less than 1912 psig:
| |
| : 1) IF letdown in service, THEN 1) IF letdown is NOT in service, THEN ALIGN aux spray USING Appendix B (FR-H.1), ALIGN USE one pzr PORV.
| |
| AUX SPRAY.
| |
| : b. BLOCK auto SI actuation signals
| |
| [68-B], and [69-B]:
| |
| : 1) NOTIFY IMs to block auto SI USING IMl-99.040, AUTO SI Block.
| |
| : 2) WHEN RCS pressure is less than 1962 psig (P-li),
| |
| THEN
| |
| * BLOCK low pzr pressure SI.
| |
| * BLOCK low steam pressure SI.
| |
| Step continued on next page Page9of37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.l Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained Ii. (continued)
| |
| : c. ENSURE high cntmt pressure SI signal CLEARED [78-G}.
| |
| : d. CHECK SI actuated. d. **
| |
| GO TO Substep 1 if.
| |
| : e. RESET SI, AND CHECK the following:
| |
| * SI ACTUATED permissive DARK.
| |
| * AUTO SI BLOCKED permissive LIT.
| |
| : f. MAINTAIN RCS pressure less than 1912 psig.
| |
| Page 10 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained NOTE e Cycling reactor trip breakers to allow MEW Isolation reset is required if SI, HI-HI S/G level, or Valve Vault Room Elooding has occurred.
| |
| . If any valid SI signal has occurred since SI reset, cycling reactor trip breakers will initiate SI.
| |
| : 12. PREPARE for MEW startup:
| |
| : a. PLACE MEW pump controllers in MANUAL, AND SET to zero.
| |
| : b. PLACE MEW reg valve controllers in MANUAL, AND SET to zero.
| |
| : c. PLACE MEW reg bypass valve controllers in MANUAL, AND SET to zero.
| |
| Step continued on next page.
| |
| Page 11 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Uniti Rev.0018 Step Action/Expected Response Response Not Obtained
| |
| : 12. (continued)
| |
| : d. CHECK FW bypass isolation d. PERFORM the following:
| |
| valves OPEN.
| |
| : 1) WHEN SI signals are blocked or cleared, THEN CYCLE reactor trip breakers to allow MEW Isolation reset.
| |
| : 2) RESET MFW isolation:
| |
| * PLACE both MEW isol reset switches to RESET
| |
| [M-3].
| |
| * ENSURE MEW isol signal clears
| |
| [M-6 Master Panel].
| |
| * PUSH MEW isol reset pushbuttons [M-3].
| |
| * ENSURE MEW bypass isol valves OPEN.
| |
| : 3) PLACE 1-HS-3-45 in LONG CYCLE RECIRC.
| |
| IF no FW bypass isolation valve can be opened, THEN GOTOStepl7.
| |
| Page 12 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Unit I Rev. 0018 Step Action/Expected Response Response Not Obtained NOTE If the standby feed pump will be used, only the hotwell pumps should be started to prevent an overpressure condition.
| |
| : 13. ESTABLISH feedwater flow:
| |
| : a. START secondary plant pumps a. IF secondary plant pumps are as necessary: NOT available, THEN
| |
| : 1) Hotwell pumps. **
| |
| GO TO Step 17.
| |
| : 2) Condensate booster pumps.
| |
| : 3) Cond Dl booster pumps.
| |
| : b. CHECK MSIVs OPEN. b. IF MSIVs can be OPENED, THEN:
| |
| : 1) OPEN MSIV bypass valves.
| |
| : 2) OPEN MSIVs as necessary.
| |
| : c. ESTABLISH MEW pump flow: c. IF MEW pump flow is NOT
| |
| : 1) START MFW pump turbine established, THEN:
| |
| or standby feed pump.
| |
| : 2) CONTROL MFW pump and
| |
| * START additional secondary bypass reg valve(s) to plant pumps as necessary.
| |
| restore S/G level(s). * **
| |
| GO TO Step 15.
| |
| Page 13 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H1
| |
| * Uniti Rev.0018 Step Action/Expected Response Response Not Obtained
| |
| : 14. CHECK secondary heat sink restored:
| |
| : a. NR level in at least one SIG a. IF feed flow established to at least greater than 29% [39% ADV]. one SIG:
| |
| * S/G Wide Range level rising, OR
| |
| * Incore TIC dropping.
| |
| THEN MAINTAIN flow to restore NR level to greater than 29% [39% ADV}.
| |
| IF feed flow NOT established to at least one SIG, THEN GOTOStepl5.
| |
| : b. RETURN TO Instruction in effect.
| |
| Page 14 of 37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Uniti Rev.0018 Step Action/Expected Response Response Not Obtained
| |
| : 15. ESTABLISH condensate flow:
| |
| : a. ENSURE condensate aligned to S/Gs:
| |
| : 1) OPEN MFW pump bypass valve 1-FCV-3-86.
| |
| : 2) THROTTLE OPEN bypass reg valves.
| |
| : b. DEPRESSURIZE at least b. IF condenser NOT available, one S/G at maximum rate THEN (25% demand) USING steam dump to condenser UNTIL USE S/G PORV(s) for at least one condensate flow established. intact S/G at maximum rate.
| |
| : c. WHEN condensate flow is established, THEN STOP S/G depressurization AND MAINTAIN S/G press (using steam dump or PORV) low enough to ensure condensate flow is maintained.
| |
| Page 15of37
| |
| | |
| WBN Loss of Secondary Heat Sink FR-H.1 Unit I Rev. 0018 Appendix A (Page 1 ofl)
| |
| Establishing MFW Following Reactor Trip 1.0 INSTRUCTIONS CAUTIONS
| |
| : 1) Rx trip breakers must be cycled to allow reset of MEW when isolated by SI or HI-HI SIG level or Valve Vault Room level switches.
| |
| : 2) If any valid SI signal has occurred since SI reset, cycling Rx trip breakers may initiate SI actuation, if signal has NOT yet been blocked by IMs.
| |
| A. ENSURE MEW reg valves controllers in MANUAL, AND SET to ZERO demand.
| |
| B. ENSURE bypass reg valves controllers in MANUAL, AND SET to ZERO demand.
| |
| C. WHEN SI signals blocked OR cleared, THEN CYCLE reactor trip breakers to allow MEW Isolation reset.
| |
| D. RESET MEW isolation:
| |
| : 1. PLACE both MEW isolation reset switches to RESET [M-3].
| |
| : 2. ENSURE MEW isolation signal clears [M-6 Master Panel].
| |
| : 3. PUSH MEW isolation reset push-buttons [M-3].
| |
| E. ENSURE MFW mode switch 1-HS-3-45 in LONG CYCLE RECIRC.
| |
| E. ENSURE MEW bypass isolation valves OPEN.
| |
| G. ENSURE standby MEW pump RUNNING, if available, AND CONTROL S/G levels with MEW bypass reg controllers.
| |
| Page 36 of 37
| |
| | |
| Scenano SHIFT TURNOVER C11ECKLIST Page 1 of2 SHIFT TURNOVER CHECKLIST Page 1 of 1 Q SM US/MCR Unit 1 UO Unit Off-going Name E1 AUO Station Q STA (STA Function) On-coming Name Part 1 Completed by off-going shift / Reviewed by on-coming shift:
| |
| RCS Cb = 747 ppm
| |
| * Abnormal equipment lineup / conditions:
| |
| lA-A CCP is tagged for motor bearing replacement. LCO 3.5.2 and TR 3.1.4 were entered 8 hours ago.
| |
| Protected equipment signs have been posted for lB-B CCP.
| |
| * SI/Test in progress/planned: (including need for conduct of evolution briefings)
| |
| * Major Activities/Procedures in progress/planned:
| |
| Train A/Channel 1 Work Week. An oil leak on the lA-MFP was identified last shift, and Operations Management is meeting to determine whether the pump will be removed from service. The TB AUO is monitoring the 1A MFP every 30 minutes. Current direction is to maintain current plant conditions.
| |
| * Radiological changes in plant during shift:
| |
| Part 2 Completed by on-coming shift prior to assuming duties
| |
| [] Review station rounds / Abnonnal reading (AUOs only)
| |
| EJ Review Narrative Logs (previous day and carry-over items)
| |
| D Current qualification status Q Review the current controlling Reactivity Management Plans (N/A for AUOs)
| |
| Q Review current TS/TRMJODCM/FPR Required Actions (N/A for AUOs) i:i Wallcdown MCR Control Boards with off-going Operator (N/A for AUOs, as applicable for SMISTAs)
| |
| D SR/PER reviews complete for previous shift (SM/US/STA)
| |
| Relief Time: Relief Date:
| |
| Part 3 Completed by on-coming shift. These items may be reviewed after assuming duties:
| |
| Q Review Operator Workarounds, Burdens and Challenges (applicable UniliStation)
| |
| Q Review applicable ODMI actions (first shift of shift week)
| |
| [] Review changes in Standing / Shift Orders (since last shift worked)
| |
| Q Review changes to TACFs issued (since last shift worked) (N/A for AUOs)
| |
| [J Review Control Room Deficiencies (first shift of shift week) (N/A for AUO5)
| |
| Q Review Component Deviation Log (N/A for AUOs)
| |
| TVA 40741 Page 1 of 1 OPDP-1-l [01-14-20111
| |
| | |
| Scenario SHIFT TURNOVER CHECKLIST Page 1 of2 SHIFT TURNOVER CHECKLIST Page 1 of 1 E] SM D US/MCR Unit 1 UO Unit Off-going Name E1 AUO Station D STA (STA Function) On-coming -Name Part I - Completed by off-going shift / Reviewed by on-coming shift: RCS Cb = 747 ppm Abnormal equipment lineup / conditions:
| |
| lA-A CCP is tagged for motor bearing replacement. LCO 3.5.2 and TR 3.1.4 were entered 8 hours ago.
| |
| Protected equipment signs have been posted for I B-B CCP.
| |
| * SI/Test in progress/planned: (including need for conduct of evolution briefings)
| |
| * Major Activities/Procedures in progress/planned:
| |
| Train A/Channel 1 Work Week. An oil leak on the 1 A-MFP was identified last shift, and Operations Management is meeting to determine a plan for repairs. The TB AUO is monitoring the 1A MFP every 30 minutes.
| |
| Current direction is to maintain current plant conditions.
| |
| * Radiological changes in plant during shift:
| |
| Part 2 Completed by on-coming shift prior to assuming duties Q Review station rounds / Abnormal reading (AUOs only)
| |
| Review Narrative Logs (previous day and carry-over items)
| |
| Current qualification status Review the current controlling Reactivity Management Plans (N/A for AUO5)
| |
| LJ Review current TS/TRMIODCM/FPR Required Actions (N/A for AUOs) 1 Walkdown MCR Control Boards with off-going Operator (N/A for AUOs, as applicable for SM/STAs)
| |
| SR/PER reviews complete for previous shift (SMJUS/STA)
| |
| Relief Time: Relief Date:
| |
| Paz-t 3 - Completed by on-coming shift. These items may be reviewed after assuming duties:
| |
| Q Review Operator Workarounds, Burdens and Challenges (applicable Unit/Station)
| |
| Q Review applicable ODMI actions (first shift of shift week)
| |
| Q Review changes in Standing / Shift Orders (since last shift worked)
| |
| Q Review changes to TACFs issued (since last shift worked) (N/A for AUOs)
| |
| Review Control Room Deficiencies (first shift of shift week ) (N/A for AUOs)
| |
| Q Review Component Deviation Log (N/A for AUOs)
| |
| TVA 40741 Page 1 of 1 OPDP-l-l [01-14-20111
| |
| | |
| IL/
| |
| 1 Watts Bar Nuclear Plant Unit I Periodic Instruction I -PI-OPS-I -MCR Main Control Room Revision 0055 Quality Related Level of Use: Reference Use Effective Date: 05-1 0-2011 Responsible Organization: OPS, Operations Prepared By: John Lovell Approved By: Brian Mcllnay
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 2 of 23 Revision Log Rev or Affected Change Effective Page Number Date Numbers Description of Revision/Change 48 12/03/09 2, 14, 22 Changed the point monitored to 10 mm average (U2II8RA) from 5 mm average (U211805) and enhanced action to take if power approaches or exceeds the Thermal Power Limit.
| |
| 49 02/09/10 2, 7, 18, Administrative change to update CSST tap changer 20 alternate alignments lAW EDC 54778-A and to update section 6.1 step E to account for DCN 52019 (PCR 4031) 50 04/12/10 2, 14, 22 Administrative change to change control to the one hour average from the 10 minute average. Added descriptive usage of the 10 minute average.
| |
| 51 04/16/10 2, 18 Administrative change to add requirement to maintain B phase of main generator 525+/-5KV.
| |
| 52 07/07/10 2, 7, 10, Added clarifying note to Section 5.3 and added 15E500-3 12, 13, to the Developmental References (PER 226140).
| |
| 20, 19, Deleted Section 5.1.2 Subsection 3 (PER 232132/PCR 25 4093).
| |
| Deleted requirement for having SQN #2 Line in service and added new incoming Mvars limit while WBN-SQN #2 and WBN-Roane lines are out of service, in Section 5.4.
| |
| Added TRO-TO-SOP-10.130 to Developmental References.
| |
| Reformatted source notes and deleted source notes 2-4.
| |
| 53 09/10/10 2, 24 Added date block to Appendix A. [PER 239770]
| |
| 54 11/12/10 2, 8, 10, Minor/editorial revision:
| |
| 16-19 Updated references of NEAD to Load Dispatcher (PCR 4556).
| |
| Updated WO to SR.
| |
| 55 05/10/11 2, 8, 17, Removed references to no longer used, Base Adjust to 18 implement DCN 52769, which replaced the Westinghouse generator voltage regulator with the digital dual channel Unitrol 5000 Excitation Control System.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 3 of 23 Table of Contents
| |
| | |
| ==1.0 INTRODUCTION==
| |
| 5 1.1 Purpose 5 1.2 Scope 5 1.3 Applicability 5 1.4 Background - Leading Edge Flow Monitor (LEFM) 5
| |
| | |
| ==2.0 REFERENCES==
| |
| 7 2.1 Performance References 7 2.2 Developmental References 7 3.0 PRECAUTIONS AND LIMITATIONS 8 4.0 PREREQUISITE ACTIONS 8 4.1 PreliminaryActions 8 4.2 Field Preparations 8 4.3 Approvals and Notifications 8 5.0 GENERAL 9 5.1 General Duties 9 5.1.1 Unit Operator Duties 9 5.1 .2 Unit Supervisor Duties 10 5.2 Reactor Thermal Power 10 5.2.1 Specific Guidance for Controlling Thermal Power 11 5.2.2 Reactor Power Limits 13 5.2.3 Monitoring ICS Instantaneous Reactor Power 14 5.2.4 Determination of Reactor Power 15 5.3 Voltage Control Monitoring 16 5.4 Monitoring Generator Loading 18 6.0 PERFORMANCE 19 6.1 UnitOperator 19 7.0 RECORDS 20 7.1 QARecords 20 7.2 Non-QA Records 20
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 4 of 23 Table of Contents (continued)
| |
| Attachment I: REACTOR THERMAL POWER (RTP) LIMITS AND ACTIONS 21 Appendix A: 500Kv And Generator Voltage Schedule 22 Source Notes 23
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Uniti Rev. 0055 Page 5 of 23
| |
| | |
| ==1.0 INTRODUCTION==
| |
| | |
| 1.1 Purpose This Instruction details selected MCR responsibilities for licensed operators.
| |
| 1.2 Scope This Instruction includes guidance to Unit Operators (UOs) for the proper method of performing Main Control Room (MCR) walk down/inspections. This Instruction also includes guidance for responding to selected off normal events identified during MCR monitoring of plant status.
| |
| 1.3 Applicability This Instruction is applicable in all Modes. Portions of this Instruction, which are only applicable in Modes I through 4, are specifically identified.
| |
| 1.4 Background Leading Edge Flow Monitor (LEFM)
| |
| NOTE LEFM and venturi-based ICS Points are listed in Table I-I.
| |
| The LEFM uses ultrasonic transducers placed in a section of Main Feedwater piping and measures transient time of ultrasonic sound waves. It is very accurate and substantially lowers the uncertainty associated with using venturi-based flow measurement in the secondary side power calorimetrics to determine thermal output of the core. The U1118 Series ICS Points measuring system consists of nozzle venturis placed in the feedwater lines to the individual steam generators. Using the venturi-based flow measurement accuracies associated with reactor power measurement have a 2% uncertainty. A revision to NRC rules provides for a core power (Rated Thermal Power) rise to 3459 MWT when LEFM Calorimetric is used because its uncertainty is 0.6%.
| |
| The core power calculation, as determined by secondary side calorimetrics, will be made using the LEFM inputs of feedwater mass flow and temperature. Control of feedwater flow will be by the existing controls from the nozzle venturis. The LEFM is backed up by the venturi-based flow monitoring system. Loss of LEFM results in reverting to the venturi-based monitoring system. Plant equipment is operated in the same manner at the 3459 MWT power level as it is at the 3411 MWT level. If the LEFM becomes unavailable for a duration that exceeds the conditions of the Tech Requirements Manual, the secondary side calorimetrics, is performed with inputs from the nozzle venturis and requires a core power adjustment toward a lower core power based on the 2% uncertainty associated with the nozzle inaccuracies.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 6 of 23 1.4 Background Leading Edge Flow Monitor (LEFM) (continued)
| |
| The LEFM System includes a fault indication alarm that is provided as part of the plant computer software. This alarm is an audible and visual alarm through the ICS.
| |
| This alarm is provided as a new stand-alone screen through ICS and indicates NORMAL, ALERT, or FAIL. Operations is alerted to this by the ICS LEFM points which will turn blue when it is in an abnormal condition.
| |
| The venturi-based measurements alarms are suppressed by the LEFM System as long as the system is NOT failed. The alarms become active when the LEFM system is failed. The UI 127 Series ICS Points display LEFM, but toggle and display venturi-based measurement upon LEFM failure. For example, should the LEFM system fail and the venturi-based measurement indicates a core power above the 3411 MWT, computer alarms will be enabled and displayed on the alarm display.
| |
| TABLE 1-I CS POINT UNITS DESCRIPTION U1118 MWT REACTORTOTALTHERMALQ Ui 11805 MWT REACTOR TOTAL THERMAL 5 MIN AVG U1118RA MWT 10 MINUTE AVERAGE OF U1118 UI 125 MWT REACTOR TOTAL THERMAL Q 1 HR AVG Ui 126 MWT REACTOR TOTAL THERMAL Q 8 HR AVG Ui 127* % PERCENT CORE THERMAL POWER U1127H* % 1HRAVGREACTORTHERMALQ UI 127H24* MW CORE THERMAL POWER 24 HR AVG UI i27H8* % 8 HR AVG REACTOR THERMAL Q UII27MWT* MW CORE THERMAL POWER Ui 127TM* % 10 MIN AVG REACTOR THERMAL Q U1254 % IOMINAVG%PWRBASEDON34IIMWT U2118 MWT LEFMREACTORTOTALTHERMALQ U2125 MWT LEFMRXTOTALTHERMALQ1 HRAVG U2126 MWT LEFM RX TOTAL THERMAL Q 8 HR AVG Ui 127 Series ICS computer points toggle between LEFM calorimetric, venturi-based calorimetric, and delta temperature % power based on plant conditions and the new RTP of 3459 MWT.
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 7 of 23
| |
| | |
| ==2.0 REFERENCES==
| |
| | |
| 21 Performance References A. Good Practice OP-206, INPO 84-030, Rev 1, Generic Round Sheets and Shift Operating Practices.
| |
| B. SPP-3.5, Regulatory Reporting Requirements.
| |
| C. OPDP-1, Conduct of Operations.
| |
| D. OPDP-8, Limiting Conditions For Operation Tracking 2.2 Developmental References A. 1-Sl-0-2 Series, Shift and Daily Surveillance Log.
| |
| B. Drawings:
| |
| 47E235-42, -45 1 5E500-3 C. WBN Grid Voltage Study, dated 4/9/01 (E32 010409 601)
| |
| D. TI-I 2.15, 161 kV Offsite Power Requirements E. Standard VAR-002-1 - Generator Operation for Maintaining Network Voltage Schedule.
| |
| F. NRC Discussion of Licensed Power Level (AITS F14580H2) Dated August 22, 1980 G. NEI Position Statement on Licensed Power Limit, Project Number 689, Dated June 23, 2008.
| |
| H. NRC Regulatory Issue Summary 2007-21, Rev 1, Adherence to Licensed Power Limits. Dated February 9, 2009 I. Watts Bar Nuclear Plant, Unit I Facility Operating License NPF-90, Docket NO 50-390. Amendment No. 81 J. DON 52019 Provide a robust flash storage drive for Thermowestronic recorders K. TRO-TO-SOP-10.130, WBN Grid Operating Guide
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 8 of 23 3.0 PRECAUTIONS AND LIMITATIONS A. This Instruction does NOT supersede or replace other documents necessary for safe operation of the plant.
| |
| B. 1-Sl-0-2 Series and other similar Surveillance Instructions will have precedence over this Instruction.
| |
| C. The Load Dispatcher is to be notified within 30 minutes when the Main Generator Voltage Regulator is NOT in automatic.
| |
| 4.0 PREREQUISITE ACTIONS None 4.1 Preliminary Actions None 4.2 Field Preparations None 4.3 Approvals and Notifications None
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 9 of 23 5.0 GENERAL 5.1 General Duties 5.1.1 Unit Operator Duties NOTE Steps in this Instruction may be performed in any sequence.
| |
| [1] WHEN assigned, THEN V
| |
| [1.1] PERFORM assigned duties and inspections, including a joint walk down of at least one Board section (This is in addition to the OPDP-1 requirements). [Cl]
| |
| [1.2] NOTIFY US of difficulties in performing assigned tasks.
| |
| [1.3] VERIFY the following:
| |
| A. Equipment labels are NOT unreadable due to Caution Tags or Hold Orders.
| |
| B. Indicating lights are energized as required.
| |
| C. Recorders are operating properly.
| |
| D. Alarm panels in Main Control Room test satisfactorily.
| |
| E. Computer is operable.
| |
| F. Floor space open, allowing access to all panels.
| |
| G. Gauges are in expected range and operating properly (i.e., no stuck indicators, no unexpected swings, no off scale high or low).
| |
| H. US/SM has performed the shiftly PER operability review.
| |
| [1.4] PERFORM handswitch alignment check (e.g., CS, AFW, EGTS, Rad Monitor Block Switch). [Cl]
| |
| [1.5] ENSURE alarms that are LIT are fully understood and expected or a corrective action is in progress.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 10 of 23 5.1.1 Unit Operator Duties (continued)
| |
| [2] WHEN problems arise, THEN:
| |
| [2.1] NOTIFY US of abnormal conditions.
| |
| [2.2] TROUBLESHOOT problems, AND DETERMI NE corrective action(s).
| |
| [2.3] INITIATE Service Request as required.
| |
| [2.4] ASSIST maintenance personnel as required.
| |
| 5.1.2 Unit Supervisor Duties NOTES
| |
| : 1) LCO Tracking Only entries are defined in OPDP-8, Limiting Conditions for Operation Tracking.
| |
| : 2) The following duties can be performed by the Unit Supervisor or his designee.
| |
| [1] REVIEW the following:
| |
| [1.1] ICS computer points out of scan
| |
| [1.2] Substituted value logs
| |
| [1.3] Alarms suppressed logs
| |
| [2] US/STA ENSURES MCR UNIT SUPV operator rounds are completed daily by day shift (0700-1 900).
| |
| 5.2 Reactor Thermal Power SEC. 5.2.1 AND SEC. 5.2.2 APPLIES AT ALL TIMES. SECTIONS 5.2.3, AND 5.2.4 SHALL BE USED, AS APPROPRIATE, TO ENSURE COMPLIANCE: IC.51 Watts Bar is authorized to operate at a Licensed Power Limit (LPL) not to exceed 3459 megawatts thermal as measured by the eight hour average. The one hour core thermal power average over eight hours shall be monitored to ensure Reactor Thermal Power is maintained at or below the LPL as measured by the eight hour average.
| |
| Reactor power limits and any required actions are also contained in Attachment 1.
| |
| This attachment may be laminated and made available for the OAC to use as a reference during operations.
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 11 of23 5.2.1 Specific Guidance for Controlling Thermal Power A. Steady State The term steady state implies that temperatures, pressures, and flows are stable such that the nominal value of reactor power remains stable, subject to statistical uncertainties and normal fluctuations (e.g., feedwater oscillations).
| |
| Normal fluctuations (i.e., automatic control system response), random processes (i.e., feedwater temperature changes), and instrument uncertainties (i.e., flow meter measurement uncertainties) may slightly affect core thermal power indications, but these affects do not result in a violation of the licensed power limit license condition when operating at steady state conditions.
| |
| B. Pre-Planned Evolutions If an evolution is expected to cause core thermal power to exceed 3459 MWT, then action should be taken to reduce power, (prior to the evolution) in order to maintain sufficient margin from the limit.
| |
| C. Compliance No actions are allowed that would intentionally raise core thermal power above the Licensed Power Limit (LPL) of 3459 MWT for any period of time. Operators may not intentionally operate or authorize operation above 3459 MWT. Small, short-term fluctuations in power that are not under the direct control of a license reactor operator (e.g., fluctuations caused by secondary-side control valve oscillations) are not considered intentional.
| |
| Prompt action is to be taken to ensure reactor thermal power is maintained with in stated limits following any un-planned event.
| |
| Closely monitor thermal power during steady state power operation with the goal of maintaining the one-hour thermal power average at or below the LPL. If the core thermal power average for a one-hour period is found to exceed the LPL, take prompt action specified in Section 5.2.2 to ensure that thermal power is less than or equal to LPL.
| |
| The eight hour average is not to exceed LPL.
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 12 of 23 5.2.1 Specific Guidance for Controlling Thermal Power (continued)
| |
| NOTES
| |
| : 1) The 5 and 10-minute average (U21 1805 and U21 18RA) can be of particular value to the operating crew when conducting operations with particular sensitivity to affecting core thermal power. The 5 and 10-minute average are used for trending purposes to allow anticipatory response to changing plant conditions. This can be used as a leading indicator in making decisions for taking prompt action to mitigate rising power conditions.
| |
| : 2) The licensed thermal power limit is NOT considered to be exceeded when short duration peaks are normal fluctuations inherent in the design of the controlling system as long as the one-hour average (iJ2l 25) is at or below the licensed thermal power limit.
| |
| D. IF core thermal power one-hour average (U2125) exceeds 3459 MWt or an increasing power trend which will exceed 3459 MWt is observed, THEN ENSURE immediate action is taken to decrease reactor power as necessary.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 13 of 23 5.2.2 Reactor Power Limits NOTE Due to the calculation method used for U2125 and U2126, the one and eight-hour averages may occasionally change by large amounts. The Operator is responsible for being aware of the entire previous periods trend.
| |
| E. REACTOR THERMAL POWER AVERAGED OVER EIGHT HOURS (U2126)
| |
| SHALL BE MAINTAINED BELOW 3459 MWT AT ALL TIMES. Routine monitoring of power indications should be used to maintain indicated power at or below 3459 MWT (U2126, NIS Power Indication).
| |
| F. If U2126 exceeds 3459 MWT the Unit SRO shall be notified, Reactor Thermal Power must be reduced below 3459 MWT and the requirements of SPP-3.5, Regulatory Reporting Requirements met.
| |
| N OTE 3457 MWT is a value based on steady state plant conditions, trended experience, and equipment operating properly. If equipment conditions change, such that normal fluctuations change in magnitude or frequency, a new target value should be discussed by shift management, (in consultation with Operations department management). Any new target will be set to ensure sufficient Operating Margin is maintained.
| |
| G. Reactor Thermal Power averaged over one hour (U2125) should be maintained at a target of 3457 MWT.
| |
| H. The Unit SRO shall be notified when Reactor Thermal Power is above 3459 MWT on one-hour average (U2125). Action shall be initiated promptly to ensure power trends to less than or equal to 3459 MWT (U2125).
| |
| I. The Operations Manager should be notified immediately if the one-hour average of core thermal power (U2125) exceeds 3475 MWT.
| |
| J. An SR shall be initiated on any unplanned transient that results in an adverse trend in thermal power. (heater string isolation, MSR isolation, FW reg valve transient, etc)
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 14 of 23 5.2.3 Monitoring ICS Instantaneous Reactor Power NOTE The inputto U1127 automaticallytoggles between U2118, U1118, and U0484, as conditions warrant.
| |
| A. The lOS Points U2118 series, U2125, U2126 and U1127 series are available for monitoring trends or to alert operators to transient conditions.
| |
| B. When U21 18 is identified as being unreliable other available lOS indications and NIS power indications must be used to monitor Reactor Power for compliance with Section 5.2.2. REFER TO Section 5.2.4 for the Determination of Reactor Power.
| |
| C. If the LEFM calorimetric is or has been unreliable:
| |
| : 1. Power operation may continue per Sections 5.2.2, at 100% RTP (3459 MWT) until the next scheduled performance of SR 3.3.1.2.
| |
| REFER TO TRM 3.3.7.
| |
| : 2. The most limiting of the indications provided in Section 5.2.4 shall be used to determine Reactor Power as applicable to this Section and Section 5.2.2.
| |
| : 3. Monitoring using Ui 127 will toggle between LEFM and venturi-based calculations, if available.
| |
| : 4. Monitoring using Uiii8 series points, U1125 and U1i26 may be available or may become available.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Uniti Rev. 0055 Page 15 of 23 5.2.4 Determination of Reactor Power The following should be used on an as needed basis to compensate for a lack of reliable computer data, or as necessary to ensure compliance with Section 5.2.2:
| |
| NOTES
| |
| : 1) Items A and B are to be used following LEFM failure until the next performance of SR 3.3.1.2.
| |
| : 2) Item C may be used for trending when LEFM is failed but the venturi-based calorimetric is good. U1125, UI 126 and U1127 series trends will trend down over time following power reduction to 341 1 MWT.
| |
| A. For reactor power at or above 50% AND Tavg within 0.5°F of Tref, the average NIS Power Range drawer meter readings [1-M-13] (if calibrated per SR 3.3.1.6) shall be used. Do not exceed 102% on valid indicated power.
| |
| B. For reactor power below 50% OR Tavg NOT within 0.5°F of Tref, the higher reading of the following shall be used:
| |
| : 1. The average NIS Power Range drawer meter readings [1 -M-l 3]
| |
| (if calibrated per SR 3.3.1.6).
| |
| OR
| |
| : 2. The average AT power readings [I-M-5].
| |
| NOTE U 1127 series points should be maintained reliable as they toggle between their applicable inputs U21 18 and UI 118, however, the LEFM data will be included in the trends.
| |
| C. Following performance of I -SI-92-l (SR 3.3.1.2), the trends for U 1118, U 1125, and Ui 126 may be available for use to maintain core power at or below 3411 MWT UI 127 series trends may also be used following LEFM data removal from the averages. (e.g. UII27TM may be used 10 minutes after LEFM failure when Ui 118 calculation is still available.)
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 16 of 23 5.3 Voltage Control Monitoring NOTES
| |
| : 1) Bus I B Phase is the point used to determine WBN 500KV switchyard voltage. This may be monitored using ICS point LGI-1219.
| |
| : 2) Transmission Voltage Schedule is to be maintained 525+/-5Kv.
| |
| [1] WHEN CSST tap changer(s) are in OFF (Pushed In), OR Voltage Schedule can NOT be maintained, THEN
| |
| [1.1] RECORD generator and 500Kv voltage readings on Appendix A.
| |
| [1.2] TRANSMIT Appendix A to Management Servies when completed.
| |
| [2] IF 500Kv voltage is outside the Transmission Voltage Schedule and cannot be returned to the required voltage, THEN PERFORM the following:
| |
| A. NOTIFY the Transmission Operator within 30 minutes.
| |
| B. DOCUMENT the notification in the narrative log.
| |
| NOTE VAR5 are to be maintained in accordance with section 5.4.
| |
| C. IF 500Kv voltage is high, THEN ENSURE Main Generator VARS are incoming.
| |
| D. IF 500Kv voltage is low, THEN ENSURE Main Generator VARS are outgoing.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 17 of 23
| |
| &3 Voltage Control Monitoring (continued)
| |
| NOTE Tap Changers are normally operated in auto but can be operated in manual at SRO discretion. Operation in manual is considered an alternate alignment with respect to the operating requirements and limitations imposed by the WBN grid operating guide.
| |
| Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-1O.130 WBN Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. The Load Dispatcher shall be notified when the alternate alignments are planned, entered, and exited.
| |
| [3] WHEN CSST tap changer(s) have been placed in any of the following alternative alignments:
| |
| * 6.9kV Common Board A or B Loads on Alternate Feeders
| |
| * 480V Turbine Building Common Board A or B on Alternate Feeder
| |
| * Common Station Service Transformer C or D Controls on Alternate Feeder
| |
| * Common Station Service Transformer C or D Load Tap Changer Loss of Power or De-energized o Common Station Service Transformer C or D Load Tap Changer in OFF or in Manual During Modes I 4 -
| |
| THEN, NOTIFY Load Dispatcher of the alternative alignment.
| |
| [4] NOTIFY Load Dispatcher within 30 minutes when Main Generator Voltage Regulator is NOT in automatic.
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Page 18 of 23 5.4 Monitoring Generator Loading NOTE If operating the generator with incoming Mvars, ICS alarm(s) should actuate at -90 (LO) and -100 Mvars (LO-LO).
| |
| [1] ADJUST 1-HS-57-22, EXCITER VOLTAGE ADJUSTER [l-M-l], to maintain 500kV bus voltage within Load Dispatcher Voltage Schedule.
| |
| [1.1] IF the generator is to be operated with incoming Mvars, THEN ENSURE the following parameters are maintained:
| |
| * Generator voltage between 22,800 and 24,400 volts.
| |
| * Mvar loading no more negative than -100 Mvars by observing 1-EI-57-8, MEGAVARS [1-M-1] or ICS point Q2823A.
| |
| * With WBN-SQN #2 and WBN-Roane Lines out of service maximum incoming Mvar limit is -50 Mvars.
| |
| [1.2] IF the generator is to be operated with zero or outgoing Mvars, THEN ENSURE the following parameters are maintained:
| |
| * Generator voltage between 22,800 and 24,400 volts
| |
| * Mvar loading within limits specified by the Generator Capability Curve (See S0I-47.02, Appendix E)
| |
| | |
| WBN Main Control Room 1-Pl-OPS-1-MCR Unit I Rev. 0055 Pagel9of23 6.0 PERFORMANCE 6.1 Unit Operator A minimum of one Control Board (e.g., 1-M-1, 1-M-6, 1-M-26) will be walked down by UO and SRO (US, STA, or SM) each shift (this is in addition to the OPDP-1 requirements).
| |
| A. If any Technical Specification or Technical Requirement parameter is out of limits, then US must be notified immediately.
| |
| B. Tours will be conducted using electronic data recorder at the frequency directed by the Operations Superintendent and when the site network is available. The data recorders contain logic to detect noted and out of limit data, and indicate audibly and visually on the data recorder when data is entered. When the data is transferred to the network database, the abnormal and out of limit data indicate visually when viewed or printed. Noted and out of limit data shall be reviewed by the Unit Supervisor and corrective action taken when required.
| |
| C. When electronic data recording is unavailable, tours will be conducted at the frequency directed by the Operations Superintendent using a printout of the tour as a guideline. The Unit Supervisor will discuss equipment status with the performer and corrective actions taken when required. When electronic recording becomes available, the data should be entered in the database, and the printout discarded.
| |
| D. Service Requests will be written as appropriate.
| |
| E. IF a video graphic recorder has not been changed to wrap mode lAW reference J, THEN REMOVE AND REPLACE with a new card when the video graphic recorders flash card is 90% full and alarms (indicated by a flashing red box on the toolbar at the bottom of the recorders display screen). The full card will be transmitted to Document Control in accordance with SPP-2.4, Records Management.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 20 of 23 7.0 RECORDS 7.1 QA Records Appendix A, 500KV and Generator Voltage Schedule 7.2 Non-QA Records None
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 21 of 23 Attachment I (Page 1 ofl)
| |
| REACTOR THERMAL POWER (RTP) LIMITS AND ACTIONS LIMIT VALUE SOURCE ACTION Eight-hour 3459 MWT U2126, NIS RTP shall be maintained at or below 3459 MWT average Reactor (100%) Power Indication (100% NIS) for any eight-hour period. Notify the Power SRO if U2126 is greater than 3459 AND reduce RTP below 3459 MWT.
| |
| One-hour RTP 3459 MWT U2125 If U2125 rises above 3459 MWT. Notify Unit SRO average and immediately initiate actions to ensure RTP trends to less than or equal to 3459 MWT.
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Uniti Rev. 0055 Page 22 of 23 Appendix A (Page 1 of 1) 500Kv And Generator Voltage Schedule Date NOTE Readings are to be taken every two hours whenever CSST tap changer hand switches are OFF, or when Voltage Schedule can NOT be maintained. For voltage schedule go to the following web site: http:lltro web. cha. tva.govNoltageSchedule/.
| |
| The four CSST C & D tap changer hand switches for X and Y windings are located on ECBs 2 and 3 and are normally aligned to PULL-FOR-AUTO.
| |
| 500 kV BUS VOLTAGE GENERATOR UNIT I 0-EI-245-CB-63B BUS1 BUS2 VOLTS SECTION 3 MW MVAR
| |
| ,E u
| |
| (B PHASE) VOLTS (0-ECB-6) (0-ECB-6)
| |
| (B_PHASE)
| |
| DAY SHIFT 0800 ET 0700 CT 1000 ET 0900 CT 1200 ET 1100 CT 1400 ET 1300 CT 1600 ET 1500 CT 1800 ET 1700 CT NIGHT SHIFT 2000 ET 1900 CT 2200 ET 2100 CT 0000 ET 2300 CT 0200 ET 0100 CT 0400 ET 0300 CT 0600 ET 0500 CT
| |
| | |
| WBN Main Control Room 1-PI-OPS-1-MCR Unit I Rev. 0055 Page 23 of 23 Source Notes (Page I of I)
| |
| Implementing Requirements Statement Source Document Statement Both EGTS P controllers were found in A-Auto WBPER96OI89 C.1 Standby position. (LER 50-390/96010)
| |
| DELETED C.2 DELETED C.3 DELETED C.4 Revise GOs to provide guidance to prevent WBPER96O26O C.5 exceeding (power) limits. Specify instruments to be monitored.
| |
| | |
| ppendix D Scenario Outline Form ES-D-1 Facility: Watts Bar October 2011 Scenario No. 6 Op Test No.: 2 Examiners: Operators: SRO RO BOP Initial Conditions: 1 x1O
| |
| % power. RCS boron concentration is 1616 ppm. Control Bank D is at98 2
| |
| steps. Train A/Channel Ill Work Week.
| |
| Turnover: Continue with startup and raise power to 1-4% using GO-2, Reactor Startup. Currently GO-2, Reactor Startup, Section 5.3, Reactor Startup, is in progress, ready for performance of Step 31.
| |
| Event MaIf. No. Event Type* Event Description No.
| |
| I n/a R-RO Continue with startup and raise power to 1-4% using GO-2, N-SRD/BOP Reactor Startup.
| |
| 2 si3la TS-SRO 1-LT-63-50, RWST LEVEL fails low.
| |
| 3 fwl 03b C-BDP 1 B Condenser Vacuum pump trips.
| |
| 4 rwlOa C-BOP 1A ERCW header leak in the Auxiliary Building.
| |
| TS-SRO 5 cvrl 8 C-RD Inadvertent dilution caused by a local valve misalignment.
| |
| 6 thO4b C-RD 1-SV-68-564, PZR SAFETY VALVE fails partially open.
| |
| 7 thO4b M-ALL 1 -SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped.
| |
| 8 siO9l C-BOP The Containment Phase A isolation signal fails to automatically siO9m occur on both trains.
| |
| 9 csOla C-BOP 1A Air Return Fan fails to start automatically. 1A Containment csO6g Spray Pump trips on instantaneous overcurrent when started.
| |
| edO6b 1 B-B 6.9 KV Shutdown Board trips on differential relay operation 4 minutes after the Phase B occurs.
| |
| (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Watts Bar Examination October2011 NUREG 1021 Revision 9
| |
| | |
| ppendix D Scenario Outline Form ES-D-1 Scenario 6 - Summary Initial Condition 1 x10
| |
| % power. RCS boron concentration is 1616 ppm. Control Bank D is at 98 2
| |
| steps. Train A/Channel Ill Work Week.
| |
| Turnover Continue with startup and raise power to 1-4% using GO-2, Reactor Startup.
| |
| Currently GO-2, Reactor Startup, Section 5.3, Reactor Startup, is in progress, ready for performance of Step 31.
| |
| Event I Raise power to 1-4% power, using GO-2, Reactor Startup, Section 5.3, Reactor Startup, Step 31.
| |
| Event 2 1 -LT-63-50, RWST LEVEL transmitter fails low. Requires entry into ARI 126-C, RWST LEVEL LO RECIRC INTLK. Requires a Tech Spec evaluation and entry into LCO 3.3.2, Condition K for Function 7.b RWST Level Low for automatic switchover to the containment sump, and LCO 3.3.3, Condition A for Function 23 Refueling Water Storage Tank Water Level.
| |
| Event 3 1 B Condenser Vacuum pump trips. Requires entry into ARI 14-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another condenser vacuum pump to be started, using SOl-2&3.O1, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established.
| |
| Event 4 1A ERCW header leak in the Auxiliary Building. Requires entry into AOl-i 3, Loss of Essential Raw Cooling Water, Section 3.3, Supply Header Rupture in Auxiliary Building; High flow on supply header AND building flood alarm LIT.
| |
| Requires a Tech Spec evaluation and entry into LCO 3.7.8.
| |
| Event 5 1-ISV-62-932, Dilution to CCP Suction is inadvertently opened locally. Requires entry into AOl-3, Malfunction of Reactor Makeup Control. Requires RO to manually stop the primary water pumps.
| |
| Event 6 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOl-6, Small Reactor Coolant System Leak. Following isolation of Charging and Letdown in AOl-6, Pzr Safety Valve leak will worsen requiring reactor trip.
| |
| Event 7 1-SV-68-564, PZR SAFETY VALVE fails fully open when the plant is tripped. This causes the PRT to rupture, resulting in containment parameters degrading.
| |
| Event 8 Containment Phase A isolation signal fails to automatically occur. Requires MANUAL initiation of Phase A isolation for both trains of equipment.
| |
| Event 9 1A Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. lB-B 6.9 KV Shutdown Board trips on differential relay operation 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Appendix D Watts Bar Examination October 2011 NUREG 1021 Revision 9
| |
| | |
| ppendix D Scenario Outline Form ES-D-1 Scenario 6 Critical Task Summary Critical Task 1 Manually initiate at least one train of Containment Phase A isolation prior to completion of E-O, Reactor Trip or Safety Injection, Appendix A, Equipment Verification Critical Task 2 Trip the RCPs prior to completing E-1, Loss of Reactor or Secondary Coolant.
| |
| Appendix D Watts Bar Examination October 2011 NUREG 1021 Revision 9
| |
| | |
| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions SIMULATOR SETUP INFORMATION ENSURE Examination Security has been established.
| |
| : 2. RESET to Initial Condition 305 by performing the following actions:
| |
| : a. Select lCManager on the THUNDERBAR menu (right hand side of Instructor Console Screen).
| |
| : b. Locate IC# 305.
| |
| : c. Right click on IC# 305.
| |
| : d. Select Reset on the drop down menu.
| |
| : e. Right click on RESET.
| |
| : f. Enter the password for IC 305.
| |
| : g. Select Yes on the INITIAL CONDITION RESET pop-up window.
| |
| : h. Perform SWITCH CHECK.
| |
| : 3. SELECT Director on the THUNDERBAR menu (right hand side of Instructor Console Screen).
| |
| : 4. ENSURE the following information appears on the Director Screen:
| |
| Key Description Type Event Delay Inserted Ramp Initial Final Value J cso6g air return fan a-a fail to start on phase b 00:00:00 00:00:00 00:00:00 Active Active rpo2b auto si initiation signal failure 00:00:00 00:00:00 00:00:00 Active Active cs06h air return fan b-b fail to start on phase b 00:00:00 00:00:00 00:00:00 Active Active cs0la containment spray system pump a trip 00:00:00 00:00:00 00:00:00 Active Active ed06b loss of 6.9 kv shutdown board bus 1 b-b 13 00:04:00 00:00:00 Active InActive th04b pzr safety valve sv-68-564 19 00:00:00 00:00:00 100 0 si3la set rwst level LT-63-50 2 0 Page 1 of 6
| |
| | |
| Wafts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions SIMULATOR SETUP INFORMATION Key Description Type Event Delay Inserted Ramp Initial Final Value
| |
| [ fwlO3b oc trip vacuum pump b 3 00:00:00 00:00:00 Active InActive
| |
| . rwl0a ercw supply header 1-a break in aux bdlg 4 00:00:00 00:00:00 50 0 cvrl8 dilution to charging pmp suction 62-932 (0=closed, 1=open) 5 00:00:00 00:00:00 1 0 th04b pzr safety valve sv-68-564 4 00:00:00 00:05:00 1 0
| |
| : 5. Place simulator in RUN and acknowledge any alarms.
| |
| : 6. -
| |
| ENSURE the Train A Week Channel 1 sign is placed on 1-M-30.
| |
| : 7. Place simulator in FREEZE.
| |
| : 8. ENSURE Wafts Bar Nuclear Plant Unit I Reactivity Briefing Book BOL (Beginning Of Life) is provided to the crew as part of the Turnover Package, and that the BOL placards are on 1-M-6, below the Boric Acid and Primary Water Integrators.
| |
| : 9. WHEN prompted by the Chief Examiner, place the Simulator in RUN.
| |
| : 10. When Charging and Letdown are isolated in AOI-6, increase severity of PZR Safety Valve Failure to 6% to require reactor trip.
| |
| Page 2 of 6
| |
| | |
| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions Exam Simulator vent Event DescriptionlRole Play No. No.
| |
| I n/a Continue with startup and raise power to 1-4% using GO-2, Reactor Startup.
| |
| ROLE PLAY: N/A 2 2 1-LT-63-50, RWST LEVEL fails low.
| |
| ROLE PLAY: If contacted as Work Control, acknowledge request to prepare a package to troubleshoot and repair 1-LT-63-50 and initiate procedure to trip associated bistables as required.
| |
| ROLE PLAY: If/when contacted as the Outside AUO, repeat back request to inspect the RWST locally. Report back that there are no obvious problems with the RWST.
| |
| 3 3 1 B Condenser Vacuum pump trips.
| |
| ROLE PLAY: When Dispatched to check pump breaker locally report Instantaneous overcurrent relay actuation at breaker.
| |
| ROLE PLAY: When dispatched to check pump locally, report odor of overheated insulation.
| |
| ROLE PLAY: If dispatched to check another Condenser Vacuum Pump ready for start, Console Operator will acknowledge request and report pump is available for start.
| |
| ROLE PLAY: If work control is contacted to troubleshoot and repair, acknowledge request and initiate work package.
| |
| ROLE PLAY: If AUO dispatched to establish drain to running Condenser Vacuum Pump acknowledge request.
| |
| ROLE PLAY: If AUO dispatched to isolate drain for tripped Condenser Vacuum Pump acknowledge request.
| |
| Page 3 of 6
| |
| | |
| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions xam Simulator vent Event DescriptionlRole Play No. No.
| |
| 4 4 1A ERCW header leak in the Auxiliary Building.
| |
| ROLE PLAY: If AUO dispatched to the Auxiliary Building to locate the leak, acknowledge the request.
| |
| Report leak from a pipe above the TDAFW Pump room.
| |
| ROLE PLAY: When AUO dispatched to Reactor MOVboards acknowledge request.
| |
| ROLE PLAY: When AUO stationed at the Reactor MOVboards is requested to manipulate breaker for 1-FCV-67-81 modify remote function rwrO3 (power to appendix r valve 67-8 1) and report breaker closed.
| |
| ROLE PLAY: If report requested from AUO in AUX Building, report that leakage is stopped after ERCW valves have been manipulated.
| |
| ROLE PLAY: if Work Control requested to prepare a troubleshooting and repair package for the ERCW leak, acknowledge the request.
| |
| ROLE PLAY: If Report requested, AUO in the Aux BdIg has determined that no additional equipment has been affected.
| |
| ROLE PLAY: If Shift manager or other personnel contacted to evaluate RISK, Acknowledge the request.
| |
| 5 5 Inadvertent dilution caused by a local valve misalignment.
| |
| ROLE PLAY: If AUO dispatched to locally stop the Waste Gas Compressors. Acknowledge the request.
| |
| ROLE PLAY: If dispatched to Locally CLOSE 1-IS V-62-933 use remote function cvrl I to close valve and report closed.
| |
| ROLE PLAY: When AUO dispatched to blender area, AUO at the blender area reports that 1-ISV-62-932 was found in the OPEN position. Workers in the area are responsible for the misalignment.
| |
| ROLE PLAY: When directed by SROIRO AUO to close valve 1-IS V-62-932, use remote function cvrl8 to close valve and report.
| |
| ROLE PLAY: If SRO contacts the Operations Duty Manager, Plant Manager, and Rx Engineering to report the inadvertent dilution acknowledge the report.
| |
| 6 6 1-SV-68-564, PZR SAFETY VALVE fails partially open.
| |
| ROLE PLAY: N/A Page4of6
| |
| | |
| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions Exam Simulator vent Event DescriptionlRole Play No. No.
| |
| 7 nla 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped.
| |
| ROLE PLAY: When contacted, acknowledge the request to shutdown the Upper and Lower Containment Radiation monitors sampling pumps, and inform the requestor that the sample pumps are off ROLE PLAY: When contacted, acknowledge the request to open the Ice Condenser AHU breakers, and inform the requestor that the breakers are open.
| |
| ROLE PLAY: When contacted, acknowledge the need to perform E-1 Appendix A, B, and C.
| |
| Use remote function sirOl to complete E-1 Appendix A (place power on CLA outlet valves).
| |
| Use remote sirl4 to complete E-1 Appendix B (place power on 1-FCV-63-1.
| |
| Use remote function sirO6 to complete El Appendix C (place power on 1-FCV-63-22).
| |
| After remote functions are entered, report that the Appendices are complete.
| |
| ROLE PLAY: If Shift Manager contacted to Refer to EPIP-1, acknowledge the request.
| |
| ROLE PLAY: If Chemistry or Rad Protection contacted to survey or sample secondary lines, acknowledge the request.
| |
| ROLE PLAY: When contacted as the Control Building AUO the Console Operator repeat back request to check low analyzer temp lights, and reports that the lights are NOT LIT.
| |
| 8 nla The Containment Phase A isolation signal fails to automatically occur on both trains.
| |
| ROLE PLAY: N/A Page 5 of 6
| |
| | |
| Watts Bar Nuclear Plant 10-2011 NRC Examination Scenario 6 Simulator Console Operators Instructions Exam Simulator Event Event DescriptionlRole Play No. No.
| |
| 9 nla 1A Air Return Fan fails to start automatically.
| |
| ROLE PLAY: N/A 1A Containment Spray Pump trips on instantaneous overcurrent when started.
| |
| ROLE PLAY: When contacted as the Control Building AUO, state that the breaker for the 18 Con tainment Spray Pump tripped due to instantaneous overcurrent.
| |
| ROLE PLAY: When contacted as the Auxiliary Building AUO, state that the IA Containment Spray pump is severely damaged, and the shaft appears to be broken.
| |
| 1 B-B 6.9 KV Shutdown Board trips on differential relay operation 4 minutes after the Phase B occurs.
| |
| ROLE PLAY: When contacted as the Control Building AUO, repeat back request to investigate the lB-B 6.9 KV Shutdown Board, report that the differential lockout relay has operated, and that there is extensive damage to the board. There is no fire.
| |
| ROLE PLAY: When contacted as Work Control, acknowledge the request to have a maintenance team go to the lB-B 6.9 KVShutdown Board to assess the damage, and to plan repairs.
| |
| ROLE PLAY: If TSC orAUO contacted to manually close valves Containment Isolation Valves without power (1-FCV-62-61 and 1-FCV-31-67) acknowledge the request.
| |
| n/a Page 6 of 6
| |
| | |
| Lppendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 1 Page 1 of 34 Event
| |
| | |
| == Description:==
| |
| Continue to startup and raise power to 1-4% using GO-2, Reactor Startup.
| |
| Time j Position Applicants Actions or Behavior The applicants will begin to raise reactor power using GO-2, Reactor Startup, Section 5.3, Reactor Startup, Step 31.
| |
| NOTE TAVG will vary as a function of reactor power until the unit is greater than 15% turbine load (C5) and the Tavg program is maintained by AUTO or manual rod control. The TAVG-TREF deviation alarm is expected as reactor power approaches 7% RTP.
| |
| RO [31] (p) ADJUST Control Rods or RCS CB to RAISE Reactor power, at a rate of less than 1 dpm, to between 1 and 4%.
| |
| CAUTION IF AFW is controlling levels in one or more SGs, THEN Reactor power must be maintained within AFW capability (less than 4%).
| |
| [32] STABILIZE Reactor power between I and 4%:
| |
| [32.1] MAINTAIN RCS Steam Dumps in Pressure Mode, set at 84% (1092 psig.), or SG PORVs set at 84%.
| |
| RO
| |
| [32.2] () FOLLOW Xenon by Rod movement or Boration to maintain control banks ABOVE the LO INSERTION LIMIT.
| |
| When power is stabilized at 1-4%, cue Console Operator to insert Event 2.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario# 6 Event# 2 Page 2 of 35 Event
| |
| | |
| == Description:==
| |
| 1 -LT-63-50 RWST LEVEL Fails low. Requires entry into ARI I 26-C, RWST LEVEL LO RECIRC INTLK. Requires a Tech Spec evaluation.
| |
| Time Position Applicants Actions or Behavior Indications:
| |
| ARI 126-C, RWST LEVEL LO RECIRC INTLK ARI 126-D, RWST Level LO-LO RO Diagnoses and announces 1-Ll-63-50, RWST LEVEL failed low.
| |
| RO May refer to annunciator response procedure.
| |
| Enters and directs actions of ARI 126-C, RWST LEVEL LO RECIRC SRO INTLK.
| |
| The following actions are taken from ARI 126-C, RWST LEVEL ARI 126-C LO RECIRC INTLK.
| |
| NOTES
| |
| : 1) Alarm may be expected if Unit is in refueling operation.
| |
| : 2) If SI has actuated and containment sump level reaches 16.1% as sensed by 2/4 level switches (See Window 127-E) in conjunction with 2/4 RWST level switches (listed above),
| |
| then 1-FCV-63-72 and 1-FCV-63-73 will begin to OPEN to permit RHR to take suction from containment sump, and l-FCV-74-3 and 1-FCV-74-21 will CLOSE. The status of the RWST and CNTMT level switches is displayed on RXTRIP SI STATUS panel 1-X)(-55-6A.
| |
| [1] CHECK RWST level on 1-Ll-63-50, 1-Ll-63-51, 1-LI-63-52, and RO 1-Ll-63-53 [1-M-6].
| |
| RO identifies 1-LT-63-50 has failed low.
| |
| [2] IF SI has occurred, THEN REFER TO ES-I .3, TRANSFER TO RO RHR CONTAINMENT SUMP.
| |
| [3] IF level increase in RWST is desired, THEN REFER TO SOI RO 62.02, BORON CONCENTRATION CONTROL.
| |
| [4] IF leak is suspected, THEN DISPATCH Operator to RO investigate.
| |
| [5] REFER TO Tech Specs.
| |
| LCO 3.3.2 (ESFAS instrumentation) Function 7.b Automatic Switchover to Containment Sump Refueling SRO Water Storage Tank (RWST) Level Low. Condition Kr LCO 3.3.3 (PAM Instrumentation) Function 23 Refueling Water Storage Tank Water Level. Condition A EXAMINER: If contacted as Work Control, console operator will repeat back request to prepare a package to troubleshoot and repair 1-LT-63-50 and initiate procedure to trip associated bistable as required.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario # 6 Event # 2 Page 3 of 35 Event
| |
| | |
| == Description:==
| |
| 1-LT-63-50 RWST LEVEL Fails low. Requires entry into ARI 126-C, RWST LEVEL LO RECIRC INTLK. Requires a Tech Spec evaluation.
| |
| Time Position I Applicants Actions or Behavior Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| When Tech Spec Evaluation complete cue console operator to insert Event 3.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 1 Page 1 of 34 Event
| |
| | |
| == Description:==
| |
| Continue to startup and raise power to 1-4% using GO-2, Reactor Startup.
| |
| Time Position N Applicants Actions or Behavior The applicants will begin to raise reactor power using GO-2, Reactor Startup, Section 5.3, Reactor Startup, Step 31.
| |
| NOTE TAVG will vary as a function of reactor power until the unit is greater than 15% turbine load (C5) and the Tavg program is maintained by AUTO or manual rod control. The TAVG-TREF deviation alarm is expected as reactor power approaches 7% RTP.
| |
| RO [31] (p) ADJUST Control Rods or RCS CB to RAISE Reactor power, at a rate of less than I dpm, to between 1 and 4%.
| |
| CAUTION IF AFW is controlling levels in one or more SGs, THEN Reactor power must be maintained within AFW capability (less than 4%).
| |
| [32] STABILIZE Reactor power between 1 and 4%:
| |
| [32.1] MAINTAIN RCS Steam Dumps in Pressure Mode, set at 84% (1092 psig.), or SG PORVs set at 84%.
| |
| RO
| |
| [32.2] () FOLLOW Xenon by Rod movement or Boration to maintain control banks ABOVE the LO INSERTION LIMIT.
| |
| When power is stabilized at 1-4%, cue Console Operator to insert Event 2.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 2 Page 2 of 34 Event
| |
| | |
| == Description:==
| |
| 1-LT-63-50 RWST LEVEL Fails low. Requires entry into ARI 126-C, RWST LEVEL LO RECIRC INTLK. Requires a Tech Spec evaluation.
| |
| Time Position Applicants Actions or Behavior Indications:
| |
| ARI 126-C, RWST LEVEL LO RECIRC INTLK ARI 1 26-D, RWST Level LO-LO RO Diagnoses and announces 1-Ll-63-50, RWST LEVEL failed low.
| |
| RO May refer to annunciator response procedure.
| |
| Enters and directs actions of ARI 126-C, RWST LEVEL LO RECIRC INTLK.
| |
| The following actions are taken from ARI 126-C, RWST LEVEL LO RECIRC INTLK.
| |
| NOTES
| |
| : 1) Alarm may be expected if Unit is in refueling operation.
| |
| : 2) If SI has actuated and containment sump level reaches 16.1 % as sensed by 2/4 level switches (See Window 1 27-E) in conjunction with 2/4 RWST level switches (listed above),
| |
| then 1-FCV-63-72 and 1-FCV-63-73 will begin to OPEN to permit RHR to take suction from containment sump, and 1-FCV-74-3 and 1-FCV-74-21 will CLOSE. The status of the RWST and CNTMT level switches is displayed on RXTRIP SI STATUS panel 1-XX-55-6A.
| |
| [1] CHECK RWST level on 1-LI-63-50, 1-LI-63-51, 1-LI-63-52, and RO 1-Ll-63-53 [1-M-6].
| |
| RO identifies 1-LT-63-50 has failed low.
| |
| [2] IF SI has occurred, THEN REFER TO ES-i .3, TRANSFER TO RO RHR CONTAINMENT SUMP.
| |
| [3] IF level increase in RWST is desired, THEN REFER TO SQl RO 62.02, BORON CONCENTRATION CONTROL.
| |
| [4] IF leak is suspected, THEN DISPATCH Operator to RO investigate.
| |
| [5] REFER TO Tech Specs.
| |
| LCO 3.3.2 (ESFAS Instrumentation) Function Tb Automatic Switchover to Containment Sump Refueling SRO Water Storage Tank (RWST) Level Low. Condition Kr LCO 3.3.3 (PAM Instrumentation) Function 23 Refueling Water Storage Tank Water Level. Condition A EXAMINER: If contacted as Work Control, console operator will repeat back request to prepare a package to troubleshoot and repair 1-LT-63-50 and initiate procedure to trip associated bistable as required.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 2 Page 3 of 34 Event
| |
| | |
| == Description:==
| |
| 1-LT-63-50 RWST LEVEL Fails low. Requires entry into ARI 126-C, RWST LEVEL LO RECIRC INTLK. Requires a Tech Spec evaluation.
| |
| Time Position Applicants Actions or Behavior Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| SRO Operations Management Typically Shift Manager.
| |
| Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| When Tech Spec Evaluation complete cue console operator to insert Event 3.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-21 Op Test No.:2 2 Scenario # 6 Event # 3 Page 4 of 34 Event
| |
| | |
| == Description:==
| |
| 16 Condenser Vacuum pump trips. Requires entry into ARI 14-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another pump to be started, using SOl-2&3.O1, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established..
| |
| Time Position Applicants Actions or Behavior Indications:
| |
| Alarm 14-E, M-1 THRU M-6 MOTOR TRIPOUT Diagnoses and announces the trip of the B Condenser Vacuum BOP pump.
| |
| May enter and direct actions of AOl-i 1, Loss of Condenser SRO Vacuum.
| |
| Enters and directs actions of SOl-2&3.O1, Condensate and SRO Feedwater System, Section 8.12, Starting CVPs With Vacuum Established.
| |
| The following actions are taken from ARI 14-E M-1 THRU M-6 ARI 14-E MOTOR TRIPOUT BOP [1] DETERMINE what equipment tripped.
| |
| BOP [2] PLACE control switch for tripped equipment in OFF.
| |
| [3] START spare equipment, as needed.
| |
| BOP If dispatched to check another Condenser Vacuum Pump ready for start, Console operator will acknowledge request and report pupmp is available for start.
| |
| [4] ADJUST plant conditions to compensate for affected equipment, BOP as necessary.
| |
| [5] REFER TO Tech Specs for operability requirements of affected BOP equipment.
| |
| EXAMINER: When Dispatched to check pump breaker locally console operator will report Instantaneous overcurrent relay actuation at breaker. When dispatched to check pump locally, console operator will report odor of overheated insulation. If work control is cntacted to troubleshoot and repair console operator will acknowledge request and initiate work package.
| |
| The following actions are taken from SOl-2&3.O1, Condensate SOI-2&3.O1 and Feedwater System, Sub section 8.12, Starting CVPs With Vacuum Established.
| |
| NOTE
| |
| : 1) This section allows for starting additional condenser vacuum pumps or swapping of vacuum pumps
| |
| : 2) Venting seal water pump is necessary if the pump and/or seal water tank has been drained.
| |
| : 3) NIA the substeps of step 8.12[2] if venting is not required.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario # 6 Event # 3 Page 5 of 34 Event
| |
| | |
| == Description:==
| |
| lB Condenser Vacuum pump trips. Requires entry into ARI 14-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another pump to be started, using SOl-2&3.01, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established..
| |
| Time I Position Applicants Actions or Behavior
| |
| [1] IF necessary, THEN VENT CVPs seal water pump(s) by performing the following: (NIA substeps if venting NOT performed)
| |
| [1 .1] REMOVE cap from the following valve(s):
| |
| * 1-VTV-2-1 045, COND VACUUM WATER PMP A VENT.
| |
| * 1-VTV-2-1 047, COND VACUUM WATER PMP C VENT.
| |
| [1.21 ENSURE water level in seal water tank is above level of vent valve.
| |
| BOP [1 .3] OPEN to vent trapped air from pump, then CLOSE the following valve(s):
| |
| * 1-VTV-2-1045.
| |
| * 1-VTV-2-1047.
| |
| [1.4] REINSTALL cap on the following valve(s)
| |
| * 1-VTV-2-1 045.
| |
| * 1-VTV-2-1047.
| |
| N/As this step.
| |
| [2] START CVP(s) as necessary [1-M-3]: (pumps NOT selected may be N/Ad)
| |
| A. 1-HS-2-171A, COND VACUUM PMP A.
| |
| BOP B. 1-HS-2-176A, COND VACUUM PMP B.
| |
| C. 1-HS-2-181A, COND VACUUM PMP C.
| |
| Will start at least one operable Condenser Vacuum Pump.
| |
| [3] ESTABLISH continuous drain for the in-service pump(s) to reduce ammonia concentration by throttling the drain valve to a position which maximizes drain flow within the capabilities of the drain system and the auto makeup:
| |
| BOP A. 1-DRV-37-572, GLAND SEAL WATER DRAIN PUMP IA. -
| |
| B. l-DRV-37-571, GLAND SEAL WATER DRAIN PUMP lB. -
| |
| C. 1-DRV-37-565, GLAND SEAL WATER DRAIN PUMP 1C. -
| |
| May dispatch AUO to establish drain. AUO will acknowledge request.
| |
| [4] STOP selected pump and place in P-AUTO: (pumps NOT selected may be NIAd)
| |
| A. 1-HS-2-171A, COND VACUUM PMP A.
| |
| BOP B. 1-HS-2-176A, COND VACUUM PMP B.
| |
| C. 1-HS-2-1 81 A, COND VACUUM PMP C.
| |
| Pump previously stopped.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 3 Page 6 of 34 Event
| |
| | |
| == Description:==
| |
| 1 B Condenser Vacuum pump trips. Requires entry into ARI I 4-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another pump to be started, using SOl-2&3.01, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established..
| |
| Time Position Applicants Actions or Behavior (5] ISOLATE continuous drain for any pumps STOPPED:
| |
| A. 1-DRV-37-572, GLAND SEAL WATER DRAIN PUMP 1A. -
| |
| B. I -DRV-37-571, GLAND SEAL WATER DRAIN PUMP I B.
| |
| BOP C. 1-DRV-37-565, GLAND SEAL WATER DRAIN PUMP 1C. -
| |
| May dispatch AUO to isolate drain for tripped pump. AUO will acknowledge request.
| |
| EXAMINER: AOl-I I may not be entered if a Condenser Vacuum pump has been placed in service and vacuum is being maintained.
| |
| If AOl-I I not entered, Cue console operator to insert Event 4 and continue to the next event actions. Otherwise, continue with AOl-I I steps below.
| |
| The following actions are taken from AOl-I I, Loss of AOl-Il Condenser Vacuum.
| |
| CAUTION If steam dumps are lost due to condenser backpressure a turbine/reactor trip from high power may result in secondary and primary safety valve actuation.
| |
| NOTE Reference Appendix A as required for Condenser Vacuum LO-LO and LO Alarm setpoints if ICS graph from Turn On code AOI1 1 is NOT available.
| |
| : 1. MONITOR condenser backpressure is, and will remain, less than 0.1 in. below the Lo-Lo Alarm using ICS Turn On code AOll 1.
| |
| Other evaluation points available:
| |
| : a. ICS Pt. P2265A (C-Zone)
| |
| BOP b. ICS Pt. P11 33A (C-Zone)
| |
| : c. ICS Pt. P2264A (B-Zone)
| |
| : d. ICS Pt. P2263A (A-Zone)
| |
| : e. Cond Back Press Rate of Rise
| |
| : f. Environmental Conditions BOP 2. ENSURE condenser vacuum breaker CLOSED.
| |
| CAUTION If loss of vacuum is due to undesirable atmospheric conditions for cooling tower operation, then use of Supplemental Condenser Circulating Water (SCCW) or operating unit at reduced load may be required.
| |
| BOP 3. ENSURE adequate number of CCW pumps in service.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 3 Page 7 of 34 Event
| |
| | |
| == Description:==
| |
| 1 B Condenser Vacuum pump trips. Requires entry into ARt 1 4-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another pump to be started, using SOI-2&3.01, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established..
| |
| Time Position Applicants Actions or Behavior
| |
| : 4. ENSURE condenser vacuum pump in service.
| |
| BOP Operator will start condenser vacuum pump if not previously performed.
| |
| BOP 5. EVALUATE use of SCCW, if time permits.
| |
| : 6. CHECK gland seal steam pressure approximately 115 psig, 1-Pl-BOP 47-1 87, STEAM SEAL SUP PRESS [1-M-2].
| |
| : 7. CHECK cooling tower basin level NORMAL (164-E DARK BOP
| |
| [1-M- 5]).
| |
| : 8. CHECK condenser air in-leakage less than 30 SCFM [computer point_F2700A].
| |
| NOTE Condenser water box T below normal indicates tube fouling or Amertap system failure. tiT and tube tP above normal indicates tube sheet fouling.
| |
| : 9. CHECK the condenser tube and water box parameters NORMAL:
| |
| : a. tiT (38°F at 100% load) 1-Tl-27-58, -68, -74, -84 [1-M-15]
| |
| : b. tP (less than 17 psid) 162-C and 163-C DARK [1 -M-1 51 NOTE If low vacuum is due to high outside air temperature, DG operability should be evaluated. 1-Sl 2-00, Shift and Daily Surveillance Log Master, may be referenced to determine DG ventilation limitations with respect to outside air temperature.
| |
| : 10. NOTIFY Unit SRO to EVALUATE the combination of high outside air temperature vs. any inopreable DG ventilation fans.
| |
| : 11. INITIATE repairs as required.
| |
| : 12. CHECK condenser backpressure less than Lo Alarm.
| |
| NOTE The following step is to avoid cycling turbine load by limiting load increases based upon plant and environmental conditions (e.g.; the weather forecast). Operations Manager approval indicates that this has been evaluated for the load increase.
| |
| : 13. IF turbine load is to be increased, THEN OBTAIN Operations Manager approval prior to increasing Turbine Load.
| |
| : 14. RETURN TO Instruction in effect.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 3 Page 8 of 34 Event
| |
| | |
| == Description:==
| |
| lB Condenser Vacuum pump trips. Requires entry into ARI 14-E, M-1 THRU M-6 MOTOR TRIPOUT. Requires another pump to be started, using SOI-2&3.O1, Condensate and Feedwater Systems, Section 8.12, Starting CVPs With Vacuum Established..
| |
| Time Position Applicants Actions or Behavior Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| Cue Console Operator to insert Event 4, if not already entered.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Lppendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 4 Page 9 of 34 Event
| |
| | |
| == Description:==
| |
| 1A ERCW header leak in the Auxiliary Building. Requires entry into Aol-I 3, Loss of Essential Raw Cooling Water, Section 3.3, Supply Header Rupture in Auxiliary Building. Requires a Tech Spec evaluation.
| |
| Time Position Applicants Actions or Behavior Indications:
| |
| 167-D, TURB/AUXIRX BLDG FLOODED.
| |
| 223-A, ERCW HDR A SUP PRESS LO.
| |
| 1-Fl-67-81 IA ERCW SUP HDR FLOW indicating rising flow.
| |
| BOP Responds to alarms and evaluates ERCW system status.
| |
| SRO Enters AOI-13 and determines that Section 3.3 is applicable.
| |
| The following actions are taken from AOl-I 3, Loss of Essential AOI-13 Raw Cooling Water, Section 3.3, Supply Header Rupture in Aux Bldg.
| |
| : 1. DISPATCH personnel to determine location of rupture.
| |
| BOP contacts an AUO and dispatches the AUO to the Auxiliary Building to locate the leak. Console operator will acknowledge BOP the request and report a leak from a pipe above the TDAFW Pump room.
| |
| BOP may contact the Outside AUO and dispatch the Outside AUO to the IPS to monitor/inspect the ERCW pumps.
| |
| : 2. DISPATCH AUO, with a radio, to the Rx MOV Bds.
| |
| BOP BOP contacts an AUO and dispatches the AUO to Reactor MOV boards.
| |
| CAUTION MOVs with power normally removed may not travel to full closed position under high flow conditions, local verification of isolation may be required.
| |
| : 3. CHECK Supply Header 2A flow at expected value for existing plant conditions.
| |
| BOP BOP determines that Supply Header 2A flow is at expected value for existing plant conditions.
| |
| : 4. CHECK Supply Header 2B flow at expected value for existing plant conditions.
| |
| BOP BOP determines that Supply Header 2B flow is at expected value for existing plant conditions.
| |
| : 5. CHECK Supply Header IA flow at expected value for existing plant conditions.
| |
| BOP BOP determines that Supply Header IA flow is NOT at expected value for existing plant conditions and SRO enters the RNO column for actions.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 4 Page 10 of 34 Event
| |
| | |
| == Description:==
| |
| IA ERCW header leak in the Auxiliary Building. Requires entry into Aol-i 3, Loss of Essential Raw Cooling Water, Section 3.3, Supply Header Rupture in Auxiliary Building. Requires a Tech Spec evaluation.
| |
| Time Position Applicants Actions or Behavior
| |
| : 5. RESPONSE NOT OBTAINED:
| |
| PERFORM the following:
| |
| : a. UNLOCK, and CLOSE bkr on Rx MDV Bd 1A2-A cI8A, 1-FCV-67-81.
| |
| BOP contacts the AUO stationed at the Reactor MOV boards and requests breaker manipulation.
| |
| : b. CLOSE 1-FCV-67-81, AB Supply Hdr 1A.
| |
| BOP locates 1-HS-67-81, AB Supply Hdr IA and rotates the switch to the left to the CLOSE position.
| |
| : c. ENSURE O-FCV-67-208, C&SS Compr Sup From Hdr I B, is OPEN.
| |
| BOP BOP locates O-FCV-67-208 C&SS Compr Sup From Hdr lB. and determines OPEN by RED indicating light LIT, GREEN indicating light DARK.
| |
| : d. ENSURE 1-FCV-67-i47, CCS Hx C Sup From Hdr 1A, is CLOSED.
| |
| BOP locates I-FCV-67-147, CCS Hx C Sup From Hdr IA, observes the PD/C tag (power disconnected closed.) BOP observes that both the RED and GREEN indicating lights are DARK.
| |
| : e. STOP ERCW pumps as required to control header pressure.
| |
| SRO may direct the BOP to stop ERCW pumps based on high header pressure.
| |
| NOTE For a rupture upstream of i-FCV-67-81, water will be coming from Ui ERCW tunnels on El 692.
| |
| IF rupture upstream of 1-FCV-67-81, THEN **GO TO Section 3.4 Step 5 to isolate yard header.
| |
| BOP GO TO Step 7.
| |
| BOP requests a report from the AUO in the Aux Bdlg. AUO reports that leakage is stopped.
| |
| : 7. REFER TO Tech Specs:
| |
| * 3.4.6, RCS Loops-Mode 4.
| |
| N/A
| |
| * 3.7.8, Essential Raw Cooling Water System (ERCW).
| |
| SRO Condition A, One ERCW train inoperable, restore ERCW train to OPERABLE status within 72 hours.
| |
| * 3.8.1, AC Sources-Operating.
| |
| N/A
| |
| : 8. CHECK ERCW flow adequate to C&SS Air Compressors.
| |
| BOP Previously verified.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario # 6 Event # 4 Page 11 of 34 Event
| |
| | |
| == Description:==
| |
| IA ERCW header leak in the Auxiliary Building. Requires entry into AOl-I 3, Loss of Essential Raw Cooling Water, Section 3.3, Supply Header Rupture in Auxiliary Building. Requires a Tech Spec evaluation.
| |
| [ Time 1 Position Applicants Actions or Behavior ]
| |
| : 9. INITIATE repair.
| |
| SRO contacts Work Control and requests a troubleshooting SRO and repair package for the ERCW leak.
| |
| If SRO requests a RISK evaluation, Console Operator will acknowledge the request.
| |
| : 10. IF ERCW to in-service CCS heat exchanger was interrupted, SRO THEN NOTIFY Duty System Engineer to initiate evaluation for effect on CCS equipment and piping.
| |
| : 11. EVALUATE flooding on equipment operability.
| |
| SRO SRO requests reports from the AUO in the Aux Bdlg and determines that no additional equipment has been affected.
| |
| : 12. REFER TO SOl-67.01, Essential Raw Cooling Water System, SRO for system realignment.
| |
| SRO 13. RETURN TO Instruction in effect.
| |
| EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| Cue Console Operator to insert Event 5, if not already entered.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Lppendix D Required Operator Actions Form ES-DJ Op Test No.:2 2 Scenario # 6 Event # 5 Page 12 of 34 Event
| |
| | |
| == Description:==
| |
| 1-ISV-62-932, Dilution to CCP Suction is inadvertently opened locally. Requires entry into AOl-3, Malfunction of Reactor Makeup Control. Requires RO to manually stop the primary water pumps.
| |
| Time I Position Applicants Actions or Behavior Indications:
| |
| Flow visible on 1-Fl-62-142 PW TO BLENDER FLOW Audible indication from 1-FQ-62-142 PW BATCH COUNTER RO Diagnoses and announces dilution flow in progress.
| |
| RO May stop the running primary water pump.
| |
| May dispatch an AUO to the blender area to determine valve RO alignment and to check for leaks.
| |
| Enters and directs actions of AOI-3, Malfunction of Reactor Makeup SRO Control, Sub section 3.2, Inadvertent Dilution.
| |
| The following actions are taken from AOI-3, Malfunction of AOI-3 Reactor Makeup Control, Sub section 3.2, Inadvertent Dilution.
| |
| : 1. PERFORM the following:
| |
| : a. CHECK PWST in normal alignment (PWST NOT in Bypass Mode).
| |
| : b. ENSURE standby Primary Water Pump HS in MAN.
| |
| RO locates 1-HS-81-7A PRIMARY WATER PUMP B and observes that the RED indicating light is DARK and the GREEN indicating light is Lfl and the handswitch is in the pushed-in MAN RO position.
| |
| : c. STOP the running Primary Water Pump.
| |
| RO locates 1-HS-81-3A PRIMARY WATER PUMP A and rotates the handswitch to the left to the STOP position.
| |
| : d. STOP the Waste Gas Compressors by placing Handswitches in STOP/PULL-TO-LOCK.
| |
| RO contacts an AUO to locally stop the Waste Gas Compressors.
| |
| : 2. ENSURE primary water flow to blender isolated:
| |
| RO a. CLOSE 1-FCV-62-143, PW To Blender.
| |
| : b. CHECK 1-Fl-62-142, PW to Blender Flow, ZERO.
| |
| EXAMINER: After Primary water pump is stopped, PW to Blender Flow will indicate ZERO however, SRO may implement RNO based on initial conditions of the event.
| |
| : 2. RESPONSE NOT OBTAINED:
| |
| Locally CLOSE 1-ISV-62-933, CVCS BA Blender PW Supply Isol RO
| |
| [A3V/71 3].
| |
| RO dispatches AUO to close Locally CLOSE 1-IS V-62-933 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 5 Page 13 of 34 Event
| |
| | |
| == Description:==
| |
| 1-ISV-62-932, Dilution to CCP Suction is inadvertently opened locally. Requires entry into AOl-3, Malfunction of Reactor Makeup Control. Requires RO to manually stop the primary water pumps.
| |
| Time I Position I Applicants Actions or Behavior CAUTION Charging flow path should be isolated when Letdown is taken out of service.
| |
| : 3. PLACE 1-HS-62-79A, Ltdn Hi Temp Divert to VC TK.
| |
| RO RO locates 1-HS-62-79A, LTDN HI TEMP DIVERT TO VC TK and rotates the handswitch to the left to the VC TK position.
| |
| NOTE A letdown temperature drop can reduce RCS boron concentration by changing the demin bed boron equilibrium.
| |
| : 4. CHECK letdown temp stable:
| |
| * 1 -Tl-70-1 91, LTDN HX RET TEMP.
| |
| RO locates 1-TI-70-191, LTDN HX RET TEMP on panel 1-M-6 and determines that temperature is stable.
| |
| * 1-Tl-62-78, Letdown HX Outlet Temp.
| |
| RO RO locates 1-Tl-62-78, Letdown HX Outlet Temp on panel 1-M-6 and determines that temperature is stable.
| |
| * 1-11-62-131, VCT Outlet Temp.
| |
| RO locates 1-Tl-62-131, VCT Outlet Temp. on panel 1-M-6 and determines that temperature is stable.
| |
| : 5. CONSIDER containment evacuation based on changing plant SRO conditions.
| |
| SRO 6. CHECK Unit in Mode 6.
| |
| : 6. RESPONSE NOT OBTAINED:
| |
| SRO GO TO NOTE prior to Step 10.
| |
| NOTE Time should be allowed for plant to stabilize after dilution event.
| |
| SRO 10. CHECK Unit in Mode 1 or 2.
| |
| : 11. CHECK dilution terminated:
| |
| * Tavg STABLE or DROPPING.
| |
| SRO
| |
| * Automatic rod insertion stopped.
| |
| * Rx power STABLE or DROPPING.
| |
| : 12. MONITOR ROD INSERTION LIMIT LO-LO, annunciator [87-B],
| |
| RO DARK.
| |
| 13; STABILIZE Tavg/Tref within 3°F:
| |
| * ADJUST RCS CB as required by plant conditions:
| |
| RO REFER TO SOI-62.02, Boron Concentration Control. OR
| |
| * ADJUST turbine load as required by plant conditions.
| |
| 2011-10 Wafts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario # 6 Event# 5 Page 14 of 34 Event
| |
| | |
| == Description:==
| |
| 1-ISV-62-932, Dilution to CCP Suction is inadvertently opened locally. Requires entry into AOl-3, Malfunction of Reactor Makeup Control. Requires RO to manually stop the primary water pumps.
| |
| Time Position Applicants Actions or Behavior SRO 14.GOTOStepl7.
| |
| : 17. REFER TO Tech Specs:
| |
| * 3.1.1, Shutdown Margin (SDM) Tavg > 200°F. not applicable SRO
| |
| * 3.1 .2, Shutdown Margin (SDM) - Tavg 200°F. not applicable
| |
| * 3.1.7, Control Bank Insertion Limits. not applicable
| |
| * 3.2.3, Axial Flux Difference (AFD). not applicable CAUTION If all RCPs are stopped, the potential for low boron concentration pockets must be evaluated per S0l-68.02 prior to the start of the first RCP.
| |
| : 18. DETERMINE cause of dilution, AND INITIATE Corrective action.
| |
| SRO AUO at the blender area reports that 1-IS V-62-932 was found in the OPEN position. Workers in the area are responsible for the misalignment.
| |
| EXAMINER: When directed by SRO/RO AUO will close valve 1-ISV-62-932 to allow restoration of primary makeup system to normal in subsequent steps.
| |
| : 19. NOTIFY Operations Duty Manager, Plant Manager, and Rx Engineering or inadvertent dilution.
| |
| SRO SRO contacts the Operations Duty Manager, Plant Manager, and Rx Engineering to report the inadvertent dilution.
| |
| : 20. WHEN repairs are complete, THEN:
| |
| : a. CHECK PWST in normal alignment (PWST NOT in Bypass Mode).
| |
| : b. START one Primary Water Pump.
| |
| RO locates 1-HS-81-3A PRIMARY WATER PUMP A or 1-HS TA PRIMARY WATER PUMP B and rotates the selected handswitch to the right to the START position.
| |
| : c. ENSURE second Primary Water Pump in MAN.
| |
| RO RO locates 1-HS-81-3A PRIMARY WATER PUMP A or 1-HS TA PRIMARY WATER PUMP B and observes that the RED indicating light is DARK and the GREEN indicating light is LIT, and the handswitch is in the pushed-in MAN position.
| |
| : d. PLACE CVCS and CVCS makeup control, CVCS Purification, and Waste Gas Disposal in normal alignment:
| |
| * REFER TO SOl-62.02, Boron Concentration Control.
| |
| * REFER TO S01-62.04, CVCS Purification System.
| |
| * REFER TO SOI-77.02, Waste Gas Disposal System.
| |
| SRO 21. RETURN TO Instruction in effect.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario# 6 Event# 5 Page 15 of 34 Event
| |
| | |
| == Description:==
| |
| 1-ISV-62-932, Dilution to CCP Suction is inadvertently opened locally. Requires entry into AOl-3, Malfunction of Reactor Makeup Control. Requires RD to manually stop the primary water pumps.
| |
| Time Position Applicants Actions or Behavior EXAMINER: The crew briefing is optional. The next event may be entered prior to the brief.
| |
| Crew Brief would typically be conducted for this event as time allows SRO prior to the next event.
| |
| Notifications should be addressed as applicable if not specifically addressed by the procedure or in the crew brief.
| |
| Operations Management Typically Shift Manager.
| |
| SRO Maintenance Personnel Typically Work Control Center (WCC).
| |
| (Note: Maintenance notification may be delegated to the Shift Manager).
| |
| Cue Console Operator to enter Event 6 if not already entered.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 6 Page 16 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOl-6, Small Reactor Coolant System Leak.
| |
| Time N Position Applicants Actions or Behavior Indications:
| |
| Alarm [89-B] PZR SAFETY LINE TEMP HI Temp Indicator 1-TI-68-329 Safety 68-564 Tailpipe Temp shows elevated temperature.
| |
| PZR Valves Acoustic Monitor 1 -XI-68-364 (68-564) indicates flow.
| |
| RO Respond to Alarm and evaluate RCS parameters.
| |
| RO Diagnose open Pressurizer Safety valve.
| |
| SRQ Enter and direct actions of AC 1-18 or AC 1-6.
| |
| EXAMINER: Crew may respond to this failure using either AOl-I 8, Malfunction Of Pressurizer Pressure Control System, Section 3.2, PZR pressure drop due to failed PORVISafety or spray valve, or AOl-6, Small Reactor Coolant System Leak. Event guide for both procedures are included.
| |
| The following actions are taken from ARI 89-B PZR SAFETY ARI 89-B LINE TEMP HI.
| |
| [1] DETERMINE which valve is lifting/leaking through by monitoring 1-TI-68-330 -329, and -328 [1-M-4], and Acoustic Monitors [O-M RO 25].
| |
| RO should identify leaking Safety valve 1-SV-68-564.
| |
| RO [2] MONITOR RCS pressure.
| |
| [3] MONITOR the following for indication of leakage:
| |
| * 1-Ll-68-300, PRT LEVEL RO
| |
| * 1-PI-68-301, PRT PRESS
| |
| * 1-TI-68-309, PRT TEMP SRO [4j REFER TO AOl-6.
| |
| [5] REFER TO Tech Specs.
| |
| SRO SRO may defer Tech Specs evaluation at this time.
| |
| AOI-18 The following actions are taken from AOl-I 8, Malfunction Of Pressurizer Pressure Control System, Section 3.2, PZR pressure drop due to failed PORVISafety or spray valve.
| |
| NOTE Step I and 2 are IMMEDIATE ACTION steps 2011-10 Watts Bar NRC Examination
| |
| | |
| I Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario# 6 Event# 6 Page 17 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOI-6, Small Reactor Coolant System Leak.
| |
| Time I Position Applicants Actions or Behavior
| |
| : 1. CHECK PZR spray valves CLOSED:
| |
| * Green indicating lights IT RO determines 1-XI-68-340B, PZR Spray Loop 2 and 1-Xl 340D, PZR Spray Loop 1 not lit (1-M-4J.
| |
| RO
| |
| * Pzr spray demand meters, 1-PIC-68-340B and 1-PIC-68-340D indicating ZERO [1-M-4]
| |
| RO determines listed PZR Spray controller demands are Zero.
| |
| : 2. CHECK PZR PORVs CLOSED
| |
| * PORV indicating lights RO observes 1-HS-68-340AA, PZRPORV34OA, CLOSED GREEN indicating light is LIT, RED indicating light is DARK.
| |
| RO observes 1-HS-68-334A, PZR PORV 334, CLOSED, GREEN indicating light is LIT, RED indicating light is DARK.
| |
| RD
| |
| * tailpipe temperature RO determines 1-Tl-68-331, PORV34OA & 334 TAILPIPE TEMP temperature is not elevated.
| |
| acoustic monitoring BOP may observe 1-Xl-68-363, PZR VALVES ACOUSTIC MONITOR indicating lights are NOT LIT for 1-XI-68-340 1-XI 334.
| |
| : 3. VERIFY actions taken in Steps I and 2 have STOPPED press RD drop.
| |
| RO determines pressurizer pressure still dropping slowly.
| |
| : 3. RESPONSE NOT OBTAINED:
| |
| ENERGIZE all PZR heaters.
| |
| RO may take PZR Press Master Control 1-PIC-68-340A to manual and decrease output to zero to energize all heaters.
| |
| IF low pressure reactor trip (1970) is IMMINENT, THEN PERFORM the following:
| |
| SRO
| |
| : a. TRIP Rx
| |
| : b. ENSURE RCP alternate bkr in MAN for the affected loop, and STOP RCP(s) supplying any stuck open spray valve.
| |
| : c. **GO TO E-0, Reactor Trip or Safety Injection.
| |
| EVALUATE continued plant operation.
| |
| SRO determines plant operation may continue at this time.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 6 Page 18 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOl-6, Small Reactor Coolant System Leak.
| |
| Time Position Applicants Actions or Behavior
| |
| : 4. CHECK PZR Safeties CLOSED:
| |
| * tailpipe temperatures RO observes 1-TI-68-329, SAFETY 68-564 TAILPIPE TEMP, temperature elevated and determines that 1-SV-68-564 is not RD closed.
| |
| * acoustic monitor BOP may also observe 1-Xl-68-363, PZR VALVES ACOUSTIC MONITOR indicating lights are LIT for 1-XI-68-364 (68-564) and determine that 1-SV-68-564 is not closed (O-M-25J.
| |
| : 4. RESPONSE NOT OBTAINED:
| |
| IF PZR press can NOT be maintained above 1 970 psig, THEN:
| |
| SRO a. TRIP Rx and INITIATE SI.
| |
| : b. **GO TO E-O, Rx Trip or Safety Injection.
| |
| EVALUATE continued plant operation.
| |
| SRO determines plant operation may continue at this time.
| |
| : 5. ENSURE PZR heaters on as required:
| |
| * Control Group on at 2220 psig RD
| |
| * Backup Groups on at 2210 psig RO turns heaters off and on manually or with master pressure controller as required.
| |
| : 6. CHECK aux spray, 1-FCV-62-84, CLOSED.
| |
| RO observes 1-HS-62-84A f1-M-6J and determines aux spray valve is closed.
| |
| : 7. CHECK PZR press STABLE or RISING.
| |
| RD RO determines pressure has stabilized.
| |
| : 8. WHEN pressurizer pressure stable and equipment status supports returned to normal, THEN ENSURE the following in AUTO:
| |
| * PZR Master controller RO
| |
| * PZR spray controllers
| |
| * All heater groups RO evaluates pressurizer pressure and may continue in auto or manual control at this time.
| |
| EXAMINER: At this point, the safety valve failure worsens, and the crew is expected to initiate a manual reactor trip and Safety injection based deceasing RCS pressure approaching Reactor Trip Setpoint and Enter E-O, Reactor Trip or Safety Injection.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario # 6 Event # 6 Page 19 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOI-6, Small Reactor Coolant System Leak.
| |
| Time Position Applicants Actions or Behavior The following actions are taken from AOI-6, Small Reactor AOI-6 Coolant System Leak.
| |
| NOTE During performance of this instruction the need for rapid load reduction or Unit trip should be continuously evaluated.
| |
| RO 1. CHECK CCP in service.
| |
| : 2. ADJUST 1 -FCV-62-89 AND 1 -FCV-62-93 as necessary to maintain pzr level on program.
| |
| RO RO may manual control of 1-FCV-62-89 and 93 to control pzr level on program.
| |
| RO 3. CHECK letdown flow in service for 75 gpm.
| |
| : 4. INCREASE VCT Auto makeup:
| |
| : a. DOUBLE setting on Boric Acid and PW water flow controllers
| |
| * 1-FC-62-.142 RO doubles the PWsetting from 35% to 70% on 1-FC-62-142.
| |
| RO
| |
| * I -FC-62-1 39 RO doubles the BA setting on 1-FC-62-139.
| |
| : b. ENSURE RED light lit on 1-HS-62-140A.
| |
| RO observes 1-HS-62-140A, MAKEUP CONTROL RED indicating light is LIT and GREEN indicating light is DARK.
| |
| SRO 5. CHECK in Modes I through 3 NOTE Pzr level must be allowed time to change following changes in charging flow.
| |
| : 6. MONITOR the following parameters:
| |
| * Pzr level STABLE or RISING.
| |
| RO
| |
| * Containment pressure STABLE or DROPPING.
| |
| * RCS pressure STABLE or RISING.
| |
| : 6. RESPONSE NOT OBTAINED:
| |
| IF any of the following occur:
| |
| * Loss of pzr level is IMMINENT,
| |
| * Containment pressure is approaching 1.5 psig.
| |
| SRO
| |
| * RCS pressure is approaching 1970 psig (dropping), THEN
| |
| : 1) (p)TRIP Rx.
| |
| : 2) INITIATE SI.
| |
| : 3) ** GO TO E-0, Rx Trip or Safety Injection.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario # 6 Event # 6 Page 20 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOl-6, Small Reactor Coolant System Leak.
| |
| Time Position Applicants Actions or Behavior
| |
| : 7. CHECK secondary plant radiation NORMAL:
| |
| * Condenser exhaust monitors.
| |
| RO
| |
| * S/G blowdown monitors.
| |
| * Main steam line monitors.
| |
| : 8. MAKE plant announcement via PA:
| |
| RO Attention plant personnel. A primary system leak has developed.
| |
| Any personnel located either inside containment or in the Auxiliary Building should exit the area immediately. (Repeat)
| |
| CAUTION Attempts to quantify leak rate should not delay performance of the remaining steps.
| |
| NOTE Appendix B may be used to estimate RCS leak rate.
| |
| : 9. INITIATE leak rate estimate:
| |
| : a. STOP any heatup/cooldown in progress.
| |
| RO
| |
| : b. ADJUST charging flow to STABALIZE pzr level.
| |
| : c. CHECK Net Charging on ICS (UFIO16)
| |
| : 10. CHECK safety valves CLOSED:
| |
| RO
| |
| * EVALUATE tailpipe temp and acoustic monitors.
| |
| RO determines that a 1-SV-68-564 is not closed.
| |
| : 10. RESPONSE NOT OBTAINED:
| |
| SRO EVALUATE need for shutdown.
| |
| REFER TO Tech Spec 3.4.10, Pressurizer Safety Valves.
| |
| : 11. CHECK PORVs CLOSED:
| |
| R
| |
| * EVALUATE tailpipe temp and acoustic monitors.
| |
| NOTE Relief valves (pzr PORVs, pzr safeties, CVCS letdown, RHR suction, and SI lines), and Rx head vent isolation valves could be leaking to the PRT. Further investigation will have to be made if PRT conditions become abnormal and leakage path is not readily identifiable.
| |
| : 12. MONITOR PRT conditions NORMAL:
| |
| * Level.
| |
| * Temperature.
| |
| RO
| |
| * Press.
| |
| RO determines that PRT conditions are not normal due to leaking pzr safety valve.
| |
| 2011-10 Watts Bar NRC Examination
| |
| * Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario# 6 Event# 6 Page 21 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails partially open. Requires entry into AOl-6, Small Reactor Coolant System Leak.
| |
| Time Position Applicants Actions or Behavior
| |
| : 12. RESPONSE NOT OBTAINED:
| |
| IF PRT diaphragm rupture is IMMINENT (85 psig), THEN:
| |
| : a. (p)TRIP Rx.
| |
| RO b. GO TO E-O, Reactor Trip or Safety Injection.
| |
| RESTORE PRT to normal:
| |
| : a. REFER TO SOl-68.O1 WHILE attempting to determine leak path into PRT.
| |
| NOTE If leak is on CVCS, pzr level should begin rising with charging and letdown isolated. Appendix B can be used to verify small leaks are isolated.
| |
| : 13. ISOLATE letdown:
| |
| * CLOSE 1FCV-62-72, (45 gpm).
| |
| CLOSE 1-FCV-62-73, (75 gpm).
| |
| RO
| |
| * CLOSE 1-FCV-62-74, (75 gpm).
| |
| . CLOSE 1-FCV-62-76, (5 gpm).
| |
| * CLOSE 1-FCV-62-69.
| |
| * CLOSE 1-FCV-62-70.
| |
| EXAMINER: After Letdown is isolated in the next step, the safety valve failure worsens, and the crew is expected to initiate a manual reactor trip based on Step 6 RESPONSE NOT OBTAINED conditions OR Step 12 RESPONSE NOT OBTAINED conditions.
| |
| : 14. ISOLATE charging:
| |
| * CLOSE 1-FCV-62-85.
| |
| RO
| |
| * CLOSE 1-FCV-62-86.
| |
| * CLOSE 1-FCV-62-90.
| |
| * CLOSE 1-FCV-62-91.
| |
| Cue console operator to increase pzr safety valve leak severity.
| |
| NOTE Normal range of seal injection flow is between 8 and 13 gpm per RCP with a minimum allowed flow of 6 gpm.
| |
| : 15. MINIMIZE RCP seal injection flow (greater than 6 gpm per RO pump), and EVALUATE pzr level trend.
| |
| NOTE Pzr level must be allowed time to change following changes in charging flow.
| |
| RO 16. CHECK pzr level DROPPING or STABLE.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I OpTestNo.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 22 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1 High Containment Pressure.
| |
| Time I Position Applicants Actions or Behavior The following actions are taken from E-O, Reactor Trip or Safety Injection.
| |
| NOTE
| |
| . Steps 1 thru 4 are IMMEDIATE ACTION STEPS.
| |
| Status Trees I SPDS should be monitored when transitioned to another instruction.
| |
| : 1. ENSURE reactor trip:
| |
| * Reactor trip and bypass breakers OPEN.
| |
| RO checks 1-52RTB, RX TRIP BKR A GREEN indicating light LIT on panel l-M-4.
| |
| RO checks 1-52RTB, RC TRIP BKR B GREEN indicating light LIT on panel 1-M-4 RO checks 1-52BYA, BYPASS BKR A lights DARK RO checks 1-52BYB, BYPASS BKR B lights dark RO
| |
| * RPIs at bottom of scale.
| |
| RO observes 1-MON 85 5000/1 CERPI Monitor I and 1-MON-85 5000/2 CERPI MONITOR 2 for indication that all SHUTDOWN and CONTROL bank rods are inserted.
| |
| * Neutron flux DROPPING.
| |
| RO observes neutron flux trending down on 1-NR-92-145, NEUTRON FLUX LEVEL RECORDER. May also observe levels decreasing on 1-Nl-92-135A, CH I NEUTRON MON % PWR, and 1-NI-92-136A, CH II NEUTRON MON % PWR.
| |
| : 2. ENSURE Turbine Trip:
| |
| * All turbine stop valves CLOSED.
| |
| RO RO observes that indicating lights on 1-XX-47-1000 EHC CONTROL for individual throttle and governor valves are GREEN.
| |
| : 3. CHECK 6.9 kV shutdown boards:
| |
| : a. At least one board energized from:
| |
| RO CSST (offsite),
| |
| OR DIG (blackout).
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 23 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. IA Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.I, High Containment Pressure.
| |
| Time I Position Applicants Actions or Behavior
| |
| : 4. CHECK SI actuated:
| |
| : a. Any SI annunciator LIT.
| |
| RO will announce that the window 70-A, SI ACTUATED is LIT.
| |
| RO
| |
| : b. Both trains SI ACTUATED.
| |
| * 1 -XX-55-6C
| |
| * 1 -XX-55-6D EXAMINER: E-O, Reactor Trip or Safety Injection, Appendixes A and B are included as Attachment 1.
| |
| EXAMINER: Critical Task I is identified in Attachment I (E-O Appendix A Step 5). If not previously performed, the BOP will complete Critical Task I during the performance of E-O Appendix A.
| |
| BOP may manually initiate at least one train of Containment Phase A isolation prior to (E 0 Appendix A Step 5).
| |
| Critical Task I Manually initiate at least one train of Containment Phase A isolation prior to completion of E-0, Reactor Trip or Safety Injection, Appendix A, Equipment Verification.
| |
| Critical 5. PERFORM Appendixes A and B, E-0, pages 16-30.
| |
| Task I BOP is assigned to perform actions contained in the Appendices.
| |
| BOP A separate copy of the Appendices is contained in this package for Examiner use.
| |
| SRO 6. ANNOUNCE reactor trip and safety injection over PA system.
| |
| : 7. ENSURE secondary heat sink available with either:
| |
| * Total AFW flow greater than 410 gpm, RO OR
| |
| * At least one SIG NR level greater than 29% [39% ADV].
| |
| : 8. MONITOR RCS temperature stable at or trending to 557°F using:
| |
| * RCS Loop T-avg with any RCP running, RO OR
| |
| * RCS Loop T-cold with RCPs out-of-service.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario 4 6 Event # 7, 8, and 9 Page 24 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 8. RESPONSE NOT OBTAINED IF temp less than 557°F, THEN ENSURE steam dumps and SIG PORVs CLOSED.
| |
| IF cooldown continues, THEN CONTROL total AFW flow to maintain greater than 410 gpm UNTIL NR level in at least one SIG greater than 29% [39% ADV].
| |
| BOP takes manual control of AFW LCVs and reduces AFW flow.
| |
| IF cooldown continues after AFW flow is controlled, THEN
| |
| * PLACE steam dump controls OFF.
| |
| * CLOSE MSIVs.
| |
| * ENSURE MSIV bypasses CLOSED.
| |
| IF RCS temp greater than 564°F, THEN ENSURE either steam dumps or SIG PORVs OPEN.
| |
| If Chemistry or Rad Protection contacted to survey or sample secondary lines, Console Operator will acknowledge the request.
| |
| : 9. ENSURE excess letdown valves CLOSED:
| |
| * 1 -FCV-62-54 RO
| |
| * 1 -FCV-62-55 RO observes GREEN indicating lights LIT on handswitches 1-HS-62-54A, EXCESS LTDN ISOL, and 1-HS-62-55A, EXCESS LTDN.
| |
| : 10. CHECK pzr PORVs and block valves:
| |
| : a. Pzr PORVs CLOSED.
| |
| : b. At least one block valve OPEN.
| |
| RO observes 1-HS-68-340AA, PZR PORV34OA, CLOSED, GREEN indicating light is LIT RED indicating light is DARK.
| |
| RO RO observes 1-HS-68-334A, PZR PORV 334, CLOSED, GREEN indicating light is LIT, RED indicating light is DARK.
| |
| RO observes 1-HS-68-333A, BLOCK VLV FOR PORV 340A, OPEN, GREEN indicating light is DARK, RED indicating light is LIT.
| |
| RO observes 1-HS-68-332A, BLOCK VLV FOR PORV 334, OPEN GREEN indicating light is DARK, RED indicating light is LIT.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 25 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. IA Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 11. CHECK pzr safety valves CLOSED:
| |
| * EVALUATE tailpipe temperatures and acoustic monitors.
| |
| RO observes response of 1-TI-68-330, SAFETY 68-563 TAILPIPE RO TEMP. 1-Tl-68-329, SAFETY 68-564 TAILPIPE TEMP and 1-TI 328, SAFETY 68-565. All are elevated, with 1-Tl-68-329, SAFETY 68-564 TAILPIPE TEMP, indicating the highest temperature.
| |
| BOP may observe 1-XI-68-363, PZR VALVES ACOUSTIC MONITOR indicating lights are LIT for 1-XI-68-364 (68-564).
| |
| : 11. RESPONSE NOT OBTAINED:
| |
| RO IF RCS pressure less than 2485 psig, AND Pzr safety valve open, THEN *k GO TO E-1, Loss of Reactor or Secondary Coolant.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 26 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous overcurrent when started. lB-B 6.9KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.l, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior EXAMINER: Status tree monitoring is in effect when transition to E-1 occurs. 4 minutes after Phase B occurs a transition to FR-Z.1 will be indicated. Refer to page 30 for applicable steps from FR-Z.1.
| |
| The following actions are taken from E-1, Loss of Reactor or Secondary Coolant.
| |
| EXAMINER: As the RCS leak becomes larger, containment temperature and pressure will rise sharply. 265-A UPPER CNTMT RE-2711272 RAD HI and 265-B LOWER CNTMT RE-2731274 are expected to alarm, since testing has shown rad monitor to give unreliable indication for up to 2 minutes following a rapid increase or decrease in containment temperature. The alarms will clear after the initial temperature transient.
| |
| NOTE Seal injection flow should be maintained to all RCPs.
| |
| : 1. CHECK if RCPs should remain in service:
| |
| RO a. Phase B DARK [MISSP].
| |
| : b. RCS pressure greater than 1500 psig.
| |
| Critical Task 2 Trip the RCPs prior to completing E-1, Loss of Reactor or Secondary Coolant.
| |
| Critical 1. b. RESPONSE NOT OBTAINED Task 2 RO ENSURE at least one Charging pump or SI pump injecting.
| |
| WHEN injection flow established, THEN STOP all RCPs.
| |
| SRO 2. REFER TO EPIP-1, Emergency Plan Classification Flowchart.
| |
| NOTE Time since initiation of event is defined by performance of Step 3.
| |
| : 3. RECORD current time to mark initiation of LOCA and SRO determination of time for hot leg recirc.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario # 6 Event# 7,8,and9 Page 27 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Fosition Applicants Actions or Behavior
| |
| : 4. CHECK SIG pressures:
| |
| * All SIG pressures controlled or rising.
| |
| RO observes PAM pressure instruments on SG I through 4 (black labels) and determines that pressures are controlled.
| |
| May also observe trends on 1-PR-1-2, SG I & 2 PRESS PSIG, RD and 1-PR-1-23, SG 3 & 4 PRESS PSIG to assess this step.
| |
| * All SIG pressures greater than 120 psig.
| |
| RO observes PAM pressure instruments on SG I through 4 (black labels) and determines that pressures are greater than 120 psig.
| |
| : 5. MAINTAIN Intact SIG NR levels:
| |
| : a. MONITOR levels greater than 29% [39% ADVJ.
| |
| RO informs the SRO that S/G narrow range levels are controlled after observing PAM narrow range level instruments (black BOP labels).
| |
| : b. CONTROL intact SIG levels between 29% and 50% [39% and 50% ADV].
| |
| RO acknowledges the need to control SG levels between 39 and 50%.
| |
| EXAMINER: The status of secondary radiation may have already been reported as normal by the BOP during performance of E-O Appendix A.
| |
| : 6. CHECK secondary radiation:
| |
| * SIG discharge monitors NORMAL.
| |
| BOP
| |
| * Condenser vacuum exhaust rad monitors NORMAL.
| |
| * SIG blowdown rad monitor recorders NORMAL trend prior to isolation.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 28 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Retum Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 7. ENSURE cntmt hydrogen analyzers in service:
| |
| * PLACE 1-HS-43-200A in ANALYZE [M-10].
| |
| BOP places 1-HS-43-200A, H2 ANALYZER A FAN to the ANALYZE position on panel 1-M-1O.
| |
| * PLACE 1-HS-43-210A in ANALYZE [M-10].
| |
| BOP places 1-HS-43-200A, H2 ANALYZER A FAN to the ANALYZE position on panel 1-M-1O.
| |
| * CHECK low flow lights NOT lit [M-I0].
| |
| BOP BOP checks 1-Xl-43-200, LO FLOW- ANAL A, WHITE indicating light is DARK.
| |
| BOP checks 1-XI-43-210, LO FLOW- ANAL B, WHITE indicating light is DARK.
| |
| * LOCALLY CHECK low analyzer temp lights NOT lit AND RESET local alarm panel. [North wall of Train A 480V SD Bd rmj.
| |
| When contacted as the Control Building AUO the Console Operator repeat back request to check low analyzer temp lights, and reports that the lights are NOT LIT.
| |
| : 8. MONITOR pzr PORVs and block valves:
| |
| : a. Pzr PORVs CLOSED.
| |
| RO observes handswitch 1-HS-68-340AA, PZR PORV34OA GREEN light is LIT and 1-HS-68-334A, PZR PORV 334 GREEN RO light is LIT.
| |
| : b. At least one block valve OPEN.
| |
| RO observes handswitch 1-HS-68-333A, BLOCK VLV FOR PORV 340A RED indicating light is LIT.
| |
| : 9. DETERMINE if cntmt spray should be stopped:
| |
| SRO a. MONITOR cntmt pressure less than 2.0 psig.
| |
| : b. CHECK at least one cntmt spray pump RUNNING.
| |
| : 9. b. RESPONSE NOT OBTAINED:
| |
| SRO IF both spray pumps stopped, THEN **
| |
| GO TO Step 10.
| |
| : 10. ENSURE both pocket sump pumps STOPPED [M-15j:
| |
| * 1-HS-77-410.
| |
| * 1-HS-77-411.
| |
| BOP BOP observes handswitch 1-HS-77-410, POCKET SUMP PMP A GREEN indicating light is LIT, and 1-HS-77-411, POCKET SUMP PMP B GREEN indicating light is LIT.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Lppendix D Required Operator Actions Form ES-D-2 OpTestNo.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 29 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. IA Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time j Position Applicants Actions or Behavior
| |
| : 11. CHECK SI termination criteria:
| |
| : a. CHECK RCS subcooling greater than 65°E [85°F ADV].
| |
| RD RO determines that subcooling is less than 85°F by observing I-TI-68-105, RCS SUBCOOLING and I-Xl-68-I15 RCS SUBCOOLING indicators.
| |
| : a. RESPONSE NOT OBTAINED:
| |
| SRO GO TO Caution prior to Step 12.
| |
| CAUTION If offsite power is lost after SI reset, manual action will be required to restart the SI pumps and RHR pumps due to loss of SI start signal.
| |
| : 12. RESET SI and CHECK the following:
| |
| RO resets SI signal by depressing I-HS-63-134A, SI RESET TR A and I-HS-63-134B one at a time.
| |
| * SI ACTUATED permissive DARK.
| |
| RD RO observes and reports that Window 70-A, SI ACTUATED is DARK.
| |
| * AUTO SI BLOCKED permissive LIT.
| |
| RO observes and reports that Window 70-B, AUTO SI BLOCKED is DARK.
| |
| : 13. DETERMINE if RHR pumps should be stopped:
| |
| : a. CHECK RCS pressure greater than 150 psig.
| |
| RD
| |
| : b. CHECK RHR suction aligned from RWST.
| |
| : c. CHECK RCS pressure stable or rising.
| |
| : c. RESPONSE NOT OBTAINED:
| |
| ENSURE CCS from RHR heat exchanger 1-FCV-70-153 and 1-FCV-70-1 56 OPEN.
| |
| BOP rotates 1-FCV-70-156 RHR HX IA OUTLET to the right to the OPEN position. Ensures RED indicating light is LIT and GREEN indicating light is DARK.
| |
| CLOSE SEP heat exchanger A CCS supply O-FCV-70-197.
| |
| BOP rotates 0-FCV-70-197, SFP HXA SUPPLY to the left to the CLOSE position. Ensures RED indicating light is DARK and GREEN indicating light is LIT.
| |
| GO TO Step 14.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 30 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous
| |
| * overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 14. CHECK pressure in all SIGs controlled or rising.
| |
| RO observes PAM pressure instruments on SG I through 4 RD (black labels) and determines that pressures are controlled.
| |
| May also observe trends on I-PR-1-2, SG I & 2 PRESS PSIG, and I-PR-1-23, SG 3 & 4 PRESS PSIG to assess this step.
| |
| : 15. CHECK RCS pressure stable or dropping.
| |
| RD RO observes RCS pressure on I-XI-68-IOO, RVLIS ICCM -
| |
| PLASMA DISPLAY and/or 1-Xl-68-IIO, RVLIS ICCM PLASMA -
| |
| DISPLAY and determines that pressure is dropping.
| |
| : 16. MONITOR electrical board status:
| |
| : a. CHECK offsite power available.
| |
| BOP b. CHECK all shutdown boards ENERGIZED by offsite power.
| |
| : c. PLACE any unloaded DIG in standby USING SOl-82 Diesel Generators.
| |
| EXAMINER: 4 Minutes after Phase B occurs, lB-B 6.9 KV shutdown board trips on differential relay therefore AOI-43 may be implemented at that time. Applicable sections from AOI-43.02, Loss of Unit I Train B Shutdown Boards, are included as Attachment 2.
| |
| May transfer Instrument Power B Rack [1-M-7] to ENERGIZED BOP feeder prior to entry into AOI-43.02.
| |
| EXAMINER: When Dispatched to check I B-B 6.9 KV shutdown board, console operator will report that the differential lockout relay has operated, and that there is extensive damage to the board. There is no fire.
| |
| 16.b. RESPONSE NOT OBTAINED:
| |
| ENERGIZE shutdown boards USING:
| |
| SRD
| |
| * SOI-21 1 Shutdown Boards OR
| |
| * AOI-43 Loss of Shutdown Boards OR
| |
| * S0I-82 Diesel Generators EXAMINER: AOI-l7, Turbine Trip, Section 3.3, BOP Realignment is contained as Attachment 3.
| |
| : 17. INITIATE BOP realignment:
| |
| * REFER TO AOl-i 7, Turbine Trip.
| |
| SRO assigns AOl-17, Turbine Trip to the BOP for performance on a not to interfere basis.
| |
| EXAMINER: E-l, Loss of Reactor or Secondary Coolant, Appendices A, B, C, and D are contained as Attachment 4.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| I Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario# 6 Event# 7, 8, and9 Page 31 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. IA Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 18. INITIATE 480V board room breaker alignments USING the following:
| |
| * Appendix A (E-1), CLA Breaker Operation.
| |
| * Appendix B (E-1), 1-FCV-63-1 Breaker Operation.
| |
| BOP
| |
| * Appendix C (E-1), 1-FCV-63-22 Breaker Operation.
| |
| When contacted as the Control Building AUO, the Console Operator will repeat back the request, and then report that Appendix A through C have been performed. Console Operator enters remote functions sirOl, sirl4 and sirO6 to perform actions.
| |
| : 19. ENSURE RHR available for cntmt sump recirculation:
| |
| * Power to at least one operable RHR pump AVAILABLE.
| |
| RO
| |
| * Cntmt sump valve 1-FCV-63-72 or 1-FCV-63-73 to operable RHR pump AVAILABLE.
| |
| : 20. EVALUATE plant equipment status:
| |
| BOP
| |
| * REFER TO Appendix D (E-1), Equipment Evaluation.
| |
| : 21. CHECK Aux Bldg radiation for loss of RCS inventory outside cntmt:
| |
| : a. Area monitor recorders 1-RR-90-1 and 0-RR-90-12A Aux BOP Bldg points NORMAL.
| |
| : b. Vent monitor recorder 0-RR-90-101 NORMAL trend prior to isolation.
| |
| SRO 22. NOTIFY Chemistry of event status and plant conditions.
| |
| : 23. DETERMINE if RCS cooldown and depressurization is required:
| |
| : a. CHECK RCS pressure greater than 150 psig.
| |
| SRO **
| |
| : b. GO TO ES-i .2, Post LOCA Cooldown and Depressurization.
| |
| SRO will transition to ES-1.2.
| |
| EXAMINER: When SRO transitions to ES-I .2 scenario may be terminated if crew has completed FR-Z.I (Next Page). If FR-Z.I has not been completed delay scenario termination until FR-Z.I is complete.
| |
| END OF SCENARIO 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I OpTestNo.:2 2 Scenario # 6 Event# 7, 8, and 9 Page 32 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. IA Air Return Fan fails to start automatically. IA Containment Spray Pump trips on instantaneous overcurrent when started. 1 B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time 0 Position Applicants Actions or Behavior The following actions are taken from FR-Z.1, High Containment FR-Z.1 Pressure.
| |
| CAUTION If ECA-1.1, Loss of RHR Sump Recirculation, is in effect, the number of cntmt spray pumps to be operated is directed in ECA-1 .1 rather than in this Instruction.
| |
| NOTE Adverse containment setpoints [ADV] should be used where provided due to Phase B actuation.
| |
| : 1. ENSURE cntmt spray operation:
| |
| : a. Cntmt spray signal ACTUATED.
| |
| : b. Cntmt spray pumps RUNNING.
| |
| RO determines that both Containment spray pumps are NOT RUNNING.
| |
| : c. Cntmt spray valves 1-FCV-72-2 and 1-FCV-72-39 OPEN.
| |
| : d. Cntmt spray pump suction valves OPEN:
| |
| RD Valves from RWST:
| |
| 1 -FCV-72-21 and 1 -FCV-72-22 OR Valves from cntmt sump:
| |
| 1 -FCV-72-44 and 1 -FCV-72-45
| |
| : e. Cntmt spray flow:
| |
| * I-F 1-72-34
| |
| * 1-Fl-72-131 1 RESPONSE NOT OBTAINED:
| |
| ESTABLISH at least one train of cntmt spray flow.
| |
| RD RO Determines that no train of containment spray flow can be established due to IA-A Spray pump trip and lB-B Spray Pump Power not available.
| |
| EXAMINER: If Dispatched to check IA-A Containment spray pump breaker locally, console operator will report I B Containment Spray Pump tripped due to instantaneous overcurrent.
| |
| If Dispatched to check IA-A Containment spray pump locally, console operator will report that the pump is severely damaged, and the shaft appears to be broken.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 Op Test No.:2 2 Scenario # 6 Event # 7, 8, and 9 Page 33 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.I, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior
| |
| : 2. ENSURE cntmt isolation:
| |
| : a. Phase A isolation:
| |
| * Train A GREEN.
| |
| * Train B GREEN.
| |
| If TSC orAUO contacted to manually close valves Containment Isolation Valves without power (1-FCV-62-61 and 1-FCV-31-67),
| |
| RD Console Operator will acknowledge the request.
| |
| : b. Cntmt vent isolation:
| |
| * Train A GREEN.
| |
| * Train B GREEN.
| |
| : c. Phase B isolation:
| |
| * Train A GREEN.
| |
| * Train B GREEN.
| |
| SRO 3. ENSURE MSIVs and bypasses CLOSED.
| |
| : 4. PLACE steam dump controls OFF:
| |
| RD
| |
| * l-HS-I-103A, STEAM DUMP FSV A
| |
| * l-HS-1-103B, STEAM DUMP FSV B RD 5. ENSURE all four RCPs STOPPED.
| |
| : 6. MONITOR EGTS operation:
| |
| SRD a. EGTS fans RUNNING.
| |
| : b. Filter bank dp between 5 and 9 inches of water.
| |
| : 7. ENSURE ABGTS operation:
| |
| : a. ABGTS fans RUNNING.
| |
| : b. ABGTS dampers OPEN:
| |
| RD
| |
| * FCD-30-146A.
| |
| * FCD-30-146B.
| |
| * FCD-30-1 57A.
| |
| * FCO-30-157B.
| |
| : 8. WHEN 10 minutes has elapsed since Phase B actuation, THEN ENSURE cntmt air return fans start.
| |
| RD RO will manually start IA Air Return Fan when 10 minutes have elapsed since Phase B actuation. lB Air Return Fan does not have power available.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Appendix D Required Operator Actions Form ES-D-2 I Op Test No.:2 2 Scenario# 6 Event# 7, 8, and 9 Page 34 of 34 Event
| |
| | |
| == Description:==
| |
| 1-SV-68-564, PZR SAFETY VALVE fails open when the plant is tripped. Automatic Containment Isolation Phase A fails to occur on both trains. 1A Air Return Fan fails to start automatically. 1A Containment Spray Pump trips on instantaneous overcurrent when started. I B-B 6.9 KV Shutdown Board trips on differential relay 4 minutes after the Phase B occurs. Requires entry into FR-Z.1, High Containment Pressure.
| |
| Time Position Applicants Actions or Behavior CAUTION
| |
| * RCS cooldown requires the availability of at least one SIG.
| |
| * If ALL SIGs are Faulted, at least a minimum detectable feed flow to each SIG is required to limit thermal stress during subsequent SIG feed operations.
| |
| : 9. CHECK SIG pressures:
| |
| SRO
| |
| * All SIG pressures controlled or rising.
| |
| * All SIG pressures greater than 120 psig.
| |
| : 10. DETERMINE if RHR spray should be placed in service:
| |
| : a. CHECK the following conditions:
| |
| * At least one hour has elapsed since the beginning of the accident.
| |
| * Cntmt pressure is greater than 9.5 psig.
| |
| * RHR suction is aligned to cntmt sump.
| |
| RO
| |
| * At least one charging pump and one SI pump running.
| |
| : b. ALIGN Train B RHR spray:
| |
| : 1) ENSURE Train B RHR pump RUNNING.
| |
| : 2) CLOSE RHR crosstie 1-FCV-74-35.
| |
| : 3) CLOSE RHR injection 1-FCV-63-94.
| |
| : 4) OPEN RHR spray 1-FCV-72-41.
| |
| RO 1 1. RETURN TO Instruction in effect.
| |
| EXAMINER: If crew has previously transitioned to ES-1.2 Scenario may be terminated when FR-Z.1 is complete.
| |
| 2011-10 Watts Bar NRC Examination
| |
| | |
| Scenario 6 Attachment I EO, Reactor Trip or Safety Injection Appendix A and B Attachments I through 5
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 1 of 9)
| |
| Equipment Verification
| |
| _Step Ltion/Expected Response Response Not Obtained ENSURE PCBs OPEN: OPEN manually.
| |
| * PCB 5084.
| |
| * PCB 5088.
| |
| : 2. ENSURE AFW pump operation: ESTABLISH at least one
| |
| * Both MD AFW pumps train AFW operation.
| |
| RUNNING.
| |
| * TDAFWpump RUNNING.
| |
| * LCVsinAUTO,OR controlled in MANUAL.
| |
| : 3. ENSURE MEW isolation: Manually CLOSE valves AND
| |
| * MEW isolation and bypass isolation valves CLOSED.
| |
| STOP pumps, as necessary.
| |
| * MFW reg and bypass reg valves CLOSED. IF any valves can NOT be closed, THEN
| |
| * MFP A and B TRIPPED.
| |
| * Standby MFP STOPPED. CLOSE #1 heater outlet valves.
| |
| * Cond demin pumps TRIPPED.
| |
| * Cond booster pumps TRIPPED.
| |
| * #3 HDT Pumps TRIPPED.
| |
| * #7 HDT Pumps TRIPPED.
| |
| Page 16 of 31
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 2 of 9)
| |
| Equipment Verification
| |
| _Step Action/Expected Response Response Not Obtained
| |
| : 4. MONITOR ECCS operation:
| |
| : a. Charging pumps RUNNING. a. Manually START charging pumps.
| |
| : b. Charging pump alignment: b. ENSURE at least one
| |
| * RWST outlets 1 -LCV-62-1 35 valve in each set aligned.
| |
| and 1-LCV-62-136 OPEN.
| |
| * VCT outlets 1 -LCV-62-1 32 and 1 -LCV-62-1 33 CLOSED.
| |
| * Charging 1-FCV-62-90 and 1-FCV-62-91 CLOSED.
| |
| : c. RHR pumps RUNNING. c. Manually START RHR pumps.
| |
| : d. SI pumps RUNNING. d. Manually START SI pumps.
| |
| : e. BIT alignment: e. ENSURE at least one valve
| |
| * aligned, and flow thru BIT.
| |
| Outlets 1 -FCV-63-25 and 1-FCV-63-26 OPEN.
| |
| * Flow thru BIT.
| |
| : f. RCS pressure greater f. ENSURE SI pump flow.
| |
| than 1650 psig.
| |
| IF RCS press drops to less than 150 psig, THEN ENSURE RHR pump flow.
| |
| Page 17 of 31
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 3 of 9)
| |
| Equipment Verification Step Action/Expected Response Response Not Obtained u***aiiiimau***imiimi***i*r
| |
| : 5. CHECK cntmt isolation: ACTUATE Phase A and
| |
| : a. Phase A isolation: : Cntmt Vent Isolation signal,
| |
| * Train A GREEN. OR
| |
| * TrainBGREEN.
| |
| Manually* CLOSE valves and
| |
| : b. Cntmt vent isolation:
| |
| * dampers as necessary.
| |
| * Train A GREEN. CRITICALTASK1 -Manually initiateat leastone *
| |
| * train of Containment Phase A isolation prior to *
| |
| * Train B GREEN.
| |
| completion of E-O,ReactorTrip or Safety lnjection,Appendix A, Equipment Verification.
| |
| *** * ** ** iiuiu*a uii* **
| |
| * Page 18 of 31
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page4of9)
| |
| Equipment Verification
| |
| _Step Action/Expected Response Response Not Obtained
| |
| : 6. CHECK cntmt pressure: PERFORM the following:
| |
| * Phase B DARK [MISSPJ. 1) ENSURE Phase B actuated.
| |
| * Cntmt Spray DARK [MISSP]. 2) ENSURE Cntmt Spray actuated.
| |
| * Cntmt press less than 2.8 psig. 3) ENSURE cntmt spray pumps running.
| |
| : 4) ENSURE cntmt spray flow.
| |
| : 5) ENSURE Phase B isolation:
| |
| * Train A GREEN.
| |
| * Train B GREEN
| |
| * Manually CLOSE valves and dampers as necessary.
| |
| : 6) STOP all RCPs.
| |
| : 7) ENSURE MSlVs and bypasses CLOSED.
| |
| : 8) PLACE steam dump controls OFF.
| |
| : 9) WHEN 10 minutes has elapsed since Phase B actuated, THEN ENSURE air return fans start.
| |
| : 10) USE adverse cntmt [ADV] setpoints where provided.
| |
| : 7. DISPATCH AUO to perform Attachment I (E-O), Ice Condenser AHU Breaker Operation.
| |
| Page 19of31
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 AppendixA (Page 5 of 9)
| |
| Equipment Verification
| |
| __Step Action/Expected Response Response Not Obtained
| |
| : 8. CHECK plant radiation NORMAL: NOTIFY Unit Supervisor IMMEDIATELY.
| |
| * SIG blowdown rad recorder 1-RR-90-120 NORMAL prior to isolation M-12].
| |
| * Condenser vacuum exhaust rad recorder 1 -RR-90-1 19 NORMAL prior to trip [M-12].
| |
| * 1-RR-90-106 and 1-RR-90-112 radiation recorders NORMAL prior to isolation [M-12].
| |
| * S/G main steamline discharge monitors NORMAL [M-30].
| |
| * Upper and Lower containment high range monitors NORMAL
| |
| [M-30].
| |
| * NOTIFY Unit Supervisor conditions NORMAL.
| |
| : 9. ENSURE all DIGs RUNNING. EMERGENCY START DIGs
| |
| : 10. ENSURE ABGTS operation:
| |
| : a. ABGTS fans RUNNING. a. Manually START fans.
| |
| : b. ABGTS dampers OPEN: b. Locally OPEN dampers.
| |
| * FCO-30-146A.
| |
| * FCO-30-146B.
| |
| * FCO-30-157A.
| |
| * FCO-30-1 57B.
| |
| Page 20 of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 6 of 9)
| |
| Equipment Verification
| |
| _Step Action/Expected Response Response Not Obtained
| |
| : 11. ENSURE at least four ERCW pumps Manually START pumps as necessary.
| |
| RUNNING, one on each shutdown board preferred.
| |
| : 12. ENSURE ERCW supply valves IF ERCW can NOT be OPEN to running DIGs. aligned to running DIG, THEN EMERGENCY STOP affected DIG.
| |
| : 13. ENSURE O-FCV-67-152, CCS HX C Manually OPEN O-FCV-67-152 ALT DISCH TO HDR B, is open to to position A.
| |
| position A.
| |
| : 14. CLOSE O-FCV-67-144, CCS HX C DISCH TO HDR A.
| |
| : 15. MONITOR EGTS operation: Manually START fans
| |
| * EGTSfansRUNNING.
| |
| AND
| |
| * ENSURE dampers OPEN OPEN dampers.
| |
| * VERIFY filter bank dp between 5 and 9 inches of water.
| |
| : 16. ENSURE CCS pumps RUNNING: Manually START pumps as necessary.
| |
| * 1A-ACCS pump.
| |
| * lB-B CCS pump.
| |
| * C-S or 2B-B CCS pump.
| |
| : 17. DISPATCH AUO to shutdown Upper and Lower CNTMT rad monitors USING SOI-90.02.Gaseous Process Radiation Monitors Page 21 of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 7 of 9)
| |
| Equipment Verification
| |
| _Step Action/Expected Response Response Not Obtained
| |
| : 18. WHEN Attachment 1 is complete (Ice Condenser AHU Breakers OPEN),
| |
| THEN ENERGIZE hydrogen igniters
| |
| [1 -M-1 0]:
| |
| * 1-HS-268-73 ON.
| |
| * 1-HS-268-74 ON.
| |
| NOTE The following equipment is located on 1-M-9.
| |
| : 19. CHECK CNTMT PURGE fans STOP fans AND STOPPED.
| |
| PLACE handswitch in PULL-TO-LOCK.
| |
| : 20. CHECK FUEL HANDLING EXH fans STOP fans AND STOPPED, Fuel and Cask loading dampers CLOSED: PLACE handswitch in PULL-TO-LOCK, THEN Manually CLOSE dampers.
| |
| : 21. ENSURE AB GEN SUPPLY and STOP fans AND EXH fans STOPPED.
| |
| PLACE handswitch in PULL-TO-LOCK.
| |
| NOTE Dampers 1-HS-30-158 and 2-HS-30-270 remain open during ABI.
| |
| : 22. ENSURE AB GEN SUP & EXH Manually CLOSE dampers.
| |
| dampers CLOSED.
| |
| Page 22 of 31
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 8 of 9)
| |
| Equipment Verification
| |
| _Step Action/Expected Response Response Not Obtained
| |
| : 23. ENSURE MCR & SPREAD RM Manually CLOSE dampers.
| |
| FRESH AIR dampers CLOSED:
| |
| * FCV-31-3.
| |
| FCV-31-4.
| |
| : 24. ENSURE at least one CB EMER CLEANUP fan RUNNING and associated damper OPEN:
| |
| e CB EMERG CLEANUP FAN A-A, Manually START fan.
| |
| OR Fan B-B RUNNING..
| |
| e FCO-31-8, OPEN. NOTIFY TSC if at least one damper NOT OPEN.
| |
| OR e FCO-31-7, OPEN
| |
| : 25. ENSURE at least one CB EMER PRESS fan RUNNING and associated damper OPEN:
| |
| * CB EMERG PRESS FAN A-A, Manually START fan.
| |
| OR FAN B-B RUNNING.
| |
| * FCO-31-6, OPEN. NOTIFY TSC if at least one damper NOT OPEN.
| |
| OR
| |
| * FCO-31-5, OPEN.
| |
| Page23of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix A (Page 9 of 9)
| |
| Equipment Verification Step Action/Expected Response Response Not Obtained
| |
| : 26. ENSURE Control Building fans STOPPED and dampers CLOSED:
| |
| * SPREADING ROOM SUPPLY Manually STOP fans AND and EXH FANS AND dampers.
| |
| NOTIFY TSC if any damper NOT
| |
| * TOILET & LKR RM EXHAUST CLOSED.
| |
| FAN AND dampers.
| |
| : 27. INITIATE Appendix B (E-O), Phase B Pipe Break Contingencies.
| |
| Page24of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Appendix B (Page 1 of 1)
| |
| Phase B Pipe Break Contingencies Step 1 Action/Expected Response Response Not Obtained
| |
| : 1. CHECK PHASE B actuated. WHEN PHASE B actuation occurs,
| |
| [MISSP 1-XX-55-6C, -6D]
| |
| - THEN GO TO step 2.
| |
| : 2. ENSURE 1 -FCV-32-1 10 CLOSED. DISPATCH AUO to perform
| |
| [CISP 1-)O(-55-6E]
| |
| - Attachment 2 (E-0).
| |
| (A-train, window 13)
| |
| : 3. ENSURE 1-FCV-67-107 CLOSED. DISPATCH AUO to perform
| |
| [CISP 1-XX-55-6E]
| |
| - Attachment 3 (E-0).
| |
| (A -train, window 43)
| |
| : 4. ENSURE 1-FCV-70-92 CLOSED. DISPATCH AUO to perform
| |
| [CISP 1-XX-55-6E]
| |
| - Attachment 4 (E-0).
| |
| (A -train, window 73)
| |
| : 5. ENSURE 1-FCV-70-140 CLOSED. DISPATCH AUO to perform
| |
| [CISP 1-XX-55-6F]
| |
| - Attachment 5 (E-0).
| |
| (B -train, window 74)
| |
| Page25of3l
| |
| | |
| WBN Reactor Trip or Safety njection E-0 Unit I Rev. 0030 Attachment I (Page 1 of I)
| |
| Ice Condenser AHU Breaker Operation OPEN the following to remove power from ice condenser air handling units AND REPORT completion to UO:
| |
| BOARD COMPT NOMENCLATURE 480 V Reactor Vent 13D 1-BKR-232-A000/13D ICE COND Board 1 A-A 1 -AHU-61 -1/4/8/12/16/20/24/28 480 V Reactor Vent 14D 1-BKR-232-A000/14D ICE COND Board lA-A 1-AHU-61-3/7/1 1/1 5/1 9/23/27 480 V Reactor Vent 1 3D 1 -BKR-232-B000/1 3D ICE COND Board 1 B-B 1-AHU-61-2/6/1 0/14/18/22/26/30 480 V Reactor Vent 14D 1-BKR-232-B000/14D ICE COND Board 1 B-B 1 -AHU-61 -5/9/13/17/21/25/29 Page26of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Attachment 2 (Page 1 of 1)
| |
| Control Air Isolation A. CLOSE O-ISV-32-1013 CONTROL AIR EL 713 AB HDR ISOL
| |
| [A6/S EL. 713] (chain operated behind Fuel and Waste Handling Bd. A).
| |
| B. IF O-ISV-32-1013 CANNOT BE CLOSED, THEN OPEN and DISCONNECT C&SS air compressor breakers:
| |
| : 1. O-BKR-32-25 [480V SD BD 1A2-A, C/3D]
| |
| : 2. O-BKR-32-26 [480V SD BD 1B1-B, C/3D]
| |
| : 3. O-BKR-32-27 [480V AUX BLDG COM BD, C/6C]
| |
| : 4. O-BKR-32-4900A [480V TURB BLDG COM BD, C/6C]
| |
| Page27of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Attachment 3 (Page 1 ofl)
| |
| ERCW Isolation UNLOCK AND CLOSE I -ISV-67-523B, LOWER CNTMT VENT CLR 1 B &1 D ERCW SUP ISOL [A2U/692] (U-I penetration room North of AB Pipe Chase Cooer lB-B in overhead)
| |
| Page 28 of3l
| |
| | |
| WBN Reactor Trip or Safety Injection E-0
| |
| * Unit I Rev. 0030 Attachment 4 (Page 1 of I)
| |
| CCS Return Isolation CLOSE i-ISV-70-700, RCP OIL COOLER CCS RETURN ISOLATION
| |
| [A4N EL. 710 U-i Penetration Room] (approximately lOft. North of Penetration Room Cooler lB-B on mezzanine above RHR Sump Valve Room)
| |
| Page29of3i
| |
| | |
| WBN Reactor Trip or Safety Injection E-0 Unit I Rev. 0030 Attachment 5 (Page 1 of 1)
| |
| CCS Supply Isolation CLOSE 1-ISV-70-516, REACTOR BUILDING CCS SUPPLY ISOLATION
| |
| [A6IT EL. 737] (Behind Elevator approximately 2 ft. west on mezzanine above A CCS Heat Exchanger)
| |
| Page3Oof3l
| |
| | |
| Scenario 6 Attachment 2 A0143.02 Loss of Unit I Train B Shutdown Boards.
| |
| Section 3.1 Initial Actions.
| |
| Section 3.4, Compensatory Actions for Loss of 6.9kVSD BD 1B-B.
| |
| | |
| I !/ Watts Bar Nuclear Plant Unit I Abnormal Operating Instruction AOI-43.02 Loss of Unit I Train B Shutdown Boards Revision 0009 Quality Related Level of Use: Continuous Use Effective Date: 03-22-2011 Responsibie Organization: OPS, Operations Prepared By: R. A. ORear Approved By: Brian Mcllnay
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change 3 08/07/03 2, 7, 17, Non-intent. Deleted reference to Vital Inverter 20, 22 2-lI for DCN-51368, stage 2. Corrected protective relay operation verification.
| |
| 4 10/02/2003 2, 7, 28 Non-intent. Changed 1-HIC-62-81A required position from 25% open to 40-50% open to be consistent with SOl-62.01. Clarified step 12 to indicated manning is for additional manpower, not do to rep classification. Operator feedback.
| |
| 5 09/17/07 2, 16 Added clarification for primary water system when in bypass mode. Operator feedback.
| |
| 6 12/03/07 2, 33 Administrative change. Changed Maintenance Supply to 6.9KV SD BD 2A-A from 6.9KV Unit BD 2B (PER 170433) 7 02/03/10 All General revision to incorporate corrective actions for PER 176605. Procedure rewritten to provide clarity and logical flow. Section 3 (Operator Actions) divided into subsections for Initial Actions, Energizing 6.9kV Shutdown Board, Restoration of 6.9kV SD BD after Energization, Compensatory Actions for Loss of 6.9kV SD BD. Added section to address loss of 480V SD BDs IB1-B or 1B2-B. Added Appendix for list of equipment affected by loss of B-Train SD BDs. Increased level of detail through-out procedure. Added steps for resetting Black-Out Relays in Section 3.3.
| |
| 8 07/12/10 2, 36, 39, Added Spare Charger 8-S (DCN 53437).
| |
| 49, 50 9 03/22/11 2, 4, 7 Added step for transferring Power Rack B to alternate to the initial actions section 3.1. [PCR 4639]
| |
| Minor editing and added page numbers to diagnostic box Page 2 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 1.0 PURPOSE This Instruction provides operator actions for:
| |
| * A loss of 6.9kV Shutdown Board 1 B-B OR
| |
| * A loss of either 480V Shutdown Board I B1-B or 1 B2-B, without a loss of 6.9kV SD BD lB-B 2.0 SYMPTOMS 2.1 Alarms A. PNL 1-M-7 BREAKER TRIP [15-El.
| |
| B. 6.9 SD BD lB-B UV/OV/CONTROL PWR FAILURE [13-B, 208-C].
| |
| C. 480 SD BD IB1-B/1B2-B FAILURE/ABN [11-D, 207-D].
| |
| D. RX MOV VENT BD TRAIN B UNDERVOLTAGE [142-D].
| |
| E. C &AVENT BD IB1-B/1B2-B UNDERVOLTAGE [142-El.
| |
| F. DG AUX BD 1B1-B/1B2-B UNDERVOLTAGE [207-El.
| |
| 2.2 Indications A. Low voltage on any Unit 1 Train B 6.9KV or 480V Shutdown Board.
| |
| B. Zero amps indicated on CSST to Shutdown Board indication.
| |
| C. Open breaker indications.
| |
| D. Failure of Shutdown Board supplied loads.
| |
| 2.3 Automatic Actions A. Diesel Generator 1 B-B starts upon loss of voltage to 6.9KV SD BD 1 B-B.
| |
| B. Designated loads are auto stripped from 6.9kV SD BD 1 B-B, 480V SD BD5 1B1-B and 1B2-B.
| |
| C. Designated loads auto sequence on when voltage is restored to 6.9kV SD BD lB-B and the Diesel Generator feeder breaker is closed.
| |
| D. Auto start for shed loads is blocked (except for SI auto start).
| |
| Page 3 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS NOTE I The loss of all onsite and offsite power is covered in the Emergency Instructions and AOI-40.
| |
| NOTE 2 A complete or partial loss of 161 kV offsite power is addressed in AOl-35.
| |
| DIAGNOSTICS CONDITION APPLICABLE PAGE PROCEDURE SECTION 6.9kV SD BD 1 B-B GO TO Section 3.3 28 ENERGIZED from DG I B-B following blackout LOSS of Power to 6.9kV SD GO TO Section 3.1 5 BD lB-B LOSS of Power to 480V SD BD GO TO Section 3.5.1 42 1B1-B WITHOUT loss of 6.9kV SD BD lB-B LOSS of Power to 480V SD BD GO TO Section 3.5.2 45 1 B2-B WITHOUT loss of 6.9kV SD BD lB-B End of Section Page4of67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.1 Initial Actions MONITOR A TRAIN 6.9KV SD BD EMERGENCY START DGs:
| |
| IA-A ENERGIZED.
| |
| * 1-HS-82-15 [1-M-1].
| |
| * 2-HS-82-15 [2-M-1].
| |
| IF BOTH Unit 1 6.9KV SD BDs still de-energized, AND
| |
| : a. IF Unit in MODE 1, 2, 3, or 4, THEN
| |
| **GO TO ECA-O.O, Loss of Shutdown Power.
| |
| : b. IF Unit in MODE 5, or 6,THEN:
| |
| : 1) TRIP RCPs
| |
| : 2) PERFORM AOI-14, Loss of RHR SD Cooling, WHILE continuing this procedure at Step2.
| |
| : 2. ENSURE Diesel Generators running: EMERGENCY START Diesel
| |
| * DG lA-A Generators:
| |
| * DG lB-B
| |
| * 1-HS-82-15 [1-M-lJ.
| |
| * DG 2A-A
| |
| * 2-HS-82-15 [2-M-1].
| |
| * DG 2B-B Step 2 continued on next page.
| |
| Page 5of67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.1 Initial Actions (continued)
| |
| Step 2 continued.
| |
| IF any DIG is NOT RUNNING, THEN EVALUATE RESETTING EMERGENCY STOP IF RESET and START of a DG is desired, THEN:
| |
| : a. PRESS and RELEASE DG AUTO SAFETY SHUTDOWN RELAY-RESET [O-M-26]:
| |
| * 1-HS-82-20 [lA-A].
| |
| * 1-HS-82-50 [lB-B].
| |
| * 2-HS-82-80 [2A-A].
| |
| o 2-HS-82-l1O [2B-B].
| |
| : b. EMERGENCY START Diesel Generator [O-M-26]:
| |
| * I -HS-82-1 6A [lA-A].
| |
| * l-HS-82-46A [1 B-B].
| |
| * 2-HS-82-76A [2A-A].
| |
| * 2-HS-82-106A [2B-B}.
| |
| Page 6 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOl-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.1 Initial Actions (continued)
| |
| NOTES Appendix A provides a list of affected equipment.
| |
| * RCPs can be operated for up to 10 minutes after a loss of CCS flow.
| |
| : 3. MONITOR RCP seal cooling MANUALLY START available:
| |
| * CCP IA-A
| |
| * Seal injection flow CCS pump IA-A.
| |
| OR IF seal cooling not restored, THEN
| |
| * CCS flow through Thermal Barrier Heat Exchangers MONITOR RCP trip criteria (reference AOI-24 as required).
| |
| : 4. EVALUATE ERCW supply on B Train IF any B Train Diesel Generator headers: running with NO ERCW cooling, THEN
| |
| : a. ENSURE at least one B Train 1) EMERGENCY STOP B Train DGs.
| |
| ERCW Pump In-service:
| |
| : 2) OPEN 1-FCV-67-65, DG lB-B
| |
| * ERCWPumpG-B ERCW SUP from Hdr. 2A, Manually [lB-B DG RM]
| |
| * ERCWPumpH-B
| |
| : 3) OPEN 2-FCV-67-65, DG 2B-B
| |
| : b. START second pump as needed.
| |
| ERCW SUP from Hdr. 2A, Manually. [2B-B DG RM]
| |
| : 4) RESET and EMERGENCY START B Train Diesel Generators Stopped due to lack of ERCW Cooling.
| |
| : 5. ENSURE Unit 1 Instrument Power B Rack selected to ENERGIZED feeder (amber light ON). [1-M-7]
| |
| Page7of67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit 1 Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.1 Initial Actions (continued)
| |
| CAUTION Further damage may occur if 86 LOCKOUT relay(s) are reset before status of board is evaluated and understood.
| |
| : 6. DISPATCH personnel to inspect the following equipment for damage, protective relay operation, and determine reason for BO:
| |
| * 6.9kV SD BD I B-B
| |
| * DG1B-B
| |
| * 48OVSDBDIB1-B
| |
| * 480V SD BD 1 B2-B 48OVSDXfmrs
| |
| : 7. NOTIFY Work Control for support and evaluation of BD.
| |
| : 8. MONITOR containment upper and START containment fans as needed:
| |
| lower compartment average air temperatures are within limits:
| |
| * CRD Mech Cooler Fans
| |
| * Lower Compartment Cooler
| |
| * S/R 3.6.5.1, Computer Point U9019 Fans
| |
| * S/R 3.6.5.2, Computer Point
| |
| * Upper Compartment Cooler U9020 Fans
| |
| : 9. ENSURE 1A Primary Water Pump in-service as required.
| |
| Page 8 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOl-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.1 Initial Actions (continued)
| |
| NOTE Operability verification of remaining AC power sources is required to be completed within one hour per LCO 3.8.1. S/R 3.8.1.1 (0-S 1-82-2)
| |
| : 10. NOTIFY Shift Manager to evaluate staffing the TSC/OSC for support,
| |
| : 11. EVALUATE Relay Operation and GO TO Section 3.4, Compensatory Damage reports, THEN Actions for Loss of 6.9kVSD BD I B-B, WHILE continuing to Evaluate DETERMINE if safe to energize 6.9kV Energizing lB-B 6.9kV SD BD.
| |
| SD BD lB-B.
| |
| : 12. GO TO Section 3.2 End of Section Page9of67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.4 Compensatory Actions for Loss of 6.9kVSD BD lB-B NOTE Appendix A provides list of Unavailable Equipment resulting from a loss of 6.9kV SD BD lB-B.
| |
| : 1. MONITOR condition of 6.9kV SD BD lB-B and supply sources, WHEN ready to energized Board from available power supply, THEN GO TO Section 3.2.1, Step 1.
| |
| : 2. DISPATCH AUO to D/G Bldg to monitor D/Gs conditions USING SOl-82 series, APPENDIX A, for operating parameters.
| |
| : 3. CHECK any charging pump PERFORM the following:
| |
| RUNNING.
| |
| : a. ISOLATE letdown:
| |
| * CLOSE letdown orifice(s).
| |
| * CLOSE 1 -FCV-62-69A.
| |
| * CLOSE 1 -FCV-62-70A.
| |
| : b. RESTORE charging and letdown using APPENDIX E ALIGNMENT OF CHARGING AND LETDOWN.
| |
| Page 38 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.4 Compensatory Actions for Loss of 6.9kVSD BD I B-B (continued)
| |
| : 4. CHECK A Train CCS flow adequate. START lA-A CCS Pump.
| |
| ENSURE one of the following CLOSED to avoid excessive flow
| |
| * RHR HTXA, 1-FCV-70-156 OR
| |
| * SEP HTX A, O-FCV-70-l 97 IF A Train CCS is lost, THEN REFER TO AOl-i 5, Loss of Component Cooling Water (CCS) FOR LOSS OF CCS FLOW.
| |
| : 5. ENSURE Thermal Barrier Booster REFER TO AOl-15, Loss of Pump lA-A in-service Component Cooling Water (CCS) FOR (SOl-70.O1). LOSS OF CCS FLOW.
| |
| : 6. EVALUATE transferring one of the following to preserve Vital Battery life:
| |
| * 480V AC Vital Transfer Switch II to Alternate power supply (S0I-235.02).
| |
| OR
| |
| * 125V Vital Batt BD II to Battery Charger 6-S or 8-S (SOI-236.02)
| |
| : 7. EVALUATE transferring 24V CAP Battery Charger 2 from Normal to Alternate (SOI-252).
| |
| : 8. ENSURE Aux Bldg General Supply and Exhaust Fans in-service as required to maintain ventilation and pressure (SOI-30.05).
| |
| Page 39 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOI-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.4 Compensatory Actions for Loss of 6.9kVSD BD I B-B (continued)
| |
| : 9. ENSURE EBR Air Conditioning Unit A-A and MCR Air Conditioning Unit A-A in-service (SOl-31 .01).
| |
| : 10. ENSURE lAAnnulus Vacuum Fan in-service (SOl-65.01).
| |
| : 11. ENSURE 6.9kV SDBR Air Conditioning Unit A-A in-service.
| |
| (SOl-30.07)
| |
| : 12. ENSURE A Train or B Train 480V Shutdown Board Room Ventilation in-service (SOl-30.07).
| |
| Page 40 of 67
| |
| | |
| WBN Loss of Unit I Train B Shutdown AOl-43.02 Unit I Boards Rev. 0009 Step Action/Expected Response Response Not Obtained 3.4 Compensatory Actions for Loss of 6.9kVSD BD lB-B (continued)
| |
| CAUTION LCO 3.8.1 is expected to require performance of S/R 3.8.1.1 (0-SI-82-2). Performers are NOT to take ANY actions which would interrupt power suppHes in-service by this AOl.
| |
| : 13. REFER TO Tech Specs:
| |
| * 3.5.2, ECCS-Operating.
| |
| o 353, ECCS-Shutdown.
| |
| * 3.8.1, AC Sources-Operating.
| |
| * 3.8.2, AC Sources-Shutdown.
| |
| * 3.8.4, DC Sources-Operating.
| |
| * 3.8.5, DC Sources-Shutdown.
| |
| * 3.8.9, Distribution Sys-Operating.
| |
| * 3.8.10, Distribution Sys-SD.
| |
| : 14. REFER TO EPIP-1, Emergency Plan Classification Flowchart.
| |
| : 15. CONTINUE MONITORING 6.9Kv SD BD I B-B supply sources using 1-EI-82-36A [O-M-26]..
| |
| WHEN Power supply AVAILABLE, THEN
| |
| **GO TO Section 3.2.1 Step 1.
| |
| End of Section Page 41 of 67
| |
| | |
| Scenario 6 Attachment 3 AOI17, Turbine Trip.
| |
| Section 3.3, BOP Realignment.
| |
| * WBN Turbine Trip AOl-17 Unit I Rev. 0048 Step Action/Expected Response Response Not Obtained 3.3 BOP Realignment CAUTION Performance of this instruction should not be allowed to delay or interfere with actions required by applicable emergency procedures or abnormal operating procedures.
| |
| NOTES
| |
| * Control room operators may initiate shutdown of pumps and equipment from the bench board immediately after a trip.
| |
| Performance of this instruction will subsequently verify proper secondary equipment alignment.
| |
| * Steps in this section and items in Attachment I may be performed out of sequence.
| |
| * Attachment 1 may be initiated as soon as Turbine has tripped while MCR completes Section 3.2. Initiation of Attachment 1 may be part of briefing for preplanned Turbine trip with performance to begin when NAUO is notified of Turbine trip by UO.
| |
| : 1. DISPATCH turbine building NAUO to perform Attachment 1.
| |
| : 2. NOTIFY condensate demineralizer NAUO prior to Operator initiated press changes in condensate.
| |
| : 3. CHECK exciter field breaker OPEN, MOMENTARITY PLACE 1-HS-57-19, as indicated by GREEN light lit for EXCITATION CONTROL, to STOP 1-HS-57-19, EXCITATION AND CONTROL. VERIFY GREEN light lit.
| |
| Page 11 of 26
| |
| | |
| WBN Turbine Trip AOI-17 Unit I Rev. 0048 Step Action/Expected Response Response Not Obtained 3.3 BOP Realignment (continued)
| |
| : 4. MONITOR main turbine:
| |
| : a. VERIFY seal oil backup pump a. ENSURE seal oil backup pump RUNNING as indicated by 1-HS-47-61D in NORMAL 1-HS-47-61A. (T7J/729 behind MTOT)
| |
| : b. ENSURE turning gear oil pump RUNNING using 1-HS-47-61A.
| |
| : c. WHEN less than 600 rpm, THEN ENSURE bearing lift oil pump RUNNING using 1-HS-47-II1A.
| |
| : d. WHEN turbine is at ZERO RPM, THEN ENSURE turbine on turning gear.
| |
| : e. MAINTAIN MTOT lube oil temp between 95°F and 100°F (may require RCW isolation if TCV has excessive leakage).
| |
| : f. MAINTAIN GENERATOR H2 (Cold Gas) temp 95°F (may require RCW isolation if TCV has excessive leakage).
| |
| : g. ENSURE Gland Steam Spillover Bypass valve is CLOSED using 1 -HS-47-1 91 A.
| |
| : 5. ALIGN MSRs:
| |
| : a. PUSH RESET on MSR control panel.
| |
| : b. CLOSE MSR HP steam and bypass isol.
| |
| : c. ENSURE MSR warming valves CLOSED.
| |
| : d. OPEN MSR startup vents.
| |
| : e. CLOSE MSR operating vents.
| |
| Page 12 of 26
| |
| | |
| WBN Turbine Trip AOI-17 Unit I Rev. 0048 Step Action/Expected Response Response Not Obtained 3.3 BOP Realignment (continued)
| |
| : 6. CHECK MSIVs OPEN. IF vacuum is to be maintained, THEN ENSURE auxiliary boiler is aligned for steam seals.
| |
| : 7. ENSURE adequate FW press:
| |
| : a. ENSURE two hotwell pumps RUNNING.
| |
| : b. IF FW isolation reset, THEN ENSURE one condensate booster pump RUNNING if needed for unit conditions.
| |
| : c. ENSURE CNDS demin pumps OFF.
| |
| : d. ENSURE #3 HDT pumps OFF, AND CLOSE the discharge valves to condensate heater strings.
| |
| : e. ENSURE #7 HDT pumps OFF, AND CLOSE the discharge valves to condensate heater strings.
| |
| : 8. SHUTDOWN any MEW pump NOT required.
| |
| : 9. SHUTDOWN any RCW pumps NOT required.
| |
| : 10. SHUTDOWN any CCW pumps NOT required.
| |
| Page 13 of 26
| |
| | |
| WBN Turbine Trip AOl-17 Unit I Rev. 0048 Step Action/Expected Response Response Not Obtained 3.3 BOP Realignment (continued)
| |
| : 11. ALIGN extraction steam valves and drain valves:
| |
| : a. CLOSE #1 and #2 Heater extraction steam valves.
| |
| : b. ENSURE turbine drain valves OPEN.
| |
| : c. OPEN MEW pump turbine drain valves.
| |
| : 12. PERFORM as required:
| |
| : a. OBTAIN switching instructions from NEAD, and OPEN main generator PCB(s)
| |
| MODs.
| |
| : b. PULL-TO-LOCK bus duct cooling fans.
| |
| : c. VERIFY MTOT and seal oil temps STABLE and trending to 95°R
| |
| : 13. IF MEW isolated to steam generators, THEN REQUEST Chem Lab sample condensate and feedwater prior to re-admitting water to S/Gs from condensate-feedwater system.
| |
| : 14. IF EGTS started, THEN SHUTDOWN one train after 1 to 2 hours and place in P-AUTO:
| |
| * REFER TO SOI-65.02, Emergency Gas Treatment System, section on Auto EGTS Actuation.
| |
| Page 14 of 26
| |
| | |
| WBN Turbine Trip AOI-17 Unit I Rev. 0048 Step Action/Expected Response Response Not Obtained 3.3 BOP Realignment (continued)
| |
| : 15. IF ABGTS started, THEN SHUTDOWN one train after I to 2 hours and place in P-AUTO:
| |
| * REFER TO SOl-30.06, Auxiliary Building Gas Treatment System, section on Auto Start of ABGTS
| |
| : 16. IF MFW required, THEN ESTABLISH MEW:
| |
| * REFER TO Attachment 2, Establishing MEW Following Reactor Trip.
| |
| : 17. CHECK S/G NR levels between 38% IF SIG level can NOT be maintained, and 50%. THEN START MD AFW pumps.
| |
| : 18. REFER TO SO 1-47.02 for EHC Pumps alignment.
| |
| : 19. RETURN TO applicable Instruction.
| |
| End of Subsection Page 15of26
| |
| | |
| Scenario 6 Attachment 4 E1, Loss of Reactor or Secondary Coolant.
| |
| Appendix A through D
| |
| | |
| WBN Loss of Reactor or Secondary Coolant E-1 Uniti Rev. 0016 Appendix A (Page 1 ofl)
| |
| CLA Breaker Operation CLOSE the following to restore power to cold leg accumulator isolation valves:
| |
| BOARD COMPT NOMENCLATURE 480 V Reactor MOV 3F2 1-BKR-63-1 18A Board IAI-A SIS CL ACCUM 1 OUT ISOL (1 -FCV-63-1 18) 480 V Reactor MOV 17F2 1-BKR-63-80A Board IAI-A SIS CL ACCUM 3 OUT ISOL (1 -FCV-63-80) 480 V Reactor MOV 3F2 1 -BKR-63-98A Board 1 B1-B SIS CL ACCUM 2 OUT ISOL (1 -FCV-63-98) 480 V Reactor MOV 16F2 1-BKR-63-67A Board 1B1-B SIS CLACCUM 4 OUT ISOL (1 -FCV-63-67)
| |
| Page 20 of 24
| |
| | |
| WBN Loss of Reactor or Secondary Coolant E-1 Unit I Rev. 0016 Appendix B (Page 1 ofl)
| |
| I -FCV-63-1 Breaker Operation CLOSE the following to restore power to 1-FCV-63-1:
| |
| BOARD COMPT NOMENCLATURE 480 V Reactor MOV 2E1 1-BKR-63-1A Board 1AI-A RWST TO RHR SUCT (1 -FCV-63-1)
| |
| Page 21 of24
| |
| * WBN Loss of Reactor or Secondary Coolant E-1
| |
| * Unit I Rev. 0016 Appendix C (Page 1 of 1) 1-FCV-63-22 Breaker Operation CLOSE the following to restore power to 1-FCV-63-22:
| |
| BOARD COMPT NOMENCLATURE 480 V Reactor MOV 2F2 1-BKR-63-22A Board IBI-B SIP COLD LEG INJECTION (1 -FCV-63-22)
| |
| SHUNT TRIP BREAKER Page 22 of 24
| |
| | |
| WBN Loss of Reactor or Secondary Coolant El Unit I Rev. 0016 Appendix 0 (Page I of I)
| |
| Equipment Evaluation A. EVALUATE plant equipment and systems needed to support long term cooling and recovery actions, as time and personnel availability permits:
| |
| : 1. Cntmt Isolation Status.
| |
| : 2. Emergency Gas Treatment System: One train in operation, REFER TO SOl-65.02.
| |
| : 3. Auxiliary Building Gas Treatment: One train in operation, REFER TO SOl-30.06.
| |
| : 4. Auxiliary Building Isolation alignment: REFER TO SOI-30.06.
| |
| : 5. Main Control Room Isolation alignment: REFER TO SOI-31 .01.
| |
| : 6. ERCW System: Both trains in operation.
| |
| : 7. Component Cooling Water System: Both trains in operation.
| |
| : 8. Ice Condenser System: AHUs energized after cntmt hydrogen concentration verified (if applicable). REFER TO SOI-61 .01.
| |
| : 9. Permanent Hydrogen Mitigation System: lgniters de-energized when no longer needed. REFER TO SOl-268.01.
| |
| Page 23 of 24
| |
| | |
| WBN Administration Of The Tl-7.012 I Reactivity Briefing Sheets And Rev. 0004 Reactivity Control Plans Page 21 of 23 Appendix A (Page 1 of 1)
| |
| Reactivity Control Plan (Example Form)
| |
| Station: WBN Unit: 1 Cycle: 10 Burnup: 150 MWD/MTU Revision: 0 Preparer: I Reviewer: I I Date RXE I Date Approver: Authorizer:
| |
| RXES or designee I Date Ops I Date RXE support required Onsite? ØYes DNo Describe: Until released by Operations Title of Reactivity Control Plan: Power Escalation (BOL)
| |
| Assumptions: 1. Reactor was tripped from 100% power.
| |
| : 2. Fuel was conditioned at 100% at the time of trip.
| |
| : 3. Reactor critical at 100 hrs after trip. (Assumed no xenon.)
| |
| : 4. Ramp from 15% power to 100% at 5%/hr.
| |
| : 5. Reactor reaches 100% RTP within 24 hours, therefore fuel should not decondition below 100% during the ramp.
| |
| Major Steps: 1. Reactor power is raised from HZP to -3.0% RTP
| |
| : 2. Unit is ramped to -15%
| |
| : 3. Unit is maintained at -15% for 4 hrs
| |
| : 4. Unit is ramped to 100% at 5%/hr.
| |
| Detailed
| |
| | |
| == Description:==
| |
| | |
| NOTE: See attached plots NOTE: Should deviations occur from the projected power ascension profile, particularly delays, there should be no major effects on the predicted parameters. Should delays occur, lower dilution rates could be expected later due to a greater amount of Xenon build-in during the delays.
| |
| Power Increase to 3.0% RTP:
| |
| : 1. WITHDRAW CBD to facilitate power rise. (The rod position to attain 3.0% RTP depends upon the CBD position at criticality.)
| |
| Ramp Up to -15% RTP:
| |
| : 1. WITHDRAW CBD to facilitate power rise.
| |
| : 2. IF CBD has been withdrawn to -155 steps and reactor power has not reached 15%
| |
| RTP, THEN DILUTE as necessary to raise power to -15% for synchronization of the Turbine-Generator. (The actual amount of dilution necessary depends upon the CBD position at criticality.)
| |
| | |
| WBN Administration Of The TI-7.012 I Reactivity Briefing Sheets And Rev. 0004 Reactivity Control Plans Page 21 of 23 Power Escalation (BOL)
| |
| Continued Operation on Steam Dumps until Generator Synchronization:
| |
| : 1. DILUTE -260 gal PW per hour to compensate for Xenon build-in.
| |
| : 2. MAINTAIN CBD position.
| |
| Ramp Up to 100%:
| |
| : 1. DILUTE -17,000 gal PW to raise power to 100% and maintain Tave matched with Tref.
| |
| : 2. INITIATE ramp up.
| |
| : 3. WITHDRAW CBD to maintain AFD near target.
| |
| Full Power Maintenance:
| |
| : 1. DILUTE as necessary to compensate for Xenon build-in. Dilution rates may start out at -1000 gal PW per hour, but will taper off as Xenon builds in.
| |
| : 2. WITHDRAW AND MAINTAIN CBD at -220 steps.
| |
| Critical Parameter Limit Required Action Rate of Power Increase As specified in TI-45 Reduce ramp rate or hold power until limits satisfied.
| |
| Activated: I Terminated: I SM or US I Date SM or US I Date
| |
| | |
| Power Ascension (BOL)
| |
| UL...
| |
| 70 60 EWI L E EHE*
| |
| EU
| |
| * 180 Power T*i U
| |
| - = CPL
| |
| -i 50 .
| |
| * 170
| |
| - - f::- -
| |
| 8)
| |
| = = = = = = -d 0 U) 40 * * . 1600 RILL0-Lo 120 06 12 394 Hours
| |
| | |
| Scenano 6 SHIFT TIJRNOVER CHECKLIST Page 1 of2 SHIFT TURNOVER CHECKLIST Page 1 of 1 D SM US/MCR Unit 1 D UO Unit Offigoing Name D AUO Station 1J STA (STA Function) On-coming Name Part 1 Completed by off-going shift / Reviewed by on-coming shift:
| |
| RCS Cb = 1616 ppm
| |
| * Abnormal equipment lineup / conditions:
| |
| * SI/Test in progress/planned: (including need for conduct of evolution briefings)
| |
| * Major Activities/Procedures in progress/planned:
| |
| Reactor Startup is in progress. GO-2, Reactor Startup, Section 5.3, Reactor Startup is complete through Step 30, and step 31 is to be perfonned after assuming the shift. RCS boron concentration is 1616 ppm, Control Bank D is at 98 steps. All Low Power Physics testing is complete, and satisfactory.
| |
| * Radiological changes in plant during shift:
| |
| Part 2 Completed by on-coming shift prior to assuming duties Q Review station rounds / Abnormal reading (AUOs only)
| |
| Q Review Narrative Logs (previous day and carly-over items)
| |
| [] Current qualification status 0 Review the current controlling Reactivity Management Plans (N/A for AUOs)
| |
| Review current TS/TRMIODCMIFPR Required Actions (N/A for AUOs)
| |
| Q Walkdown MCR Control Boards with off-going Operator (N/A for AUOs, as applicable for SM/STAs)
| |
| C SR/PER reviews complete for previous shift (SM/U S/STA)
| |
| Relief Time: Relief Date:
| |
| Part 3 Completed by on-coming shift. These items may be reviewed after assuming duties:
| |
| L Review Operator Workarounds, Burdens and Challenges (applicable Unit/Station)
| |
| Review applicable ODMI actions (first shift of shift week)
| |
| C] Review changes in Standing / Shift Orders (since last shift worked)
| |
| C] Review changes to TACFs issued (since last shift worked) (N/A for AUOs)
| |
| [] Review Control Room Deficiencies (first shift of shift week ) (N/A for AUOs)
| |
| C] Review Component Deviation Log (N/A for AUOs)
| |
| TVA 40741 Page 1 of 1 OPDP-1-1 [01-14-2011]
| |
| | |
| Scenario b SHifT TURNOVER ChECKLIST Page 1 of2 SHIFT TURNOVER CHECKLIST Page 1 of I D SM
| |
| [] US/MCR Unit 1 UO Unit Off-going Name AUO Station Q STA (STA Function) On-coming Name Part 1 Completed by off-going shift! Reviewed by on-coming shift:
| |
| RCS Cb = 1616 ppm
| |
| * Abnormal equipment lineup / conditions:
| |
| * SI/Test in progress/planned: (including need for conduct of evolution briefings)
| |
| * Major Activities/Procedures in progress/planned:
| |
| Reactor Startup is in progress. GO-2, Reactor Startup, Section 5.3, Reactor Startup is complete through Step 30, and step 31 is to be performed after assuming the shift. RCS boron concentration is 1616 ppm, Control Bank D is at 98 steps. All Low Power Physics testing is complete, and satisfactory.
| |
| * Radiological changes in plant during shift:
| |
| Part 2 Completed by on-coming shift prior to assuming duties Review station rounds / Abnormal reading (AUO5 only)
| |
| Q Review Narrative Logs (previous day and carry-over items)
| |
| C] Current qualification status C] Review the current controlling Reactivity Management Plans (N/A for AUOs)
| |
| C] Review current TS/TRM/ODCM/FPR Required Actions (N/A for AUOs)
| |
| C] Walkdown MCR Control Boards with off-going Operator (N/A for AUOs, as applicable for SM/STAs)
| |
| C] SR/PER reviews complete for previous shift (SM!US/STA)
| |
| Relief Time: Relief Date:
| |
| Part 3 Completed by on-coming shift. These items may be reviewed after assuming duties:
| |
| C] Review Operator Workarounds, Burdens and Challenges (applicable Unit/Station)
| |
| C] Review applicable ODMI actions (first shift of shift week)
| |
| C] Review changes in Standing / Shift Orders (since last shift worked)
| |
| C] Review changes to TACFs issued (since last shift worked) (N/A for AUO5)
| |
| [] Review Control Room Deficiencies (first shift of shift week ) (N/A for AUOs)
| |
| C] Review Component Deviation Log (N/A for AUO5)
| |
| TVA 40741 Page 1 of I OPDP-l-l [01-14-2011]
| |
| | |
| I !/i Watts Bar Nuclear Plant Unit I System Operating Instruction SOI-85M1 Control Rod Drive And Indication System Revision 0040 Quality Related Level of Use: Continuous Use Effective Date: 06-14-2011 Responsible Organization: OPS, Operations Prepared By: R. C. Davidson Approved Ey: Steve Smith
| |
| | |
| WBN Control Rod Drive And Indication SOI-85M1 Unit I System Rev. 0040 Page2of43 Revision Log Revision Affected or Change Effective Page Number Date Numbers Description of Revision/Change 35 06/23/06 2, 4, 36 Added Section 8.3 for instructions on updating ICS Rod Bank Demand position and changed 3 TOC to reflect the addition per operator feed back.
| |
| Corrected Section number for Section 5.5 in TOC.
| |
| 36 08/30/07 All This procedure has been converted from W95 to Word 2002 (XP) by the WBN Conversion Team.
| |
| 37 02/26/08 2, 8, 25 Incorporated DCN 52265 for new bank overlap.
| |
| 9 Added Mode 4 to P&L R (Operator Feedback).
| |
| 38 10/10/08 2, 17, 19, Implemented corrective action for PER 146695 20, 21, 39 by adding steps to ensure phase control cards failure LEDs are not lit. Implemented corrective action for PER 150740 by clarifying the method used to balance M-G sets. Clarified when alarm 85A will clear.
| |
| 39 10/09/09 2,23,24,25, Incorporated DCN 52957 to reflect change in 27,33,34, system description, power supplies and MCR 44 monitor-PLC pairing. Updated discussion section to reflect changes.
| |
| Corrected CERPI name to Westinghouse terminology.
| |
| Minor format change, converted listed monitor items to bulleted items, with checkmarks for place keeping.
| |
| 40 06/14/11 2,8, 17, Minor/Editorial Revision to correct key number 19, 27-30, for Control Rod Logic Cabinet on Pg 27 (SR#
| |
| 32, 34, 39 372087), and minimize the redundant use of the symbols +/- and when following the word within. (PCR 5485) Converted Checklist 1 to External Attachment-i P. Added End of Section designations where appropriate.
| |
| | |
| WBN Control Rod Drive And Indication SOI-85.01 Unit I System Rev. 0040 Page 3 of 43 Table of Contents
| |
| | |
| ==1.0 INTRODUCTION==
| |
| 5 1.1 Purpose 5 1.2 Scope 5
| |
| | |
| ==2.0 REFERENCES==
| |
| 6 2.1 Performance References 6 2.2 Developmental References 6 3.0 PRECAUTIONS AND LIMITATIONS 8 4.0 PREREQUISITE ACTIONS 11 4.1 PreliminaryActions 11 4.2 Field Preparations 11 4.3 Approvals and Notifications 11 5.0 STARTUP 12 5.1 Placing Initial Control Rod Drive Motor Generator Set in Service 12 5.2 Startup of Second Motor Generator Set for Parallel Operation 15 5.3 Closing Reactor Trip Breakers A and B 18 5.4 Shutdown Banks Withdrawal 22 5.5 Control Bank Withdrawal 26 5.6 Manual Rod Control With Reactor At Power 28 5.7 Transfer from Manual to Automatic Rod Control 29 5.8 Transfer From Automatic to Manual Rod Control 31 6.0 NORMAL OPERATION 32 6.1 Control Rod Drive and Motor Generator Sets 32 6.2 Rod Position Indicators 32 6.3 Analog Rod Position System 33 6.4 Shutdown Rod Positioning 34 6.5 Rod Insertion Limits 34 6.6 Rod Control In puts 34 7.0 SHUTDOWN 35
| |
| - 7.1 Removing Motor Generator Sets From Service 35 8.0 INFREQUENT OPERATIONS 37
| |
| | |
| WBN Control Rod Drive And Indication SOI-85.01 Unit I System Rev. 0040 Page4of43 Table of Contents (continued) 8.1 Opening Reactor Trip Breakers A and B 37 8.2 Balancing MG set A & B Current While Control Rods are Withdrawn 38 8.3 ICS Rod Bank Update 41 9.0 RECORDS 42 9.1 QARecords 42 9.2 Non-QA Records 42 Source Notes 43 ATTACHMENTS Attachment iF: Control Rod Drive And Indication System Power Checklist 85.01-IF
| |
| | |
| WBN Control Rod Drive And Indication SOI-85.01 Unit I System Rev. 0040 Page 5 of 43
| |
| | |
| ==1.0 INTRODUCTION==
| |
| | |
| 1.1 Purpose To provide instructions for Operation of the Control Rod Drive and Indication System in Manual and Auto.
| |
| 1.2 Scope This Instruction includes operation of the following subsystems:
| |
| A. Control Rod Drive (CRD) Motor Generator (MG) Sets B. Control Rod Drive System C. Rod Position Indicators (RPIs)
| |
| | |
| WBN Control Rod Drive And Indication SOl-85.01 Unit I System Rev. 0040 Page 6 of 43
| |
| | |
| ==2.0 REFERENCES==
| |
| | |
| 2.1 Performance References A. SPP-2.2, Administration of Site Technical Procedures B. SPP-1O.4, Reactivity Management Program C. 1-SI-99-4-A, Trip Actuating Device Operational Test of Reactor Trip P-4 ESFAS Interlock Train A D. l-SI-99-4-B, Trip Actuating Device Operational Test of Reactor Trip P-4 ESFAS Interlock Train B E. 1-TRI-85-1, Reactivity Control Systems Movable Control Assemblies (Modes 3, 4, and 5)
| |
| F. Tl-85.005, Quarterly Resetting of Full Out Rod Positions 2.2 Developmental References A. N3-85-4003, System Description for Control Rod Drive B. NOB SHEET A-7, Full Out Rod Position C. Tech Spec 3.1.3, Core Reactivity D. Tech Spec 3.1.5, Rod Group Alignment Limits E. Tech Spec 3.1.7, Control Bank Insertion Limits F. Tech Specification 3.2.3 Axial Flux Difference G. Tech Specification 3.4.5 RCS Loops- Mode 3 H. Tech Specification 3.1.8 Rod Position Indication I. Tech Requirements 3.1.7 Position Indication Sys, Shutdown J. TVA Drawings:
| |
| 45W600-99-1 45W760-85-1 45W708-1, -2 45W747-1, -2 45N1624-1 thru 11 45W1 646-4 45Nt&74-2
| |
| | |
| WBN Control Rod Drive And Indication SOI-8&01 Unit I System Rev. 0040 Page7of43 2.2 Developmental References (continued)
| |
| K. WA Vendor Manual WBN-VTM-W120-2414 Reactor Trip Circuit Breakers L. Westinghouse Drawings and Vendor Manuals:
| |
| Westinghouse () Full Length Rod Control, Vol I, II & Ill
| |
| [D&VMN 0855, 0856, 0857]
| |
| W 1051E05 Sheet 1-14(Contract54ll4-1)
| |
| W 6056D35 Sheet 20 W Rod Position Indication Sys (D&VMN 0853, Contract 54114-1)
| |
| W KD-8805-76 1-6 RC W 1054E62 W 1045F60 W 0092-2-1 thru 4 (Contact 54114-1)
| |
| Electrical Equipment for AC Power Supply System for Nuclear Reactor Rod Controller (D7VMN 0860)
| |
| | |
| WBN Control Rod Drive And Indication SOl-85.01 Unit I System Rev. 0040 Page8of43 3.0 PRECAUTIONS AND LIMITATIONS
| |
| * Work in a Radiological Control Area (RCA) requires the use of existing RWPs, and may require additional ALARA Preplans. Failure to follow posted Rad control requirements can cause unnecessary radiation exposure. Rad Con should be notified of work having the potential to change radiological conditions.
| |
| Before Closing Reactor Trip Breakers (RTB5), the Bank Select Switch should be in MANUAL.
| |
| Before shutting down both MG Sets, all rods should be in the core, and RTBs should be OPEN.
| |
| Group Demand Position Indicators shall be OPERABLE and capable of determining within 2 steps the demand position for each Shutdown or Control Rod that is NOT fully inserted.
| |
| Rod Banks must be operated in prescribed sequence. For WITHDRAWAL sequence is Shutdown Bank A, B, C, D; then Control Bank A, B, C, D (maintaining Control Bank Overlap). The INSERTION sequence is the reverse of withdrawal.
| |
| 1 For Manual Bank sequencing, prescribed withdrawal/insertion sequence must be followed. Correct bank position is monitored by checking the Step Counters and Rod Position Indicators.
| |
| 1 Control Bank overlap must be maintained to ensure Acceptable Core Power distributions. When Control Bank A reaches 116 Steps, verify Control Bank B 1
| |
| begins to move. Observe this relationship for Control Banks C and D.
| |
| With ROD BANK SELECTOR SWITCH in BANK SELECT, Control Bank Overlap is NOT maintained (Control Banks may be moved in Bank Select for Physics Testing).
| |
| Control Banks must be maintained above their insertion limits (Lo-Lo Alarm) to ensure:
| |
| Adequate shutdown margin Maximum Ejected Rod reactivity limits are maintained Acceptable Core Power distributions In Modes 1 and 2, before withdrawing a Rod Bank from the FULLY INSERTED position, the banks Group Step Counters must be at zero steps, and RPIs for that bank must be indicating +/- 12 steps. During withdrawal, they must remain
| |
| - wfthil9 12tepsfaeh other.
| |
| 1 -
| |
| | |
| WBN Control Rod Drive And Indication SOI-85.01 Unit I System Rev. 0040 Page9of43 3.0 PRECAUTIONS AND LIMITATIONS (continued)
| |
| . After CONTROL ROD URGENT FAILURE Alarm [86A], 1-RCAR, ROD CONTROL ALARM RESET, is NOT to be reset UNTIL cause of alarm is determined. Some failures may cause rods to step in if reset before correcting problem.
| |
| . Control Banks must be operated to maintain Axial Flux Difference within limits, AND to stay above insertion limits. Boron dilution should be completed BEFORE Control Rods reach the upper limit (Bank D, 220 steps).
| |
| 1 When possible, the Shutdown (SD) Banks should be FULLY WITHDRAWN during the conditions below to provide trippable reactivity.
| |
| \
| |
| All Shutdown Banks must be 225 steps when positive reactivity is being inserted by Boron changes, Xenon changes, RCS temp changes, or motion of control rods; except when:
| |
| Reactor Coolant System (RCS) temp and Boron Conc (CB) are maintained at HOT STANDBY XENON FREE CONDITION (Mode 4), or RCS is borated to COLD SHUTDOWN CB (Mode 5).
| |
| * If the Shutdown Banks cannot be withdrawn, the RCS must be borated as conditions require and the CB confirmed by sampling.
| |
| * During Power Operation, all RPIs and Power Range Channels must be monitored for Rod misalignment and abnormal power tilts.
| |
| The Rod Control System must be switched to MANUAL if the Power ismatch Channel or any necessary input is out of service.
| |
| Do NOT operate with more than one Control Rod inoperable (See AOI-2 for inoperable Control Rod).
| |
| * If in Mode 3 or 4, ensure at least two RCS loops in operation with two RCPs in service BEFORE closing Reactor Trip Breakers.
| |
| If STEP COUNTER goes below 0 while driving Control and Shutdown rods in:
| |
| Realign affected groups Step Counter using UP push button to raise counter output to 000.
| |
| Ensure Plant C rnputerpointslor Rod Bai,1 Step Posftiun are correct.
| |
| | |
| WBN Control Rod Drive And Indication SOI-85.01 Unit I System Rev. 0040 PageIOof43 3.0 PRECAUTIONS AND LIMITATIONS (continued)
| |
| Instrument Maintenance should be notified to ensure required instrumentation will be placed in service as necessary to support system operation.
| |
| A delay of 2 to 5 seconds should be allowed between demands for rod motion when using 1-FLRM, ROD MOTION CONTROL switch, to step rods to avoid arcing/welding of contacts.
| |
| * RCS pressure must be 100 psig, OR the RCS has been vacuum filled in accordance with GO-b, prior to energizing the CRDM coils.
| |
| | |
| WBN Control Rod Drive And Indication 501-85.01 Unit I System Rev. 0040 Page 26 of 43 TODAY Date Initials 5.5 Control Bank Withdrawal Control Bank withdrawal is a principal reactivity control evolution. Requirements of SPP-10.4, Reactivity Management Program, shall be complied with at all times. The following step should be accomplished by holding crew briefings which includes applicable sections of SPP-10.4.
| |
| ENSURE SPP-10.4 REVIEWED by personnel performing this Section. TBD CU N During withdrawal of Control Banks, observe prop r overlap. When Control Bank A reaches 116 Steps, Control Bank B should begin to move. When Bank B reaches 116 Steps, Bank C should move. When Bank C reaches 116 Steps, Bank D should move.
| |
| Once Control Bank A is above 20 Steds at Bottom alarm light on 1 -XA-55-4B J4\[87D] can be reset. Once Control Bank A gets above 20 steps and Rods are driven in, the alarm will come in. Once Control Banks B, C, and D get above 35 Steps, then drop below 20 Steps, the alarm will come back in.
| |
| Once criticality has been established, proceed with the performance of GO-2.
| |
| jActuaI Full Out Position of the Control Banks will be determined by the quarterly X\performance of Tl-85.005. Full Out position will vary from 225 Steps to 231 Steps and can be verified on NOB Sheet A-7 Full Out Rod Position.
| |
| () If it is necessary to rotate 1-RBSS, ROD BANK SELECTOR SWITCH, through the
| |
| _Y\AUTO position, Tavg-Tref should be matched within 1 °F to avoid unexpected rod motion. Group demand step counters should be verified to indicate the same position before and after the switch is rotated.
| |
| ENSURE all Control Bank Group Step Counters and Control Bank Rod Position Indicators at zero prior to initiation of Rod TBD Withdrawal.
| |
| ENSURE SctiQn 54 compIet, nd all Shutdown Rods are fully withdrawn.
| |
| | |
| WBN Control Rod Drive And Indication SOl-85.01 Unit I System Rev. 0040 Page 27 of 43 TODAY Date Initials 5.5 Control Bank Withdrawal (continued)
| |
| X\
| |
| Key # 19 may be needed to peorm the fogtep.
| |
| ENSURE the following displayed in the Control Rod Logic Cabinet [control rod drive room, el 782, panel 1-L-122]:
| |
| IF Bank Overlap Counter does NOT display 000, THEN PRESS RESET Button [top button just left of the Bank TBD Overlap Display].
| |
| IF Master Cycler Logic (MCL) card A105 [fifth slot from left in top row with three LEDs] does NOT display bottom LED lit, THEN PRESS Master Cycler +1 Button [at top center of logic panel] until at least the bottom LED, of the three LEDs, is lit. TBD ENSURE ROD BANK SELECT SWITCH, 1-RBSS, in TBD MANUAL.
| |
| TBD OBTAIN SRO Approval to withdraw Control Banks.
| |
| PLACE the 1-FLRM, IN-HOLD-OUT SWITCH, to OUT to begin intermittent programmed withdrawal of Control Banks to obtain criticality.
| |
| [8] MONITOR the following as the Control Banks are withdrawn:
| |
| Group Step Counters
| |
| * RPls D
| |
| * Rod Speed (48 Steps/Minute) D
| |
| * Proper Bank Overlap D End of Section
| |
| | |
| WBN Control Rod Drive And Indication SOl-85.01 Unit I System Rev. 0040 Page28of43 Date Initials 5.6 Manual Rod Control With Reactor At Power NOTE The manipulation of Control Rod position to maintain required parameter(s) is a continuous action by a Licensed Reactor Operator. The following is exempt from the Continuous Use requirements of SPP-2.2.
| |
| [I] ENSURE ROD BANK SELECT SWITCH (1-RBSS) in MANUAL.
| |
| [2] POSITION Control Rods as necessary to maintain Tavg with Tref using 1-FLRM, IN-HOLD-OUT SWITCH (maximum Tavg-Tref deviation <3.0°F).
| |
| [3] WHEN AUTOMATIC Rod control is desired, THEN ENSURE Tavg is within 1.0°F of Tref to avoid immediate rod movement on transfer.
| |
| CAUTION Allowing at least 5 minutes between any rod control input (i.e., T-avg, T-ref, or NIS) change and placing rods in AUTO, will help prevent undesired control rod movement.
| |
| [4] ENSURE zero demand on control rod position indication
| |
| [1 -M-4].
| |
| [5] PLACE ROD BANK SELECT SWITCH (1-RBSS) in AUTO.
| |
| [6] WHEN Rod Control is in AUTO, THEN ENSURE the following:
| |
| A. Tavg and Tref within 1.5°F B. Step counters and RPls within 12 steps C. Bank Overlap maintained D. Power distribution within limits, AFD/QPTR End of Section
| |
| | |
| I 17i Watts Bar Nuclear Plant Unit I General Operating Instructions GO-2 Reactor Startup Revision 0039 Quality Related Level of Use: Continuous Use Effective Date: 04-20-201 1 Responsible Organization: OPS, Operations Prepared By: R. A. ORear Approved By: Brian Mcllnay
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 2 of 43 Revision Log Revision Affected or Change Effective Page Number Date Numbers Description of RevisionlChange 34 10/08/09 3, 18 Revised step 5.3[8.3j to direct continued performance at step 5.3[21]. Pet 201 no longer directs the actions equivalent to steps 5.3[21 ]-[22]
| |
| 34/ UC-1 10/19/09 9 Made step 4.0[7] conditional, there is no requirement the TDAFWP be running.
| |
| 35 10/20/09 9 Incorporate UC-1 to revision 34.
| |
| 36 06/02/10 31 Incorporate UC-1 to revision 35 37 10/20/10 2, 6, 7, 10, Minor editorial changes including corrective 12-15, 17, action for PER 231487-001 and ODM-23 18, 20-22, required changes.
| |
| 24-26, 28-30, 34, 35, 37 38 02/16/11 2, 6, 7,10, Clarified the sequence for resetting TDMFWPs 12, 13, 16, and the notes associated with LCO 3.3.2 AFW 23, 27, 33, start signals, in preparation for startup of initial 34, 35, 38, MEW pump. [PER 288125, 231519]
| |
| 42 Added step to address the need for a local reset of CERPI Trouble Alarm and referenced IMI-85.001 [PCR 4810] Added note to Appendix A allowing out of sequence performance and formatted numbers to match.
| |
| Minor changes including correcting format of Notes and Cautions.
| |
| 39 04/20/1 1 2, 18 Minor/Editorial change to reflect deletion of administrative requirement, formerly in Tech Specs, to have the Hydrogen Recombiners Operable for Mode 2. Rev 94 of Tech Specs deleted LCO 3.6.7 Hydrogen Recombiners TS requirement.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page3of43 Table of Contents
| |
| | |
| ==1.0 INTRODUCTION==
| |
| 4 1.1 Purpose 4 1.2 Scope 4
| |
| | |
| ==2.0 REFERENCES==
| |
| 4 2.1 Performance References 4 2.2 Developmental References 6 3.0 PRECAUTIONS AND LIMITATIONS 6 3.1 Precautions 6 3.2 Limitations 7 4.0 PREREQUISITES 9 5.0 INSTRUCTIONS 11 5.1 General 11 5.2 Actions Performed Before Reactor Startup 12 5.3 Reactor Startup 19 6.0 RECORDS 37 6.1 QARecords 37 6.2 Non-QA Records 37 Appendix A: Mode 3-To-Mode 2 Review And Approval 38 Source Notes 43
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page4of43
| |
| | |
| ==1.0 INTRODUCTION==
| |
| | |
| 1.1 Purpose This Instruction provides actions to perform unit startup from Hot Standby (Mode 3) at normal operating temperature and pressure to between 1 and 4% Reactor Power (Mode 2).
| |
| 1.2 Scope This GO contains the following:
| |
| 5.0 Instructions 5.1 General 5.2 Actions Performed Before Reactor Startup 5.3 Reactor Startup
| |
| | |
| ==2.0 REFERENCES==
| |
| | |
| 2.1 Performance References A. 0-Sl-03, Weekly Log B. 1-Sl-0-10, Shutdown Margin C. 1-Sl-0-11, Estimated Critical Position D. 1-SI-0-2 Series, Shift And Daily Surveillance Log E. 1-S1-0-903, Primary Pressure Boundary Isolation Valve Leak Test (Safety Injection Primary Check Valves)
| |
| F. 1-SI-47-76, Trip Actuating Device Operational Test (TADOT) Turbine Trip Low Fluid Oil Pressure Channels I, II, Ill G. 1-S1-47-77, Trip Actuating Device Operational Test (TADOT) Turbine Trip Turbine Stop Valve Closure H. 1-S1-68-34, Minimum Temperature for Criticality TAVG-TREF Deviation Alarm not Reset I. 1-51-85-1, Rod Drop Time Measurement by Simultaneously Dropping All Rod Banks
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page5of43 2.1 Performance References (continued)
| |
| J. 1-SI-85-I0, Rod Drop Time Measurement Using CERPI Rod Drop Test Corn puter.
| |
| K. 1-SI-85-2, Reactivity Control Systems Moveable Control Assemblies (Mode 1 and 2)
| |
| L. 1-SI-85-3, Calibration (Cold/Hot) of Rod Position Indication Channels and Full Range Verification M. 1-Sl-92-8, Power Range Overpower Trip High Range Bistable Adjustment for Limiting Condition for Operation N. 1-SI-99-4-A and (B), Trip Actuating Device Operational Test of Reactor Trip P-4 ESFAS Interlock Train A (B)
| |
| : 0. 1-TRI-85-1, Reactivity Control Systems Movable Control Assemblies (Modes 3,4, and 5)
| |
| P. CM-3.01, System Chemistry Specifications Q. CM-5.08, Startup Primary Chemistry Control R. Nuclear Operating Book (NOB)
| |
| S. OPDP-1, Conduct of Operations T. PET-I 07, Mode 3 Physics Testing U. PET-201, Initial Criticality and Low Power Physics Testing V. SO 1-2 & 3.01, Condensate And Feedwater System W. SOI-62.02, Boron Concentration Control
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page6of43 2.1 Performance References (continued)
| |
| X. SOl-62.04, CVCS Purification System Y. SOl-68.02, Reactor Coolant Pumps Z. S0l-85.O1, Control Rod Drive And Indicating System AA. TI-127, Reactor/Turbine Trip Report, Event Critique, Root Cause Analysis BB. Tl-34.04, Loose Parts Monitoring Gain and Alarm Setpoints CC. O-Pl-OPS-1.1, Jumper Control Process\
| |
| DD. IMI-99.040A, Maintain Source Range in Service Jumpers EE. lMl-85.OO1, Computer Enhanced Rod Position Indication (CERPI) Diagnostics 2.2 Developmental References A. WBN, Technical Specifications B. GOl-7, Generic Equipment Operating Guidelines 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions NOE If a precaution cannot be complied wit the SM shall initial, date, and write a brief explanation. Precautions that contain must, shall, or will, must be adhered to.
| |
| This instruction may be entered from a partial shutdown. N/A section(s) not applicable and annotate reason.
| |
| During power changes, maximize letdown flow when possible to minimize crud induced power shifts.
| |
| nticipate criticality anytime reactivity changes are being made. [0.1]
| |
| In Hot Standby, maintain Shutdown Margin to account for positive reactivity addition from Xenon transients to ensure inadvertent criticality does not occur.
| |
| ( The SR and lR instruments must be closely observed to detect an inadvertent criticality during Xenon transients.[c.1]
| |
| Pzr-RCS C 8 difference should be less than or equal to 50 ppm and is maintained by use of Pzr heaters and spray.
| |
| | |
| WBN Reactor Startup GQ-2 Unit I Rev. 0039 Page 7 of 43 3.1 P ecautions (continued)
| |
| Contact Reactor Engineering for guidance on core operating recommendations during unusual power maneuvers such as startup at End of Life (EOL).
| |
| refueling, NIS indications may be inaccurate until calibrated at higher power levels. NIS calibration procedures will adjust PRM trip setpoints lower After than normal to ensure excore detectors protect against an overpower condition.
| |
| In Mode 2 (less than or equal to 5%), sudden temperature decreases, or C 3 changes greater than 10 ppm, should be avoided. The operator should be alert to secondary steam flow to avoid cooling the RCS below the Minimum Temperature for Criticality of greater than or equal to 555°F, and/or causing a spurious Safety lnjection.[c.2]
| |
| All jumper installation and removal shall be in accordance with 0-Pl-OPS-1 .1, Jumper Control Process.
| |
| In Mode 2, the trip function of all Turbine Driven Main Feedwater Pumps (TDMFWP) is required when one or more (TDMFWP) is supplying feedwater to the Steam Generators. During the process of placing the first TDMFW pump in service, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is deenergized to prevent inadvertent AFW auto-start during rollup trip testing and overspeed trip testing. Once the operating TDMFW pump has established sufficient feed flow to maintain SG level, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is placed in the trip condition, and the AFW pumps secured. Refer to Tech Spec 3.3.2 Table 3.3.2-1, Function 6.e. and B 3.3.2.6.e.
| |
| The Greek symbol (p) precedes steps associated with direct reactivity manipulations, for example: RCS dilution and boration, control rod manipulations.
| |
| 3.2 Limitations In Mode 2 with Keff less than 1.0, or in Mode 3 or 4, Shutdown Margin shall be maintained greater than or equal to 1600 pcm (T.S. 3.1.1).
| |
| $ In Mode 3, at least two RCPs shall be operable with two loops in operation when the Rod Control System is capable of rod withdrawal, and at least one RCP shall be in operation when the Rod Control System is not capable of rod withdrawal (T.S. 3.4.5).
| |
| () SOURCE RANGE HI FLUX AT SHUTDOWN alarm shall be in operation any time the Reactor is shutdown with fuel in the Reactor vessel.
| |
| D. Simultaneous reactivity addition by rod withdrawal and dilution is not allowed while in the Source range.
| |
| /
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page8of43 3.2 Lijritations (continued)
| |
| If at any point during the approach to criticality, ONE of the two Source Ranges 7 shows an unexplained rise in count rate by a factor of 5 or greater, or if BOTH Source ranges show an unexplained rise in count rate by a factor of 2 or greater, the approach to critical shall be SUSPENDED IMMEDIATELY and the control rods FULLY INSERTED (i. e., rod withdrawal and/or boron dilution shall be terminated). Further positive reactivity changes shall not be resumed UNTIL an evaluation is performed, Plant Manager approval obtained, and the SM authorizes resuming the approach to critical.Ic1]
| |
| ) A member of Operations Management Staff, not assigned to the operating crew, shall be present in the control room during the approach to criticality.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 9 of 43 Date 7 d c
| |
| 1ui lnitiaIs--
| |
| 4.0 PREREQUISITES STARTUP No.
| |
| Throughout this Instruction where an EN statement exists, the step should be marked N/A if the condition does not xist.
| |
| / - ) Prerequisite Actions may be complied with in any order. If a prerequisite cannot be complied with, the SM shall initial, date, and write a brief explanation. Prerequisites that contain must, shall, or will, cannot be N/Ad.
| |
| Throughout this instruction, Concurrent Verification (CV) for breaker or fuse manipulations may be marked N/A if no manipulations are performed.
| |
| MAINTAIN Pzr pressure in the normal operating band with Pzr C_K heaters and spray.
| |
| MAINTAIN Pzr level at greater tha or equal to 25%.
| |
| N9E TAVG will vary as a function of reactor power ntil the unit is greater than 15% turbine load (C5) and the Tavg program is maintained by AUTO or manual rod control. The TAVG-TREF deviation alarm should be expected as reactor power approaches 7% RTP.
| |
| MAINTAIN Steam Dumps in the Pressure Mode set at 84%
| |
| (1092 psig), or with SG PORVs set at 84%.
| |
| IF SG PORVs are in service, THEN ENSURE adequate CST level. J,4 IF A-A AFW Pump is NOT running, THEN MANUALLY START the A-A AFW Pump in accordance with SOl-3.02, Section 8.1.1.
| |
| IF B-B AFW Pump is NOT running, THEN MANUALLY START the B-B AFW Pump in accordance with SOl-3.02, Section 8.1.2.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 10 of 43 1
| |
| Date 7dzy Initials_1-4.0 PREREQUIlTES (continued)
| |
| IF TD AFW Pump is desired to be running AND TD AFW Pump is NOT running, THEN MANUALLY START the TD AFW Pump in accordance with SOl-3.02, Section 8.1.3.
| |
| MAINTAIN SG levels on program with AFW pumps.
| |
| ENSURE all RCPs are operating per SOl-68.02, ENSURE Source Range jumpers have been removed per lMl-99.040A, Maintain Source Range in Service jumpers.
| |
| (Ref GO-I>.
| |
| WHEN Source Range jumpers are removed, THEN REMOVE Caution Order from Reactor trip breakers, RTA and RTB.
| |
| (ØT CERPI Trouble Alarm can be diagnosed a d reset using lMl-85.OO1
| |
| [1] CONTACT Work Control for assistance to reset CERPI Trouble Alarm. (N/A if alarm previously reset)
| |
| CONDUCT a pre-evolution briefing in accordance with OPDP-1, stressing the following:
| |
| 0 Management Expectations 0 Limitations/Precautions Communications Chain of Command Conservative actions when repositioning control rods during approach to critical SM
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page II of43 Date________ Initials_-
| |
| 5.0 INSTRUCTIOI(JS 5.1 General (N53E Section 5.0 steps must be performed seqiially, unless specifically stated otherwise.
| |
| Prior SM approval is required to deviate from this sequence.
| |
| [1 ENSURE Section 4.0, Prerequisites, COMPLETE. ----4---.--
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 12 of 43 Date________ Initials --
| |
| 5.2 Actions Performed Before Reactor Startup
| |
| ()
| |
| 7 INITIATE Appendix A, Mode 3-To-Mode 2 Review and Approval, while continuing with this instruction.
| |
| ituji)
| |
| ENSURE Daily Scheduling Supervisor has issued the Mode 3-to-Mode 2.1 surveillance reports to the applicable department sections to prepare for Mode 2 entry. -*
| |
| ([ ENSURE Daily Scheduling Supervisor has issued the Reactor
| |
| -1 Trip Breaker reports to the applicable department sections before closing the Reactor trip breakers (N/A if previously performed).
| |
| The Pzr steam space sample should remain in service per CM-5.08 for removal of
| |
| [ncondensable gases.
| |
| COORDINATE with Chemistry to establish primary and secondary startup chemistry controls per CM-5.08, and CM-3.O1.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 13 of 43 DateZ7day Initials_____
| |
| 5,2 Actions Performed Before Reactor Startup (continued)
| |
| CN Failure to have all three AFW pumps in option with the FWPs tripped and trip buses energized will initiate an ESF Actuation.
| |
| Step 5.2[5.1] may be initiated early tore the first Turbine Driven Main Feedwater (TDMFW) pump will be ready to start when Reactor power is 1 to 4%.
| |
| In Mode 2, trip function of all TDMFW pumps is required when one or more TDMFW pumps are supplying feedwater to the Steam Generators. Refer to Tech Spec 3.3.2 Table 3.3.2-1, Function 6.e .[c.53 During the process of placing the first TDMFW pump in service, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is deenergized to prevent inadvertent AFW auto-start during rollup trip testing and overspeed trip testing. Once the operating TDMFW pump has established sufficient feed flow to maintain SG level, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is placed in the trip condition, and the AFW pumps secured. B 3.3.2.6.e PERFORM the following to prepare first TDMFW pump to support startup:
| |
| SELECT which MFW pump will be started first: (N/A non-selected pump) 1AMFW pump.
| |
| lBMFWpump.
| |
| .2] IF 1A MFW pump will be started first, THEN PERFORM the following:
| |
| 5 .1 ENSURE 1-BKR-46-36, 250V DC FDR FOR MFPT B TRIP BUS UNIT I in OFF. [250V Bat Bd 1, Panel 3]
| |
| cv ENSURE 1-BKP,-46-9, 250V DC FDR FOR MFPT A TRIP BUS UNIT 1 in ON. [250V Bat Bd 1, Panel 3]
| |
| cv
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Pagel4of43 Date________ Initials 1 52 Actions Performed Before Reactor Startup (continued)
| |
| PERFORM the applicable steps of SOl-2&3.O1 to RESET and WARM 1A MFW pump.
| |
| IF 1 B MEW pump will be started first, THEN PERFORM the following:
| |
| [5. .1 ENSURE l-BKR-46-9, 250V DC FDR FOR MFPT A TRIP BUS UNIT I in OFF. [250V Bat Bd 1, Panel 3)
| |
| ENSURE 1-BKR-46-36, 250V DC FDR FOR MFPT B TRIP BUS UNIT 1 in ON. [250V Bat Bd 1, Panel 3)
| |
| PERFORM the applicable steps of SOI-2&3.O1 to RESET and WARM lB MFW pump.
| |
| IF an SI signal has occurred, THEN CYCLE the Reactor Trip Breakers.
| |
| SELECT HIGHEST reading Source Range and Intermediate
| |
| -1 Range channels to record on 1NR-92-145.
| |
| ENSURE the following:
| |
| .1 Audio Count Rate channel IN OPERATION.
| |
| Audio Count Rate selected to the highest Source Range.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 15 of 43 Date7fctf Initials_--
| |
| 5.2 Actions 4 Per o rmed Before Reactor Startup (continued)
| |
| The Loose Parts Monitoring System (LPMS alarm is required to be in service prior to Mode 2 entry (TR 3.3.6.)
| |
| PERFORM Step 5.2[9.1] OR 5.2[9.2) below
| |
| ,NIA option NOT used):
| |
| 5 (J IF the plant has been maintained in Hot Standby Conditions, THEN PERFORM the following to enable LPMS alarm
| |
| [O-R-139 panel, Aux lnstr. Rm]
| |
| PRESS SYSTEM RESET switch [1433P module]
| |
| PLACE ANNUNCIATOR INHIBIT switch
| |
| [1439 module] in DOWN position.
| |
| PLACE TAPE RECORDER INHIBIT AUTO START switch [1436WB module] in DOWN position.
| |
| IF plant has NOT been maintained in Hot Standby Conditions, THEN ENSURE Instrument Maintenance (IM) completes performance of TI-34.04 to place LPMS in service.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Pagel6of43 Date______ Initials 5.2 Actions Perlormed Before Reactor Startup (continued)
| |
| IF Shutdown Rods are INSERTED, THEN 1 .1 CHECK Reactor Trip Breaker Closure Surveillance report (generated by scheduling) is COMPLETE.
| |
| ENSURE Shutdown Bank withdrawal criteria MET per 1 i-si-o-io.
| |
| ENSURE all Reactor Trip first-out alarms RESET.
| |
| [1 .4 PLACE Control Rod Drive System in service per SOl-85.O1.
| |
| .5] CLOSE Reactor Trip Breakers per SOl-85..O1.
| |
| .6j PERFORM the following concurrently:
| |
| A. 1-TRI-85-1 (N/A if performed within the previous 31 days),
| |
| B. 1-Sl-85-2 (N/A if in frequency).
| |
| PERFORM the ROD BANK UPDATE on the ICS Computer (ROD BANK UPDATE is on NSSS Screen).
| |
| ENSURE Instrument Maintenance (IM) has completed 1-Sl-99-4-A and -B, to check P-4 lnterlock.[c.3]
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 17 of 43 Date________ Initials 5.2 Actions Performed Before Reactor Startup (continued)
| |
| Procedure PET-I 07, Mode 3 Physics Testing, satisfies low power physics testing if completed satisfactorily. If not, further testing will be required per PET-201 Initial Criticality, and Low Power Physics Testing, which includes steps to withdraw Shutdown Banks. IF PET-107 and PET-201 are not required, N/A Step 5.2[I0.9]
| |
| COORDINATE with Reactor Engineering to ensure PET-107, Mode 3 Physics Testing and PET-201, Initial Criticality and Low Power Physics Testing completed as necessary for startup.
| |
| Control Room activities not related to reactor startup should be minimized during the approach to criticality. SM should not permit shift turnover or other distractions to cause insufficient attention to the approach to criticalitY.
| |
| NIS Instruments must be monitored iipation of unplanned reactivity changes,[ci)
| |
| During rod withdrawal, Rod Position Indicators (RPls) should be compared with the Step Counters to determine if rod misalignment or other rod related problems exist.[c.4]
| |
| j) COLR defines fully withdrawn as a band for the Shutdown and Control banks.
| |
| F Computer constant K0015 provides current value, and should be consistent with NOB Sheet A-7, Monthly Full Out Rod Position.
| |
| (p) WITHDRAW Shutdown Rods to fully withdrawn per SOI-85.0l.
| |
| () TIF refueling EN was performed OR Reactor Vessel Head removed, 1 .1 ENSURE Rod Drop Timing per either 1-Sl-85-10 OR l-Sl-85-1 has been performed.
| |
| : 17. ENSURE IM has performed RPI Calibration per
| |
| / l-Sl-85-3.
| |
| | |
| WBN Reactor Startup GO-2 Uniti Rev. 0039 Page 18 of 43 Date________ Initials --
| |
| 5.2 Actions Performed Before Reactor Startup (continued)
| |
| ENSURE Permanent H 2 Mitigation System operable prior to
| |
| -i going to Mode 2 per SOl-268.O1 Checklist 1 2, & 3.
| |
| lNATE Appendix A, Mode 3-To-Mode 2 Review And
| |
| / Approval, to ensure ALL restraints to Mode 2 entry are resolved, and approvals for mode change granted.
| |
| SM
| |
| ,4i ENSURE IM has completed the following Trip Actuating
| |
| /- Device Operational Tests (TADOTs) (N/A if performed within 31 days before reactor startup):
| |
| 1-S 1-47-76 (4 1 -Sl-47-77 7 End of Section
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 19 of 43 Date 1cc/_y Initials_-
| |
| 5.3 Reactor Startup Anytime reactivity changes are made, criticality must be anticipated. Closely monitor NIS instruments.
| |
| ENSURE Section 5.2, Actions Performed Before Reactor Startup, have been COMPLETED.
| |
| BLOCK SOURCE RANGE HI FLUX AT SHUTDOWN alarm by placing both HI FLUX AT SHUTDOWN switches at SR NIS
| |
| / panels [M-13] to BLOCK.
| |
| () CHECK Alarm 81 C, SOURCE RANGE HI FLUX AT 7 SHUTDOWN ALM BLOCKED, is LIT.
| |
| CHECK greater than or equal to 0.5 cps on highest SR NIS.
| |
| glTE)
| |
| The following count rates may be used as ference during the approach to criticality.
| |
| RECORD SR NIS count rates from 5.3[5.1) or 5.3[5.2] (N/A NIs not used):
| |
| 51]
| |
| 1-M-4 (preferred) 1-Nl-92-131A . cps 1-Nl-92-132A 9 5 cps SR NIS Drawer [1 -M-1 31 1-Nl-92131D P/A cpsl-Nl-92-132E cps
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 20 of 43 Date iday InitiaIs 5.3 Reactor Startup (continued>
| |
| After refueling, NIS indications may be ina&urate until calibrated at higher power. NIS calibrations will adjust PRM setpoints lower than normal to ensure excore detectors protect against an overpower condition. Redundant indications of Reactor Power should be used until confidence is established in the PR indicators (Kf, Turbine Power, NIS).
| |
| 1N N TE ICRR plot being performed should be refer ed.to until criticality is achieved.
| |
| IF initial startup following refueling, OR activities occurred which could cause non-conservative NIS response, THEN ENSURE Power Range high-flux trip setpoints are reduced to less than or equal to 85% RTP.
| |
| IM Date Time
| |
| ?UM9i Avoid operations that could produce sudd?perature changes or unplanned CB
| |
| [anges during approach to critical or at low power.
| |
| MONITOR Source and Intermediate range Nls during approach to critical to identify potential reactivity anomalies.
| |
| IF initial startup is following refueling, THEN PERFORM the following:
| |
| ENSURE Shift and Daily Surveillance Log Mode 2, 1 -SI-0-2A-02 (1900-0700) or 1 -Sl-0-2B-02 (0700-1900) is COMPLETE for Mode 2 entry.
| |
| Initials Date Time NOTIFY on-site personnel of Reactor startup over P/A.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 21 of 43 Date________ InitiaIs-5.3 Reactor Startup (continued)
| |
| LSR 3i.7.1 (ECP above insertion limit), is triggered in PET-201.
| |
| IF dilution to critical is required, THEN ENSURE Appendix A, Mode 3-To-Mode 2 Review And Approval, COMPLETE, AND, LOG Mode 2 entry in Narrative Log as directed by PET-201, Initial Criticality and Low Power Physics Testing.
| |
| GO TO step 5.3[21].
| |
| If performing PET-201, steps 5.3[9} through 5.3[22] are N/A.
| |
| CALCULATE Estimated Critical Position (ECP) per I -Sl-0-1 1.
| |
| icdty Initials Date Time 0] IF actual C8 is NOT within 5 ppm of Estimated Critical C, THEN RECALCULATE ECP OR PERFORM the following:
| |
| 3j (p) DILUTE/BORATE per SOl-6202, to the Estimated Critical CB.
| |
| EQUALIZE RCS-Pzr C 3 (within 50 ppm) using Pzr heaters and spray.
| |
| WHEN sufficient mixing has occurred (30 minutes),
| |
| THEN REQUEST Chemistry to obtain a new C 3 sample.
| |
| * WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 22 of 43 Date________ Initials -1--
| |
| 5.3 Reactor Startup (continued)
| |
| ENSURE actual CB is within 5 ppm of estimated CB.
| |
| 1] RECORD the 750 pcm Upper and Lower rod position limits, Estimated Critical Position, and the Mode 2 rod position, determined by 1-Sl-0-11:
| |
| If ARO on Control Bank D (CBD) is less than 750 pcm, the upper limit rod position is the
| |
| [ontrol Bank D ARO number.
| |
| Upper rod position limit (1..6steps on bank D Lower rod position limit 7 steps on bank ..D Estimated Critical Position: /O steps on bank Mode 2, -1000 pcm rod position I$ steps on bank .D CHECK Shutdown Banks fully withdrawn, AND CHECK the ROD BANK UPDATE was updated on the ICS Computer (ROD BANK UPDATE is on NSSS Screen).
| |
| ENSURE the following are COMPLETE as required:
| |
| Tl-34.04, Loose Parts Monitoring System Gain and Alarm Setpoints (Ref. Step 5.2[9]) /V 1-SI-47-76, Trip Actuating Device Operational Test (TADOT) Turbine Trip Low Fluid Oil Pressure Channels I, II, Ill (N/A if performed within 31 days before reactor startup) (Ref. Step 5.2[14j) 1-Sl-47-77, Trip Actuating Device Operational Test (TADOT) Turbine Trip Turbine Stop Valve Closure (N/A if performed within 31 days before reactor startup) (Ref. Step 5.2[14])
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 23 of 43 Date 7cay Initials 5.3 Reactor Startup (continued) 4] ENSURE a member of Operations Management Staff who is NOT a member of the operating crew, is present in the control room during the approach to criticality.
| |
| SRO
| |
| /fj) ANNOUNCE Reactor startup over PJA (N/A if previously performed).
| |
| __iI___-.
| |
| U ONS Do NOT exceed a steady startup rate + I DPM.
| |
| If the approach to criticality is suspended or delayed, the core shall be maintained sufficiently subcritical to avoid inadvertent criticality.
| |
| 6 INITIATE Reactor Startup by performing the following:
| |
| 1 .1 INITIATE Inverse Count Rate Ratio monitoring (ICRR).
| |
| 1 .2 RECORD both SR NIS readings for ICRR base counts:
| |
| 1-Nl-92-131A £4 . cps 1-Nl-92132A 1.tS cps (0;
| |
| NIS Instruments shall be monitored in ant?tion of unplanned reactivity rate of change[c.1)
| |
| (p) START Control Bank withdrawal per SOI-85.O1.
| |
| 1 .4] ENSURE RPls and step counters are within +/- 12 steps prior to rods reaching Mode 2 rod position specified in Step 5.3[11]D.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 24 of 43 Date________ Initials_____
| |
| 5.3 Reactor Startup (continued)
| |
| IF rods reach MODE 2 rod position specified in Step 5.3[1 lID, THEN ENSURE Appendix A, Mode 3-To-Mode 2 Review And Approval, COMPLETE, AND LOG Mode 2 entry in Narrative Log.
| |
| Initials Date Time Step 5.3[16.6] and 5.3[17] may be signed off at the completion of the last full 50 steps before critical.
| |
| I 1 .6 STOP at 50 steps, AND PERFORM ICRR.
| |
| [1 IF ICRR predicts criticality will not occur in next 50 steps, THEN (p) WITHDRAW RODS an additional 50 steps, AND RETURN to step 5.3[16.4].
| |
| IF ICRR plot predicts criticality in the next 50 steps AND yithin the tolerance in step 5.3[1 1], THEN 1 .1 CHECK Estimated Critical Position above Insertion Limit, per 1 -Sl-0-1 1, within 4 hours of achieving criticality (SR 3.1.7.1).
| |
| L
| |
| .2 ENSURE TAVG greater than or equal to 555°F.
| |
| .3] (p) WITHDRAW RODS to achieve criticality, establish a positive startup rate, THEN GO TO step 5.3{21].
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 25 of 43 Date 7dy Initials____
| |
| 5.3 Reactor Startup (continued)
| |
| \4Start of Critical Step(s)
| |
| I
| |
| [1 IF criticality cannot be achieved within step 5.3[11] tolerance,
| |
| \THEN FORM the following:
| |
| [19.1] \TOP rod withdrawal.
| |
| [19.2] (cN.JSERT ALL Control Banks fully.
| |
| [19.3] LOG ode 3 entry in Narrative Log.
| |
| Initials Date Time
| |
| [19.4] RECALCULAT Shutdown Margin per 1-Sl-0-10.
| |
| Initials Date Time End of Critical Step(s)
| |
| I
| |
| [20] IF all control rods were inserted per 3[19], THEN
| |
| [20.1] ENSURE Reactor Engineering eluates AND Initiates a Service Request (SR).
| |
| [20.2] RECALCULATE ECP per 1-Sl-0-11.
| |
| Initials Date Time
| |
| [20.3] OBTAIN permission to proceed from Plant Mana[.
| |
| [20.4] (p) DILUTE/BORATE per SOl-62.02 to Estimated Critical CB.
| |
| [20.5] EQUALIZE RCS-Pzr CB (within 50 ppm) using Pzr heaters and spray.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 26 of 43 Date 70 IntiaIs-5.3 Reactor Startup (continued)
| |
| \ [20.6] WHEN sufficient mixing has occurred (30 minutes),
| |
| \ THEN
| |
| \ REQUEST Chemistry to obtain a new C 6 sample.
| |
| [20.7]\ ENSURE actual CB is within 5 ppm of ECP boron for eactor Startup.
| |
| NOTE If ARC on Control Bank D (CBD) is less than 750 pcm, the upper limit rod position is the Control Bank D ARO number.
| |
| [20.8] RECORD the pcm Upper and Lower rod position limits, Estimated {JticaI Position, and Mode 2 rod position, determined. 1-Sl-0-11:
| |
| A. Upper rod positioniiçt steps on bank\
| |
| B. Lower rod position limit steps on bank C. Estimated Critical Position Nsteps on bank CBD.
| |
| D. Mode 2, -1000 pcm rod position \eps on bank
| |
| [20.9] ENSURE 1-SI-0-2, COMPLETE for Mode 2 entry.
| |
| Initials Date Time
| |
| [20.10] INITIATE Inverse Count Rate Ratio monitoring (ICRR).
| |
| [20.11] RECORD both SR NIS readings for ICRR base counts:
| |
| 1-NI-92-131A 1-NI-92-132A
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev, 0039 Page 27 of 43 Date 1SCC4/
| |
| 5.3 Reactor Startup (continued)
| |
| \
| |
| NOTE NIS lnstrnts shall be monitored in anticipation of unplanned reactivity rate of change.[c.11
| |
| [20.12ART Control Bank withdrawal per SOl-85.Ol.
| |
| [20.13) ENS E RPIs and step counters are within +/- 12 steps prior tolqds reaching Mode 2 rod position specified in Step 5.3[2]D
| |
| [20.14) IF control rodcach Mode 2 rod position specified in Step 5.3[20.8]D IEN ENSURE Appendix\Mode 3-To-Mode 2 Review And Approval, COMPLETE, ND LOG Mode 2 entry in Narrative Log.
| |
| lnitial\ Date Time NOTE Step 5.3[20.15][ and 5.3[20.16] may be signed off after last 50t,ps before criticality.
| |
| [20.15) STOP at 50 steps, AND PERFORM ICRR.
| |
| [20.16] IF ICRR predicts criticality will NOT be achieved in net 50 steps, THEN (p) WITHDRAW RODS an additional 50 steps, AND RETURN to step 5.3[20.13)
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 28 of 43 f
| |
| 6 Date7tz Initials 1\--
| |
| 5.3 Reactor Startup (continued)
| |
| N2o.17] IF ICRR plot predicts criticality in next 50 steps AND within the tolerance in step 5.3[20.8], THEN
| |
| [b7.1j CHECK Estimated Critical Position above Insertion
| |
| \ Limit, per 1-SI-0-1 1, within 4 hours of achieving
| |
| \criticality (SR 3.1.7.1).
| |
| [20.17.2] NSURE T 0 greater than or equal to 555°F.
| |
| [20.17.3] (p)IHDRAW rods to achieve Reactor criticality, establish positive startup rate, THEN GO TO ste[21].
| |
| Start of Critical Step(s)
| |
| IF criticality cannot be achieved within s 5.3[20.8] tolerance, THEN
| |
| [20.17.4] STOP rod withdrawal.
| |
| [20.17.5] (p) INSERT ALL Control Bankfqy.
| |
| [20.17.6] LOG Mode 3 entry in Narrative Log\
| |
| Initials Dte Time
| |
| [20.17.7] ENSURE Reactor Engineering evaluates AND Initiates a SR.
| |
| [20.17.8] INITIATE new GO-2, if startup to continue.
| |
| SRO Date i e End of Critical Step(s)
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 29 of 43 Date_______ lnitiaIs-l-\-
| |
| 5.3 Reactor Startup (continued) c u/b The Reactor will trip at cps on one So ange, if the trip is not blocked.
| |
| Start of Critical Step(s)
| |
| [2 WHEN 1/2 IR monitors reach 1.66 X 1O%, THEN
| |
| .1 CHECK Permissive 65 D, P-6 INTERMED RANGE PERMISSIVE, is LIT.
| |
| (p) ADJUST Control Rods to maintain 0 SUR at less than i0 5 cps on SR monitors.
| |
| RECORD both SR NIS readings:
| |
| 1-NI-92-131A cps, 1-Nl-92-132A 9O cps 21 4] RECORD both IR NIS readings:
| |
| 1-NI-92-135AlX 16%
| |
| 1-N 1-92-1 36A L x
| |
| 7 End of Critical Step(s)
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 30 of 43 Date________ Initials 5.3 Reactor Startup (continued)
| |
| / 3) Blocking SRM Reactor Trip disables detectors outputs and removes audio count rate signal.
| |
| J Steps 5.3[21 .5) through 5.3[21 .7) should, at SM discretion, be performed at the time of criticality. However, SR Trip must be BLOCKED conservatively before the SR Trip setpoint of 1 cpa.
| |
| Start of Critical Step(s) 7 2 .5] BLOCK SR Trip by placing handswitches 1-N33A, SR TRIP TR A RESET-BLOCK P-6, and l-N33B, SR TRIP TR B RESET-BLOCK P-6, to BLOCK.
| |
| [d of Critical Step(s)
| |
| [ .6] CHECK the following:
| |
| 21 .1 Permissive 64 C, SOURCE RANGE TRIP BLOCKED, LIT.
| |
| /
| |
| 21 .2 Alarm 81 C, SOURCE RANGE HI FLUX AT SHUTDOWN ALM BLOCKED, NOT LIT.
| |
| ) SELECT 1-NR-92-145 to record the highest indicating IR channel and one PR channel.
| |
| ANNOUNCE Reactor Criticality on the P/A.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 31 of 43 Date_______ Initials____
| |
| 5.3 Reactor Startup (continued)
| |
| Start of Critical Step(s)
| |
| IF initial startup following refueling AND criticality can NOT be if achieved per PET-201 THEN
| |
| [2 .1 (p) INSERT ALL Control Banks fully.
| |
| [2 .2 LOG Mode 3 entry in Narrative Log.
| |
| Initials Date Ti e EVALUATE the discrepancy and INITIATE new GO-2, if startup to continue.
| |
| SM Date Time End of Critical Step(s)
| |
| IF TAVO is less than 561°F AND Alarm 94 A, TAVG-TREF DEVIATION, is LIT, THEN INITIATE 1 -Sl-68-34 (SR 3.4.2.1).
| |
| (p) ADJUST Control Rods and/or boron concentration to RAISE Reactor power, at a rate of less than 1 dpm, tolXlO 2 .
| |
| STABILIZE Reactor power at 1 X 10.2
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 32 of 43 Date Initials_____
| |
| 5.3 Reactor Startup (continued)
| |
| [27] RECORD CRITICAL DATA:
| |
| Power Level: Xi % f.O :(o %
| |
| I NI 92 135A 1 NI 92 136A Rod Position: RCS CB /6J 7 PPM Bank Steps Loop TAVG 551 F 55 53i F S5I °F 1 TI 68 2E 1 TI 68 25E 1 TI 68 44E 1 TI 68 67E L
| |
| , Initials Date Time
| |
| ,fhj) IF Actual Critical Rod Position is between 500 and 750 pcm 7 from ECP,THEN ENSURE Reactor Engineering evaluates AND initiates a SR.
| |
| IF Mode 2 physics testing required, THEN ENSURE that the Mode 2 and Mode 3 Surveillances are in effect during the duration of rod worth measurements (approx 8 hours) per PET-201, AND ENSURE Reactor Engineering has performed the applicable low-power physics tests per PET-201.
| |
| (tE2 If AFW is in service, Reactor power must be maintained within the capability of AFW to maintain SG levels.
| |
| EVALUATE closing AFW Pumps Recirc Valves (refer to if SOl-3.02, Section 8.9).
| |
| | |
| WBN Reactor Startup GO-2 Unitl RevOO39 Page 33 of 43 Date________ Initials 1---
| |
| 5.3 Reactor Startup (continued)
| |
| TAVO will vary as a function of reactor power until the unit is greater than 15% turbine load (CS) and the Tavg program is maintained by AUTO or manual rod control. The TAVG-TREF deviation alarm is expected as reactor power approaches 7% RTP.
| |
| (p) ADJUST Control Rods or RCS C to RAISE Reactor power, at a rate of less than 1 dpm, to between 1 and 4%.
| |
| CAUTION IF AFW is controlling levels in one or more SGs, THEN Reactor power must be maintained within AFW capability (less than 4% power).
| |
| [32] STABILIZE Reactor power between 1 and 4%:
| |
| [32.1] MAINTAIN RCS Steam Dumps in Pressure Mode, set at 84% (1092 psig.), or SG PORVs set at 84%.
| |
| [32.2] (p) FOLLOW Xenon by Rod movement or Boration to maintain control banks ABOVE the LO INSERTION LIMIT.
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 34 of 43 Date________ Initials_____
| |
| 5.3 Reactor Startup (continued)
| |
| CAUTION Tripping both MFW pumps after an AFW pump is shutdown will initiate an ESF Actuation.
| |
| NOTES
| |
| : 1) In Mode 2, the trip function of all Turbine Driven Main Feedwater (TDMFW) pumps is required when one or more TDMFW pump(s) is supplying feedwater to the Steam Generators. During the process of placing the first TDMFW pump in service, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is deenergized to prevent inadvertent AFW auto-start during rollup trip testing and overspeed trip testing. Once the operating TDMFW pump has established sufficient feed flow to maintain SG level, the anticipatory AFW auto-start channel for the non-operating TDMFW pump is placed in the trip condition, and the AFW pumps secured. Refer to Tech Spec 3.3.2 Table 3.3.2-1, Function 6.e. and B 3.3.2.6.e.
| |
| : 2) Power to trip bus may be verified using lights on TRIP-RESET handswitch.
| |
| [33] IF the IA MFW pump was selected to start first in Step5.2[5.1],
| |
| THEN
| |
| [33.1] ENSURE IA MFW pump RESET using 1-HS-46-9A, MFPT A TRIP-RESET.(RED light Lit) cv
| |
| [33.2] ENSURE 1-BKR-46-36, 250V DC FDR FOR MFPT B TRIP BUS UNIT 1 ON. [250V Bat Bd 1, Panel 3]
| |
| cv
| |
| [33.3] RESET MFPT B using 1-HS-46-36A, MFPT B TRIP-RESET.
| |
| Cv
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 35 of 43 Date_________ Initials_____
| |
| 5.3 Reactor Startup (continued)
| |
| [34] IF lB MFW pump was selected to start first in Step 5.2[5.1],
| |
| THEN
| |
| [34.1] ENSURE lB MFW pump RESET using 1-HS-46-36A, MFPT B TRIP-RESET.(RED light Lit)
| |
| CV
| |
| [34.2] ENSURE 1-BKR-46-9, 250V DC FDR FOR MFPT A TRIP BUS UNIT 1 ON. [250V Bat Bd 1, Panel 3]
| |
| CV
| |
| [34.3] RESET MFPT A using 1-HS-46-9A, MFPT A TRI P-RESET.
| |
| CV
| |
| [35] START the MFW pump selected in Step5.2[5.1] by continuing in SOI-2 & 3.01.
| |
| [36] Perform SOI-3.02. Section 8.11, Transfer SG Level from Auxiliary to Main Feedwater.
| |
| [37] ENSURE SG level and FW controls are maintaining SG level at program.
| |
| [38] PLACE the HS for NON-running MFPT to TRIP (N/A running M F PT):
| |
| MFPT Handswitch Initials IV MFPT A TRIP-RESET 1 -HS-46-9A MFPT B TRIP-RESET 1-HS-46-36A
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 36 of 43 Date________ Initials_____
| |
| 5.3 Reactor Startup (continued)
| |
| [39] SHUTDOWN the AFW Pumps (AFW pumps may be shutdown in any order).
| |
| [39.1] SHUTDOWN the A-A AFW pump in accordance with SOI-3.02, Section 8.1.4.
| |
| A. ENSURE SG level and FW controls are maintaining SG level at program.
| |
| [39.2] SHUTDOWN the B-B AFW pump in accordance with SOl-3.02, Section 8.1.5.
| |
| A. ENSURE SG level and FW controls are maintaining SG level at program.
| |
| [39.3] SHUTDOWN the Turbine Driven AFW pump in accordance with SOl-3.02, Section 8.1.6.
| |
| A. ENSURE SG level and FW controls are maintaining SG level at program.
| |
| [40] IF startup is to continue, THEN GO TO GO-3, Unit Startup From Less Than 4% Reactor Power To 30% Reactor Power.
| |
| [41] IF startup NOT to continue AND cooldown is desired, THEN GO TO GO-5, Unit Shutdown From 30% Reactor Power To Hot Standby.
| |
| End of Section
| |
| | |
| WBN Reactor Startup GO-2 Uniti Rev. 0039 Page 37 of 43 6.0 RECORDS 6.1 QA Records The following documents are QA records and handled in accordance with the Document Control and Records Management (DCRM) program:
| |
| Completed Data Package 6.2 Non-QA Records None
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 38 of 43 Appendix A (Page 1 of 5)
| |
| Mode 3-To-Mode 2 Review And Approval Date________ Initials_____
| |
| STARTUP No.
| |
| NOTE Steps 1 through 6 may be performed in any sequence.
| |
| [1] IF startup is following a Reactor Trip, THEN ENSURE TI-I 27 is COMPLETED, AND restart AUTHORIZATION GRANTED.
| |
| Time Date
| |
| [2] ENSURE I-Sl-O-2B-02, Shift and Daily Surveillance Log, COMPLETE for Mode 2 entry.
| |
| Time Date
| |
| [3] ENSURE O-SI-O-3, COMPLETE for Mode 2 entry.
| |
| Time Date
| |
| [4] ENSURE 1-SI-O-903 (Safety Injection Primary Check Valves.) COMPLETE (N/A if not required this startup.)
| |
| Time Date
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 39 of 43 Appendix A (Page 2 of 5)
| |
| Mode 3-To-Mode 2 Review And Approval Date________ Initials_____
| |
| STARTUP No.
| |
| [5] SRO shall review I perform the following:
| |
| * Hold Orders.
| |
| Time Date o TACFs.
| |
| Time Date
| |
| * LCO Tracking Log.
| |
| Time Date
| |
| * DISCUSS with Site Engineering the Functional Evaluations of SRs and PERs identified as GL 91-18 issues for outstanding Required Actions for Operability affecting mode change.
| |
| Site Engineering Time Date
| |
| * Annunciator and Computer Disablement Log.
| |
| Time Date
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 40 of 43 Appendix A (Page 3 of 5)
| |
| Mode 3-To-Mode 2 Review And Approval Date________ Initials_____
| |
| STARTUP No.
| |
| SRO shall review I perform the following:(continued)
| |
| NOTE The Mode 3 to Mode 2, 1 surveillance report contains two sections, Out of frequency section and Reactor Trip Report. Both sections are addressed in the following steps.
| |
| a OBTAIN and REVIEW the Mode 2 entry sections of the Mode 3-to-Mode 2, 1 of surveillance report from the responsible departments, and ENSURE required surveillance testing for Mode 2 entry COMPLETE.
| |
| Time Date a ATTACH the completed Mode 2 entry sections of the Mode 3-to-Mode 2, 1 surveillance report to this procedure.
| |
| * MCR Board walkdown COMPLETE.[c.5]
| |
| Time Date
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 41 of 43 Appendix A (Page 4 of 5)
| |
| Mode 3-To-Mode 2 Review And Approval Date________ Initials STARTUP No.
| |
| [6] SHIFT MANAGER (SM) HOLD POINT
| |
| * Tech Spec and non-Tech Spec work related activities are COMPLETE, or will not prohibit entry or impact continued operation in Mode 2.
| |
| Outage Mgr Time Date Mods Mgr Time Date
| |
| * There are no open DCN and EDCs that prohibit Mode 2 entry.
| |
| Maint Mgr Time Date
| |
| * Limitations in FP LCO program are not exceeded.
| |
| Fire Prot Mgr Time Date
| |
| * Primary and secondary chemistry is acceptable for Mode 2.
| |
| Chemistry Mgr Time Date
| |
| * Risk assessment completed for Mode 2 entry using LCO 3.0.4.
| |
| Corp Probabilistic Time Date Risk Asmt Group
| |
| * TI-i 2.18 complete for Mode 2 entry.
| |
| SM Time Date
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 42 of 43 Appendix A (Page 5 of 5)
| |
| Mode 3-To-Mode 2 Review And Approval Date________ Initials_____
| |
| STARTUP No.
| |
| [7] FINAL APPROVAL FOR MODE CHANGE TO MODE 2:
| |
| [7.1] SM HAS confirmed that:
| |
| * All applicable LCOs are satisfied OR
| |
| * A Risk Assessment has been completed in accordance with TI-i 33 and required Risk Management Actions are in place consistent with TS 3.O.4.b.
| |
| SM Time Date
| |
| [7.2] SM has reviewed this Appendix, and grants approval for Mode 2 entry.
| |
| SM Time Date
| |
| [7.3] PLANT MANAGER HOLD POINT Plant Manager concurs and grants approval to proceed to Mode 2.
| |
| Plant Manager Time Date
| |
| | |
| WBN Reactor Startup GO-2 Unit I Rev. 0039 Page 43 of 43 Source Notes (Page 1 ofl) implementing Requirements Statement Source Document Statement Premature criticality events during SOER 88-002, SER 89-022 C.1 reactor startup 3.IA, 3.1B, 3.2E 5.2[10.10]
| |
| 5.3[1 6.3]
| |
| 5.3[20.1 2]
| |
| Inadvertent Si During Cooldown IE Circular 78-05 0.2 3.IF Verify P-4 contacts after any condition NCO 920043242 C.3 requiring opening of Rx trip breakers 5.2[10.8]
| |
| Control Rod Mispositioning SOER 84-02, Rec 8 C.4 5.2[10.10]
| |
| Rx Startup was performed with TS WBPER960353 (See also C.5 3.3.2 not met. NC0960031005 and LER 50-390/96-017)
| |
| Note.priorto 5.2[5.1] and Appendix A}}
| |
