ML112790093
| ML112790093 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/04/2011 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-259/11-302, 50-260/11-302, 50-296/11-302 | |
| Download: ML112790093 (254) | |
Text
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 1 - Op-Test No.:
Examiners:______________________ Operators: SRO:_______________________
ATC:_______________
BOP:_________________
Initial Conditions: SLC pump 2B and EECW Pump A3 out of service. HPCI surveillance testing has just been completed and Torus cooling is to be secured. Reactor Power is 76%.
Turnover: Secure RHR Pump 2A from Torus cooling. Commence a power increase to 100%.
Event Malf. No. Event Type* Event Description No.
N-BOP 1 Secure Torus Cooling lineup lAW 2-01-74 Section 8.6 N-SRO C-BOP 2 SW3j RFIRISW pump C3 trip TS-SRO R-ATC 3 Commence a power increase with rods R-SRO fic 11 C-ATC CRD Controller Failure 0-100(L) C.SRO C-BOP 5 Batch file Steam Packing Exhauster failure C-SRO 6 Batch file I-ATC Loss of Feedwater Flow Signal inputs I-SRO M-ALL 7 PC14 Non-isolable leak on torus TS-SRO 8 IOR C HPCI minimum flow valve will not open C All SRVs except 3 fail to open for Emergency 9 IOR Depressurization
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor N
NY\
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Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#1-When Suppression Pool level cannot be maintained above 11.5 feet the US determines that Emergency Depressurization is required; RO initiates Emergency Depressurization as directed by US.
- 1. Safety Significance:
Precludes failure of Containment
- 2. Cues:
Procedural compliance Suppression Pool level trend
- 3. Measured by:
Observation US determines (indicated by announcement or observable transition to C-2) that Emergency Depressurization is required before Suppression Pool level drops below 11.5 feet.
AND Observation RO opens at least 6 SRVs during performance of Emergency Depressurization actions.
- 4. Feedback:
RPV pressure trend Suppression Pool temperature trend SRV status indication CT#2-When Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage.
- 1. Safety Significance:
Prevent failure of Primary Containment from pressurization of the Suppression Chamber
- 2. Cues:
Procedural compliance Suppression Pool Level indication
- 3. Measured by:
Observation HPCI Auxiliary Pump placed in Pull to Lock
- 4. Feedback:
HPCI does not Auto initiate No RPM indication on HPCI
Appendix D Scenario Outline Form ES-D-1 Events
- 2. EECW Pump C3 trip, BOP will align RHRSW Pump Cl for EECW and start Cl Pump to restore EECW flow to the south header, lAW ARPs and 0-01-67 EECW System section 8.3.
The SRO will evaluate Technical Specification 3.7.2 and Condition A. When the Cl RHRSW Pump is aligned for EECW, then evaluate Technical Specification 3.7.1 and Condition A.
- 3. ATC will commence to raise power with control rods
- 5. Steam Packing Exhauster will trip and the STBY Exhauster will Start but the discharge damper will fail to open. The BOP will open the Steam Packing Exhauster discharge damper and restore Steam Packing Exhauster operation lAW with ARPs.
- 6. Feedwater Flow Transmitters will fail the crew will respond lAW ARPs and 2-AOl-3-1 Loss of Reactor Feedwater. The ATC will report that Feedwater Level Control failed to transfer to single element and will transfer to single element. Reactor Level will stabilize after the initial transient.
- 7. An unisolable Torus leak will commence. Suppression Pool level will start to lower and continue to lower. The SRO will enter E0I-3 on flood alarms and eventually E0I-2 on Suppression Pool Level. The crew will place HPCI in pull to lock prior to Torus level lowering to less than 12.75 feet.
The SRO will determine that Suppression Pool level cannot be maintained above 11.5 feet and enter E0I-1 to scram the reactor and then transition to Emergency Depressurize. SRO will evaluate Technical Specification 3.6.2.2 Condition A
- 8. 2-FCV-73-30, HPCI MIN FLOW VALVE will fail to open. The crew will open the RCIC CST SUCTION VALVE and RCIC PUMP MIN FLOW VALVE to establish makeup to the Torus.
- 9. 11 SRVs fail on ED, with less than 4 MSRVs open the crew will try to rapidly depressurize the RPV with systems listed in C2-12 of 2-EOI-2-C-2, Emergency RPV Depressurization.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained
Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 1 7 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOl entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 1 EQI Contingencies used: List (0-3) 75 Validation Time (minutes) 2 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER Q SRQ Shutdown Suppression Pool Cooling RO U-92B-NO-05 219000A4.O1 3.8 3.7 EECW Pump Trip RO U-067-NO-12 400000A2.01 3.3 3.4 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 CRD Controller Failure RO U-085-AB-03 201001A3.01 3.0 3.0 Steam Packing Exhauster Trip RO U-47C-AL-2 271000A1.01 3.3 3.2 SRO S-047-AB-3 Feedwater Flow Transmitter Failure RO U-003-NO-12 259002A2.02 3.3 3.4 SRO S-003-AB-01 Torus Leak RO U-000-EM-7 295030EA2.01 4.1 4.2 RO U-000-EM-17 RO U-000-EM-83 SRO S-000-EM-07 SRO S-000-EM-15
1 Pag.6ot30. -
Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 2-01-74 Residual Heat Removal System Rev 156 2-ARP-9-20A, W35 EECW South HDR DG Section Pressure Low Rev 25 0-01-67 Emergency Equipment Cooling Water System Rev 91 Emergency Equipment Cooling Water (EECW) System TS 3 7 2 Amd 254
. and_Ultimate_Heat_Sink_(UHS)
Residual Heat Removal Service Water (RHRSW)
TS 3.7.1 Amd 254 System and Ultimate Heat Sink (UHS) 2-G0l-100-12 Power Maneuvering Rev 12 2-01-85 Control Rod Drive System Rev 128 2-ARP-9-5A, W10 CRD Accumulator Charging Water Header Pressure Hi Rev 48 2-ARP-9-7A, W12 Steam Packing Exhauster Vacuum Low Rev 27 2-0I-47C Seal Steam System Rev 24 2-ARP-9-6C, W14 RFWCS Input Failure Rev 19 Loss of Reactor Feedwater or Reactor Water Level 2-A0I-3-1 Rev 20 High/Low 2-ARP-9-3B, Wi 5 Suppression Chamber Water Level Abnormal Rev 28 2-E0I-3 Secondary Containment Control Rev 12 2-E0l-2 Primary Containment Control Rev 12 2-E0l-App-18 Suppression Pool Water Inventory Removal and Rev 8 2-E0l-1 RPV Control Flowchart Rev 12 2-A0l-100-1 Reactor Scram Rev 95 2-EOl-App-5A Injection Systems Lineup Condensate/Feedwater Rev 9 2-EOl-App-6A Injection Subsystems Lineup Condensate Rev 4 2-E0I-2-C-2 Emergency RPV Depressurization Revision 6 2-E0I-App-1 1 H Pressure Control Systems Main Rev 6 EPIP-1 Emergency Classification Procedure Revision 46 EPIP-4 Site Area Emergency Revision 32
1 Pa9e7Qf30 Console Operator Instructions A. Scenario File Summary 110801 Preference File F3 bat NRC/110801 F4 imf sw03j F5 mrf swO6 close F6 trg! Eli F7 bat NRC/110202-l F8 imf pcl4 100 360 10 110801 Batch File 10 SRV overrides Trg e3 NRC/singleelement Trg e3 = bat NRC/i 10202-4 br zdihs7330a close Oir ypobkrrhrswpa3 fail_ccoil br zlohs2385a[1j off ior zlohs6635a[i] on ior ypomtrspea (eli 0) fail_control_power ior ypovfcv6635 (eli 0) fail_power_now trg 10 NRC/spe trg 10= bat NRC/i 10801-1
- Steam packing blower trip 110801-1 Dor ior ypovfcv6635 Dor ior ziohs6635a TRG SPE Zdihs6635a[3]. Eq. 1 Scenario I DESCRIPTION/ACTION Simulator Setup manual Reset to IC 90 Simulator Setup Load Batch RestorePref NRC/i 10801 Simulator Setup manual F3 Simulator Setup manual Tag SLC pump B and EECW pump A3 Simulator Setup Verify file loaded RCP required (76% 100% with control rods and flow) and RCP for Urgent Load Reduction Provide marked up copy of 2-GOI-iOO-12
1 Page8of3Q Simulator Event Guide:
Event 1 Normal: Secure Torus Cooling lineup SRO Directs securing Torus cooling lineup lAW 2-01-74, section 8.6 BOP Secures Torus cooling lineup 8.6 Shutdown of Loop 1(11) Suppression Pool Cooling NOTE
- 1) All operations are performed at Panel 2-9-3 unless otherwise noted.
- 2) RHR flow should be monitored while in operation with multiple flow paths (e.g., LPCI and Suppression Pool Cooling together, etc.). During any evolution, total system flow as indicated on RHR SYSTEM 1(11) FLOW, 2-Fl-74-50(64), should remain between 7,000 to 10,000 gpm for 1 pump operation or between 10,000 and 20,000 gpm for 2-pump operation.
[1] VERIFY Suppression Pool Cooling in operation. REFER TO Section 8.5.
[2] REVIEW the precautions and limitations in Section 3.0.
[3] NOTIFY Radiation Protection of Suppression Pool Cooling loop removed from service. RECORD name and time of Radiation Protection representative notified in NOMS narrative log.
Driver As Radiation Protection 3 acknowledge removing Suppression Pool Cooling frorr service BOP CAUTIONS
- 1) To prevent draining an RHR Loop, at least one of the RHR System test valves must be closed before stopping RHR Pumps in the associated loop.
2> To prevent excessive vibration, RHR pumps should not be allowed to operate for more than 3 minutes at minimum flow.
- 3) When closing throttle valve RHR SYS 1(11) SUPPR POOL CLGJTEST VLV, 2-FCV-74-59 and 2-FCV-74-73 from the control room, the handswitch should be held in the close position for approximately 6 seconds after the red light extinguishes.
Failure to completely close these valves could provide a leak path to the suppression pool from the RHR discharge piping.
[4] IF both RHR Pumps in Loop 1(11) are in operation AND one pump is to be removed from service due to reduced heat load, THEN:
[4.1] THROTTLE RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, 2-FCV 59(73), to obtain a flow of between 7,000 to 10,000 gpm and Blue light illuminated as indicated on RHR SYS 1(11) FLOW, 2-Fl-74-50(64).
[4.2] STOP RHR PUMP 2A(2B) or 2C(2D) using 2-HS-74-5A(28A) or 1 6A(39A).
1 Page 9 Qf 30 Simulator Event Guide:
Event 1 Normal: Secure Torus Cooling lineup BOP [4.3] CLOSE associated RHR HX 2A(2B) or 2C(2D) RHRSW OUTLET VALVE, 2-FCV-23-34(46) or 40(52).
[4.4] IF RHRSW for the Heat Exchanger removed from service is not required to support other unit operations, THEN STOP RHRSW pump for the Heat Exchanger removed from service.
Driver When contacted as other unit, RHRSW HX is not required BOP [5] CLOSE RHR SYS 1(11) SUPPR POOL CLGITEST VLV, 2-FCV-74-59(73).
[6] WHEN RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, 2-FCV-74-59(73) is CLOSED, THEN STOP RHR PUMPS 2A(2B) or 2C(2D) using 2-HS-74-5A(28A) and/or 16A(39A).
[7] CLOSE RHR SYS 1(11) SUPPR CHBR/POOL ISOL VLV, 2-FCV-74-57(71).
[8] CLOSE RHR HX(s) 2A(2B) and 2C(2D) RHRSW OUTLET VLV(s), 2-FCV 34(46) and 40(52).
[9] IF RHRSW for RHR Heat Exchanger(s) A(B) and C(D) is not required to support other unit operations, THEN STOP RHRSW Pump(s) for the Heat Exchanger(s) removed from service.
[10] CHECK RHR System discharge header pressure is greater than TRM 3.5.4 limit as indicated on 2-PI-74-51(65), RHR SYS 1(11) DISCH PRESS.
Driver When contacted as other unit, AHRSW HX is not required Drivesi At NRC directo insert E4 (imf swO3j), EECW pump 03 trip
1 PagelUof3 Simulator Event Guide:
Event 2 Component: EECW pump C3 trip BOP Respond to alarm 20A-35.
20A-35 EECW SOUTH HDR DG SECTION PRESS LOW B. CHECK Panel 2-9-3 for status of North header pump(s) breaker lights and pump motor amps normal.
C. NOTIFY UNIT SUPERVISOR, Unit 1 and Unit 3.
D. START standby EECW Pump for affected header, if available.
H. IF pump failure is cause of alarm, THEN REFER TO Tech Spec 3.7.2.
Driver. If contacted, as Unit 3 Operator, inform that 4KV SD 80 3EB received aMot Overload or Trip alarm If contacted as Unit 1 operator, you did not secure the C3 EECW Pump 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3)
CAUTION Only one RHRSW pump in a given RHRSW pump room may be counted toward meeting Technical Specification 3.7.2 requirements for EECW pump operability.
NOTES
- 1) RHRSW Pump Cl may be aligned for service by this section when:
- It is used to meet the minimum number of Tech. Spec. operable pumps; or
. At the discretion of the Unit Supervisor, it is needed to replace another pumps operation; or
. At the discretion of the Unit Supervisor, it is needed to assist in supplying header flow/pressure demand.
- 2) If used to meet EECW requirements. RHRSW pump Cl must be aligned to EECW, the pump started, and should remain running. RHRSW Pump Cl does NOT have the same auto start signals as RHRSW Pump C3.
- 3) The RHRSW pump control switches and amp meters are located at Control Room Panel 9-3, Unit 1, 2, and 3.
- 4) When RHRSW Pump Cl is aligned for EECW, its RHRSW function required by the Safe Shutdown Program (Appendix R) is inoperable. Appendix R program equipment operability requirements of FPR-Volume 1 shall be addressed.
[1] To line up RHRSW Pump Cl for EECW System operation, PERFORM the following:
[1.11 VERIFY EECW System is in prestartup/standby readiness alignment in accordance with Section 4.0.
[1.2] REVIEW all precautions and limitations in Section 3.0.
[1.3] VERIFY RHRSW Pump Cl is in standby readiness in accordance with 0-01-23.
1 Page Itof 30 Simulator Event Guide:
Event 2 Component: EECW pump C3 trip 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3) (contd)
[1 .4] VERIFY RHRSW Pump Cl upper and lower motor bearing oil level is in the normal operating range.
[1.5] UNLOCK and CLOSE RHRSW PMP Cl & 02 CROSSTIE, 0-23-544 at RHRSW C Room.
[1.6] OPEN RHRSW PMP Cl CROSSTIE TO EECW, 0-FCV-67-49 using one of the following:
[1.7] REQUEST a caution order be issued to tag RHRSW Pump Cl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Cl pump should remain running to be operable for EECW.
[2] To start RHRSW (EECW) Pump Cl, PERFORM the following:
[2.1] START RHRSW Pump Cl using one of the following:
- RHRSW PUMP Cl, 0-HS-23-8A11 on Unit 1
- RHRSW PUMP Cl, 0-HS-23-8A12 on Unit 2
- RHRSW PUMP Cl, 0-HS-23-8A13 on Unit 3
[2.2] VERIFY RHRSW Pump Cl running current is less than 53 amps using one the following:
- RHRSW PUMP Cl AMPS, 0-El-23-8/1 on Unit 1
- RHRSW PUMP Cl AMPS, 0-El-23-8/2 on Unit 2
- RHRSW PUMP Cl AMPS, 0-El-23-8/3 on Unit 3
[2.3] VERIFY locally, RHR SERVICE WATER PUMP Cl breaker charging spring recharged by observing amber breaker spring charged light is on and closing spring target indicates charged.
[2.4] VERIFY RHRSW Pump Cl upper and lower motor bearing oil level is in the normal operating range.
Driver Lf dispatched to check C3 EECW pump breaker, report breaker tripped on o rlçad arid breaker smells burnt but no visible smoke or flames (3EB 4kv SD BD)
1 Paelaof3o Simulator Event Guide:
Event 2 Component: EECW pump C3 trip 8.3 Operation of RHRSW Pump Cl (for EECW in place of C3) (contd)
[2.5] NOTIFY Chemistry of running RHRSW (EECW) pump(s).
[2.6] VERIFY a caution order has been issued to tag RHRSW Pump Cl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Cl pump should remain running to be operable for EECW.
Driver When chemistry contacted, acknowledge report When contacted as Work Control for Caution Order, acknowledge direction and inform will begin working on a GautionOrderl When dispatched as intake AUO to check[OiI Levels and close 0-23-644 valve wait 2 minutes and insert E mrf swOB close), thn report oil levels are normal and the 0-23-544 valve is closed When contacted to check breaker charging spring recharged for the Cl EECW pump, wait 2 minutes and inform amber breaker spring charged light is on an closing spring target indicates charged.
When contacted aslntak AUO for second Oti LéveI check, eport Oil Levels are normai SRO Evaluate Technical Specification 3.7.2 before the Cl EECW Pump is aligned Condition A: One required EECW pump inoperable.
Required Action A.1: Restore the required EECW pump to OPERABLE status.
Completion Time: 7 days SRO Evaluate Technical Specification 3.7.1 after the Cl EECW Pump is aligned Condition A: One required RHRSW pump inoperable Required Action A.1: Restore required RHRSW pump to OPERABLE status.
Completion Time: 30 days
1 Page l3of3Q Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase.
Direct Power increase using control rods per 2-GOl-100-12.
[21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes. REFER TO 2-SR-3.3.5(A) and 2-01-68.
- MONITOR Core thermal limits using ICS, and/or 0-TI-248 ATC Raise Power with Control Rods per 2-01-85, section 6.6. Control Rods 30-23, 38-31, 30-39, 22-31 from 00 to 12 22-39, 38-39, 38-23, and 22-23 from 00 to 16 30-31 from 00 to 48 1 4-31, 30-47, 46-31, and 30-15 from 00 to 16.
6.6.1 Initial Conditions Prior to Withdrawing Control Rods
[2] VERIFY the following prior to control rod movement:
- CRD POWER, 2-HS-85-46 in ON.
- Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).
6.6.2 Actions Required During and Following Control Rod Withdrawal
[4] OBSERVE the following during control rod repositioning:
- Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.
- Nuclear Instrumentation responds as control rods move through the core. (This ensures control rod is following drive during Control Rod movement.)
[5] ATTEMPT to minimize automatic RBM Rod Block as follows:
- STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 2-9-5 and PERFORM Step 6.6.2[6].
[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:
[6.1] PLACE CRD POWER, 2-HS-85-46 in the OFF position to deselect the Control Rod.
[6.2] PLACE CRD POWER, 2-HS-85-46, in the ON position.
1
- Page 14 of 30 Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods ATC 6.6.3 Control Rod Notch Withdrawal
[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 2-XS-85-40.
[2] OBSERVE the following for the selected control rod:
- CRD ROD SELECT pushbutton is brightly ILLUMINATED.
- White light on the Full Core Display ILLUMINATED.
- Rod Out Permit light ILLUMINATED.
[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.
[4] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH, and RELEASE.
[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE liqht extinQuishes.
Dnver At NRCdiréctiónMánuaIIy enter CR01-f controller faiIur fió-85-1 I O-1OO(L
1 Pane 15f 30 Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal
[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 2-XS-85-40.
[2] OBSERVE the following for the selected control rod:
- CRD ROD SELECT pushbutton is brightly ILLUMINATED.
- White light on the Full Core Display ILLUMINATED.
- Rod Out Permit light ILLUMINATED.
[3] VERIFY Rod Worth Minimizer operable and LATCHED into correct ROD GROUP when the Rod Worth Minimizer is enforcing.
[4] VERIFY Control Rod is being withdrawn to a position greater than three notches.
[5] IF withdrawing the control rod to a position other than 48, THEN PERFORM the following: (Otherwise N/A)
[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.
[5.2) PLACE AND HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.
[5.3] WHEN control rod reaches two notches prior to the intended notch, THEN RELEASE CRD NOTCH OVERRIDE, 2-HS-85-47 and CRD CONTROL SWITCH, 2-HS-85-48.
[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.
Driver At NRC drection. Manuafly enter CROK controller failure fic-85-1 1 0-100(L)
1 Pane lQf 30 Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods
= ATC 6.6.4 Continuous Rod Withdrawal (Continued)
[5.5] WHEN control rod settles into the intended notch, THEN CHECK the following.
- Four rod display digital readout and the full core display digital readout and background light remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does NOT alarm.
[5.6] CHECK the control rod settles at intended position and ROD SETTLE light extinguishes.
[6] IF continuously withdrawing the control rod to position 48 and performing the control rod coupling integrity check in conjunction with withdrawal, THEN PERFORM the following: (Otherwise N/A)
[6.1] PLACE and HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRRIDE.
[6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.
[6.3] MAINTAIN the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position, with the control rod at position 48.
[6.4] CHECK control rod coupled by observing the following:
- Four rod display digital readout and the full core display digital readout and background light remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator, 2-XA-55-5A, Window 14, does not alarm.
[6.5] RELEASE both CRD NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48.
Driver At NRC direction, Manually enterGRDK controller failure fic-851 I O-1OO(L
1 Pagei7otao Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods ATC [6.6] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.
[6.7] IF control rod coupling integrity check fails, THEN REFER TO 2-AOl-85-2.
ATC 6.6.5 Return to Normal After Completion of Control Rod Withdrawal
[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:
[1.1] PLACE CRD POWER, 2-HS-85-46, in OFF.
[1.2] PLACE CRD POWER, 2-HS-85-46, in ON.
Driver At NRC direction, Manually enter CRDH cortroller failure fic-85-1 1 0-1 OO(L
1 Page 18 Qf 3Q Simulator Event Guide:
Event 4 Component: CRDH Controller Failure ATC Report Alarm 5A-10 CRD ACCUM CHG WTR HDR PRESS HIGH A. VERIFY pressure high on CRD ACCUM CHG WTR HDR 2-Pl-85-13A, B. CHECK 2-FCV-85-1 1A (B) in service.
C. IF in-service controller has failed, THEN REFER TO 2-01-85.
D. IF pressure is still greater than 1510 psig after verifying proper controller operation, THEN THROTTLE PUMP DISCH THROTTLING, 2-THV-085-0527, to maintain between 1475 and 1500 psig.
ATC Report CRD controller has failed in Automatic, takes manual control and restores CRD Parameters ATC Continues to withdraw control rods Driver At. NRC directiqr, insert f (tigger1 1) to enter Steam Packing. Exhauster Failur
1
- Pag19f3Q Simulator Event Guide:
Event 5 Component: Steam Packing Exhauster failure BOP Responds to Alarm 7A-12, Steam Packing Exhauster Vacuum Low.
7A-12, Steam Packing Exhauster Vacuum Low Automatic Action: Alternate SPE fan starts and discharge damper opens, and the running fans trips.
A. CHECKS the following:
- 1. Alternate STEAM PACKING EXHR BLOWER 2B, 2-HS-66-51A started.
- 2. 2B DISCHARGE VLV, 2-HS-66-35A opens.
BOP Determines that Alternate Blower started, but discharge damper fails to open.
Opens 2B DISCHARGE VLV, 2-HS-66-35A to restore SPE Vacuum.
NRC NOTE: SPE B Blower indication will have Red and Green lights. In order for Red only indication, the crew would have to stop the A SPE lAW 2-01-470 Driver When dispatched, wait 5 minutes and report no obvious problems at SPE or Breaker NRC When ready Lass, of Feedwater Flow Signal 1nput DriVer: Upon Lead examiner direction, insert Ei(bat NRC/i 1Q202-1) to enter Las of Flow Signal Inputs
1 Pag2QofSQ Simulator Event Guide:
Event 6 Instrument: Loss of Feedwater Flow Signal Inputs Respond to alarm 6C-14 RFWCS INPUT FAILURE.
A. VERIFY RFWCS continues to maintain Reactor Water level.
B. IDENTIFY bad/invalid signal by checking Control Room instrumentation and/or ICS. REFER TO ATTACHMENT 1, on next page, for list of RFWCS ATC instrumentation. REFER TO ICS RX FW LVL CONTROL SYS display (FWLCS).
C. REQUEST assistance from Site Engineering.
D. BYPASS the bad/invalid sianal with Unit Suoervisor aooroval.
ATC Report Feedwater Flow signal has failed LOW for FW Line A.
ATC Report FW Line B Feedwater Flow signal failing HIGH.
SRO Enter 2-AOI-3-1, Loss of Feedwater or Reactor Water Level High/Low.
4.1 Immediate Actions None 4.2 Subsequent Actions
[2] IF Feedwater Flow signal fails (Fl-3-78A, FI-3-78B), THEN PERFORM the following:
A. With SROs permission, REFER TO 2-01-3 and BYPASS failed Feedwater Flow Instrument in Unit 1&2 Computer Room; or Unit 2 Aux Instrument Room.
[2.11 IF both Feedwater Flow Instruments fail, THEN VERIFY level control transfers to SINGLE ELEMENT.
ATC Verifies Reactor Level control in single element, level control failed to transfer to single element; Operator depresses single element pushbutton to transfer.
[6] IF Reactor Water Level continues to rise, THEN TRIP REP, as necessary.
[7] IF REPs in automatic control, THEN VERIFY 2-LIC-46-5 lowers flow of operating REPs.
ATC Verifies RFPTs maintain water level.
Drive If crew inserts manual Reactor Scram on rising Reactor Water Level then obtahi NRC concurrence and enter (imf pcI4 100 360 10)to enter nori-isolable leak oi torus Driver When directed by NRC, insert f. (irrif pci 4 100 360 10) to enter non-isolable leak on torus
1
..Page21 of3O Simulator Event Guide:
Event 7 Major: Non-Isolable Leak on Torus ATC/BOP Respond to alarm multiple Pump Room Flood Level alarms and SUPPR CHAMBER WATER LEVEL ABNORMAL ATC/BOP Reports lowering suppression pool water level. 9-3B W15 A. CHECK level using multiple indications.
B. IF level is low, THEN DISPATCH personnel to check for leaks.
C. IF level is high, THEN D. REFER TO 2-01-74, Sections 8.2, 8.3, and 8.4.
E. REFER TO Tech Spec Section 3.6.2.2.
F. IF level is above (-) 1 or below (-) 6.25 inches, THEN ENTER 2-E0I-2 Flowchart.
Driver When dispatched, waft 4 minutes and report, Water lve1 is 4 inches and rising in the Southeast Quad Water is flowingjn from the Torus Area. Unable to determine sourceot the leak.
SRO Enters EOl-3 on Flood Alarms EOI-3 Secondary Containment Temp Monitor and Control Secondary CNTMT Temp Answers No to Is Any Area Temp Above Max Normal EOI-3 Secondary Containment Radiation Monitor and Control Secondary CNTMT Radiation Levels Answers_No_to_Is_Any_Area_Radiation_Level_Above_Max_Normal
1 Pagfi 22 ot3Q Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus SRO Enters EOl-3 on Flood Alarms EOl-3 Secondary Containment Level Monitor and Control Secondary CNTMT Water Level Answers Yes to Is Any Floor Drain Sump Above 66 inches Answers Yes to Is Any Area Water Level Above 2 inches Restore and Maintain Water Levels using all available sump pumps Answers No to Can All Water Levels be Restore and Maintained Below Isolate all systems that are discharging into the area except systems required to:
- Be operated by EOls OR
- Suppress a Fire Answers No to Will Emergency Depressurization Reduce Discharge Into Secondary Containment.
SRO Enters EOI-2 on Low Suppression Pool Level SRO Enter EOl-2 on Low Suppression Pool Level Monitor and Control Suppression Pool Level Between (-) 1 inch and (-) 6 inches. (Appendix 18)
Answers NO to: Can Suppression Pool Level Be Maintained Above (-) 6 inches?
Answers YES to: Can Suppression Pool Level Be Maintained Below (-) 1 inch?
CT #2 SRO Sets a Value for HPCI to place in Pull to Lock, prior to 12.75 feet.
CT #2 ATC/BOP Places HPCI in Pull to Lock, before Suppression Level lowers to 12.75 feet.
1 Page2aof 30 Simulator Event Guide:
Event 8 Major: NonIsolable leak on Torus
= SRO Directs Appendix 18 BOP Appendix 18
- 6. IF Dfrected by SRO to add water to suppression pool, THEN MAKEUP water to Suppression Pool as follows:
- b. OPEN 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE
- c. IF HPCI is NOT available for Suppression Pool makeup, THEN MAKEUP water to Suppression Pool using RCIC as follows:
21 OPEN 2-FCV-71-34. RCIC PUMP MIN FLOW VALVE.
BOP Attempts to makeup water to the Suppression Pool using H PCI; 2-FCV-73-30 will not open. Utilizes RCIC to makeup water to the Suppression Pool and dispatches personnel to investigate 2-FCV-73-30.
Driver. 2FCV-73-3O fails closed ,hen the Torus leak is inserted crew will dispatch personnel to investrgate. Acknowledge investigation and provide noftthei information:
CT #1 SRO Determines atrigger value for inserting a Reactor Scram on lowering Suppression Pool Water Level and enters EO1-1, Scrar sReactor before Suppression Pool level reaches 11.5 feet:
SRO Determines that Emergency Makeup to the Suppression Pool using Standby Coolant is required and directs BOP to line up Standby Coolant to the Suppression Pool per Appendix 18.
BOP Appendix 18
- 5. IF Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN CONTINUE in this procedure at Step 9.
- 9. IF Directed by SRO to Emergency Makeup to the Suppression Pool using Standby Coolant Supply, THEN MAKEUP water to the Suppression Pool as follows:
- a. VERIFY CLOSED the following valves:
. 2-FCV-74-61, RHR SYS I DW SPRAY INBD VALVE
. 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VALVE
. 2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE
. 2-FCV-74-52, RHR SYS I LPCI OUTBD INJ VALVE
. 2-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VALVE
. 2-FCV-23-52, RHR HX 2D RHRSW OUTLET VALVE
1 Page.24 of 30 Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus
= BOP Appendix 18 (continued)
- b. PLACE VERIFY RHR Pumps 2A and 2C are NOT running.
- c. START RHRSW Pumps Dl and D2.
NOTE: 2-BKR-074-0100, RHR SYS I U-i DISCH XTIE Breaker compartment is maintained in the OPEN position as an Appendix R requirement
- d. NOTIFY Unit 1 Operator to perform the following
- 2) OPEN i-FCV-23-57, STANDBY COOLANT VALVE FROM RHRSW (Unit 1, Panel 1-9-3).
- 3) DISPATCH personnel to place 2-BKR-074-0100, RHR SYS I U-i DIXCH XTIE in ON (480V RMOV BD 1B, Compartment 1 9A).
Driver When personnel dispatched to close 2-BKRO74O1OO, wait 6 minutes then close breaker and report, delete override for breaker control power. When requested 1 -FV-23-52 is closed. When requested to open 1 -FCV-23-57 insert remote function sWO9 open and report BOP Appendix 18 (continued)
- e. NOTIFY Unit 3 Operator to VERIFY CLOSED 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV (Unit 3, Panel 3-9-3).
Driver When equested 3FCV-23 ..2 is closed
1 Page 25 of 30 Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus BOP Appendix 18 (continued)
- f. INJECT Standby Coolant into the Suppression Pool as follows:
- 1) OPEN 2-FCV-74-100, RHR SYS I U-i DISCH XTIE.
CT #1 SRO Enters EOl-1 at pre-determined trigger value and directs Core Flow Runback and Reactor Scram based on EOl-2 step SP/L-7.
SRO Enters EOl-1 from EOI-2 step SPIL-7 Verify Reactor Scram EOI-1 RC/L Monitor and Control RPV Water Level Verify as Required:
- PCIS Isolations (Groups 1 ,2 and 3)
. RCIC Restore and maintain RPV water level +2 to +51 inches using Condensate and Feedwater in accordance with App 5A EOI-1 RC/Q Monitor and Control Reactor Power
- Crew will exit RC/Q and enter 2-AOl-100-1 based on RC/Q-2.
SRO May Anticipate Emergency Depressurization and Rapidly Depressurize using Bypass valves based on EOI-1 step RC/P-3 BOP Verifies and reports PCIS isolations and, if directed, opens all Bypass Valves to Rapidly Depressurize RPV irrespective of cooldown rate. Maintains Reactor Water Level +2 to +51 inches using Condensate and Feedwater per App 5A ATC Initiates Core Flow Runback and Manual Reactor Scram and performs Immediate Actions of 2-AOl-100-1 SRO EOl-1 RC/P Monitor and Control RPV pressure When Emergency Depressurization is required Exits RC/P and enters C-2, Emergency RPV Depressurization, based on Override step RCIP-4.
1 Page26 of 30 Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus
= ATC 2-AOI-100-1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS 5AJS3B, on Panel 2-9-5.
[2] IF scram is due to a loss of RPS, THEN PAUSE in START & HOT STBY mode for approximately 5 seconds before going to REFUEL. (Otherwise N/A)
[3] REFUEL MODE ONE ROD PERMISSIVE light check:
[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL.
[3.2] CHECK REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, illuminates.
[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, is not illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A)
[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN position.
[5] IF all control rods CAN NOT be verified fully inserted, THEN INITIATE ARI by Arming and Depressing, (Otherwise N/A)
- ARI Manual Initiate, 2-HS-68-119B
[6] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
- Power level
[71 US REPEAT back status to UO, eve contact is not necessary.
BOP Performs necessary actions of 2-EOl-App-5A to maintain RPV water level in band 2-EOl-App-5A
- 13. ADJUST RFPT speed as necessary to control injection using the methods of step 12.
- 14. WHEN RPV level is approximately equal to desired level AND automatic level control is desired, THEN PLACE 2-LIC-46-5, REACTOR WATER LEVEL CONTROL, in AUTO with individual 2-SIC-46-8(9)(1 0), RFPT 2A(2B)(2C) SPEED CONTROL in AUTO.
1 Pag27ot3D Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus SRO When RPV pressure has decreased to approximately Condensate Injection Pressure directs ATC to maintain RPV Water Level +2 to +51 inches per App 6A ATC Maintains RPV Water Level in band with 2-EOl-App-6A 2-EOI-App-6A
- 1. VERIFY CLOSED the following feedwater heater return valves:
2-FCV-3-71, HP HTR 2A1 LONG CYCLE TO CNDR 2-FCV-3-72, HP HTR 2B1 LONG CYCLE TO CNDR 2-FCV-3-73, HP HTR 2C1 LONG CYCLE TO CNDR.
- 2. VERIFY CLOSED the following REP discharge valves:
- 2-FCV-3-19, REP 2A DISCHARGE VALVE
- 2-FCV-3-12, RFP 2B DISCHARGE VALVE
- 2-ECV-3-5, REP 2C DISCHARGE VALVE.
- 3. VERIFY OPEN the following drain cooler inlet valves:
- 2-FCV-2-72, DRAIN COOLER 2A5 CNDS INLET ISOL VLV
- 2-ECV-2-84, DRAIN COOLER 2B5 CNDS INLET ISOL VLV
- 2-FCV-2-96, DRAIN COOLER 2C5 CNDS INLET ISOL VLV.
- 4. VERIFY OPEN the following heater outlet valves:
- 5. VERIFY OPEN the following heater isolation valves:
- 6. VERIFY OPEN the following REP suction valves:
- 2-ECV-2-83, REP 2A SUCTION VALVE
- 2-FCV-2-95, REP 2B SUCTION VALVE
- 2-ECV-2-108, REP 2C SUCTION VALVE.
- 7. VERIFY at least one condensate pump running.
- 8. VERIFY at least one condensate booster pump running.
- 9. ADJUST 2-LIC-3-53, REW START-UP LEVEL CONTROL, to control injection (Panel 2-9-5).
1 Page 28 of-3a Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus SRO When Emergency Depressurization is required exits RC/P and enters C-2, Emergency RPV Depressurization Determines Emergency Depressurization is required and enters C-2 Answers Yes to will the reactor remain subcritical under all conditions.
Answers No to is DW pressure above 2.4 psig Answers Yes to is Suppression Pool Level above 5.5 ft Directs All ADS Valves opened Answers No to can Six ADS Valves be opened Directs BOP to open additional MSRVs as necessary to establish 6 MSRVs open Answers No to are at least 4 MSRVs open Answers Yes to is RPV pressure 80 psi or more above Suppression Chamber Pressure Directs BOP to Rapidly Depressurize the RPV to less than 80 psi above Suppression Chamber pressure with one or more of the systems listed on C2-12
1 Pa 2 of Simulator Event Guide:
Event 8 Major: Non-Isolable leak on Torus SRO Directs BOP to Rapidly Depressurize the RPV to less than 80 psi above Suppression Chamber pressure utilizing App 11 H BOP 2-EOI-App-1 1 H
- 2. VERIFY Main Condenser Off-Gas is aligned to the stack as follows:
- b. VERIFY OPEN 2-FCV-66-28, OFFGAS SYSTEM ISOLATION VALVE (Panel 9-53).
- 3. VERIFY SJAE 2A or 2B in service and aligned to Main Condenser (Panel 9-7).
- 5. IF ANY Main Steam Line is NOT isolated, THEN CONTINUE in this procedure at Step 12.
CAUTION Offsite release rate limits may be exceeded.
- 12. OPEN Turbine Bypass valves as necessary to rapidly depressurize RPV.
SRO Classify the Event Event Classification is 2.1-S
1 Page 3Q of 30 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:
SLC pump 2B and EECW pump A3 out of service.
Operations/Maintenance for the Shift:
HPCI surveillance testing has just been completed and Torus cooling is to be secured. Reactor Power is 76%. Secure RHR Loop II from Torus cooling. Commence a power increase to 100%.
Units 1 and 3 are at 100% power Unusual Conditions/Problem Areas:
None
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 2 Op-Test No.: 1108 Examiners:______________________ Operators: SRO:_______________________
ATC:_______________
BOP:________________
Initial Conditions: 86% power, CCW pump 3A is ready to return to service.
Turnover: Return to service Condenser Circulating Water pump 3A per 3-01-27 section 8.2. Raise power to 100%
Event Maif. No. Event Type* Event Description No.
N-BOP Returning to service Condenser Circulating Water Pump 3A, 1
N-SRO lAW 3-01-27 section 8.2 R-ATC 2 Commence power increase with flow R-SRO C-BOP 3 RCO2 Inadvertent start of RCIC TS-SRO C-ATC 4 RDO1a CRD Pump 3A trip C-SRO RDO7 C-ATC Control Rod 46-19 drifts in to position 40 46-19 TS-SRO Steam leak in the RCIC room C-BOP 6 RC1O RCIC Steam line isolation valves 3-FCV-71-2 and 3 will not TS-SRO auto isolate.
MSO6A MSL A Break in Reactor BLDG with MSL A valves failing to 7 MSO6B M-ALL close TH35A 8 RPO7 I RPS Fails to de-energize, ART inserts all Rods 9 HPO15 C HPCI flow controller failure in Auto to 10%
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Critical Tasks Four CT#1-With reactor at power and with a primary system discharging into the secondary containment, manually scram the reactor before any area exceeds the maximum safe operating level.
- 1. Safety Significance:
Scram reduces the decay heat energy that the RPV may be discharging into the secondary containment
- 2. Cues:
Procedural compliance Secondary containment area temperature, level, and radiation indication Field reports
- 3. Measured by:
Observation With a primary system discharging into secondary containment, a reactor scram is initiated before a maximum safe condition is reached.
OR Observation With a primary system discharging into secondary containment, US transitions to EOI- 1 and RO initiates scram upon report that a maximum safe condition has been reached.
- 4. Feedback:
Control rod positions Reactor power decrease CT#2-With a primary system discharging into the secondary containment, when two or more areas are greater than their maximum safe operating values for the same parameter, RO initiates Emergency Depressurization as directed by US.
- 1. Safety Significance:
Places the primary system in the lowest possible energy state, rejects heat to the suppression pool in preference to outside the containment, and reduces driving head and flow of system discharging into the secondary containment.
- 2. Cues:
Procedural compliance Secondary containment area temperatures, level, and radiation indication Field reports
- 3. Measured by:
Observation US transitions to C-2 and RO opens at least 6 SRVs when two or more areas are greater than their maximum safe operating values for the same parameter.
- 4. Feedback:
RPV pressure trend SRV status indications
Appendix D Scenario Outline Form ES-D-1 CT#3-With a primary system discharging into the secondary containment, take action to manually isolate the break.
- 1. Safety Significance:
Isolating high energy sources can preclude failure of secondary containment and subsequent radiation release to the public.
- 2. Cues:
Procedural compliance Area temperature indication
- 3. Measured by:
With the reactor at pressure and a primary system discharging into the secondary containment, operator takes action to manually isolate the break.
- 4. Feedback:
Valve position indication In field reports CT#4-With a reactor scram required and the reactor not shutdown, take action to reduce power by initiating ARI to cause control rod insertion.
- 1. Safety Significance:
Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.
Correct reactivity control
- 2. Cues:
Reactor power indication Procedural compliance
- 3. Measured by:
Observation ARI pushbuttons armed and depressed to cause control rod insertion.
- 4. Feedback:
Reactor power trend Rod status indication
Appendix D Scenario Outline Form ES-D-1 Events
- 1. BOP returns the Condenser Circulating Water Pump 3A to service lAW 3-01-27 Condenser Circulating Water System, section 8.2
- 2. ATC commences power increase 100% using recirculation flow.
- 3. Inadvertent start of RCIC. BOP will attempt to trip RCIC, RCIC trip pushbutton fails BOP will close FCV-71-9 Valve and SRO will detennine RCIC System inoperable, Technical Specification 3.5.3 Condition A
- 4. CRDH pump 3A trips ATC will perform 3-A0I-85-3 actions to start the Standby CRD Pump and restore CRD parameters.
- 5. When CRD Pump 3B is started Control rod 46-19 will drift in to position 40. ATC will respond lAW 3-A0I-85-5 Control Rod Drift In. ATC will fully insert Control Rod 46-19.
SRO will determine Control Rod 46-19 is Inoperable Technical Specification 3.1.3 Condition C.
The BOP will isolate RCIC. The SRO will determine RCIC Isolation Valves inoperable Technical Specification 3.6.1.3 Condition A.
- 7. MSL break in Reactor Building with MSL A valves failing to close, with small fuel failure on scram. SRO will enter E0I-3 and transition to E0I-1 and Scram the Reactor Crew will monitor secondary containment radiation levels. Eventually the SRO will determine that ED on Radiation Levels is required.
- 8. On the Scram RPS will fail to de-energize, ATC will initiate ARI to insert control rods
- 9. RFPTs will trip on the scram, HPCI is available for level control but the HPCI flow controller will fail in Auto at 10%. Crew will take manual control to restore and maintain reactor level.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained
Appendix D Scenario Outline Form ESD-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2 9 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOl Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER RO SRO Condenser Circ Water Pump Start RO U-027-NO-5 400000A4.O1 3.1 3.0 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-1 38 2.1.23 4.3 4.4 RCIC Inadvertent Start RO U-071-NO-5 217000A2.01 3.8 3.7 RCIC Steam Leak RO U-071-.AL-20 217000A2.15 3.8 3.8 SRO S-000-EM-1 2 CRD Pump Trip RO U-085-AL-07 201 001 A2.01 3.2 3.3 SRO S-085-AB-03 Control Rod Drift RO U-085-AL-12 201 003A2.03 3.4 3.7 SRO S-085-AB-5 Secondary Containment High Radiation RO U-090-AL-4 295033EA2.01 3.8 3.9 SRO S-000-EM-15 SRO S-000-EM-10
2 Page 7 of 32 Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 3-01-27 Condenser Circulating Water System Rev 58 3-GOI-100-12 Power Maneuvering Rev 35 3-01-68 Reactor Recirculation System Rev 80 3-ARP-9-3B, W27 RCIC Gland Seal Vacuum Tank Pressure High Rev 20 TS 3.5.3 RCIC System Amd 244 3-A0l-85-3 CRD System Failure Rev 10 3-A0l-85-5 Rod Drift In Rev 10 TS 3.1.3 Control Rod Operability Amd 212 3-ARP-9-3A, W22 Reactor Building Area Radiation High Rev 43 3-ARP-9-3D, W10 RCIC Steam Line Leak Detection Temperature High Rev 28 3-EOI-3 Secondary Containment Control Rev 10 TS 3.6.1.3 Primary Containment Isolation Valves Amd 212 3-ARP-9-3D, W24 Main Steam Line Leak Detection Temperature High Rev 28 3-AOl-i 00-1 Reactor Scram Rev 53 3-EOI-i RPV Control Rev 8
. Restoring Refuel Zone and Reactor Zone Ventilation 3-EOI-Appendix-8F Rev 2 Fans Following Group 6 Isolation Bypassing Group 6 Low RPV level and High Drywell 3-E0l-A 1JD endIX -8E R ev 1 Pressure Isolation Interlocks 3-EOl-Appendix-11A Alternate Pressure Control Systems MSRVs Rev 2 3-EOI-Appendix-5D Injection System Lineup HPCI Rev 5 3-E0I-3-C-2 Emergency RPV Depressurization Rev 8 3-EOl-Appendix-6A Injection Subsystems Lineup Condensate Rev 2 3-EOI-2 Primary Containment Control Rev 8 3-EOl-Appendix-17A RHR System Operation Suppression Pool Cooling Rev 5 EPIP-i Emergency Classification Rev 46
2
- PageSof32 Simulator Instructor - IC-199
- RCIC inadvertent start imf rcO2 (e5 0) ior zdihs7l9a[1] null
- RCIC steam leak imf rclO br zdihs7l 2a[2] auto imf rcO9 (e6 0)50 120 10 icr zdihs7l 9a{1] null
- CR 46-19 drift in imf rdola (elO 0) imf rd07r4619 (e12 0)
- MSL A break inside containment imf th35a (el 5 0) 3 600 0 imf ms06a imf ms06b imf hpO3 (e15 0)10 br xa557c[8] alarm_off imf rpO7
- Fuel failure imf th23 (e20 120)4600 1 icr zdihsO3l25[1] (e20 10) trip icr zdihsO3l5l[1] (e20 10) trip icr zdihs03l 76[1] (e20 10) trip Scenario 2 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 199 Simulator Setup Load Batch bat nrcl 108-2 Simulator Setup manual Simulator Setup Verify file loaded Simulator_Setup RCP required (86% 100% with control rods and flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12
2 Paga9of 32 Simulator Event Guide:
Event 1 Normal: Returning to service Condenser Circulating Water Pump 3A SRO Directs CCW Pump 3A returned to service lAW 3-01-27, section 8.2 BOP 8.2 Returning a CCW Pump to Service
[1] COLLECT the Amertap system balls. REFER TO 3-Ol-27B.
[2] SECURE the Radwaste Discharge. REFER TO 0-Ol-77B.
[3] CHECK the CCW Pump to be started for operational readiness as follows:
[3.1] CHECK for visible oil level in the CCW pump motor upper and lower bearing reservoir level indicators.
[3.2] VERIFY motor cooling water flow for CCW Pump 3A (3B) (3C) by ensuring that pressure is greater than 20 psig, as indicated by local gauges, 3-Pl-025-001 2(001 3)(001 4).
NOTE Normal bearing cooling water differential pressure is 8 psid.
[3.3] VERIFY bearing cooling water flow for CCW Pump 3A(3B)(3C) by ensuring that pressure is greater than 5 psid and less than 11 psid, as indicated by CCW PMP 3A(3B)(3C) BRG LUBE WTR FLOW DP HI/LOW, 3-PDIS-025-0004(0006)(0008), on 3-LPNL-925-01 34C.
Driver When contacted report: there is visible oil in the upper arid lower bearing reservoi level indicators, motor cooling water flow pressure is greater than ?O psig, and bearing cooling water pressure is 8 psic
2 PagetUof32 -
Simulator Event Guide:
Event 1 Normal: Returning to service Condenser Circulating Water Pump 3A
= BOP [4] VERIFY CLOSED the CCW PUMP 3A(3B)(3C) DISCH ISOL VALVE, 3-FCV-27-13(21)(29), on Panel 3-9-20.
CAUTiONS
- 1) Capacitor bank fuses are subject to clearing when the unit boards are being supplied from the 161kV source and large pumps are started. Unit Supervisors should evaluate placing the Capacitor Banks in Manual prior to starting RHR, CS or CCW pumps.
- 2) When returning a pump to service with at least one pump already in operation, the pump being placed in service may experience perturbations in flow and motor amps. It may be necessary to throttle Condenser Water Box Discharge Valves as stated in Section 6.1 to stabilize pump.
[6] START CCW PUMP 3A(3B)(3C) using 3-HS-27-1 OA(1 8A)(26A) on Panel 3-9-20 and VERIFY the respective CCW PUMP 3A(3B)(3C)
DISCH ISOL VALVE, 3-FCV-27-1 3(21 )(29), automatically travels to the fUll open position.
BOP Verifies CCW Pump 3A Discharge valve closed and starts COW Pump 3A, verifies COW Pump 3A discharge valve automatically travels open
2 Page -LI of 32 Simulator Event Guide:
Event 2 Reactivity: Power increase with Recirc Flow SRO Notifies ODS of power increase.
Directs Power increase using Recirc Flow, per 3-GOI-1 00-1 2.
[211 WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes.
REFER TO 3-SR-3.3.5(A) and 3-01-68.
ATC Raise Power w/Recirc, lAW 3-01-68, Section 6.2
[1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;
- Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM), 3-HS 1 5A(1 5B).
AND/OR
- Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM), 3-HS 16A(16B).
[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A &
38 using the following push buttons as required:
RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32 NRO When satisfied with Reactivity Manipulation, inadvertent start of )3010 Dri,er When directed by NRC, triger 5 inadvertent start of RCId
2 Page 12 of 32 Simulator Event Guide:
Event 3 Component: Inadvertent start of RCIC BOP Responds to alarm 9-3B, Window 27, RCIC Gland Seal Vacuum Tank Pressure High A. VERIFY RCIC VACUUM PUMP, 3-HS-71 -31 A, running.
B. VERIFY RCIC VACUUM TANK CONDENSATE PUMP, 3-HS 29A, running.
C. VERIFY the following valves open:
- RCIC VACUUM PUMP DISCHARGE VLV, 3-HCV-71-32 BOP While responding to alarm determines RCIC is running and reports to SRO Verifies by multiple indications that initiation signal is not valid and reports it to SRO SRO Directs BOP to trip RCIC BOP Attempts to trip RCIC, recognizes RCIC failed to trip with the Trip Pushbutton.
Operator performs actions that should have automatically occurred when tripped Operator shuts the 71-9 and 71 -34. Operator recognizes turbine is now shutting down, however, the RCIC Mm Flow Valve will not remain shut because an inadvertent initiation signal is sealed in, BOP reports this to SRO SRO Directs BOP to close RCIC Mm Flow Valve and have operator in field open breaker ATC Reports power/level/pressure stable after RCIC secured BOP Dispatches personnel to RCIC Mm flow valve breaker at 250V RMOV BD 3B, Compt 5D to open breaker when valve is closed BOP Dispatches Instrument Mechanics to investigate inadvertent initiation signal Driver Acknowledge dispatch to breakers wait 3 minutes and report on station at 250V RMOV BD 3B, Compt 5D, when directed insext override to open breaker for 71-34 valve: br ypovtcv7l34 lab Lpower_nov Acknowledge dispatch_as_Instrument_Mechanió BOP Reports to SRO that 71-34 valve is closed and breaker is open SRO Evaluates Technical Specification 3.5.3 Condition A: RCIC system inoperable Required Action A.1: Verify by administrative means HPCI system is operable Required Action A.2: Restore RCIC system to operable status Completion Time A. 1: Immediately Completion Time A.2: 14 days NRC When ready, CRD. Pump 3A trip When directed by NRC insert trigger CRD Pump 3A trip Driver
2
- Pagal3oi32 Simulator Event Guide:
Event 4 Component: CRD Pump 3A trip ATC Reports Trip of CRD Pump 3A.
SRO Announces entry into 3-AOl-85-3, CRD System Failure.
4.1 Immediate Actions
[1] IF operating CRD PUMP has failed AND the standby CRD Pump is available, THEN PERFORM the following at Panel 3-9-5:
[1.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-11, in MAN at minimum setting.
[1.21 START associated standby CRD Pump using one of the following:
- CRD PUMP 38, using 3-HS-85-2A
[1.3] ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, to establish the following conditions:
- CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, between 40 and 65 gpm.
[1.4] BALANCE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, and PLACE in AUTO or BALANCE.
If Dispatched to CR0 Pump 3A, pump is extremely hot to toUchi CR0 Pump 3B oil levels in band, pump ready for start, conditions normal after th start, CR0 3A report breaker tripped on over current, Electrical Maint called.
NRC When ATO begins to restore CRD parameters, Control Rod 461 9 drifts in t 40 When directed by NRC, insert trigger 12, Control Rod 46-1 drifts irt When rod Dtive to position 42 on Full Core Display delete the rod drifti
2
-PageI4of 32 Simulator Event Guide:
Event 5 Component: Control Rod 46-19 drifts in to position 40 ATC Report Control Rod Drift Alarm 5A-28, reports Control Rod 46-19 drifting in.
SRO Enter 3-AOl-85-5 Rod Drift In.
ATC 4.1 Immediate Actions
[1] IF multiple rods are drifting into core, THEN MANUALLY SCRAM Reactor.
Refer to 3-AOl-i 00-1.
SRO 4.2 Subsequent Actions
[1] IF a Control Rod is moving from its intended position without operator actions, THEN INSERT the Control Rod to position 00 using CONTINUOUS IN.
[2] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.
[3] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 3-AOl-i 00-1.
ATC Reports rod 46-19 stopped drifting at position 40 ATC Inserts Control Rod 46-19 to position 00.
[4] CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).
[5] ADJUST control rod pattern as directed by Reactor Engineer and CHECK Thermal Limits on ICS (RUN OFFICIAL 3D).
Crew Dispatch AUO to check scram valves.
Driver As Reactor engineer acknowledge rod drift, if asked for rqd pattern adjustment,,
inform crew thatyou are working on it As AUO after dispatched report scram valves are normaL SRO Evaluate Tech Spec 3.1.3 Condition C One or more control rods inoperable for reasons other than Condition A or B Required Action C.1 Fully Insert inoperable control rod Completion Time 3 Hours AND Required Action C.2 Disarm the associated CRD Completion Time 4 Hours NRC When ready, Steam leak in the RCIC room RCIC Steam line isolation valves 3-FCV-71-2 and 3 will not auto isolate When directed by..NRC insert trigger 6 for BCIC room steam leak Driver
2 Page 15ofS2 Simulator Event Guide:
Event 6 Component: Steam leak in the RCIC room RCIC Steam line isolation valves 3-FCV-71 -2 and 3 will not auto isolate BOP Respond to Annunciator RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 3-9-1 1. (Alarm on Panel 3-9-1 1 will automatically reset if radiation level lowers below setpoint.)
C. NOTIFY RADCON.
D. IF the TSC is NOT manned and a VALID radiological condition exists, THEN USE public address system to evacuate area where high airborne conditions exist.
BOP Determine RCIC Area Radiation Monitor is in Alarm and report, Evacuate affected area and notify radiation protection.
BOP Respond to annunciator RCIC STEAM LINE LEAK DETECTION TEMP HIGH If temperature continues to rise it will cause isolation of the following valves at steam line space temperature of 165°F Torus Area or 165°F RCIC Pump Room.
. RCIC STEAM LINE INBD ISOLATION VLV, 3-FCV-71-2
. RCIC STEAM LINE OUTBD ISOLATION VLV, 3-FCV-71-3 A. CHECK RCIC temperature switches on LEAK DETECTION SYSTEM TEMPERATURE indicator, 3-Tl-69-29 on Panel 3-9-21.
B. IF RCIC is NOT in service AND 3-FI-71-1A(B), RCIC STEAM FLOW indicates flow, THEN ISOLATE RCIC and VERIFY temperatures lowering.
C. IF high temperature is confirmed, THEN ENTER 3-EOI-3 Flowchart.
D. CHECK CS/RCIC ROOM El 519 RX BLDG radiation indicator, 3-RI-90-26A on Panel 3-9-11 and NOTIFY RADCON if rising radiation levels are observed.
E. DISPATCH personnel to investigate.
2 PagGi6of 32- -
Simulator Event Guide:
Event 6 Component: Steam leak in the RCIC room RCIC Steam line isolation valves 3-FCV-71-2 and 3 will not auto isolate BOP Reports rising temperature in RCIC, reports RCIC failed to auto isolate and isolates RCIC Steam Line BOP Reports 3-FCV-71-2 failed to close manually, 3-FCV-71-3 is closed SRO Enter EOl-3 on Secondary Containment Area Radiation Drivei If dispatched to RCIC area report a..fter 5 minutes that cannot access area at th time.
SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. Then verify isolation of Reactor Zone or Refuel Zone and verify SGTS initiates If above 72 mr/hr direct Operator to verify isolation of ventilation system and SGTS initiated ATC/BOP Verifies Reactor Zone and Refuel Zone Ventilation Systems isolated and SGTS initiated SRO If Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels are below 72 mr/hr Then Restart Reactor Zone and Refuel Zone Ventilation per Appendix 8F If ventilation isolated and below 72 mr/hr directs Operator to perform Appendix 8F SRO Enters EOI-3 on High Secondary Containment Temperature Secondary Containment Temperature Monitor and Control Secondary Containment Temperature Operate available ventilation per Appendix 8F Is Any Area Temp Above Max Normal YES -
Isolate all systems that are discharging into the area except systems CT#3 required to:
. Be operated by EOls OR
- Suppress a Fire CT#3 BOP Isolates RCIC Steam Lines and reports Temperatures and Radiation Levels lowering SRO Evaluates Technical Specification 3.6.1.3 Condition B Condition B One or more penetration flow paths with two PCI Vs inoperable except due to MSIV leakage not within limits.
Required Action B.1 Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.
Completion Time 1 Hour
2 Page 17 of 32 Simulator Event Guide:
Event 6 Component: Steam leak in the RCIC room RCIC Steam line isolation valves 3-FCV-71-2 and 3 will not auto isolate SRO Enters EOl-3 on High Secondary Containment Temperature (continued)
Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels Is Any Area Radiation Level Max Normal NO -
Isolate all systems that are discharging into the area except systems required to:
- Be operated by EQIs OR
- Suppress a Fire Ensures no systems are still discharging to Secondary Containment, remains in EOl-3 until entry conditions are cleared.
SRO Enters EOl-3 on High Secondary Containment Temperature (continued)
Secondary Containment Level Monitor and Control Secondary Containment Water Levels Is Any Floor Drain Sump Above 66 inches NO -
AND Is Any Area Water Level Above 2 inches NO -
NRC When ready, MSL A Break in Reactor BLOGw h .MSL A vaIvas failing to close Driver When directed by NRC, insert trigger 15 for MSL A Break in Rx Building
2 Page laofaa -
Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close BOP Respond to Rx Building Area Radiation High Alarm, 3A-22 A. DETERMINE area with high radiation level on Panel 2-9-1 1. (Alarm on Panel 2-9-11 will automatically reset if radiation level lowers below setpoint.)
D. NOTIFY RAD PRO.
E. IF the TSC is NOT manned and a VALID radiological condition exists, THEN USE public address system to evacuate area where high airborne conditions exist.
G. MONITOR other parameters providing input to this annunciator frequently as these parameters will be masked from alarming while this alarm is sealed in.
J. For all radiation indicators except FUEL STORAGE POOL radiation indicator, 2-Rl-90-30. ENTER 2-EOI-3 Flowchart.
BOP Determines Suppression Pool Area ARM,90-29A, is in alarm and several other ARMs on Panel 9-1 1 are showing elevated radiation Uses Public Address System to evacuate the affected area(s)
Reports to SRO the current Radiological conditions and trends and reports EOI-3 entry conditions SRO Enters EOI-3 on Secondary Containment Radiation BOP Responds to Main Steam Line Leak Detection Temperature High alarm, 3D-24 A. CHECK the following temperature indications:
- MN STEAM TUNNEL TEMP temperature indicator, 3-TIS-1-60A on Panel 3-9-3.
BOP Determines Main Steam Tunnel Temperature on 3-TIS-1-60A is rising and reports to SRO CT #1 SRO Determines leak is in the Main Steam Tunnel from a MSL and determines a trigger value for Rx Scram and MSIV isolation before Main Steam Tunnel temperature reaches 189F Driver When ATCarms and depresses ARI, insert trigger 20 for RFPT trips and Fuel Failure
2 Pe1of32 - -
Simulator Event Guide:
Event 7 Major: MSL A Break n Reactor BLDG with MSL A valves failing to close Driver After ATC has armed and depressed ARI, insert trigger 20 for RFPT trips and Fuel Failure CT #1 SRO Directs ATC to insert manual Rx Scram prior to MSIV isolation at a Steam Tunnel Temperature of 1 89F.
Directs BOP to shut MSIVs after Scram and prior to MSIV isolation at a Steam Tunnel Temperature of 189F CT #1 ATC Inserts Manual Rx Scram and performs immediate actions of 3-AOl-i 00-1, Reactor Scram
[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5AIS3A and 3-HS 5AJS3B, on Panel 3-9-5.
[2] IF scram is due to a loss of RPS, THEN PAUSE in START & HOT STBY mode for approximately 5 seconds before going to REFUEL. (Otherwise N/A)
[3] Refuel Mode One Rod Permissive Light check
[3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-Si, in REFUEL.
[3.2] CHECK REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46, illuminates.
[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-Xl-85-46, is NOT illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A)
[4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in the SHUTDOWN position.
[5] IF all control rods CAN NOT be verified fully inserted, THEN INITIATE ARI by Arming and Depressing: (Otherwise N/A)
- ARI Manual Initiate, 3-HS-68-119B
[6] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Water Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
- Power level
[7] US REPEAT back status to UO, eye contact is not necessary.
CT #4 ATC Depresses Reactor Scram A and B pushbuttons, places the Mode Switch in Shutdown, and reports No Rod Motion.
Initiates ARI by Arming and Depressing one of the ARI Manual Initiate collars and
- 4 pushbuttons then reports I have rod motion.
Verifies all rods insert and makes Scram Report to the SRO
2 Paga2ftof32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Provides repeat back of Scram Report with All Rods Inserted and enters EOI-1 on Low_Reactor_Water_Level_after_Scram BOP After Reactor Scram and Turbine Trip, Shuts all MSIVs to isolate the leak Reports to the SRO that the A MSL MSlVs failed to isolate manually or automatically SRO Enters EOI-3 on High Secondary Containment Temperature or Radiation SRO IF Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels are below 72 mr/hr THEN Restart Reactor Zone and Refuel Zone Ventilation, per Appendix 8F. Defeat isolation interlocks if necessary, Appendix 8E.
If ventilation isolated and below 72 mr/hr, directs Operator to perform Appendix 8F.
ATC/BOP 3-EOI Appendix 8F
- 1. VERIFY PCIS Reset.
- 2. PLACE Refuel Zone Ventilation in service as follows (Panel 3-9-25):
- a. VERIFY 3-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch is in OFF.
- b. PLACE 3-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch to SLOW A (SLOW B).
- c. CHECK two SPLY/EXH A(B) green lights above 3-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.
- d. VERIFY OPEN the following dampers:
- 3-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR
- 3-FCO-64-6, REFUEL ZONE SPLY INBD ISOL DMPR
- 3-FCO-64-1O, REFUEL ZONE EXH INBD ISOL DMPR CT #3 BOP Dispatches personnel to investigate A MSIVs and manually close Outboard MSIVs Driver If requested, wait 3 minutes and report Appendix BE complete, enter bat appO8e If dispatched for A MSIVs, acknowledge dispatch
2 Page 21 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Monitor and Control Secondary Containment Temperature.
Operate available ventilation, per Appendix 8F.
Is Any Area Temp Above Max Normal? YES -
Isolate all systems that are discharging into the area except systems required to:
. Suppress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment?
CT#2 -YES Proceeds to the STOP sign Before any area temp rises to Max Safe (table 5)
Continue:
CT #1 Enters EOI-1 RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe. (Reactor Scram already conducted to prevent automatic Scram from occurring when MSIVs isolated on High Temperature)
CT#2 Stops at Stop sign When temperatures in two or more areas are above Max Safe, Then Emergency Depressurization is required.
Crew Monitors for Max Safe Temperatures SRO EOl-3 Secondary Containment (Level)
Monitor and Control Secondary Containment Water Levels.
Is Any Floor Drain Sump Above 66 inches? NO Is_Any_Area_Water_Level_Above_2_inches?_-_NO
2 Page.22Qf32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO EOI-3 Secondary Containment (Radiation)
Monitor and Control Secondary Containment Radiation Levels.
Is Any Area Radiation Level Above Max Normal? - YES Isolate all systems that are discharging into the area except systems required to:
- Suppress a Fire (MSIV5 have already been shut to prevent automatic isolation, however MSL A MSIVs did not shut)
Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES-Before any area radiation rises to Max Safe (table 4) Continue and enter EOl-1 (Ed-i has already been entered after Reactor Scram)
CT #2 Stops at Stop sign When radiation levels in two or more areas are above Max Safe, Then Emergency Depressurization is required.
Crew Monitors for Max Safe Radiation and reports (Suppression Pool Area,90-29A, and CRD West,90-20A, will be the first two Max Safe Radiation Areas in that order)
ATC Reports that RFPTs tripped after Reactor Scram and Reactor Water Level and pressure are lowering
2 Page23of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Enters EOl-1 on Low Reactor Water Level after Scram SRO Reactor Pressure Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig ?- NO IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate.
May Answer YES; during Scenario and direct Bypass Valves opened to Depressurize through the open MSIVs on the A MSL.
IF Emergency Depressurization is required, THEN exit RC/P and enter C2 Emergency Depressurization.
Answers YES; when two area radiation levels have reached MAX Safe.
IF RPV water level cannot be determined? - NO Is any MSRV Cycling? - NO IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3? - NO IF Suppression Pool level cannot be maintained in the safe area of Curve 4? - NO IF Drywell Control air becomes unavailable? - NO IF Boron injection is required? NO SRO Directs a Pressure Band, however, Reactor Pressure will be slowly lowering due to leak on the A MSL. If ED is not anticipated directs Reactor Pressure controlled using SRVs, if necessary, using 3-EOl-Appendix-1 1A ATC/BOP Controls Reactor Pressure as directed and if ED anticipated opens Bypass Valves to Rapidly Depressurize the RPV irrespective of cooldown rate.
Driver If ED anticipated, Fuel Failure may haye to be increased to force crew to ED on two Max Safe Had Levels
2 Pag& 24 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close BOP Maintains prescribed pressure band per 3-EOI-Appendix-1 1 A, if necessary
- 1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.
- 2. IF Suppression Pool level is at or below 5.5 ft, THEN CLOSE MSRVs and CONTROL RPV pressure using other options.
- a. 1 3-PCV-1-179 MN STM LINE A RELIEF VALVE.
- b. 2 3-PCV-1 -1 80 MN STM LINE D RELIEF VALVE.
- c. 3 3-PCV-1-4 MN STM LINE A RELIEF VALVE.
- d. 4 3-PCV-1-31 MN STM LINE C RELIEF VALVE.
- e. 5 3-PCV-1-23 MN STM LINE B RELIEF VALVE.
- f. 6 3-PCV-1-42 MN STM LINE D RELIEF VALVE.
- g. 7 3-PCV-1-30 MN STM LINE C RELIEF VALVE.
- h. 8 3-PCV-1-19 MN STM LINE B RELIEF VALVE.
9 3-PCV-1-5 MN STM LINE A RELIEF VALVE.
- j. 10 3-PCV-1-41 MN STM LINE D RELIEF VALVE.
- k. 11 3-PCV-1-22 MN STM LINE B RELIEF VALVE.
I. 12 3-PCV-1-18 MN STM LINE B RELIEF VALVE.
- m. 13 3-PCV-1-34 MN STM LINE C RELIEF VALVE.
SRO Reactor Level Monitor and Control Reactor Level Verify as required PCIS isolations SRO IF It has not been determined that the reactor will remain subcritical? NO IF RPV water level cannot be determined? NO -
IF PC water level cannot maintained below 105 feet? - NO Restores and Maintains RPV Water Level between +2 and +51 inches, with one of the following injection sources:
Directs a Level Band of (+) 2 to (+) 51 inches with HPCI, 3-EOI-Appendix-5D.
2 Paga 25 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close
= BOP Maintains the prescribed level band, per 3-EOl-Appendix-5D.
- 1. IF Suppression Pool level drops below 12.75 ft during HPCI operation, THEN TRIP HPCI and CONTROL injection using other options.
- 2. IF Suppression Pool level CANNOT be maintained below 5.25 in., THEN EXECUTE EOI Appendix 16E concurrently with this procedure to bypass HPCI High Suppression Pool Water Level Suction Transfer Interlock.
- 3. IF BOTH of the following exist:
- High temperature exists in the HPCI area, AND
- a. EXECUTE EOl Appendix 1 6L concurrently with this procedure.
- b. RESET auto isolation logic using 3-XS-73-58A(B) HPCI AUTO-ISOL LOGIC A(B) RESET pushbuttons.
CAUTION
- Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.
- Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.
- 4. VERIFY 3-IL-73-18B, HPCI TURBINE TRIP RX LVL HIGH amber light extinguished.
- 5. VERIFY at least one SGTS train in operation.
- 6. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.
NOTE HPCI Auxiliary Oil Pump will NOT start UNTIL 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, starts to open.
2 PagB 26 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close BOP Maintains the prescribed level band, per 3-EOI-Appendix-5D (contd).
- 7. PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in START.
- 8. PLACE 3-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, handswitch in START.
- 9. OPEN the following valves:
- 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE
- 3-FCV-73-44, HPCI PUMP INJECTION VALVE.
- 11. CHECK proper HPCI operation by observing the following:
- a. HPCI Turbine speed accelerates above 2400 rpm.
- b. 3-FCV-73-45, HPCI TESTABLE CHECK VLV, opens by observing 3-ZI-73-45A, DISC POSITION, red light illuminated.
- d. 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds 1200 gpm.
- 12. VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly.
- 13. WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in AUTO.
- 14. ADJUST 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control iniection.
BOP Reports to SRO that HPCI Flow Control Valve has failed in automatic control Takes manual control of HPCI Flow Control Valve and controls injection to maintain prescribed level band
2 Page2Z o132 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Reactor Power Monitor and control Reactor Power If the Reactor is Subcritical and no Boron has been injected then exit RCIQ and enter_3-AOl-i 00-1,_Reactor_Scram_-_YES ATC When time permits performs subsequent actions of 3-AOI-100-1 CT #2 SRO Enters 3-C-2, Emergency Depressurization when two Max Safe Rad levels are reached Will the Reactor Remain Subcritical Without Boron Under All Conditions ?- YES Is Drywell Pressure Above 2.4 psig? NO-Is Suppression Pool Level Above 5.5 feet? YES Directs All ADS Valves Open.
CT #2 ATC/BOP Opens 6 ADS Valves.
SRO Can 6 ADS Valves Be Opened? - YES SRO Directs Level Control transitioned to Condensate per 3-EOI-Appendix-6A ATC Maintains prescribed level band per 3-EOI-Appendix-6A
- 1. VERIFY CLOSED the following Feedwater heater return valves:
. 3-FCV-3-71, HP HTR 3A1 LONG CYCLE TO CNDR
. 3-FCV-3-72, HP HTR 3B1 LONG CYCLE TO CNDR
. 3-FCV-3-73, HP HTR 3C1 LONG CYCLE TO CNDR
- 2. VERIFY CLOSED the following RFP discharge valves:
. 3-FCV-3-19, RFP 3A DISCHARGE VALVE
. 3-FCV-3-12, REP 3B DISCHARGE VALVE
. 3-FCV-3-5, REP 3C DISCHARGE VALVE
- 3. VERIFY OPEN the following drain cooler inlet valves:
. 3-FCV-2-72, DRAIN COOLER 3A5 CNDS INLET ISOL VLV
. 3-FCV-2-84, DRAIN COOLER 3B5 CNDS INLET ISOL VLV
. 3-ECV-2-96, DRAIN COOLER 3C5 CNDS INLET ISOL VLV
- 4. VERIFY OPEN the following heater outlet valves:
. 3-FCV-2-124, LP HEATER 3A3 CNDS OUTL ISOL VLV
. 3-FCV-2-125, LP HEATER 3B3 CNDS OUTL ISOL VLV
.__3-FCV-2-126,_LP_HEATER 3C3 CNDS_OUTL_ISOL VLV
2
- Page 28 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close ATC Maintains prescribed level band per 3-EOI-Appendix-6A (contd)
- 5. VERIFY OPEN the following heater isolation valves:
- 6. VERIFY OPEN the following REP suction valves:
- 3-FCV-2-83, REP 3A SUCTION VALVE
- 3-FCV-2-95, REP 3B SUCTION VALVE
- 3-FCV-2-108, RFP 3C SUCTION VALVE
- 7. VERIFY at least one condensate pump running.
- 8. VERIFY at least one condensate booster pump running.
- 9. ADJUST 3-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 3-9-5).
- 10. VERIFY REW flow to RPV.
ATC Verifies REP discharge valves are closed prior to Reactor Pressure dropping below condensate system discharge pressure to prevent overfeeding the Reactor
2 Page29 of-22 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Enters EOl-2 on High Suppression Pool Temperature EOl-2 (Drywell Temperature)
SRO Monitor and Control DW Temp Below 160°F, using available DW Cooling.
Can Drywell Temp Be Maintained Below 160°F? YES -
SRO Verify H202 Analyzers placed in service, Appendix 19.
BOP Places H202 analyzers in service, lAW Appendix 19.
SRO EOl-2 Primary Containment (Pressure)
Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary. (Appendix 12)
Can Primary Containment pressure be maintained below 2.4 psig? YES-SRO EOI-2 Suppression Pool (Temperature)
Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary. (Appendix 1 7A)
Can Suppression Pool Temperature Be Maintained Below 95°F? NO -
Operate all available suppression pool cooling, using only RHR Pumps not required to assure adequate core cooling by continuous injection. (Appendix 17A)
BOP/ATC Places RHR in Suppression Pool Cooling, (lAW Appendix 17A)
2 Paga3O of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO EOl-2 Suppression Pool Level Monitor and Control Suppression Pool Level between -1 inch and -6inch, (Appendix 18).
Can Suppression Pool Level be maintained above -6 inches Yes-Can Suppression Pool Level be maintained below -1 inches Yes-BOP Places RHR in Suppression Pool Cooling lAW 3-EOl-Appendix-17A
- 1. IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD INJ VLV BYPASS SEL in BYPASS.
- 2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:
- c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- d. IF Directed by SRO, THEN PLACE 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.
- e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLACE 3 XS-74-121 (129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.
- f. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11)
- h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.
2 Page 31 of 32 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close
= BOP Places RHR in Suppression Pool Cooling lAW 3-EOl-Appendix-17A (contd)
CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage.
THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-FI-74-50(64), RHR SYS 1(11) FLOW:
- Between 7000 and 10000 gpm for one-pump operation.
- At or below 13000 gpm for two-pump operation.
- j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
I. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.
- m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional RHR and RHRSW pumps in service using Steps 2.b through 2.1.
SRO Emergency Plan Classification 3.2-S
2 e32of3Z -
SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
None Operations/Maintenance for the Shift:
Return Condenser Circulating Water Pump 3A to service, JAW 3-01-27 section 8.2. All Amertap balls have been collected JAW 3-0I-27B, Radwaste discharge is not in progress, and Cooling Towers are not in service.
Commence a power increase to 100%
Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas:
None
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.:
Examiners: Operators: SRO:
ATC:
BOP:
Initial Conditions: 100% power. HPCI is out of service.
Turnover: Transfer 4kV Unit board 3A from USST to Start Bus 1A 0-OI-57A section 8.15.1. Lower reactor power to 90% using recirc for surveillance testing Event Maif. No. Event Type* Event Description No.
N-BOP Transfer 4KV UB-3A from USST 3B to Start Bus 1A JAW 1
N-SRO 0-OJ-57A section 8.15.1 R-ATC 2 Power decrease with flow R-SRO Batch TS-SRO Core Spray Loop 1 Inoperable failed FCV-75-25 File C-BOP 4 EGO2 Stator Water Cooling Pump Trip C-SRO C-BOP 5 EGO3 Turbine Generator Voltage Regulator Failure C-SRO C-ATC LOCA Recirculation Pump B Inboard and Outboard seal 6 R-ATC TH1O/1 lb failure TS-SRO C-BOP 7 TC lOb EFIC Pressure Transducer Failure C-SRO 8 M-ALL ATWS, without MSIVs 9 RCO8 C RCIC steam supply valve fails to auto open 10 IOR C CRD Controller Fails Low (FIC-85-1 1)
(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 CRITICAL TASKS Three -
CT#1 -With a reactor scram required and the reactor not shutdown, initiate action to reduce power by injecting boron (If still critical with challenge to BuT) and inserting control rods.
- 1. Safety Significance:
Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.
- 2. Cues:
Procedural compliance Suppression Pool temperature
- 3. Measured by:
Observation If operating lAW EOI-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool.
AND RO places SLC A I B Pump control switch in ON, when directed by US.
AND Control Rod insertion commenced in accordance EOl Appendices.
- 4. Feedback:
Reactor Power trend Control Rod indications SLC tank level CT#2 RPV Level maintained above -162 inches, RCIC has been manually initiated.
- 1. Safety Significance:
Maintaining adequate core cooling
- 2. Cues:
RPV level indication
- 3. Measured by:
RCIC injecting at 600 gpm
- 4. Feedback:
RPV level trend RCIC injection valve open
Appendix D Scenario Outline Form ES-D-1 CT#3 With reactor scram required and the reactor not shutdown, to prevent an uncontrolled RPV depressurization and subsequent power excursion, inhibit ADS.
- 1. Safety Significance:
Precludes core damage due to an uncontrolled reactivity addition
- 2. Cues:
Procedural compliance
- 3. Measured by:
ADS logic inhibited prior to an automatic initiation unless all required injection systems are Terminated and Prevented.
- 4. Feedback:
RPV pressure trend RPV level trend ADS annunciator status
Appendix D Scenario Outline Form ES-D-1 Events
- 2. ATC lowers power with flow
- 3. Core Spray Loop #1 FCV-75-25 Loss of Power in Close position. SRO will determine Technical Specification 3.5.1 Condition A and D is applicable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore HPCI or Core Spray Loop 1 to Operable.
- 4. Stator Water Cooling Pump trip, BOP operator starts standby pump and restores stator water cooling prior to a turbine trip.
- 5. Turbine Generator Voltage Regulator will fail high in automatic and not transfer to manual.
BOP will respond according to ARPs and transfer the voltage regulator to manual and restore Generator MVAR loading to normal.
- 6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate B RR Pump JAW with 3-AOI-68-JA. ATC will insert control rods to exit Region 2 of the power to flow map. SRO will determine Technical Specification 3.4.1 Condition A, is applicable with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to establish single loop conditions. Can follow up with RCS Operational Leakage Technical Specification prior to RR Loop isolation, Technical Specification 3.4.4 Condition A.
- 7. EHC Pressure Transducer Failure non-operating pressure regulator takes control. This results in slowly decreasing reactor pressure. ATC inserts a scram and the BOP operator closes the MSIVs prior to reactor pressure lowering to less than 900 psig lAW 3-AOI-47-2.
- 8. ATWS exists on the scram the crew will enter EOI-1, EOI-2 and EOI-C-5. Crew will insert control rods, control reactor pressure on SRVs, initiate SLC.
- 9. RCIC steam supply valve will not auto-open on initiation signal, level will degrade until RCIC is manually started. Once started RCJC will maintain level above TAF.
Appendix D Scenario Outline Form ES-D-1 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are being inserted Reactor Level is being maintained Reactor Pressure Controlled on SRVs
Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 7 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER K/A RO SRO Transfer 4KV Unit Board RO U-57A-NO-1 262001A4.05 3.3 3.3 SRO S-57A-NO-4 Lower Power with Recirc Flow RO U-068-N0-17 SRO S-000-NO-138 2.1.23 4.3 4.4 Stator Water Cooling Pump Trip RO U-35A-AL-2 245000A4.03 2.7 2.8 SRO S-070-AB-1 Turbine Generator Voltage Regulator Failure RO U-47-AL-20 262001A2.09 3.1 3.4 SRO S-57A-AB-4 RR Pump Seal Failure RO U068-AL-9 203000A4.02 4.1 4.1 SRO S-068-AB-1 EHC Pressure Transducer Failure RO U-047-AB-2 241000A2.03 4.1 4.2 SRO S-047-AB-2 ATWS RO U-000-EM-35 295015AA2.O1 4.1 4.3 SRO S-000-EM-1 SRO S-000-EM-2 SRO S-000-EM-3
4 Pag8 of 35 Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 0-Ol-57A Switchyard and 4160V AC Electrical System Rev.141 3-GOI-100-12 Power Maneuvering Rev. 35 3-01-68 Reactor Recirculation System Rev. 80 3-ARP-9-3C Alarm Response Procedure Rev. 26 3-TSR BEN-UNIT 3 Tech Spec 3.5-1 124OO3 3-ARP-9-7A Alarm Response Procedure Rev. 22 3-ARP-9-8A Alarm Response Procedure Rev. 34 3-ARP-9-4B Alarm Response Procedure Rev. 42 Drywell Pressure and/or Temperature High, or 3 AOI-64-l R ev. 3 Excessive Leakage Into Drywell Recirc Pump Trip/Core Flow Decrease OPRMs 3-A0 I i A Rev. 6 Operable 3-AOI-47-2 Turbine EHC Control System Malfunctions Rev. 6 3-EOl-i RPV CONTROL FLOWCHART Rev. 8 INSERT CONTROL RODS USING REACTOR 3-EOI APPENDIX-i D Rev. 2 MANUAL CONTROL SYSTEM 3-EOI-2 PRIMARY CONTAINMENT CONTROL FLOWCHART Rev.7 3-EOI APPENDIX-5C INJECTION SYSTEM LINEUP RCIC Rev. 3 3-EOl APPENDIX-i F MANUAL SCRAM Rev. 2 3-EOI APPENDIX-2 DEFEATING ARI LOGIC TRIPS Rev. 4 3-EOI-C-5 LEVEL-POWER CONTROL FLOWCHART Rev. 9
4 Page 9of5 Simulator Instructor IC-204
- HPCI tagout bat nrchpcito
- Tech Spec call SRO Core Spray System #1 br ypovfcv7525 (el 0) faiLnow br xa553c[27] (el 0) crywolf
- B stator water pump trip irf eg02 (e5 0) off or ypobkrscwpa (e5 0) fail_ccoil ior zdihs3535a[2j (e5 0) stop ior zIohs3535a[1 j (e5 0) off
- Turbine Generator Voltage Regulator failure imf egO3 (elO 0)
- B Recirc pump seal failures imf thl2b (e15 0) imfthl0b (e15 0)100 imfthllb(e15 180)100600
- B EHC Pressure transducer failure bat atws70 ior zdihs0l 1 6[1 I (e20 0) select or zdihs472O4[1] (e20 0) null br zlohsoll6[1] off br zlohs472O4[1] on bmf tclob (e20 0)86 120079
- RCIC steam supply valve fails to auto open bmf rcO8 trg 25 = bat sdv trg 26 = bat atws-1 trg 27 = bat appOif trg 28 = bat appo2 trg 29= bat appo8ae
- After Scram manually insert under DI Overide
- 3-FIC-85-1 1 0-100(L)
Scenario 4 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 204 Simulator Setup Load Batch bat nrcl 108-4 Simulator Setup manual Verify file loaded Simulator Setup Manual Hang clearance on HPCI
4 Simulator Event Guide:
Event 1 Normal: Transfer 4KV UB-3A from USST 3B to Start Bus 1A Directs Transfer 4KV UB-3A from USST 3B to Start Bus 1A per 0-OI-57A section SRO 8.15.1 8.151 Transfer 4kV Unit Board 3A from USST to Start Bus BOP
[1] REVIEW all Precautions and Limitations in Section 3.0.
CAUTIONS
- 1) This board transfer can cause a power interruption causing a loss of Computer Rooms and Communication Battery Board ACU, Computer UPS ACU, and Communication rooms ACU.
- 2) Capacitor bank fuses are subject to clearing when Unit Boards are supplied from the 161 source and large pumps are started. Unit Supervisors should evaluate placing the Capacitor Banks in Manual prior to starting Condensate, CBP, RHR, CS or CCW pumps.
- 3) If 4kV Unit Board 3A is fed from the Alternate Power Supply (Start Bus), then Auto transfer must be blocked for:
- 4kV COM BD A and B. (3-45E721 QPL)
- 4) If either 4kV UNIT BD 1A. 1 B, 2A or 2B is aligned to a Start Bus, prior to aligning UNIT BD 3A to the Start Bus, check Technical Specifications 3.8.1 .a and 3.82.a to determine operability of qualified AC circuits between the offsite transmission network and the onsite Class 1 E Electrical Power Distribution System.
NOTES 1> All procedural steps are performed from Control Room Panel 3-9-8, unless specified.
- 2) This procedure section contains actions ensure electrical load restrictions are not exceeded when 4kV UNIT BD 3A is placed on Alternate Supply (Start Bus).
[2] Ensure the 4kV Start Busses are aligned Normal.
[2.1] On Panel 9-23-2, VERIFY 4kV Start Bus 1A ALT FDR BKR 1518 OPEN.
[2.2] On Panel 9-23-2, VERIFY 4kV Start Bus 1 B ALT FDR BKR 1414 OPEN.
[3] RE-ALIGN 4kV Auto Transfers to met Load Restrictions
[3.1] On Panel 1-9-8, PLACE 1-XS-57-4, 4kV UNIT BD 1A MAN/AUTO SELECT switch to MAN.
[3.2] On Panel 1-9-8, PLACE 1-XS-57-7, 4kV UNIT BD lB MAN/AUTO SELECT switch to MAN.
4 Page1tot35 Simulator Event Guide:
Event 1 Normal: Transfer 4KV UB-3A from USST 3B to Start Bus 1A
[3.3] On Panel 1-9-8, PLACE 1-XS-57-10, 4kV UNIT BD 1C MAN/AUTO SELECT switch to MAN.
[3.4] On Panel 3-9-8, PLACE 3-XS-57-4, 4kV UNIT BD 2A MAN/AUTO SELECT switch to MAN.
[3.5j On Panel 3-9-8, PLACE 3-XS-57-7, 4kV UNIT BD 2B MAN/AUTO SELECT switch to MAN.
[3.6] On Panel 3-9-8, PLACE 3-XS-57-10, 4kV UNIT BD 2C MAN/AUTO SELECT switch to MAN.
[3.7] On Panel 0-9-23-3, PLACE 0-43-203-A, 4kV COM BD A MAN/AUTO SELECT switch to MAN.
[3.8] On Panel 0-9-23-4, PLACE 0-43-203-B, 4kV COM BD B MAN/AUTO SELECT switch to MAN.
IVER DRIVER When requestedto REAL1GN 4kV UNIT 80 Auto Transfer SGhemó. Report switches for 4 KV Unit Boards 1 A 1 B, 1 C 2A, 2B, 2C, AND Common Boards A and B have been placein MANUAL
[4] TRANSFER 4kv UNIT BD 3A to the ALT FD
[4.1] PLACE 3-XS-57-4, 4kV UNIT BD 3A MAN/AUTO SELECT switch to MAN.
[4.2] PLACE 3-XS-202-1, 4kV BD/BUS/XFMR VOLTAGE SELECT switch to START BUS 1A.
[4.3] CHECK START BUS 1A Voltage on 3-EI-57-28 is between 3950 and 4400 Volts.
[4.4] PLACE and HOLD 3-HS-57-5, 4kV UNIT BD 3A ALT FDR BKR 1432 switch to CLOSE.
[4.5] PLACE 3-HS-57-3, 4kV UNIT BD 3A NORM FDR BKR 1312 switch to TRIP.
[4.6] CHECK CLOSED the 4kV UNIT BD 3A, ALT FDR BREAKER 1432.
4 Pagei2of3 Simulator Event Guide:
Event 1 Normal: Transfer 4KV UB-3A from USST 3B to Start Bus 1A (continued)
[4.7] CHECK OPEN the 4kV UNIT BD 3A, NORM FDR BREAKER 1312.
[4.8] RELEASE BKRs 1432 and 1312 control switches.
[4.9] PLACE 3-XS-202-1, 4kV BD/BUS/XFMR VOLTAGE SELECT SWITCH TO UNIT BD 3A.
[4.10] CHECK 4kV UNIT BD 3A voltage is between 3950 and 4400 Volts.
[4.11] VERIFY LOCALLY 4kV BKR 1432 closing spring target indicates charged and the amber breaker spring charged light is on.
[4.12] As directed by the Unit Supervisor, PLACE a Caution Order on the Condensate, CBP, CS, RHR or CCW Pump stating, Evaluate the need to place CAP Banks in Manual prior to starting Pump.
[4.13] RETURN the Computer Rooms, Communication Battery Board, Computer UPS, and Communication rooms ACUs to service per 0-01-31.
When requested, acknowledge that a Caution Order will need to be placed on the Condensate,. CBP, CS, RH or CW Pump statipg, Evaluate the needto plac bIVER DRIVER CAP Banks in Manual prior to starting Pump When requested, acknowledge that the Computer Rooms, Communication Battery.
Board, Computer UPS, and Communication rooms ACUs are to be returned to DRIVER DRIVER service per 0-01-31 There are no simulator actions requird to complete this step.
4 Page 13ot3 Simulator Event Guide:
Event 2 Reactivity: Lower Reactor Power with Recirc Flow SRO Notify ODS of power decrease Direct Power Reduction using Recirc Flow per 3-GOI-100-12:
[9] REDUCE reactor power by a combination of control rod insertions and core flow changes, as recommended by Reactor Engineer.
REFER TO 3-SR-3.1.3.5(A) and 3-01-68. (N/A if entering 3-GOl-100-12 to recover from Recirc Pump Trip)
ATC Lowers Power w/Recirc using 3-01-68, section 6.2
[1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following; (Otherwise N/A)
. Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM),
3-HS-96-1 5A(1 5B). (Otherwise N/A)
- Lower Recirc Pump3A using SLOW (MEDIUM) (FAST),
3-HS-96-1 7A(1 7B)(1 7C). (Otherwise N/A)
AND/OR
- Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM),
3-HS-96-1 6A(1 6B). (Otherwise N/A)
. Lower Recirc Pump 3B using SLOW (MEDIUM) (FAST),
3-HS-96-1 8A(1 8B)(1 8C). (Otherwise N/A)
[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required:
RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32 LOWER SLOW, 3-HS-96-33 LOWER MEDIUM, 3-HS-96-34 LOWER FAST, 3-HS-96-35
4
.Pag.e.t4of3 -
Simulator Event Guide:
Event 3: Core Spray Loop 1 Inoperable failed FCV-75-25 NRC NRC When satisfied with Reactivity manipulation- Move on to Core Spray Loop 1 failed FCV-75-25 DRIVER DRIVER TRIGGER I to cause.a loss of power on 3-FCV-25-75 BOP Responds to annunciator on panel 3-9-3 Window 27 Recognizes 3-FCV-25-75, CORE SPRAY SYS I INBD INJECT VALVE, does not have indication.
. When dispatched o theck th breaker for 3-FGV-25-75, GORE SPRAY SYS I DRIVER DRIVEJ INBD INJECT VALVE, report thatthe. control power fue is blown and there is some charred wirin in the bréaker SRO References Tech Spec 3.5.1 and enters Conditions A and D.
3.5 EMERGENCY CORE COOLING SYSTEMS IECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injectioftspray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.
APPLICABILITY: MODE 1.
MODES 2 and 3. except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.
ACTIONS
-- ---NOTE ----
LCO 3.0.4.b is not applicable to HPCI.
CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.l Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem(s) to OPERABLE status.
OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.
D. HPCI System inoperable. D.I Restore HPCI System to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.
AND OR Condition A entered.
D.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injectlon.spray subsystem to OPERABLE status.
4 Rage 15 of 35 Simulator Event Guide:
Event 4 Component: Stator Water Cooling Pump Trip t NRC direction Insert TRIGGER 5 to cause a trip of the B stator water cooling DRIVER DRIVER ump and insert TRIGGER 10 to cause the Turbine Generator Voltage Regulator to all high in automatic (This failure will take approximately 9 minutes before the first annunciator is received)
BOP Responds to annunciator 3-9-7A window 22, GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action of this ARP 3-XA-55-9-8A Window 1 alarms THEN,
- 1. VERIFY all available Stator Cooling Water Pumps running.
- 2. Attempt to RESET alarm
- 3. IF alarm fails to reset, AND reactor power is above turbine bypass valve capability THEN SCRAM the Reactor B. VERIFY a stator cooling water pump is running and CHECK stator temperature recorder, 3-TR-57-59, Panel 3-9-8.
Operator starts the standby Stator Water Cooling Pump and restores Stator Water BOP Cooling.
DRIVER When/If dispatched to investigate the B Stator Water Cooling. Pump, waitS minutes.
DRIVER and report uhable to determine cause of trip
4 Page 16 of 35 Simulator Event Guide:
Event 5 Component: Turbine Generator Voltage Regulator Failure iv TRIGGER 10 that causes the Turbine Generator Voltage Regulator to fail high iii:
is already in progressJ This failure will take approximately 9 minutes before the first annun lator is NRC received Reports the following alarms:
GENERATOR EXCTR PWR RECTIFIER TEMP HIGH BOP GEN VOLTS PER CYCLE HIGH GEN HYDROGEN SYSTEM ABNORMAL GEN VOLTS PER CYCLE HIGH, 3-9-8A window 9 A. VERIFY VOLTAGE REG TRANSFER switch in MANUAL.
B. At Panel 3-9-8, ADJUST EXCITER FIELD VOLTAGE 70P MANUAL (3-HS-57-25) to maintain the following:
- 1. GENERATOR VOLTS, 3-EI-57-39, between 20,900V and 23,100V.
- 2. GENERATOR MVARS, 3-El-57-51, within the generator capability curve.
REFER TO 3-01-47, Illustration 6.
C. IF Turbine/Generator trips and power is less than -30%, THEN VERIFY Bypass Valves Controlling Reactor Pressure. REFER TO 3-AOI-47-1.
D. IF Reactor scrams, THEN REFER TO 3-AOl-i 00-1.
BOP Takes Voltage Regulator to Manual Crew Make notifications, Must notify Load Dispatch when voltage regulator not in Auto Qri. Driver Acknowledge notifications Driver Driver At NRC direction initiate trigger 15 for Reactor Reciro Pump B Seal Failure
4 Page 17 of 35 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure
[IVER DRIVE At NRC direction, insert TRIGGER 15 to cause the BRécircpürnéeals to fall]
ATC Reports failure of the #1 Reactor Recirc Pump B Seal RECIRC PUMP B NO. 1 SEAL LEAKAGE ABN, 3-9-4B Window 25:
A. DETERMINE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 3-9-4 or ICS.
. Plugging of No. 1 RO No. 2 seal cavity pressure indicator drops toward zero.
. Plugging of No. 2 RO No. 2 seal pressure approaches no. 1 seal pressure.
. Failure of No. 1 seal No. 2 seal pressure is greater than 50%
of the pressure of No. 1.
- Failure of No. 2 seal no. 2 seal pressure is less than 50% of the No. 1 seal.
NOTE
- 1) Possible indications of dual seal failure include:
. Window 18 on this panel alarming in canjuncton with this window.
. Rising drywell pressure andlor temperature.
a Increased leakage into the drywell sump.
a Increased vibration of the recirc pump.
ATC Identifies that the #2 seal is also failed/failing.
D. IF dual seal failure is indicated, THEN
- 1. SHUTDOWN Recirc Pump 3B by DEPRESSING RECIRC DRIVE 3B SHUTDOWN, 3-HS-96-20.
- 2. VERIFY TRIPPED, RECIRC DRIVE 3B NORMAL FEEDER, 3-HS-57-1 4.
4 Page 8of 35 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure
- 3. VERIFY TRIPPED, RECIRC DRIVE 3B ALTERNATE FEEDER, 3-HS-57-1 2.
- 4. CLOSE Recirculation Pump 3B suction valve.
- 5. CLOSE Recirculation Pump 3B discharge valve.
- 6. REFER TO 3-AOl-68-1A or 3-AOI-68-1B AND 3-01-68.
- 7. DISPATCH personnel to SECURE Recirculation Pump 3B seal Water Enters:
3-AOI-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable, SRO 3-AOl-64-1, Drywell Pressure and/or Temperature High, or Excessive Leakage Into Drywell.
3-AOI-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable
[2] IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve.
[3] IF Region I or II of the Power to Flow Map is entered, THEN (Otherwise N/A)
IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. REFER TO O-TI-464, Reactivity Control Plan Development and Implementation.
[4] RAISE core flow to greater than 45%. REFER TO 3-01-68.
[5] INSERT control rods to exit regions if not already exited. Refer to O-Tl-464, Reactivity Control Plan Development and Implementation.
NOTE The remaining subsequent action steps apply to a single Reactor Recirc Pump trip.
[6] MAINTAIN operating Recirc pump flow less than 46,600 gpm.
REFER to 3-01-68.
[7] WHEN plant conditions allow, THEN, (Otherwise N/A)
MAINTAIN operating jet pump loop flow greater than 41 x 106 Ibm/hr (3-FI-68-46 or 3-Fl-68-48).
4 Page 19 of 35 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure ATC Inserts Rods per Emergency shove sheet to get below 95% loadline When less than 95% load line raises core flow SRO AOI-64-1 Directs BOP to Vent the Drywell 3-AOI-64-1 Drywell Pressure and/or Temperature High, or Excessive Leakage Into Drywell
[3] VENT Drywell as follows:
[3.1] CLOSE SUPPR CHBR INBD ISOLATION VLV 3-FCV-64-34 (Panel 3-9-3).
[3.2] VERIFY OPEN, DRYWELL INBD ISOLATION VLV, BOP 3-FCV-64-31 (Panel 3-9-3).
[3.3] VERIFY 3-FIC-84-20 is in AUTO and SET at 100 scfm (Panel 3-9-55).
[3.4] VERIFY Running, required Standby Gas Treatment Fan(s) SGTS Train(s) A, B, C (Panel 3-9-25).
[3.5] IF required, THEN REQUEST Unit 1 Operator to START Standby Gas Treatment Fan(s) SGTS Train(s) A, B. (Otherwise N/A)
DRIVER DRIVER When requested to start a standby gas fan remote function peOla or b or
4 Pag2of-3&
Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure SRO Evaluates Tech Spec 3.4.1 and enters Condition A 3.4.1 Recirculation Loops Operating LCO 3.4.lTwo recirculation loops with matched flows shall be in operation OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:
- a. LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;
- b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR), single loop operation limits specified in the COLR;
- c. LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power
- High), Allowable Value of Table 3.3.1.1.1 is reset for single loop operation; APPLICABILITY: MODES 1 and 2.
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO not met. requirements of the LCO.
B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.
OR No recirculation loops in operation.
4 Rage 21 of 35 -
Simulator Event Guide:
Event 7 Component: EHC Pressure Transducer Failure At NRC direction, insert TRIGGER 20 to âause the 6 EHC Piessure tranéducero DRIVER DRIVER fail, Verify tclOb initial setpoint prior to inserting trigger ATC Recognizes lowering Reactor Pressure and generator megawatts SRO Directs entry into 3-AOI-47-2.
3-AOl-47-2 Turbine EHC Control System Malfunctions
[1j IF Reactor Pressure lowers to or below 900 psig, THEN MANUALLY SCRAM the Reactor and CLOSE the MSIVs.
SRO Directs manual scram, closing of the MSIVs, and entry into 3-AOl-i 00-1.
ATC Manually scrams the reactor.
After Scram manually insert under Di Override 3-FIC-85-11 1pQ(L)anØinser DRIVER bApjR: TRIGGER 25 to enter bat S BOP Closes the MSIVs.
SRO Enter 3-EOI-1, RPV Control.
SRO EOl-1 (Reactor Pressure)
Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? NO -
IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate ?- NO IF Emergency Depressurization is required THEN exit RC/P and enter C2 Emergency Depressurization? NO -
IF RPV water level cannot be determined? NO -
Is any MSRV Cycling? YES -
IF Steam cooling is required? NO-IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3? - NO
4 Page 22 of 35.
Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs SRO 3-EOI-1 (Reactor Pressure)
IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO IF Drywell Control air becomes unavailable? NO.
THEN crosstie CAD to Drywell Control Air, Appendix 8G.
IF Boron injection is required? NO SRO Direct a Pressure Band of 800 to 1000 psig, Appendix 1 1 A.
ATC/BOP Maintain directed pressure band, lAW Appendix hA.
EOI-1 RPV Pressure Augment RPV Pressure control as necessary with one or more of the following depressurization systems: HPCI Appendix 1 1 C, RCIC Appendix 11 B, RFPTs on minimum flow Appendix 11 F, Main Steam System Drains SRO Appendix 1 1 D, Steam Seals Appendix 1 1 G, SJAEs Appendix 1 1 G, Off Gas Preheater Appendix 1 1 G, RWCU Appendix 11 E.
ATC/BOP Pressure Control lAW Appendixl 1A, RPV Pressure Control SRVs
- 1. IF Drywell Control Air is NOT available, THEN:
EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.
- 2. IF Supp.ression Pool level is at or below 5.5 ft, THEN:
CLOSE MSRVs and CONTROL RPV pressure using other options.
- a. 3-PCV-1-179 MN STM LINE A RELIEF VALVE
- b. 3-PCV-1-180 MN STM LINE D RELIEF VALVE.
- c. 3-PCV-1-4 MN STM LINE A RELIEF VALVE
- d. 3-PCV-1-31 MN STM LINE C RELIEF VALVE
- e. 3-PCV-1-23 MN STM LINE B RELIEF VALVE
- f. 3-PCV-1-42 MN STM LINE D RELIEF VALVE
- g. 3-PCV-1-30 MN STM LINE C RELIEF VALVE
4 Page 2 of 35 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs ATC/BOP Pressure Control lAW Appendixl 1 A, RPV Pressure Control SRVs (continued)
- h. 3-PCV-1-19 MN STM LINE B RELIEF VALVE.
- i. 3-PCV-1-5 MN STM LINE A RELIEF VALVE.
- j. 3-PCV-1-41 MN STM LINE D RELIEF VALVE
Monitor and Control Reactor Level.
Verify as required PCIS isolations group (1 ,2 and 3), ECCS and RCIC, Directs group 2 and 3 verified.
ATC/BOP Verifies Group 2 and 3 isolation.
SRO IF it has not been determined that the reactor will remain subcritical, THEN Exit RC/L; ENTER C5 Level I Power Control.
If Emergency Depressurization is required? NO -
RPV Water level cannot be determined? NO The reactor will remain subcritical without Boron under all conditions? NO PC water level cannot be maintained below 105 feet OR Suppression Chamber pressure cannot be maintained below 55 psig? NO -
CT#3 SRO Directs ADS Inhibited.
CT#3 ATC/BOP Inhibits ADS.
SRO Is any Main Steam Line Open?- NO
4 Pae24of35 -
Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs SRO C5 Level I Power Control IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND a MSRV is open or cycling OR drywell pressure is above 2.4 psig AND RPV water level is above -162 inches? NO Is Reactor Power above 5% ?- YES Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC (Appendix 4).
WHEN RPV Level drops below -50 inches; THEN Continue:
SRO Direct Terminate and Prevent lAW Appendix 4.
IF Suppression Pool Temperature is above 110°F AND Reactor Power is above 5% AND a MSRV is open or cycling OR drywell pressure is above 2.4 psig AND RPV water level is above -162 inches IF YES?
Stop and Prevent all injection into the RPV except from RCIC, CRD, and SLC; irrespective of any consequent reactor power or reactor water level oscillations.
WHEN RPV Level drops below -50 inches and any of the following exist:
- Power drops below 5% OR
. All MSRVs remain closed and DW pressure remains below 2.4 psig OR
- Water level reaches -162 inches THEN Continue:
ATCIBOP Terminate and Prevent lAW Appendix 4 BOP/ATC Appendix 4
- 1. PREVENT injection from HPCI by performing the following:
- a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURBINE TRIP push-button.
- b. WHEN HPCI Turbine is at zero speed, THEN PLACE 3-HS 47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 3-HS-73-18A, HPCI TURBINE TRIP push button.
- 3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.
4 Pag25 of 3&
Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs
- 4. PREVENT injection from LPCI SYSTEM I by performing the following:
NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.
- a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.
- b. BEFORE RPV pressure drops below 450 psig,
AND
- 5. PREVENT injection from LPCI SYSTEM II by performing the following:
NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.
- a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.
- b. BEFORE RPV pressure drops below 450 psig,
AND
- 6. PREVENT injection from CONDENSATE and FEEDWATER by performing the following:
- a. IF Immediate injection termination from a reactor feedwater pump is required, THEN PERFORM step 6.d for the desired pump.
4 Page 2Sof 35.
Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs Appendix 4 (continued)
- c. CLOSE the following valves BEFORE RPV pressure drops below 500 psig:
- 3-FCV-3-19, REP 2A DISCHARGE VALVE
- 3-ECV-3-12, REP 2B DISCHARGE VALVE
- 3-FCV-3-5, RFP 2C DISCHARGE VALVE
- 3-LCV-3-53, RFW START-UP LEVEL CONTROL
- d. TRIP REPTs as necessary to prevent injection by DEPRESSING the following push-buttons:
- 3-HS-3-125A, RFPT 3A TRIP
- 3-HS-3-1 51 A, RFPT 3B TRIP
- 3-HS-3-176A, RFPT 3C TRIP.
WHEN RPV Level drops below -50 inches THEN Continue:
OR SRO WHEN RPV Level has dropped below -50 inches AND Power is below 5% OR CT#2 Reactor Level reaches -162 inches, THEN Continue:
Directs a Level Band with RCIC.
4
- Pae2Zof35 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs (continued)
SRO EOI-1 (Power Control)
Monitor and Control Reactor Power.
Will the reactor will remain sub subcritical without boron under all conditions? NO If the reactor subcritical and No boron has been injected?- NO Verify Reactor Mode Switch in Shutdown.
Initiate ARt.
SRO Verify Recirc Runback ( pump speed 480 rpm).
ATC Verifies Recirc Runback.
SRO Is Power above 5%? YES -
Directs tripping Recirc Pumps.
ATC Trips Recirc Pumps.
CT#1 SRO Before Suppression Pool temperature rises to 110°F, continue:
Insert Control Rods Using one or more of the following methods:
. Appendix iF
- Appendix 1 D DRIVER WHEN directed to perform Appendix 1 F and Appendix 2, wait 4 mihutes and inert TRIGGER 27 and TRIGGER 28 THEN report appendix 2 complete and field action for appendix 1 F cornpIet WHEN the Scram has been reset THEN insert TRIGGER 26 to enter bat ATWS1 CT#1 ATC Inserts Control Rods, lAW Appendix 1 D and 1 F.
4
.Page2aof3 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs ATC Insert Control Rods, lAW Appendix 1 F.
- 3. VERIFY OPEN Scram Discharge Volume vent and drain valves.
- 4. DRAIN SDV UNTIL the following annunciators clear:
- WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 1)
- EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 29).
- 5. DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SHUTOFF.
- 6. WHEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.
- 7. CONTINUE to perform Steps 1 through 6, UNTIL ANY of the following exists:
- ALL control rods are fully inserted, OR
- NO inward movement of control rods is observed, OR
- SRO directs otherwise.
4 Page 2ot35 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs CT#1 BOP/ATC Initiate SLC lAW Appendix 3A
- 1. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 2A12B, control switch in START-A or START-B position.
- 2. CHECK SLC System for injection by observing the following:
- Selected pump starts, as indicated by red light illuminated above pump control switch.
- SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 20).
- System flow, as indicated by 3-IL-63-1 1, SLC FLOW, red light illuminated on Panel 3-9-5.
- SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 14).
- 3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.
- 4. VERIFY RWCU isolation by observing the following:
- RWCU Pumps 2A and 2B tripped.
- 3-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.
- 5. VERIFY ADS inhibited.
- 6. MONITOR reactor power for downward trend.
- 7. MONITOR 3-Ll-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is dropping approximately 1% per minute.
4 Page 3OQf3E Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs SRO ENTER 3-EOl-2, Primary Containment Control EOI-2 (Drywell Temperature)
SRO Monitor and Control DW Temp Below 160°F using available DW Cooling.
Can Drywell Temp Be Maintained Below 160°F? YES -
SRO EOl-2 (Primary Containment Hydrogen)
If PCIS Group 6 isolation exists? YES THEN DIRECTS:
- 1. Place analyzer isolation bypass keylock switches to bypass.
- 2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.
BOP 1. Place analyzer isolation bypass keylock switches to bypass.
- 2. Select Drywell or suppression chamber and momentarily pull out select switch handle to start sample pumps.
SRO EOl-2 (Suppression Pool Temperature)
Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary (Appendix 1 7A)
Can Suppression Pool Temperature Be Maintained Below 95°F? NO Operate all available Suppression pool cooling, using only RHR Pumps not required to assure adequate core cooling by continuous injection, Appendix 17A.
ATC/BOP Place an RHR System in Pool Cooling, when directed lAW Appendix 1 7A.
SRO Before Suppression Pool Temperature rises to 110°F Continue in EOl-1 RPV Control Can Suppression Pool temperature and level be maintained within a safe area of curve 3? YES SRO EOI-2 (Suppression Pool Level)
Monitor and Control Suppression Pool Level between -1 inch and -6 inches, (Appendix 18).
Can Suppression Pool Level be maintained above -6 inches? YES Can Suppression Pool Level be maintained below -1 inch? YES
4
.Page31of35 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs SRO EOI-2 (Primary Containment Pressure)
Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary, (Appendix 12)
SRO Can Primary Containment pressure be maintained below 2.4 psig? YES ATC Place Suppression Pool Cooling in service, lAW Appendix 1 7A.
- 1. IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock AS NECESSARY:
- 2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:
- c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.
- e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLACE 3-XS-74-121 (129), RHR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.
4 Page 32 of 3S Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs
- f. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.
- h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.
I. THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-FI-74-50(64), RHR SYS 1(11) FLOW:
- Between 7000 and 10000 gpm for one-pump operation.
- At or below 13000 gpm for two-pump operation.
- j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
4 FaeaQf 35 Simulator Event Guide:
Event 9 Component: RCIC steam supply valve fails to auto open ATC/BOP Recognize that 3-FCV-71-8, RCIC TURBINE STEAM SUPPLY VLV fails to open CT#2 on a RCIC automatic initiation signal.
Manually starts RCIC.
ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C..
- 3. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRIP/THROT VALVE RESET.
- 4. VERIFY 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 600 gpm.
- 5. OPEN the following valves:
. 3-FCV-71-39, RCIC PUMP INJECTION VALVE
. 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE
. 3-FCV-71-25, RCIC LUBE OIL COOLING WTR VLV.
- 6. PLACE 3-HS-71-31A, RCIC VACUUM PUMP, handswitch in START.
- 8. CHECK proper RCIC operation by observing the following:
- a. RCIC Turbine speed accelerates above 2100 rpm.
- c. 3-FCV-71-40, RCIC Testable Check Vlv, opens by observing 3-Zl-71-40A, DISC POSITION, red light illuminated.
- d. 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE, closes as flow rises above 120 gpm.
- 9. IF BOTH of the following exist? NO
- 10. ADJUST 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller as necessary to control injection.
4 Page 34 of 35 Simulator Event Guide:
Event 10 Component: CRD Controller Fails Low (FIC-85-1 1)
ATC Recognizes CRD flow controller 3-FIC-85-1 1 has failed to control in automatic.
Takes manual control of 3-FIC-85-1 1 and restores CRD flow.
CT#1 ATC Insert Control Rods lAW Appendix 1 D
- 1. VERIFY at least one CRD pump in service.
- 2. IF Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SHUTOFF (RB NE, El 565).
- 3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.
- 4. BYPASS Rod Worth Minimizer.
- 5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:
- a. SELECT control rod.
- b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.
- c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.
- 6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SHUTOFF (RB NE, El 565 ft).
DRIVER WHEN dispatched to close Charging Water Shutoff, wait 2nnutes and reporta SHV-085-0586 closed, (rnrf rdO6 cIose WHEN asked to open Charging Water Shutoff wait 2 minutes and eport3SHV 085-0586 open. (mn rdO6 open)
REP classification is 1 .2-S Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are being inserted Reactor Level is being maintained Reactor Pressure Controlled on SRVs
4 Page 35Qt 3 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:
None Operations/Maintenance for the Shift:
100% power. HPCI is out of service.
Transfer 4kV Unit board 3A from USST to Start Bus 1A 0-OI-57A section 8.15.1. Lower reactor power to 90% using recirc for surveillance testing Unit 1 and 2 at 100% Power Unusual ConditionsIProblem Areas:
None
Appendix D Scenario Outline Form ES-D-1 icility: Browns Ferry NPP Scenario No.: NRC 6- Op-Test No.: jj Examiners: Operators: SRO:_
ATC:
BOP:
Initial Conditions: 80% power. RCIC is out of service and Breaker 1624 Alternate Feed to SD BD C.
Turnover: Place RFPT A in service from 600 RPM in accordance with 2-0I-3section 5.7 and then raise power to 100%
Event Malt. No. Event Type* Event Description No.
N-BOP Place RFPT A in service from 600 RPM in accordance with 2-1 N-SRO 01-3 section 5.7 R-ATC 2 Raise Power with Control Rods R-SRO C-ATC 3 RD06r3016 CR 30-15 Difficult to withdraw at position 00 C-SRO C-BOP 4 OGO4a Loss of SJAE A C-SRO 5 C Shutdown Board Supply Breaker trips DG C fails to auto start
. C-ATC RBCCW pump B trips, RBCCW sectionalizing valve fails to 6 Batch file TS-SRO auto close OGO5a 7 M-ALL Explosion in 0ff-gas system, Loss of condenser vacuum OGO1 8 TH21 C LOCA, Loss of SD BD C I
9 IOR RHR Sys 1 Containment Spray Valve select switch failure TS-SRO (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor
\
Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#1 When Suppression Chamber Pressure exceeds 12 psig, initiate Drywell Sprays while in the safe region of the Drywell Spray initiation Limit(DSIL) curve and prior to exceeding the PSP limit.
- 1. Safety Significance:
Precludes failure of containment
- 2. Cues:
Procedural compliance.
High Drywell Pressure and Suppression Chamber Pressure
- 3. Measured by:
Observation - US directs Drywell Sprays lAW with EOl Appendix 17B AND Observation - RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment OR CT#1 Before Drywell temperature rises to 280°F, initiate Drywell Sprays while in the safe region of the Drywell Spray Initiation Limit(DSIL) curve.
- 1. Safety Significance:
Precludes failure of containment
- 2. Cues:
Procedural compliance.
High Drywell Pressure and Suppression Chamber Pressure
- 3. Measured by:
Observation - US directs Drywell Sprays lAW with EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment
Appendix D Scenario Outline Form ES-D-1 CT#2 Terminate Drywell/Suppression Chamber Sprays before Drywell/Suppression Chamber pressure drops below 0 psig.
- 1. Safety Significance:
Precludes failure of containment
- 2. Cues:
Procedural compliance.
Drywell Pressure at or below 1.0 psig
- 3. Measured by:
Observation - US directs Drywell Sprays secured lAW with EOI Appendix 17B AND Observation - RO secures Drywell Sprays
- 4. Feedback:
RHR flow to containment lowering RHR Sprays Valves closed REP Classification is an Alert. EAL 2.1-A
Appendix D Scenario Outline Form ES-D-1 Events
- 2. ATC increases power with Control Rods
- 3. Control Rod 30-19 difficult to withdraw. ATC refers to 2-01-85 CRD System section and determines double clutching is to be used initially. Double clutching will work to withdraw rod 30-19.
- 4. Loss of SJAE A, BOP operator swaps to B SJAE JAW 2-A0I-47-3 Loss of Condenser Vacuum.
- 5. Maintenance work in the area of Shutdown Board C will cause the Normal Supply Breaker to trip. Diesel Generator C will fail to automatically start and tie to the shutdown board. The BOP will respond and start DG C and tie to the shutdown board. The SRO will evaluate Technical Specifications and determine TS 3.8.1 Condition B is entered.
Since the Alternate Feeder Breaker is also out of service for SD BD C, Condition G is also entered and Shutdown Board C is declared Inoperable. The SRO will then evaluate Technical Specification 3.8.7 and Condition A is entered.
- 6. RBCCW Pump will trip and the sectionalizing valve will fail to close automatically. ATC will take actions lAW 2-A0I-70-1 and trip RWCU Pumps and close the sectionalizing valve for RBCCW. SRO to evaluate TRM 3.4.1 and inform Chemistry that Reactor Coolant Sampling will for conductivity will have to be performed every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- 7. Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum. Crew scrams the reactor and Enter E0I-1. Bypass valves are unavailable for pressure control and HPCI is the only high pressure system available for level control.
- 8. LOCA will develop and crew enters E0I-2 to control degrading Containment parameters. Loss of SD BD C occurs.
- 9. RHR System 1 Containment Spray/Cooling Valve Select will fail. RHR Loop 2 is available for Drywell Spray. The drywell will be sprayed and drywell sprays will be secured when drywell pressure lowers to 1.0 psig. SRO to evaluate Technical Specification for RHR System 1 Select Logic Failure, Technical Specification 3.6.2.5 Condition B.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are inserted Drywell has been sprayed Reactor Level is restored and maintained
Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 6 8 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Run Time (minutes) 2 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Page 6 of 38 Scenario Tasks TASK NUMBER K/A RO SRO Place RFPT A in Service RO U-003-NO-4 259002A4.03 3.8 3.6 Raise Power with Control Rods RO U-085-NO-7 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod Difficult to Withdraw RO U-085-NO-19 201003A2.O1 3.4 3.6 Loss of SJAE A RO U-066-NO-7 295002AA2.O1 2.9 3.1 SRO S-047-AB-3 DG C Auto Start Failure RO U-082-AL-7 264000A4.04 3.7 3.7 SRO S-000-AD-27 Loss of RBCCW RO U-070-AL-3 206000A2.17 3.9 4.3 SRO S-070-AB-1 LOCA RO U-000-EM-1 295024EA1.11 4.2 4.2 RO U-000-EM-5 SRO S-000-EM-i SRO S-000-EM-2 SRO S-000-EM-5
- Page 7 of 38 Procedures Used/Referenced:
Procedure Number Procedure Procedure Title Revision 2-01-3 Reactor Feedwater System Revision 136 2-GOl-100-12 Power Maneuvering Revision 40 2-01-85 Control Rod Drive System Revision 128 2-01-3 Reactor Feedwater System Revision 136 2-ARP-9-5A Alarm Response Procedure Panel 2-9-5A Revision 48 2-AOl-47-3 Loss of Condenser Vacuum Revision 19 2-ARP-9-53 Alarm Response Procedure Panel 2-9-53 Revision 36 ODCM Offsite Dose Calculation Manual Revision 20 TS 3.8.1 AC Sources Operating Amendment 269 TS 3.8.7 AC Distribution Amendment 269 2-AOI-66-1 Off-Gas H2 High Revision 19 2-AOl-100-1 Reactor Scram Revision 95 2-EOI-1 RPV Control Flowchart Revision 12 2-EOl-2 Primary Containment Control Flowchart Revision 12 2-EOl-2-C-1 Alternate Level Control Flowchart Revision 9 2-EOI-2-C-2 Emergency RPV Depressurization Revision 6 2-EOI Appendix-6D Injection Subsystems Lineup Core Spray System I Revision 7 2-EOI-APPENDIX-17A RHR System Operation Suppression Pool Cooling Revision 12 2-EOI Appendix-5C Injection System Lineup RCIC Revision 5 2-EOl Appendix-7B Alternate RPV Injection System Lineup SLC System Revision 6
Page 8of38 -
Procedures Used/Referenced Continued:
Procedure Number Procedure Procedure Title Revision 2-EOI Appendix-i 1 A Alternate RPV Pressure Control Systems MSRVs Revision 4 2-EOl Appendix-i2 Primary Containment Venting Revision 4 2-EOl Appendix-5B Injection System Lineup CRD Revision 3 2-EOl Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode Revision 8 2-EOl Appendix-i 7C RHR System Operation Suppression Chamber Sprays Revision 1 1 EPIP-1 Emergency Classification Procedure Revision 46 EPIP-5 General Emergency Revision 41
- Page9of38 Console Operator Instructions A. Scenario File Summary Batch File 1108-6 Pref File 110806 Imf dg0ic Imf dg03c F3 bat NRC/i l08rcicto Trg e4 NRC/dgstart F4 Trg e4 = dmf ed09c F5 bat NRC/i 10806 br zdi0hs2 ii 0c02a[ 1] trip F6 imfrdO6r3Ol5 br zloOhs2iiOcO2a[i] off F7 dmfrdO6r3Ol5 br zloOhs2l iOcO2a[2] on F8 imf og04a br zlohs7O48a[21 on F9 imf ed09c br zlohs7O48a[i] off FlO imf sw02b br xa554c19 alarm_off Fl 1 mrf swO2 align Trg el 7048-1 F12 imfogOl Trg el = bat NRC/i 10806-1 Si imfogo5a8o 1200 100 br zlohs66la[1] on S2 ior zdihs66ia open br zlohs66ia[2] off br zdixs74l2l[i] reset Trg e2 modesw bmfth2i (e2 180) 0.5 600 0.1 Imf dgo3c (e2 0)
Scenario 6 DESCRI PTION/ACTION Simulator Setup manual Reset to IC 92 Simulator Setup Load Batch RestorePref NRC/i 10806 Simulator Setup manual F3 and F5 Simulator Setup Verify file loaded
Page 1Oof3S Simulator Event Guide:
Event 1 Normal: Place RFPT A in service from 600 RPM in accordance with 2-01-3 section 5.7 SRO Directs Placing RFPT A in service from 600 rpm.
BOP Places RFPT A in service from 600 rpm.
2-01-3 section 5.7 Placing the Second and Third RFP/RFPT In Service CAUTIONS
- 1) FAILURE to monitor SJAEIQG CNDR CNDS FLOW, 2-Fl-2-42, on Panel 2-9-6 for proper flow (between 2 x 10 b
and 3 x 106 lbm/hr) may result in SJAE isolation.
- 2) Changes in Condensate System flow may require adjustment to SPE CNDS BYPASS, 2-FCV-002-01 90.
NOTE Placing RFP 2A(2B)(2C) MIN FLOW VALVE, 2-HS-3-20(13)(6) in OPEN position will lock it open, preventing minimum flow valve oscillations at ow flow.
[1] NOTIFY Radiation Protection that an RPHP is in effect for the impending action to place RFPT 2A(2B)(2C) in service. RECORD time Radiation Protection notified in NOMS Narrative Log.
[1 .1) VERIFY appropriate data and signatures recorded on Appendix A per Appendix A instructions
[3] VERIFY REP 2A MIN FLOW VALVE, 2-HS-3-20, in OPEN position.
. CHECK OPEN MIN FLOW VALVE, 2-FCV-3-20.
[4] SLOWLY RAISE speed of RFPT, using RFPT 2A SPEED CONT RAISE/LOWER, 2-HS-46-8A, to establish flow to vessel and maintain level.
[5] IF discharge valve was not opened in Step 5.6[2.2.8J AND RFPT discharge pressure is within 250 psig of Reactor pressure, THEN (Otherwise N/A)
OPEN RFP 2A DISCHARGE VALVE, 2-FCV-3-19.
[6] SLOWLY RAISE RFPT speed, using RFPT 2A SPEED CONT RAISE/LOWER switch, 2-HS-46-8A, to slowly raise REP discharge pressure and flow on the following indications (Panel 2-9-6):
[7j WHEN sufficient flow is established to maintain REP 2A MIN FLOW VALVE, 2-FCV-3-20, in CLOSED position ( 2 x 106 lbm/hr), THEN PLACE REP 2A MIN FLOW VALVE, 2-HS-3-20, in AUTO.
Page ilof 38 Simulator Event Guide:
Event 1 Normal: Place RFPT A in service from 600 RPM in accordance with 2-01-3 section 5.7 BOP [8] OBSERVE lowering of speed and discharge flows on other operating RFPs.
NOTE Steps 5.7[9] and 5.7[1 0] transfers control of RFPT from MANUAL GOVERNOR to individual RFPT Speed Control PDS.
[9] PULL RFPT 2A SPEED CONT RAISE/LOWER switch, 2-HS-46-8A, to FEEDWATER CONTROL position.
- CHECK amber light at switch extinguished.
[10] PERFORM the following on RFPT 2A SPEED CONTROL (PDS), 2-SIC-46-8 (Panel 2-9-5):
[10.1] SELECT Column 3.
[10.2] VERIFY PDS in MANUAL.
NOTE Performance of Steps 5.7[1 1] through 5.7[13] will transfer control of RFPT to REACTOR WATER LEVEL CONTROL PDS, 2-LIC-46-5.
[11] VERIFY REACTOR WATER LEVEL CONTROL (PDS), 2-LIC-46-5 functioning properly and ready to control second or third RFP.E
[12] SLOWLY RAISE RFP speed.
- CHECK discharge flow and discharge pressure rise.
[13] WHEN REP speed is approximately equal to operating REP(s) speed, THEN on REPT 2A SPEED CONTROL (PDS), 2-SIC-46-8:
[13.1] PLACE PDS in AUTO.
[13.2] VERIFY Column 3 selected.
[14] WHEN REP is in automatic mode on REACTOR WATER LEVEL CONTROL, (PDS) 2-LIC-46-5, THEN CLOSE the following valves:
Driver When called report 2PCV-6-163 and 2-FCV 6-154 closed
Page 12 of 38 Simulator Event Guide:
Event 1 Normal: Place RFPT A in service from 600 RPM in accordance with 2-01-3 section 5.7 BOP [15] VERIFY CLOSED the following valves on first REP started in Section 5.5:
[16] VERIFY both RFPT Main Oil Pumps running.
Driver When called report 2-FVG-6156/158 are closeØ
[17] IF desired to stop Turning Gear for in service RFPT, THEN PLACE appropriate handswitch in STOP and RETURN to AUTO:
- RFPT2ATURNING GEAR MOTOR, 2-HS-3-1O1A
[18] GO TO Section 6.0.
[18.1] CONTROL and MONITOR RFW System operation.
Page 13 Of 38 -
Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods SRO Direct Power Increase lAW RCP SRO Notify ODS of power increase Direct Power increase using Recirc Flow per 2-GOl-1 00-1 2.
[20] IF desired to raise power with only two (2) Reactor feedpumps in service, THEN RAISE Reactor power, as desired, maintaining each Reactor feedpump less than 5850 RPM.
ATC Raise Power with Control Rods per 2-01-85, section 6.6. Control Rods 14-31, 30-47, 46-31, and 30-15 from 00 to 24, 30-31 from 00 to 48 6.6.1 Initial Conditions Prior to Withdrawing Control Rods
[2] VERIFY the following prior to control rod movement:
- CRD POWER, 2-HS-85-46 in ON.
- Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).
6.6.2 Actions Required During and Following Control Rod Withdrawal
[4] OBSERVE the following during control rod repositioning:
- Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.
- Nuclear Instrumentation responds as control rods move through the core. (This ensures control rod is following drive during Control Rod movement.)
[5] ATTEMPT to minimize automatic RBM Rod Block as follows:
- STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 2-9-5 and PERFORM Step 6.6.2[6].
[6) IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:
[6.1] PLACE CRD POWER, 2-HS-85-46 in the OFF position to deselect the Control Rod.
[6.2] PLACE CRD POWER, 2-HS-85-46, in the ON position.
Pagel4ofS8 -
Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods Event S DRIVER Insert Malfunction to F6 (irnf RDO6r3O15) to stick.rod 3O-1 ATC Withdraw control rods lAW 2-01-85 AT Responds to annunciator 9-5A Window 7, CONTROL ROD WITHDRAWAL BLOCK.
Operator A DETERMINE initiating condition from corresponding rod withdrawal Action: block alarm(s) and REFER TO operator action for alarm(s). 0 ATC Responds to annunciator 9-5A Window 24, RBM HIGH/INOP.
Operator A IF moving control rods for start-up or power maneuvering, THEN Action: PERFORM the following: (otherwise NIA)
I, VERIFY correct control rod selected El
- 2. VERIFY Rod Out Permit light is not illuminated to ensure selected rod withdrawal s inhibited. El I CHECK annunciator LPRM HIGH (1-xa-55-5a, Window 12) and matrix light. Panel 1-9-5 to determine if the alarm is due to high flux. El 4.. DESELECT then RESELECT the desired Control Rod to reset the alarm and reinitialize the RBM back to normalized 100%. El
Page- 15 of 38-Simulator Event Guide:
Event 3 Component: Control Rod Difficult to Withdraw NOTE Control Rod 30-15 will fail to withdraw from position 00 ATC Report Control Rod 30-15 fail to withdraw from position 00 SRO Direct 2-01-85 Section 8.15 ATC 8.15 Control Rod Difficult to Withdraw
[1] VERIFY the control rod will not notch out. Refer to Section 6.6.
[2j REVIEW all Precautions and Limitations in Section 3.0 CAUTION INER!CJ Never pull control rods except in a deliberate, carefully controlled manner, while closely monitoring the Reactors response. wa SOER-8-1
[3] [NRC/C] IF RWM is enforcing, THEN VERIFY RWM is operable and LATCHED in to the correct ROD GROUP. [NRC-IR 84-02]
NOTES
- 1) Steps 8:1 5[4] through 8 15[6j should be used when the control rod is at Position 00 while Step 81 5[7 should be used when the control rod is at OR between Positions 02 and 46
- 2) Double clutching of a control rod at Position 00 will place the rod at the overtraveI in stop, independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet fingers from engaging the 00 notch.
- 3) Step 8.1 5[4] may be repeated as necessary until it is determined that this method Will not free the control rod.
Page l6of 38 Simulator Event Guide:
Event 3 Component: Control Rod Difficult to Withdraw (continued)
DRIVER Delete malfunction F7 (dmf rdO6r3Ql 5) when double dutch is used.,
AT
[4] IF the control rod problem is not believed to be air in the hydraulic system, THEN PERFORM the following to double clutch the control rod at Position 00:
[4.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 2-HS-85-47, in EMERG ROD IN, for several seconds.
[4.2] CHECK the control rod full in indication (double green dashes) on the Full Core Display for the associated control rod.
[4.3] SIMULTANEOUSLY PLACE CRD NOTCH OVERRIDE, 2-HS-85-47, in NOTCH OVERRIDE AND CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.
[4.4] WHEN EITHER of the following occur:
- Control rod begins to move, OR
- It is determined the rod will not move, THEN RELEASE 2-HS-85-47 AND 2-HS-85-48.
[4.5] IF the control rod successfully notches out, THEN PROCEED TO Section 6.6 and WITHDRAW the control rod to the appropriate position.
[4.6] IF Desired, THEN REPEAT Steps 8.15[4.1] through 8.15[4.5] several times prior to raising drive water pressure in Step 8.15[5].
- Page17 of 38 Simulator Event Guide:
Event 4 Component: Loss of SJAE A Event 3 DRIVER Insert Malfunction to F8 (imf OGO4a)to cause a loss of SJAE A SRO Enters AOI-47-3 Loss of Condenser Vacuum.
BOP Offgas Panel 9-53 Alarms:
Window 4, OG HOLDUP LINE INLET FLOW LOW:
Operator action:
VERIFY OPEN, FCV-66-28, off-gas system isolation valve.
VERIFY that SJAE auto isolation has NOT occurred.
Window 10, H2 WATER CHEMISTRY ABNORMAL:
Operator action:
None at this time Window 20, H2 WATER CHEMISTRY SHUTDOWN:
Operator action:
None at this time BOP Swaps to B SJAE lAW 2-AOI-47-3 Loss of Condenser Vacuum.
4.2 Subsequent Actions (continued)
[11] IF a failure of the in-service SJAE is indicated, THEN PLACE the standby SJAE in service as follows:
NOTES
- 1) This section may be used to return either SJAE to service following a shutdown or an isolation.
- 2) Potential causes of PCV valve closure are:
. Condensate pressure from SJAE A(B) less than 60 psig, 2-Pl-2-34(40),
Panel 25-105.
. SJAE 2A(2B) CONDENSATE INLET VALVE closed at 2-HS-2-31A(36),
Panel 2-9-6.
. SJAE 2A(2B) CONDENSATE OUTLET VALVE closed at 2-HS-2-35A(41A),
Panel 2-9-6.
e STEAM TO SJAE A(B) STAGE I & Il, 2-Pl-1-150(152), Panel 25-1 05 is less than 155 psig. (disabled for the SJAE selected by 2-HS-001-0375)
. Loss of l&C bus A(B), power is required to be restored to return the SJAE to service.
- 3) 2HS-001-0375, SJAE TRAIN PERMISSIVE, should be placed in the position for the SJAE being placed in service. This switch will normally be in the position of the standby SJAE.
- Pagel8of-3&
Simulator Event Guide:
Event 4 Component: Loss of SJAE A (continued)
[11.11 PLACE SJAE TRAIN PERMISSIVE 2-HS-001 -0375 in the position for the SJAE being placed in service. This switch will normally be in the position of the Standby SJAE. (Panel 925-105 on junction box 8595) (N/A if Placing the standby SJAE in service)
[11 .2] VERIFY off gas isolation is reset, using OG OUTLET/DRAIN ISOLATION VLVS, 2-HS-90-155, Panel 2-9-8.
[11 .3] VERIFY the following valves are OPEN:
- SJAE 2A(2B) INLET VALVE, 2-HS-66-1 1(15),
Panel 2-9-8
- STEAM TO SJAE 2A(2B), 2-HS-1-155A(156A),
Panel 2-9-7 BOP [11.4] VERIFY SJAE 2A(2B) OG OUTLET VALVE, 2-HS 14(18), AUTO/OPEN (Panel 2-9-8)
[11.5] PLACE SJAE 2A(2B) PRESS CONTROLLER 2-HS-1-1 50(1 52) in CLOSE and then in OPEN at Panel 2-9-7.
[11 .6] VERIFY the following valves OPEN (red lights illuminated) at Panel 2-9-7.
- STEAM TO SJAE 2A(2B) STAGES 1,2, AND 3, 2-PCV-1-151/166 (153/167).
- SJAE 2A(2B) INTMD CONDENSER DRAIN 2-FCV-1 -150(152).
[11.7] MONITOR hotwell pressure as indicated on HOTWELL PRESS AND TEMP recorder, 2-XR-2-2 (Panel 2-9-6).
[11.8] For the SJAE not being placed in service,
- VERIFY CLOSED SJAE 2B(2A) PRESSURE CONTROLLER, 2-HS-1 -1 52(1 50) (Panel 2-9-7)
[11.9] VERIFY SJAE TRAIN PERMISSIVE, 2-HS-001 -0375, in the position for the SJAE selected for Standby operation SJAE A(SJAE B). (Panel 925-1 05 on junction box 8595)
Pagelaof 38-Simulator Event Guide:
Event 5 Component: DG C Auto Start Failure DRIVER Insert malfunction F9 (imf EDO9c to cause a loss of Shutdown Board C when operator start_DG_C_ensure_EDQ9C_is deleted.
BOP Recognizes Loss of Shutdown Board C failure of to DG C start, and Manually Starts_DG_C_and_close_DG_Supply_Breaker BOP Reports Loss of Shutdown Board C, failure of DG C to start, and manual start of DG Cto SRO.
DRIVER When requested t investigate the shutdowh board jeport that while MaintenancG was moving the Shutdown Board C Alternate breaker it bumped the racking toot shutter door for the Shutdown Board C Normal breaker. This caused the Normal breaker to open SRO Evaluates Tech Specs 3.8.7 (condition A) and 3.8.1 (condition B)
TSR 3.8.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TI ME A. One Unit 1 and 2 416 kV NOTE Shutdown Board Enter applicable CondItions and inoperable. Required Actions of Condition B, C, D, and G when Condition A results in no power source to a required 480 volt board.
A,l Restore the Unit I and 2 5 days 4.16 kV Shutdown Board to OPERABLE status. NP
. 12 days from discovery of failure to meet
- LCO AND A.2 Declare associated diesel Immediately generator inoperable.
(continued)
Page20 of 38 Simulator Event Guide:
Event 5 Component: DG C Auto Start Failure SRO TSR 3.8.1 B. One required Unit I and 2 B.1 Verify power availability 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> DG inoperable, from the offsite transmission network.
Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND B. (continued) B.2 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from feature(s), supported by discovery of the inoperable Unit 1 and Condition B 2 DG, inoperable when concurrent with the redundant required inoperability of feature(s) are inoperable, redundant required feature(s)
AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Unit I and 2 DG(s) are not inoperable due to common cause failure.
OR B.3.2 Perform SR 3.8.1.1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE Unit 1 and 2 DG(s).
AND BA Restore Unit 1 and 2 DG 7 days to OPERABLE status.
AND 14 days from discovery of failure to meet LCO
- Page2lof38 Simulator Event Guide:
Event 6 Component: Loss of RBCCW CR1 VEF Insert malfuncflon (swO2b) to cause a loss of RBCcW.
Responds to alarm 4C-12, RBCCW PUMP DISCH. HDR PRESS LOW BOP/ATC Report Trip of RBCCW Pump 2B.
BOP/ATC Automatic Action: Closes 2-FCV-70-48, non-essential loop, closed cooling water sectionalizing MOV.
A. VERIFY 2-FCV-70-48 CLOSI NG/CLOSED.
B. VERIFY RBCCW pumps A and B in service.
C. VERIFY RBCCW surge tank low level alarm is reset.
D. DISPATCH personnel to check the foNowing:
- RBCCW surge tank level locally.
- RBCCW pumps for proper operation.
E. REFER TO 2-AOI-70-1, for RBCCW System failure and 2-01-70, for starting spare pump.
SRO Enters 2-AOI-70-1.
ATC Closes 2-FCV-70-48 and report the sectionalizing valve failed to close automatically BOP Dispatch Personnel to investigate RBCCW Pump 2B trip ATC 2-AOI-70-1 4.1 Immediate Actions
[1] IF RBCCW Pump(s) has tripped, THEN Perform the following
- SECURE RWCU Pumps.
ATC Secures RWCU Pumps and Closes 2-FCV-70-48.
4.2 Subsequent Actions
[1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, SRO AND core flow is above 60%,THEN: (Otherwise N/A):
[2] IF any E0I entry condition is met, THEN ENTER appropriate EOI(s) (Otherwise N/A).
Page 22 of 3&
Simulator Event Guide:
Event 6 Component: Loss of RBCCW (continued)
Steps 1 and 2 are NA
[3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (Otherwise N/A):
[3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions.
[3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s).
DRIVER When dispatched, report RBGOW Pump 2B breaker is trapped There is also a smell of burnt wiring arid charring on the breaker.
SRO [4] IF unable to restart a tripped pump, THEN PLACE Spare RBCCW Pump in service. REFER TO 2-01-70. Direct Unit 1 to place Spare RBCCW Pump in service DRIVER When called to place spare RBCCW Pump in services wait 3 minutes (IRE SWO align). THEN inform Unit? Operato that spare RBCOW Pump is in seMce SRO [5] IF RBCCW flow was restored to two pump operation by placing the Spare RBCCW pump in service in the preceding step, THEN PERFORM the following:
[5.1] REOPEN RBCCW SECTIONALIZING VLV, 2-HS-70-48A.
[5.2] RESTORE the RWCU system to operation. (REFER TO 2-01-69)
Directs ATC or BOP to Open Sectionalizing Valve and Restore RWCU.
ATC Opens Sectionalizing Valve, 2-FCV-70-48.
Page 2Sof 38 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
DRIVER Insert P12 (irnf ogOl) and Shift Fl, to cause High Offgas ydrogen BOP Responds to alarm the following alarms:
HIGH OFFGAS % H2 TRAIN A (2-XA-55-53, Window 3)
HIGH OFFGAS % H2 TRAIN B (2-XA-55-53, Window 13)
OFFGAS MONITOR PANEL TROUBLE,(2-XA-55-589, Window 07)
BOP Reports a rise in hydrogen concentration on OFF GAS HYDROGEN ANALYZER (OH 1-Analyzer 2A, OH 2-Analyzer 2B) recorder, 2-H2R-66-96, Panel 9-53.
SRO Enters 2-AOI-66-1, Off-Gas H2 High.
DRIVER Insert Shift F2 when many alarms are received on OFF GAS panel (ior zdihs66ia open),_opens condenser vacuum_breaker BOP Responds to alarm 9-53-Window 14 OG HOLDUP LINE INLET FLOW HIGH.
ATC Report degrading condenser Vacuum.
ATC Inserts Reactor Scram when directed; and places mode switch in shutdown.
ATC Recognizes reactor scram. Verifies rods inserted. Reports Scram announcement.
SRO Enters EOI-1 and EOI-2.
SRO EOI-1 (Reactor Pressure)
Monitor and Control Reactor Pressure IF Drywell Pressure Above 2.4 psig? YES, but action Not Required.
IF Emergency Depressurization is Anticipated and the Reactor will remain subcritical without boron under all conditions, THEN Rapidly depressurize the RPV with the Main Turbine Bypass Valves irrespective of cooldown rate? NO-IF Emergency Depressurization is required THEN exit RO/P and enter C2 Emergency Depressurization? NO -
IF RPV water level cannot be determined? NO -
Directs Manually open MSRVs until RPV Pressure drops to the pressure at which all turbine bypass valves are open. (Appendix hA)
IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area j of Curve 3?- NO
Page 24 of 38 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
(continued)
IF Suppression Pool level cannot be maintained in the safe area of Curve 4? NO -
IF Drywell Control air becomes unavailable? - NO IF Boron injection is required? - NO SRO Directs a Pressure Band with SRVs, lAW Appendix hA.
Should begin to lower Reactor Pressure, not to exceed 1 00°F/hr cooldown.
ATC Control Reactor Pressure in assigned band, lAW Appendix 1 1 A.
ATC/BOP Pressure Control lAW AppendixilA, RPV Pressure Control SRVs.
NA 1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.
NA 2. IF Suppression Pool level is at or below 5.5 ft, THEN CLOSE MSRVs and CONTROL RPV pressure using other options.
- a. 2-PCV-1 -1 79 MN STM LINE A RELIEF VALVE
- b. 2-PCV-1 -1 80 MN STM LINE D RELIEF VALVE
- c. 2-PCV-1-4 MN STM LINE A RELIEF VALVE
- d. 2-PCV-1-31 MN STM LINE C RELIEF VALVE
- e. 2-PCV-1-23 MN STM LINE B RELIEF VALVE
- f. 2-PCV-1-42 MN STM LINE D RELIEF VALVE
- g. 2-PCV-1-30 MN STM LINE C RELIEF VALVE
- h. 2-PCV-1-19 MN STM LINE B RELIEF VALVE.
- i. 2-PCV-1-5 MN STM LINE A RELIEF VALVE.
- j. 2-PCV-1-41 MN STM LINE D RELIEF VALVE
- m. 2-PCV-1-34 MN STM LINE C RELIEF VALVE
Page 2Sof 38 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
(continued)
ATC/BOP Pressure Control lAW Appendixi 1A RPV Pressure Control SRVs NA 4. IF Drywell Control Air header supplied from CAD System A, shows indications of being depressurized, as determined by Appendix 8G, THEN OPEN MSRVs supplied by CAD System B; using the following sequence to control_RPV_pressure;_as_directed_by_SRO:
NA 5. IF Drywell Control Air header supplied from CAD System B, shows indications of being depressurized, as determined by Appendix 8G, THEN OPEN MSRVs supplied by CAD System A; using the following sequence to control_RPV_pressure;_as_directed_by_SRO:
EOI1 RPV Pressure Augment RPV Pressure control, as necessary; with one or more of the following depressurization systems:
. HPCI Appendix 11C
. RCIC Appendix 11 B
. RFPTs on minimum flow Appendix 1 1 F SRO
- Main Steam System Drains Appendix 11 D
- Steam Seals Appendix 1 1 G
- SJAEs Appendix 1 1 G
- Off Gas Preheater Appendix 11 G
- RWCU Appendix 1 1 E.
ATC/BOP Augments RPV Pressure Control, if directed by SRO.
SRO EOI-1 (Reactor Level)
Monitor and Control Reactor Water Level.
Directs Verification of PCIS isolations.
ATC/BOP Verifies PCIS isolations.
SRO Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with one or more of the following injection sources. (H PCI, Appendix 5D)
ATC Maintains the prescribed level band, lAW Appendix 5D.
- 1. IF Suppression Pool level drops below 12.75 ft during HPCI operation, THEN_TRIP_HPCI_and_CONTROL_injection_using_other_options.
- 2. IF Suppression Pool level CANNOT be maintained below 4.25 in., THEN EXECUTE EOl Appendix 1 6E concurrently with this procedure to bypass HPCI High Suppression Pool Water Level Suction Transfer Interlock.
Page 26 ef 38 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
(continued)
- 3. IF BOTH of the following exist:
. High temperature exists in the HPCI area, AND
. SRO directs bypass of HPCI High Temperature Isolation interlocks, THEN PERFORM the following:
- a. EXECUTE EOI Appendix 1 6L concurrently with this procedure.
- b. RESET auto isolation logic using 2-XS-73-58A(B) HPCI AUTO-ISOL_LOGIC A(B)_RESET_pushbuttons.
CAUTION Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.
Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.
- 4. VERIFY 2-IL-73-18B, HPCI TURBINE TRIP RX LVL HIGH amber light extinguished.
- 5. VERIFY at least one SGTS train in operation.
- 6. VERIFY 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.
NOTE HPCI Auxiliary Oil Pump will NOT start UNTIL 2-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, starts to open.
- 7. PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in START.
- 8. PLACE 2-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, handswitch in START.
- 9. OPEN the following valves:
- 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE
- 2-FCV-73-44, HPCI PUMP INJECTION VALVE.
[
- Page27o38 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
(continued)
- 11. CHECK proper HPCI operation by observing the following:
- a. HPCI Turbine speed accelerates above 2400 rpm.
- b. 2-FCV-73-45, HPCI TESTABLE CHECK VLV, opens by observing 2-Zl-73-45A, DISC POSITION, red light illuminated.
- d. 2-FCV-73-30, HPCI PUMP MIN FLOW VALVE, closes as flow exceeds 1200 gpm.
- 12. VERIFY HPCI Auxiliary Oil Pump stops and the shaft-driven oil pump operates properly.
- 13. WHEN HPCI Auxiliary Oil Pump stops, THEN PLACE 2-HS-73-47A, HPCI AUXILIARY OIL PUMP, handswitch in AUTO.
- 14. ADJUST 2-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller as necessary to control injection.
Page 28 of 38 Simulator Event Guide:
Event 8 Component: LOCA, Loss of SD BD C SRO Enters EOl-2, all legs.
EOl-2 (Drywell Temperature)
SRO Monitor and Control DW Temp Below 160°F, using available DW Cooling.
Can Drywell Temp Be Maintained Below 160°F? NO -
SRO Directs H202 Analyzers placed in service, lAW Appendix 19.
BOP Places H202 analyzers in service, lAW Appendix 19.
SRO EOl-2 (Primary Containment Pressure)
Monitor and Control PC Pressure Below 2.4 psig, Using the Vent System As Necessary. (Appendix 12)
SRO Directs venting of Primary Containment, per Appendix 12.
BOP Vents Primary Containment, lAW Appendix 12.
- 1. VERIFY at least one SGTS train in service.
- 2. VERIFY CLOSED the following valves (Panel 2-9-3 or Panel 2-9-54):
. 2-FCV-64-31, DRYWELL INBOARD ISOLATION VLV
. 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE
. 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV
.__2-FCV-64-32,_SUPPR_CHBR VENT_INBD_ISOL VALVE Steps 3, 4, 5 and 6 are If I Then steps that do not apply.
- 7. CONTINUE in this procedure at:
Step 8 to vent the Suppression Chamber through 2-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.
Page29of 3 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure
- 8. VENT the Suppression Chamber using 2-FIC-84-1 9, PATH B VENT FLOW CONT, as follows:
- a. PLACE keylock switch 2-HS-84-35, DW/SUPPR CHBR VENT SQL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).
- b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).
- c. PLACE 2-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).
- d. PLACE keylock switch 2-HS-84-1 9, 2-FCV-84-1 9 CONTROL, in OPEN (Panel 2-9-55).
- e. VERIFY 2-FIC-84-1 9, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.
- f. CONTINUE in this procedure at step 12.
SRO Can PC Pressure Be Maintained Below 2.4 psig? NO -
SRO Directs Suppression Chamber Sprays per Appendix 1 7C NOTE Sprays are unavailable on Loop I of RHR due to faiied Select Logic ATC/BOP Sprays the Suppression Chamber per Appendix 17C
- 1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.
- 2. IF Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
- 3. IF Directed by SRQ to spray the Suppression Chamber using Standby Coolant Supply, THEN ... CONTINUE in this procedure At Step 7 using RHR Loop I OR At Step 8 using RHR Loop II.
- Page3Gof38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
- 4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN ... CONTINUE in this procedure at Step 9.
- 5. INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN...PLACE keylock switch 2-XS-74-1 22(1 30), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- c. MOMENTARILY PLACE 2-XS-74-121 (129), RHR SYS 1(11)
CTMT SPRAY/CLG VLV SELECT, switch in SELECT.
- d. IF 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN...VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11)
OUTBD INJECT VALVE.
- e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spray.
- g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
- h. IF RHR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN...CONTINUE in this procedure at Step 5.k.
- i. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
- j. RAISE System flow by placing the second RHR System 1(11) pump in service as necessary.
I. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
Page 31 of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
- 6. WHEN ... EITHER of the following exists:
- Before Suppression Pool pressure drops below 0 psig, OR
- Directed by SRQ to stop Suppression Chamber Sprays, THEN ... STOP Suppression Chamber Sprays as follows:
- a. CLOSE 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
- b. VERIFY CLOSED the following valves:
- 2-FCV-74-100, RHR SYS I U-i DISCH XTIE
- 2-FCV-74-1O1, RHR SYS II U-3 DISCH XTIE.
- c. IF RHR operation is desired in ANY other mode, THEN...EXIT this EQI Appendix.
- d. STOP RHR Pumps 2A and 2C (2B and 2D).
Page 32 of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
SRO EOI-2 (Suppression Pool Level)
Monitor and Control Suppression Pool Level between (-) 1 inch and (-) 6 inches.
(Appendix 18)
Can Suppression Pool Level Be Maintained above (-) 6 inches? - YES Can Suppression Pool Level Be Maintained below (-) 1 inch? - YES BOP Places H202 analyzers in service, lAW Appendix 19.
- 1. IF A Group 6 PCIS signal exists, THEN PLACE 2-HS-76-69, H2102 ANALYZER ISOLATION BYPASS switch in BYPASS (Panel 2-9-54).
- 2. DEPRESS 2-HS-76-91, H2/02 ANALYZER ISOLATION RESET.
- 3. IF H2102 Analyzer is to sample the Suppression Chamber, THEN ALIGN Analyzer as follows (Panel 2-9-54):
- a. PLACE 2-HS-76-1 10, H2102 ANALYZER DW/SUPPR CHBR SELECT in SUPPR CHBR position.
- b. VERIFY SUPPR CHBR SMPL VLVS 2-FSV-76-55/56 OPEN using 2-IL-76-49-1.
- c. VERIFY OPEN SMPL RTN VLVS 2-FSV-76-57158 using 2-IL-76-49-3.
- 4. IF H2/O2 Analyzer is to sample the Drywell, THEN ALIGN Analyzer as follows (Panel 2-9-54):
- a. PLACE 2-HS-76-1 10, H2/O2 ANALYZER DW/SUPPR CHBR SELECT in DRYWELL position.
- b. VERIFY OPEN DRYWELL SMPL VLVS 2-FSV-76-49/50 using 2- I L-76-49-2.
- c. VERIFY OPEN SMPL RTN VLVS 2-FSV-76-57/58 using 2-IL-76-49-3.
Page33of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
BOP Places H202 analyzers in service, lAW Appendix 19.
- 5. IF H2/02 Analyzer is in STANDBY at 2-MON-76-1 10 (Panel 2-9-55), THEN PLACE H2/02 Analyzer in service at as follows:
- a. TOUCH 2-MON-76-1 10 display screen.
- b. DEPRESS Go To Panel PROCESS VALUES soft key.
- c. DEPRESS Go To Panel MAINT MENU soft key.
- d. DEPRESS LOG ON soft key.
- e. ENTER password 1915 on soft keypad.
- f. DEPRESS ENT soft key on keypad.
- g. DEPRESS STANDBY MODE ON soft key to enable sample pump operation.
- h. VERIFY soft key reads STANDBY MODE OFF.
DEPRESS Go To Panel PROCESS VALUES soft key.
- j. DEPRESS Go To Panel MAIN soft key.
- k. VERIFY STANDBY MODE is NOT displayed.
- 6. VERIFY H2/O2 ANALYZER SAMPLE PUMP running using 2-XI-76-1 10 (Panel 2-9-55).
- 7. VERIFY red LOW FLOW indicating light extinguished at 2-MON-76-1 10, H2/02 ANALYZER (Panel 2-9-55).
- 8. WHEN H2/O2 Analyzer has been aligned and sampling for 10 minutes or greater, THEN OBTAIN H2 and 02 readings from 2-XR-76-1 10 H2/O2 CONCENTRATION recorder (Panel 2-9-54).
SRO EOI-2 (Suppression Pool Temperature)
Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Cooling As Necessary. (Appendix 1 7A)
Can Suppression Pool Temperature Be Maintained Below 95°F? - NO ATC Places Suppression Pool Cooling in service, lAW Appendix 17A using Loop I of Residual Heat Removal.
Page34of38 -
Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
ATC/BOP Places Suppression Pool Cooling in service, lAW Appendix 17A.
- 1. IF Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, THEN BYPASS LPCI injection valve auto open signal as necessary; by PLACING 2-HS-74-155B, LPCI SYS II OUTBD INJVLV BYPASS SEL in BYPASS.
- 2. PLACE RHR SYSTEM II in Suppression Pool Cooling as follows:
- c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- d. IF Directed by SRO, THEN PLACE 2-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.
- e. IF LPCI INITIATION Signal exists, THEN MOMENTARILY PLACE 2-XS-74-129, RHR SYS II CTMT SPRAY/CLG VLV SELECT in SELECT.
- f. IF 2-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE.
- h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.
- i. THROTTLE 2-FCV-74-73, RHR SYS II SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 2-Fl-74-64, RHR SYS II FLOW:
- Between 7000 and 10000 gpm for one-pump operation.
- At or below 13000 gpm for two-pump operation.
- j. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II MIN FLOW VALVE.
Page3S of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
SRO When Suppression Chamber Pressure exceeds 12 psig, determines that Drywell Sprays are required.
Directs Loop II of RHR to be placed in Drywell Sprays per EOI Appendix 17B.
ATC/BOP Drywell Sprays per appendix 17B
- 1. IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
. PLACE 1-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.
- 2. VERIFY Recir Pumps and Drywell Blowers shutdown.
THEN CONTINUE in this procedure at Step 6 using RHR Loop 1(11).
NOTE Step 6 is performed ONLY if directed by Step 3 to spray the Drywell using RHR Loops 1(11).
- 6. INITIATE Drywell Sprays using RHR Loop 1(11) as follows:
- a. BEFORE drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 9.
- c. IF EITHER of the following exists:
. LPCI Initiation signal is NOT present, OR
. Directed by SRO, THEN PLACE keylock switch 1-XS-74-122(130),
RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- Page 36 of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
- d. MOMENTARILY PLACE 1-XS-74-1 21 (1 29), RHR SYS 1(11)
CTMT SPRAY/CLG VLV SELECT, switch in SELECT.
- e. IF 1-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 1-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.
- f. VERIFY OPERATING the desired System 1(11) RHR pump(s) for Drywell Spray.
- g. OPEN the following valves:
. 1-FCV-74-60(74), RHR SYS 1(11) DW SPRAY OUTBDVLV
. 1-FCV-74-61(75), RHR SYS 1(11) DW SPRAY INBD VLV.
- h. VERIFY CLOSED 1-FCV-074-0007(0030), RHR SYSTEM 1(11)
MIN FLOW VALVE.
- i. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System 1(11) RHR Pump in service.
I. THROTTLE the following in-service RHRSW outlet valves to obtain between 1 ,350 and 4,500 gpm RHRSW flow:
. 1-FCV-23-34, RHR HX 1A RHRSW OUTLET VLV
. 1 -FCV-23-46, RHR HX 1 B RHRSW OUTLET VLV
. 1-FCV-23-40, RHR HX 1C RHRSW OUTLET VLV
. 1 -FCV-23-52, RHR HX 1 D RHRSW OUTLET VLV.
Page37 of 38 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
- 9. WHEN EITHER of the following exists:
- Before drywell pressure drops below 0 psg, OR
- Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:
- a. VERIFY CLOSED the following valves:
- 1-FCV-74-101, UNITS 1-2 DISCHARGE CROSSTIE
SRO REP Classification is an Alert. EAL 2.1-A
Page 3&oN3&
SHIFT TURNOVER SHEET The unit is at approximately 80% power.
Equipment Out of Service/LCOs:
RCIC is out of service.
Breaker 1624 Alternate Feed to SD BD C.
Operations/Maintenance for the Shift:
Place RFPT A in service from 600 RPM in accordance with 2-OI-3section 5.7 and then raise power to 100% in accordance with the RCP.
Units 1 and 3 are at 100% Power Unusual Conditions/Problem Areas:
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 7- Op-Test No.: 1108 Examiners:_____________________ Operators: SRO:_____________________
ATC:______________
BOP:_______________
Initial Conditions: 95% power. Loop 2 Core Spray is tagged out.
Turnover: Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 and then raise reactor power to 100% with Recirculation.
Event Maif. No. Event Type* Event Description No.
N-BOP Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 1
TS-SRO section 5.2, Relative Humidity heater fails for TS action R-ATC 2 Raise Power with Flow R-SRO R-ATC 3 ADO1a TS-SRO ADS SRV 1-5 fails open C-BOP C-ATC VFD Cooling Water Pump 2B trips with failure of the standby 4 TH18d C-SRO pump to auto start R-ATC Loss of FW Heating 5 FWO5b CBOP 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to C-SRO isolate 6 FW3Oa RFPT A Flow Controller failure 7 Batch File M-ALL Earthquake, Loss of All High Pressure injection 8 C Loss of LPCI MG sets 9 C Loss of ALL Level Control Systems Steam Cooling
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor J
Appendix D Scenario Outline Form ES-D-1 Critical Tasks Four CT#1 With NO injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -200 inches, initiate Emergency Depressurization.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
Water level trend.
- 3. Measured by:
Observation At least 6 SRVs must be opened when RPV level lowers to -200 inches.
- 4. Feedback:
RPV pressure trend.
SRV status indications.
OR CT#1 With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, initiate Emergency Depressurization before RPV level lowers to -180 inches.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
Water level trend.
- 3. Measured by:
Observation At least 6 SRVs must be opened before RPV level lowers to -180 inches.
- 4. Feedback:
RPV pressure trend.
SRV status indications.
Appendix D Scenario Outline Form ES-D-1 CT#2 With RPV pressure below the Shutoff Head of the available Low Pressure system(s), operate available Low Pressure system(s) to restore RPV water level above TAF.
- 1. Safety Significance:
Maintaining adequate core cooling.
- 2. Cues:
Procedural compliance.
Pressure below low pressure ECCS system(s) shutoff head.
- 3. Measured by:
Operator manually starts initiates at least one low pressure ECCS system and injects into the RPV to restore water level above TAF.
- 4. Feedback:
Reactor water level trend.
Reactor pressure trend.
CT#3-To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
- 3. Measured by:
ADS logic inhibited prior to an automatic initiation.
- 4. Feedback:
RPV pressure trend.
RPV level trend.
ADS ADS LOGIC BUS A/B INHIBITED annunciator status.
Appendix D Scenario Outline Form ES-D-1 CT#4 With a SRV(s) open due to failure or incorrect automatic actuation, initiate action to close the SRV(s).
- 1. Safety Significance:
Preclude exceeding Tech. Spec limit.
Degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
SRV OPEN annunciator status.
- 3. Measured by:
Observation SRV closed when the MSRV Inhibit Switch placed in OFF.
- 4. Feedback:
Suppression Pool temperature trend.
SRV status indications.
Appendix D Scenario Outline Form ES-D-1 EVENTS
- 1. BOP starts SBGT Fan C and aligns to Reactor Bldg JAW 0-01-65 section 5.2. The relative humidity heater will fail to start and the SRO will evaluate Technical Specification 3.6.4.3 and determine Condition A is entered.
- 2. ATC raises Power with flow
- 3. ADS SRV 1-5 will fail open. ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to close SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F
- 4. The VFD Cooling Water Pump for the B Reactor Recirc VFD will trip and the standby pump will fail to start. The ATC will start the standby VFD Cooling Water Pump to restore cooling water preventing a VFD and Reactor Recirc Pump trip.
- 5. A tube leak on High Pressure Feedwater Heater B2 results in isolation of Extraction Steam to the heater. The crew will respond in accordance with 2-A0I-6-1A or 1C. The ATC will lower reactor power by 5%. The BOP Operators refers to 2-AOI-6-1A or 1C and determine that all automatic actions failed to occur and will isolate Heater B2.
- 6. RFPT A flow controller will slowly fail high, level will remain unchanged, RFPT A speed will continue to increase until the ATC or Crew notices. The controller will fail to respond until the ATC takes manual control with handswitch. The Operator will be able to restore RFPT A speed in manual. SRO should direct entry into 2-AOI-3-1.
- 7. Earthquake and Feedwater line Break Loss of High Pressure Injection. On the scram, a feedwater line will break requiring the crew to isolate feedwater and HPCI. The crew will respond JAW EOI-l, EOI-2 and EOI-3.
- 8. Loss of LPCI MG Sets Loss of RHR and Core Spray Pumps. Electrical faults will result in all injection to the core being lost. The SRO will transition to C-i, at -180 inches the SRO will transition to Steam Cooling. Once steam cooling is entered repairs will be completed to one electrical bus and an ECCS low pressure system will be restored for vessel injection. The SRO will transition to C-2, direct Emergency Depressurization and level restored to +2 to +51 inches.
- 9. Loss of All injection sources When crew enters steam cooling, one LPCI MG set will be restore to service, Crew will ED and restore reactor level.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.
All Control Rods are inserted.
Emergency Depressurization is complete Reactor Level is restored and maintained
Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 7 10 Total Malfunctions Inserted: List (4-8) 6 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 3 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER K/A RO SRO Manual Initiation of SBGT Fan C RO U-065-NO-02 261000A4.07 3.1 3.2 SRO S-000-AD-27 Raise Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 ADS SRV Fails Open RO U-OO1-AB-1 239002A2.03 4.1 4.2 SRO S-0001-AB-1 VFD Cooling Water Pump Failure RO U-068-AL-33 202001A2.22 3.1 3.2 SRO 5-068-AB-Ol Loss of Feedwater Heating RO U-006-AB-01 2.1.43 4.1 4.3 SRO 5-006-AB-Ol Reactor Feed Pump Turbine Governor Failure RO U-003-AL-9 259002A4.01 3.8 3.6 SRO S-003-AB-1 Steam Cooling RO U-000-EM-15 29503 1EA2.04 4.6 4.8 SRO S-000-EM-15 SRO T-000-EM-16
Scenario 7 Page8of35- --
Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 0-01-24 Standby Gas Treatment System Revision 53 TS 3.6.4.3 Containment Systems Amendment 290 2-GOI-100-12 Power Maneuvering Revision 40 2-01-68 Reactor Recirculation System Revision 138 2-ARP-9-3C Alarm Response Procedure Panel 2-9-3C Revision 20 2-AOl-i-i Relief Valve Stuck Open Revision 26 2-01-74 RHR System Revision 157 2-ARP-9-4B Alarm Response Procedure Panel 2-9-4B Revision 39 2-ARP-9-4C Alarm Response Procedure Panel 2-9-4C Revision 30 2-ARP-9-7A Alarm Response Procedure Panel 2-9-7A Revision 27 2-ARP-9-6A Alarm Response Procedure Panel 3-9-6A Revision 28 High Pressure Feedwater Heater String/Extraction Steam 2-A0I-6-1A . Revision 17 Isolation High and Low Pressure Feedwater Heater String/Extraction 2-AOI-6-1C Revision 14 Steam Isolation 2-01-6 Feedwater Heating and Misc Drains System Revision 84 2-ARP-9-5A Alarm Response Procedure Panel 3-9-5A Revision 48 2-ARP-9-6C Alarm Response Procedure Panel 3-9-6C Revision 19 TS 3.5.1 ECCS Operating Amendment 269 Loss of Reactor Feedwater or Reactor Water Level 2-AOI-3-1 . Revision 20 High/Low 0-AOI-100-5 Earthquake Revision 33 2-AOI-100-1 Reactor Scram Revision 95 2-EOI-1 RPV Control Flowchart Revision 12 2-EOI-2 Primary Containment Control Flowchart Revision 12 2-EOI-2-C-1 Alternate Level Control Flowchart Revision 9 2-EOI-2-C-2 Emergency RPV Depressurization Revision 6 2-EOI Appendix-6D Injection Subsystems Lineup Core Spray System I Revision 7 2-EOI-APPENDIX-17A RHR System Operation Suppression Pool Cooling Revision 12 2-EOI Appendix-SC Injection System Lineup RCIC Revision 5
Scenario 7 Page9of35-.
Procedure Number Procedure Title Procedure Revision 2-EOI Appendix-5B Injection System Lineup CRD Revision 3 2-EOI Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode Revision 0 2-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System Revision 6 2-EOI Appendix-i IA Alternate RPV Pressure Control Systems MSRVs Revision 4 2-EOI Appendix-12 Primary Containment Venting Revision 4 2-EOI Appendix-17C RHR System Operation Suppression Chamber Sprays Revision 0 EPIP- 1 Emergency Classification Procedure Revision 46 EPIP-5 General Emergency Revision 41
Scenario 7 Page- 1-0-of -35 Console Operator Instructions Batch 1108-7 Preference File 110807 Trg ii NRC/msrvinhibit F3 bat NRC/i 10807 Trg ii =dmfad0la F4 bat csloop2to br zlohs682b2a[i1 on F5 imfadola70 br zlohs682b2a[2] off F6 ior zdihs682bla[1] off Mrf thi8d trip F7 imf fw05b 100 300 75 Trg 15 NRC/bvfd F8 Trg 15 =batNRC/110807-i F9 th22 100 5:00 50 Trg 1 modesw FlO dmfedl2a Trg 1 =batNRC/110807-4 Fl 1 mrf edO9 norm br zdihs858a[1j close F12 mrf rpO2 reset Trg 17 NRC/rcic Si mrfsl0l align Imfrc09 (e17 1:00) 100 1:00 S2 imf adOla 100 Trg 10 NRC/rfptamaual Trg 10 = dmf fw30a br ypovfcvo52l fail_power_now br zlohs052la[21 on Trg 5NRC/fwheating zdihs052 la[ 1] .eq. 1 Trg 5 = bat NRC/i 10807-2 br ypomtrsbgtrrh fail_control_power Batch 110807-4 Batch 110807-2 Imf slOla and slOib br ypovfcv052l fail_power_now Imfedllaandedllb(el 4:00) br zlohsO52la[2] on Imf edl2a and edl3a (el 1:00)
Irnf edi ic and edi id (el 5:00) Batch 110807-2 bmffwl9 (el 0)100 3:00 mrfthl8d close Imfth2l (el 8:00) .25 1200 dor zlohs682b2a[ 1]
Imfrd0la(el 3:00) dor zlohs682b2a[2]
Manually Enter FW3OA Scenario 7 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 93 Simulator Setup Load Batch RestorePref NRC/1l0807 Simulator Setup manual Tag out Core Spray Loop 2 Simulator Setup manual F3 and F4 Simulator Setup Verify file loaded RCP required (95% 100% with flow) and RCP for Urgent Load Reduction Provide marked up copy of 2-GOI-100-12
Scenario 7 Page 11 of 35 Simulator Event Guide:
Event 1 Normal: Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 SRO Directs Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 section 5.2 BOP Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 5.2 Standby Gas Treatment System Manual Initiation
[1] VERIFY the following requirements are satisfied:
- SGT Train A(B)(C) in standby readiness.
- Main Stack Radiation Monitoring in Service.
[2] REVIEW the Precautions and Limitations in Section 3.0.
[3] VERIFY suction path is aligned to SGT System as follows:
[3.2] IF alignment to Reactor Zone Ventilation suction path is desired, THEN VERIFY OPEN the following dampers for the desired unit(s) to be aligned.
[4] START SGT FAN C as follows:
[4.2] IF starting SGT FAN C from Panel 2-9-25, THEN PLACE SGTS FAN C, 0-HS-65-69A12 in START.
[5] CHECK SGT TRAIN C INLET DAMPER as follows:
[5.3] IF SGT FAN C was started, THEN CHECK OPEN SGTS TRAIN C INLET DAMPER, 0-HS-65-5 1A indicates OPEN on Panel 2-9-25.
[6] CHECK SGT TRAIN C RH CONTROL HTR as follows:
[6.2] IF SGT FAN C was started, THEN CHECK ENERGIZED SGTS TRAIN C RH CONTROL HTR, 0-HS-65-60 on Panel 2-9-25.
[7] RECORD start time and filter bank differential pressure for SGT Train as follows:
[7.2] IF SGT FAN C was started, THEN RECORD start time and FILTER BANK DIFFERENTIAL PRESSURE, 0-PDI-65-53 on Panel 2-9-25, in the Narrative Log.
[8] DISPATCH Operator to the Standby Gas Treatment building as soon as time allows to check for abnormal conditions (i.e. belt tightness, rubbing or vibration noises).
[9] MONITOR Standby Gas Treatment Train operation. REFER TO Section 6.0.
BOP Reports failure of RH Heater
Scenario 7 Page 12 of 35 Simulator Event Guide:
Event 1 Normal: Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 section 5.2 SRO SRO Evaluate Technical Specification 3.6.4.3 LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
Condition A One SOT subsystem inoperable Required Action A. 1 Restore SGT subsystem to OPERABLE status Completion Time 7 Days
Scenario 7
- Page 13of35-Simulator Event Guide:
Event 2 Reactivity: Raise Reactor Power with flow SRO Direct Power Increase JAW RCP SRO Notify ODS of power increase Direct Power increase using Recirc Flow per 2-GOI-100-12
[211 WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes.
REFER TO 2-SR-3.3.5(A) and 2-01-68.
ATC Raise Power w/Recirc lAW 2-01-68, section 6.2
[1J IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;
- Raise Recirc Pump 2A using, RAISE SLOW (MEDIUM),
2-HS 1 5A( 1 5B).
AND/OR
- Raise Recirc Pump 2B using, RAISE SLOW (MEDIUM),
2-HS I 6A( 1 6B).
[21 WHEN desired to control Recirc Pumps 2A and/or 2B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 2A & 2B using the following push buttons as required:
RAISE SLOW, 2-HS-96-3 1 RAISE MEDIUM, 2-HS-96-32 NRC NRC When satisfied with Reactivity Marnpülation ADS SRV Pails Open requiring power to be lowered to less than 90%
Driver Drivd At lead Iloor instructor direction , for failure of ADS SRV 1$
Scenario 7 Page 14of35 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open BOP Report alarm MAIN STEAM RELIEF VALVE OPEN (2-9-3C Window 25)
A. CHECK MSRV DISCHARGE TAiLPIPE TEMPERATURE, 2-TR-1-l, on Panel 2-9-47 and SRV Tailpipe Flow Monitor on Panel 2-9-3 for raised temperature and flow indications.
B. REFER TO 2-AOl-i-i.
SRO Enters 2-AOl- i-i BOP 4.1 Immediate Action
[ii IDENTIFY stuck open relief valve by OBSERVING the following:
- MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR-1-l on Panel 2-9-47.
BOP Identifies ADS SRV 1-5 open ATC [21 IF relief valve transient occurred while operating above 90% power, THEN PERFORM the following (Otherwise N/A):
[2.1] INITIATE a load reduction to 90% power with recirc flow.
ATC Lowers reactor power to 90% with recirc flow.
BOP [3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch several times as required:
- CLOSE to OPEN to CLOSE positions 4.2 Subsequent Action 4.2.2 Attempt to close valve from Panel 9-3:
[11 PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the OFF position.
[2] PLACE the SRV TAILPIPE FLOW MONITOR POWER SWITCH in the ON position.
[3] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (Otherwise N/A)
[4] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT, 2-XS-l-202 in INHIBIT:
CT#4 Observe and report when 2-XS-l-202 is placed in Inhibit, ADS SRV 1-5 closes.
Scenario 7 Page 15 of 35 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open BOP [51 IF relief valve closes, THEN OPEN breaker or PULL fuses as necessary using V Attachment 1 (Unit 2 SRV Solenoid Power Breaker/Fuse Table).
[6] PLACE MSRV AUTO ACTUATION LOGIC INHIBIT 2-XS-1-202, in AUTO.
Operator Does NOT perform step 6 until Breaker opened or fuses pulled iiriver V If MSRV AUTO Actuation Logic lhhibit Switch is returned to Auto prior to pulling fuses insert imf adOla (shift F21 V . V V SRO Evaluate Tech Spec 3.5.1 Condition E One ADS valve inoperable Required Action E. 1 Restore ADS Valve to OPERABLE status Completion Time 14 Days AND Condition F One ADS valve inoperable AND Condition A entered Required Action F. 1 Restore ADS Valve to OPERABLE status Completion Time 72 Hours OR Required Action F.2 Restore low pressure ECCS spray subsystem to OPERABLE status Completion Time 72 Hours BOP Directs AUO to Remove Power from SRV 1-5 REMOVE the power from 2-PCV- 1-5 by performing one of the following:
A. OPEN the following breakers (Preferred method)
- 2C 250V RMOV, compartment 8A
- Battery Board 1, breaker 727 OR B. In Panel 2-25-32 PULL the following fuses as necessary
Scenario 7 Page-16 of 35 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open Driver Driver When directed to remove power from SRV 1-5, insert irf adOla OUT in two nirnutes SRO May direct Suppression Pool Cooling placed in service lAW 2-01-74 BOP If Directed places Suppression Pool Cooling in Service Loop 1
[6] VERIFY at least one RHRSW Pump is operating on each EECW Header.
[7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows:
[7.11 START an RHRSW Pump to supply RHR Heat Exchanger A(C).
[7.21 ESTABLISH RHRSW flow by performing one the following:
[7.2.21 THROTTLE OPEN RHR HX 2A(2C) RHRSW OUTLET VLV, 2-FCV-23-34(40), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 2-FI-23-36(42), RHR HTX 2A(2C) RHRSW FLOW.
[7.31 VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 2-FCV-74-53.
[7.41 VERIFY CLOSED RHR SYS I SUPPR POOL CLG/TEST VLV, 2-FCV-74-59.
[7.51 VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-58.
[7.61 VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60.
[7.7] VERIFY OPEN RHR SYS I SUPPR CHBRIPOOL ISOL VLV, 2-FCV-74-57.
[7.9] START RHR PUMP A(C) using 2-HS-74-5A(16A).
[7.101 THROTTLE RHR SYS I SIJPPR POOL CLG/TEST VLV, 2-FCV-74-59, to maintain RHR flow within limits, as indicated on RHR SYS I CTMT SPRAY FLOW, 2-FI-74-56.
RHR Pumps in 1 2 Operation Loop flow 7000 to <13.000 gpm &
i0000 gpm & Blue Blue light light flumthated iltuminatd
[7.11] IF desired to raise Suppression Pool Cooling flow and only one Loop I pump is in service, THEN PLACE the second Loop I RHR Pump and Heat Exchanger in service_by_REPERFORMING_Step_8.5[7]_for the_second_pump.
Scenario 7 PageEl7of35 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open BOP If Directed places Suppression Pool Cooling in Service Loop 2
[10] PLACE RHR Pump and Heat Exchanger B(D) in service as follows:
[10.1] START_an_RHRSW_Pump_to_supply_RHR_Heat Exchanger B(D).
[10.2] ESTABLISH RHRSW flow by performing one the following:
[10.2.2] THROTTLE OPEN RHR HX 2B(2D) RHRSW OUTLET VLV, 2-FCV- 23-46(52), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 2-FI-23-48(54), RHR HTX 2B(2D) RHRSW FLOW.
[10.3] VERIFY CLOSED RHR SYS II LPCI 1NBD INJECT VALVE, 2-FCV-74-67.
[10.4] VERIFY CLOSED RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73.
[10.5] VERIFY CLOSED RHR SYS II SUPPR CHBR SPRAY VALVE, 2-FCV-74-72.
[10.6] VERIFY CLOSED RHR SYS II DW SPRAY OUTBD VLV, 2-FCV-74-74.
[10.71 VERIFY OPEN RHR SYS II SUPPR CHBR/POOL ISOL VLV, 2-FCV-74-7 1.
[10.9] START RHR PUMP B(D) using 2-HS-74-28A(39A).
[10.10] THROTTLE RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73, to maintain RHR flow within limits, as indicated on RHR SYS II CTMT SPRAY FLOW, 2-FI-74-70.
RHR Pumps in 1 2 0peraton Leop flow 7000 to <13000 pm &
10000 gpm & Bue 3lue light light ilftjmnated illuminated
[10.11] IF desired to raise Suppression Pool Cooling flow and only one Loop II pump is in service, THEN PLACE the second Loop II RHR Pump and Heat Exchanger in service by REPERFORMING Step 8.5[10] for the second pump.
Driver Dnvei At lead floor instructor direction for trip of 2B4 VFD Cooling Pump
Scenario 7 Page 18 of 35 Simulator Event Guide:
Event 4 Component: VFD Cooling Water Pump 2-B-i Failure ATC Reports the following annunciators 4B-12, 28 and 32 RECIRC DRIVE 2B COOLANT FLOW LOW, RECIRC DRIVE 2B PROCESS ALARM, and RECIRC DRIVE 2B DRIVE ALARM ATC Reports the 2-B-i VFD Cooling Water Pump for the B Recirc Pump, has tripped.
ATC Reports Standby Recirc Drive Cooling Water Pump2-B-2, failed to auto start.
ATC RECIRC DRVIE 2B COOLANT FLOW LOW STARTS RECIRC DRIVE cooling water pump and DISPATCHES personnel to the RECIRC DRiVE, to check the operation of the Recirc Drive cooling water system.
SRO Concurs with start of Standby VFD Pump.
BOP RECIRC DRIVE 2B DRIVE ALARM A. REFER TO ICS Group Display GD @VFDBDA and determine cause of alarm.
B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system.
C. IF the problem is conductivity in the cooling water system, THEN VERIFY demineralizer is in service.
D. IF a problem with power supplies is indicated, THEN VERIFY all the low voltage supply breakers are CLOSED/ON.
E. For all other alarms, or any problems encountered CONTACT system engineering.
Crew Verifies Standby pump started by pulling up ICS displays.
BOP Dispatches personnel to VFD.
IRJVER Wait 4. minutes after dispatched, THEN report tripped VFD Pump is hot to touchy internal bkr closed. 480 volt bkr tripped (480 V SI) BD 2A-.5D).
DRIVER Upon Lead examiner direction for Loss of Feedwater Heating
Scenario 7 Page 1of5 Simulator Event Guide:
Event 5 Component: Loss of Feedwater Heating and 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate DRIVER When directed by NRC insert for Loss of Feedwater Heating and 2-FCV-521, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate.
ATCJBOP Announces BYPASS VALVE TO CONDENSER NOT CLOSED and refers to 2-ARP-9-6A, window 18.
A. CHECK heater high or low level or moisture separator high or low level alarm window illuminated on Panel 2-9-6 or 2-9-7 to identify which bypass valve is opening.
B. CHECK ICS to determine which bypass valve is open.
C. DISPATCH personnel to check which valves light is extinguished on junction box.
DRIVER Acknowledge dispatch, wait 1-2 minutes and report 2-LCV-6-22B light is out on junction box 34-21:
ATCIBOP Announces HEATER B2 LEVEL HIGH and refers to 2-ARP-9-6A window 9.
A. CHECK the following indications:
- Condensate flow recorder 2-29, Panel 2-9-6. Rising flow is a possible indication of a tube leak.
- Heater B2 shell pressure, 2-PI-5-22 and drain cooler B5 flow, 2-FI-6-34, Panel 2-9-6. High or rising shell pressure or drain cooler flow is possible indication of a tube Leak.
B. CHECK drain valve 2-FCV-6-95 open.
C. CHECK level on ICS screen, FEEDWATER HEATER LEVEL (FWHL).
- IF the 2B2 heater indicates HIGH (Yellow), THEN VERIFY proper operation of the Drain and Dump Valves.
- DISPATCH personnel to local Panel 2-LPNL-925-562C to VERIFY and MANUALLY control the level.
D. IF a valid HIGH HIGH level is received, THEN GO TO 2-AOI-6-IA or 2-AOI-6-1C.
ATCIBOP Checks condensate flow recorder, Heater B2 shell pressure and Drain Cooler B5 flow for indications of a tube leak Checks drain valve 2-FCV-6-95 open Checks 2B2 Heater level on ICS and dispatches personnel to verify and manually control level DRIVER Acknowledge order to verify and manually control level on B2 Heater. Wait 6 minutes and report unable to take manual control of B2 Heater
Scenario 7
--Page2Qf35*
Simulator Event Guide:
Event 5 Component: Loss of Feedwater Heating and 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate ATC/BOP Announces B 1 and B2 High Pressure Heater Extraction Isolation SRO Directs crew to enter 2-AOI-6-1A or 2-AOI-6-1C ATC/BOP 2-AOI-6-1A High Pressure Feedwater Heater StringIExtraction Steam Isolation 4.1 Immediate Actions
[1] REDUCE Core Thermal Power to? 5% below initial power level to maintain thermal margin.
4.2 Subsequent Actions
[1] REFER TO 2-01-6 for turbine/heater load restrictions.
[2] REQUEST Reactor Engineer EVALUATE and ADJUST thermal limits, as required.
[3] ADJUST reactor power and flow as directed by Reactor Engineer/Unit Supervisor to stay within required thermal and feedwater temperature limits. REFER TO 2-GOT- 100-12 or 2-GOI-100-12A for the power reduction.
[4] ISOLATE heater drain flow from the feedwater heater string that isolated by closing the appropriate FEEDWATER HEATER B-2 DRAIN TO HTR B-3, 2-FCV-6-95.
[5] IF a tube leak is indicated, THEN PERFORM manual actions of Attachment 1 for affected heaters.
[6] VERIFY automatic actions occur. REFER TO Attachment 1.
[7] MONITOR TURB THRUST BEARING TEMPERATURE, 2-TR-47-23, for rises in metal temperature and possible active/passive plate reversal.
[8] DETERMINE cause which required heater isolation and PERFORM necessary corrective action.
Scenario 7 Page 21 of 35 Simulator Event Guide:
Event 5 Component: Loss of Feedwater Heating and 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate ATC/BOP 2-AOI-6-1A High Pressure Feedwater Heater String[Extraction Steam Isolation (continued) 4.2 Subsequent Actions (continued)
[91 WHEN the condition which required heater isolation is no longer required, THEN RESTORE affected heater._REFER TO 2-01-6.
ATC Lower Reactor Power greater than 5% below initial power level using Recirc Pump flow adjustments BOP Refers to 2-01-6 for turbine/heater load restrictions Contacts Reactor Engineer to evaluate and adjust Thermal Limits, if needed Isolates heater drain flow B2 Heater Drain to B3 Heater by shutting 2-FCV-6-95 SRO Directs isolating FW to B HP heater string based on indications of tube leak by performing manual actions of Attachment 1 and verifying automatic actions occur 2-AOI-6-1A Attachment 1 B 1 or B2 The following valves must be manually closed:
2-FCV-3-31, HP HTR 2B2 FW INLET ISOL VALVE 2-FCV-3-76, HP HTR 2B1 FW OUTLET ISOL VALVE The following valves AUTO Isolate 2-FCV-5-9, NP HEATER 2B1 EXTR ISOL VLV 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV 2-FCV-6-74, MOISTURE SEP LC RES B 1 ISOL VLV 2-FCV-6-172, MOISTURE SEP LC RES B2 ISOL VLV Directs power reduction to 920 MWe (79%) power (Power Reduction with RCP flow or Control Rods) per 2-01-6, Illustration 1 2-01-6 Illustration 1 HEATERS OUT (Tube and Shell Side) **
One HP string 920 MWe (79%)
One LP string 920 MWe (79%)
One HP and LP string 920 MWe (79%)
Enters 2-GOl- 100-12, Power Maneuvering Notifies Rx Eng. And ODS of Feedwater Heater isolation and power reduction
Scenario 7
- Page 22 of 35 Simulator Event Guide:
Event 5 Component: Loss of Feedwater Heating and 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate BOP 2-AOI-6-1A Attachment 1 Closes the following Feedwater Valves Manually 2-FCV-3-31, HP HTR 2B2 FW INLET ISOL VALVE 2-FCV-3-76, HP HTR 2B1 FW OUTLET ISOL VALVE Verifies the following valves close automatically 2-FCV-5-9, HP BEATER 2B 1 EXTR ISOL VLV 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV 2-FCV-6-74, MOISTURE SEP LC RES B 1 ISOL VLV 2-FCV-6-172, MOISTURE SEP LC RES B2 ISOL VLV Takes action to manually shut 2-FCV-5-21 upon determining the valve did not automatically close and reports to SRO Recognizes HTR level lowers as a result of isolating the Condensate side of 2B HP HTR string (i.e. tube leak) and reports to crew DRIVER After HS for 2-FCV-5-21 taken to closed, verify Trigger 5 goes active As Reactor Engineer when contacted direct crew to follow the guidance of urgent load reduction and 2-01-6 ATC Lower Reactor Power to <920 MWe/<79% power by lowering recirc flow.
SRO Direct ATC to insert the first group of control rods on the Emergency Shove Sheet per Reactor Engineer recommendation.
ATC Inserts the first group of rods on the Emergency Shove Sheet using a peer check as directed by Rx Engineer & Unit Supervisor
Scenario 7 Page 23 of 35 Simulator Event Guide:
Event 6: Feedwater Pump 3A Governor Drifts Up DRIVER When NRC directs, insert imf fw30a check current setting of fw30a and then ramp to 100 over 20 minutes for Feedwater Pump Governor Failure. When operator takes the RFPT Governor to manual the malfunction is automatically deleted, therefore, IF the operator pulls the Governor control knob back out, the malfunction rnustbe manually reinserted and deleted when the operator returns the Governor control knob back down to force th operatortq control level manually.
ATC Report Rising Reactor Water Level and RFPT is not responding.
SRO Direct manual control of operating RFPT and Enter 2-AOI 1.
4.2 Subsequent Actions
[1] VERIFY applicable automatic actions.
6.0 HIGH REACTOR WATER LEVEL
[2] IF Feedwater Control System has failed, THEN PERFORM the following:
[2.11 PLACE individual RFPT Speed Control Raise/Lower switches in MANUAL GOVERNOR (depressed position with amber light illuminated).
[2.21 ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain level.
[6] IF level continues to rise, THEN TRIP a RFP, as necessary.
[81 IF RFPs are in manual control, THEN LOWER speed of operating RFPs.
[9] EXPECT a possible Reactor power rise due to a rise in moderation.
[10] IF unit remains on-line, THEN PERFORM the following:
- RETURN Reactor water level to normal operating level of 33 (normal range).
- REQUEST Nuclear Engineer check core limits.
ATC Take MANUAL GOVERNOR control of RFPT and maintain Reactor Water Level Manually in the Normal Level Band. Operator may attempt to control RFPT with PDS.
PDS will not respond.
DRIVER If a scram is mserted or at NRC direction initiate E for LOCA and make Earthquake calls
Scenario 7
-Page-24of 35 Simulator Event Guide:
Event 7 Major: Earthquake Driver Driver Report confirmed earthquake Unit 1 is handling O-AOI-100-5, Earthquake ATC/BOP Reports rising Drywell pressure SRO Establishes Drywell Pressure to insert a Reactor Scram ATC Insert Manual SCRAM when directed SRO Enters 2-AOI-100-l, EOI-1 and EOI-2 on High Drywell Pressure ATC 2-AOI-100-1
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS-99-5A/S3B, on Panel 2-9-5
[2] IF scram is due to a loss of RPS, THEN (Otherwise N/A)
[31 REFUEL MODE ONE ROD PERMISSIVE light check:
[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-Sl, in REFUEL.
[3.21 CHECK REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-.46, illuminates.
[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46, is not illuminated, THEN CHECK all control rod positions at Full-In Overtravel, or Full-In. (Otherwise N/A)
[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN position.
Driver Driver Ensure trIgger 1 goes active onMODESWITCH
Scenario 7 Page 25 of 35 Simulator Event Guide:
Event 7 Major: Earthquake Feedwater Line Break Driver Driver Report confirmed earthquake Unit 1 is handling O-AOI-100-5, Earthquake ATC Determines Feedwater Leak on the A Feedwater Line due to Feedwater Line A Flow high and Feedwater line B flow lowering to 0 and Reactor Feed Pump Flows Increasing with a Lowering Reactor Water Level.
SRO Directs Reactor Feed Pumps to be tripped, Reactor Feed Pump Discharge Valves shut, and Condensate Booster Pumps then Condensate Pumps secured. (Isolate and stop leak) Also directs HPCI locked out due to Feedwater Line Break on the A line.
ATC Trips Reactor Feed Pumps and shuts Reactor Feed Pump Discharge Valves.
Secures Condensate Booster Pumps then Condensate Pumps.
BOP Trips HPCI if running and places HPCI Aux Oil Pump in PTL when HPCI speed lowers to 0 rpm.
SRO Enters EOI-1 on Low Reactor Water Level and High Drywell Pressure RC/Q Monitor and Control Reactor Power.
Directs Exit of EOI- 1 RC/Q Leg, after ATC reports All Rods In on Scram Report.
RC/P Monitor and Control RPV Pressure.
Answers NO to: Is any MSRV cycling?
Directs BOP to maintain RPV Pressure 500 -1000 psig using Appendix hA..
RC/L Monitor and Control RPV Water Level.
Verify as Required:
- PCIS Isolations (Groups 1, 2 and 3)
- RCIC Directs level band of +2 to +51 inches, with Appendix 5C, 5B and/or 7B.
Scenario 7 Page26 of 35 Simulator Event Guide:
Event 7 Major: Earthquake Feedwater Line Break ATC/BOP Pressure Control JAW Appendix hA, RPV Pressure Control SRVs I. IF Drywell Control Air is NOT available, THEN:
EXECUTE EOI Appendix 8G, CROSST1E CAD TO DRY WELL CONTROL AIR,_CONCURRENTLY with this procedure.
- 2. IF Suppression Pool level is at or below 5.5 ft, THEN:
CLOSE_MSRVs_and_CONTROL_RPV_pressure_using_other_options.
- a. 2-PCV-1-179 MN STM LINE A RELIEF VALVE
- b. 2-PCV-i-180 MN STM LINE D RELIEF VALVE.
- c. 2-PCV-1-4 MN STM LINE A RELIEF VALVE
- d. 2-PCV- 1-31 MN STM LINE C RELIEF VALVE
- e. 2-PCV-1-23 MN STM LINE B RELIEF VALVE
- f. 2-PCV-1-42 MN STM LINE D RELIEF VALVE
- g. 2-PCV-1-30 MN STM LINE C RELIEF VALVE
- h. 2-PCV-1-19 MN STM LINE B RELIEF VALVE.
- i. 2-PCV-i-S MN STM LINE A RELIEF VALVE.
- j. 2-PCV-1-41 MN STM LINE D RELIEF VALVE
- k. 2-PCV-1-22 MN STM LINE B RELIEF VALVE
- 1. 2-PCV-1-18 MN STM LINE B RELIEF VALVE
- m. 2-PCV-1-34 MN STM LINE C RELIEF VALVE
Scenario 7
- Page27 of 35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling NOTE NOTE When RCIC is started,. a break will occur on the RCTC Steam Line prior to FCV 71 ..8.
ATC/BOP Reports alarm RCIC STEAM LINE LEAK DETECTION TEMP HIGH and rising temperatures in RCIC SRO Directs RCIC Isolation verified ATC/BOP Verifies RCIC automatically isolates.
Attempt to align SLC per Appendix 7B. Recognize and report trip of both SLC Pumps.
Report trip of CRD Pump 2A and inability to align CRD Pump lB due to 2-85-8A will not open.
CREW Recognizes loss of all High Pressure Injection sources ATC/BOP Report loss of 480 V RMOV Bd 2A I RMOV Bd 2E / RMOV Bd 2D CREW Recognizes loss of all Injection sources SRO EOI-1 (cont)
Answers NO to: Can water level be Restored and Maintained above (+) 2 inches?
Maintain RPV Water Level above (-) 162 inches.
CT#3 Directs ADS inhibited when RPV Water Level drops below -120 inches.
Augments RPV Water Level Control with SLC, per Appendix 7B.
Answers NO to: Can RPV Water Level be maintained above (-) 162 inches?
Exits RC/L and enters C-i, Alternate Level Control.
CT#3 ATC/BOP Inhibits ADS
Scenario 7 Page 28 of 35-Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling SRO Enters C-i, Alternate Level Control Verifies ADS Inhibited Directs lineup of Injection Systems Irrespective of Pump NPSH and Vortex limits (LPCI and CS) per Appendix 6B and 6D Answers NO to can 2 or more CNDS, LPCI or CS Injection Subsystems be aligned with pumps running When RPV Water Level drops to 162 inches, Then continues Answers NO to is any CNDS, LPCI or CS Injection Subsystem aligned with at least one pump running Before RPV Water Level drops to -180 inches continue Answers NO to are pumps running that can restore and maintain RPV Water Level above -180 inches after Emergency Depressurization When RPV Water Level drops to -180 inches continue Answers NO to is any CNDS Injection Source aligned with at least one pump running Steam Cooling is Required DriveE Driver Once steam. cooling is entered insert jj.Q (dmf edi2a). Then close normal feede breaker to RMOV Bd 2A insert Fil (mifed09 norm). Notify crew that RMQV Ed 2A is restored. Then insertZ (mrf rpO2 reset) to reset RIS B.
NOTE NOTE Restoration of RMOV Bd 2A makes Core Spray Loop I available SRO C-i, Alternate Level Control (Cont.)
If any Injection Source aligned with at least one pump running and Reactor Level is < -180 inches continue CT#1 Emergency Depressurization is required Enters C-2
Scenario 7 Page 29 of 35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling SRO C-i, Alternate Level Control (Cont.)
CT#i If RPV Water Level drops to -200 inches continue Emergency Depressurization is required Enters C-2 Directs maximizing RPV Injection from all available sources irrespective of pump NPSH and Vortex Limits Enters C-2, Emergency RPV Depressurization Answers Yes to will the Reactor remain subcritical without Boron under all conditions Answers Yes to is Drywell Pressure above 2.4 psig Does not prevent Injection from any Core Spray or LPCI pumps because they are all needed to assure adequate core cooling Answers Yes to is Suppression Pool Level above 5.5 feet CT#i Directs opening of all ADS Valves Answers NO to can 6 ADS Valves be opened Open additional MSRVs as necessary to establish 6 MSRVs Open Answers YES to are at least 4 MSRVs Open CT#1 BOP/ATC Open 5 ADS Valves and one additional SRV due to Inoperable ADS SRV CT#2 BOP/ATC With RPV pressure below the Shutoff Head of the available Low Pressure system(s),
operate available Low Pressure system(s) to restore RPV water level above TAF.
Scenario 7 Page 3&of3&
Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling BOP/ATC Appendix 6D, Loop I Core Spray
- 1. VERIFY OPEN the following valves:
- 2-FCV-75-2, CORE SPRAY PUMP 2A SUPPR POOL SUCT VLV
- 2-FCV-75-1 1, CORE SPRAY PUMP 2C SUPPR POOL SUCT VLV
- 2-FCV-75-23, CORE SPRAY SYS I OUTBD INJECT VALVE.
- 2. VERIFY CLOSED 2-FCV-75-22, CORE SPRAY SYS I TEST VALVE.
- 3. VERIFY CS Pump 2A and/or 2C running.
- 4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 2-FCV-75-25, CORE SPRAY SYS I INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.
SRO C-i, Alternate Level Control (Cont.)
Answers Yes to can RPV Water Level be restored and maintained above -180 inches Exits C-i and enters EOI- 1, RPV Control at step RCIL- 1 SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below i6OF using available Drywell cooling Answers No to can Drywell Temperature be maintained below 160F Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor (this will already be complete at this time)
Before Drywell Temperature rises to 280F continue Answers Yes to is Suppression Pool Level below 18 Feet Answers Yes to are Drywell Temperatures and Pressures within the safe area of curve 5 Directs Shutdown of Recirc Pumps and Drywell Blowers (should leave Drywell Blowers running due to being unable to spray because adequate core cooling is not assured)
Scenario 7 Page 31 of 35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling SRO Enters EOI-2 on High Drywell Pressure (cont)
PC/P Monitor and control Primary Containment pressure below 2.4 psig using the Vent System (Appendix 12) as necessary Direct Appendix 12 ATC/BOP Vent Containment JAW Appendix 12 VERIFY at least one SGTS train in service.
- 2. VERIFY CLOSED the following valves (Panel 2-9-3 or Panel 2-9-54):
2-FCV-64-31, DRYWELL INBOARD ISOLATION VLV, 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE.
Steps 3, 4, 5 and 6 are If! Then steps that do not apply
- 7. CONTINUE in this procedure at:
Step 8 to vent the Suppression Chamber through 2-FCV-84-19, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.
- 8. VENT the Suppression Chamber using 2-FIC-84-19, PATH B VENT FLOW CONT, as follows:
- a. PLACE keylock switch 2-HS-84-35, DW/SUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).
- b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).
- c. PLACE 2-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).
- d. PLACE keylock switch 2-HS-84-19, 2-FCV-84-19 CONTROL, in OPEN (Panel 2-9-5 5).
- e. VERIFY 2-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.
- f. CONTINUE in this procedure at step 12.
Scenario 7
- Page32of35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling BOP Vents Primary Containment JAW Appendix 12
- 9. VENT the Suppression Chamber using 2-FIC-84-20, PATH A VENT FLOW CONT, as follows:
- a. VERIFY OPEN 2-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 2-9-3).
- b. PLACE keylock switch 2-HS-84-36, SUPPR CHBRJDW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).
- c. VERIFY OPEN 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV (Panel 2-9-54).
- d. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).
- e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-55).
- f. VERIFY 2-FJC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.
- g. CONTINUE in this procedure at step 12.
- 12. ADJUST 2-FJC 19, PATH B VENT FLOW CONT, or 2-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:
Stable flow as indicated on controller, AND 2-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:
iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 iiCils AND 0-SJ-4.8.B.1.a.1 release fraction of 1.
DRIVER Acknowledge Notification
Scenario 7 Page33of35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling SRO Enters EOI-2 on High Drywell Pressure (cont)
PC/P Monitor and control Primary Containment pressure below 2.4 psig using the Vent System (Appendix 12) as necessary Direct Appendix 12 Answers No to can Primary Containment Pressure be maintained below 2.4 psig Before Suppression Chamber Pressure rises to 12 psig Initiate Suppression Chamber Sprays using only those pumps not required for Adequate Core Cooling Directs Drywell Spray ATC/BOP Initiate Suppression Chamber Sprays per Appendix 17C
- 1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.
- 2. IF Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
Scenario 7 Page 34 of 35 Simulator Event Guide:
Event 8 Major: Loss of all injection Steam Cooling ATC/BOP 5. INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN...PLACE keylock switch 2-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRiDE.
- d. IF 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN.. VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT VALVE.
- e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spray.
- g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
- h. IF RI-JR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN...CONTINUE in this procedure at Step 5.k.
- i. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
- j. RAISE System flow by placing the second RHR System 1(11) pump in service as necessary.
ATC/BOP 1. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
SRO The Emergency classification is 1.1 -G 1
Scenario 7
- Page 35 of 35 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
Core Spray Loop 2 is out of service and tagged out, Technical Specifications have been addressed Operations/Maintenance for the Shift:
Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 section 5.2 Once completed raise reactor power to 100% with Recirculation.
Units 1 and 3 are at 100% power.
Unusual Conditions/Problem Areas:
None
Appendix D Scenario Outline Form ES-D-1 cility: Browns Ferry NPP Scenario No.: NRC 8 - Op-Test No.:
Examiners: Operators: SRO:
ATC:
BOP:
Initial Conditions: 3% power. 3-G0I-100-1A Section 5.4 Step 63.3 Turnover: Warm RFPT B lAW 3-01-3 Section 5.6 step 20. Continue to pull rods for Mode Change.
Event Maif. No. Event Type* Event Description No.
N-BOP 1 Warm RFPT B 3-01-3 5.6 Step 20 NSRO R-ATC 2 Raise power with Control Rods, Group 34 18-43 at 08 RSRO C-ATC Control Rod stuck 26-27, after drive water pressure raised rod 3 RDO6 TS-SRO drifts out ATC will insert control rod 1-BOP 4 OGO5a Hydrogen Water Chemistry Malfunction TS.SRO C-ALL 5 RPO1A Loss of RPS A TS-SRO C-ATC 6 FW15C RFPT C Trip C-SRO 7 EDO1 M-ALL Loss of Offsite Power, LOCA, ATWS 8 C Shutdown Boards B, C and D fail to energize 9 DGO3a C DG A fails to tie to SD BD (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or
Appendix D Scenario Outline Form ES-D-1 Critical Tasks Two CT#1- With a loss of all Off Site Power and NO Diesel Generators powering their respective Shutdown Boards ties Diesel Generator A to 4 KV SD BD A.
- 1. Safety Significance:
Provides Power for Control Rod Insertion and ECCS Systems
- 2. Cues:
Procedural compliance 4 KV Shutdown Board Energized
- 3. Measured by:
Observation RO closes in DG Supply Breaker to 4 KV SD BD A AND Observation 4 KV SD BD A indicates energized
- 4. Feedback:
Powerto4KVSDBDA Power to 4 KV RMOV BD A CT#2 -With a reactor scram required and the reactor not shutdown, initiate action to reduce power by inserting control rods.
- 1. Safety Significance:
Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.
- 2. Cues:
Procedural compliance CRD Pump B operating
- 3. Measured by:
Observation Control Rod insertion commenced in accordance EOI Appendixes.
- 4. Feedback:
Reactor Power trend.
Control Rod indications.
Appendix D Scenario Outline Form ES-D-1 VENTS
- 2. ATC Continues Power ascension for Mode Change
- 3. Control Rod 26-27 will not withdraw with normal drive water pressure from position 8, The ATC will take action JAW 3-01-85 for control difficult to withdraw. When drive water pressure is increased rod 26-27 will withdraw and continue out. SRO enters AOl for rod drift out. The ATC will insert control rod 26-27, when drive in is released the control rod will drift out, the ATC will have to maintain drive in until the control rod is scrammed. The SRO will declare the Control Rod Inoperable Technical Specification 3.1.3 condition C.
- 4. The Hydrogen Water Injection system will malfunction resulting in high hydrogen concentration in Off Gas. The BOP Operator will respond JAW with ARPs and 3-AOI-66-1 and shutdown the Hydrogen Water Chemistry System. The SRO will evaluate TRM 3.7.2 and enter Condition A.
- 5. Loss of RPS A the SRO will direct entry into 3-A0I-99-1, 3-AOI-70-1 and 3-A0I-64-2D. The crew will restore RPS, PCIS, RBCCW and other systems lAW 3-01-99. The SRO will evaluate TRM 3.4.1 and direct Chemistry to sample in order to satisfy TSR 3.4.1.1.
- 6. RFPT C Trips, ATC increases the speed of RFPT B to maintain level in the normal level band. SRO will enter 3-AOI-3-1.
- 7. The crew will respond to a LOOP, LOCA and ATWS; the SRO will direct entry to 0-AOI-57-1A, EOI-l, EOI-C-5, and E0I-2. The crew will discover that HPCI has failed to operate. RCIC is available for level control. SRVs for pressure control. RHR Pump A is available to spray the Drywell. Crew will call for LOOP action and for RPS to be reset.
- 8. Shutdown Boards B, C and D fail to energize
- 9. DO A fails to tie to Shutdown board A, an operator manually closes in DO output breaker to energize SD BD A.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.
All Control Rods are inserted Reactor Level is maintained Drywell has been Sprayed
Appendix D Scenario Outline Form ES-D-1 3CENARIO REVIEW CHECKLIST SCENARIO NUMBER: 8 8 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 0 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 2 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix D Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER K/A RO SRO Warm RFPT B RO U-003-NO-23 259001A4.02 3.9 3.7 Raise Power with Control Rods RO U-085-NO-07 SRO S-000--AD-31 2.2.2 4.6 4.1 Control Rod difficult to Withdraw RO U-085-AB-7 201002A2.02 3.2 3.3 SRO S-085-AB-7 Hydrogen Water Chemistry Malfunction RO U-066-AL-1O 271000A1.13 3.2 3.7 SRO S-066-AB-1 Loss of RPS A RO U-099-AB-1 212000A2.02 3.7 3.9 SRO S-099-AB-1 RFPT C Trip RO U-003-AB-1 25900 1A2.01 3.7 3.7 SRO S-003-AB-1 LOOP RO U-57A-AB-1 295003AA1.03 4.4 4.4 RO U-082-AL-7 SRO S-57A-AB-1 SRO S-000-EM-i SRO T-000-EM-18 SRO S-000-AD-27
8 Page7of46 Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 3-01-3 Reactor Feedwater System Rev 82 3-GOI-100-12 Power Maneuvering Rev 35 3-01-85 Control Rod Drive System Rev 70 3-ARP-9-5B, W10 REACTOR PROT 120V PWR SYS ABNORMAL Rev 17 3-AOI-99-1 Loss of Power to One RPS Bus Rev 16 3-01-99 Reactor Protection System Rev 47 3-ARP-9-6C, W29 RFPT TRIPPED Rev 12 3-ARP-9-6C, W15 RFPT C ABNORMAL Rev 12 Power (161 and 500 KV)/Station 3-AOl 1 A Rev 79 Bckout 3-ARP-9-5A Control Rod Drift Rev 41 3-A0I-85-6 Rod Drift Out Rev 9 TS 3.1.3 Control Rod Operability Amd 212 TRM 3.4.1 Coolant Chemistry Rev 21 TRM 3.7.2 Airborne Effluents Rev 0 3-A0l-66-1 Off Gas H2 High Rev 6 3-AR P-9-53 Panel 9-53 Rev 24 3-AOI-3-1 Loss of Reactor Feedwater or Reactor Water Level Rev 9 3-AOl-64-2D Group 6 Ventilation System Isolation Rev 16 3-EOI-3-C-5 Level/Power Control Rev 9 3-EOl-1 RPV Control Rev 8 3-AOl-i 00-1 Reactor Scram Rev 53 3-E0l-App-2 Defeating ARI Logic Trips Rev 4 3-EOI-App-1 F Manual Scram Rev 2 3-E0l-App-1 1A Alternate RPV Pressure Control MSRVs Rev 2
8 Page8of 46 Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision Insert Control Rods using Reactor Manual Control 3-EOI-App-1 D Rev 2 3-EOl-2 Primary Containment Control Rev 8 3-EOI-App-17A RHR System Operation Suppression Pool Cooling Rev 5 3-EOI-App-17B RHR System Operation Drywell Sprays Rev 5 3-EOI-App-1 7C RHR System Operation Suppression Chamber Sprays Rev 6 3-EOI-App-1 2 Primary Containment Venting Rev 4 3-EOI-App-5C Injection System Lineup RCIC Rev 3 EPIP-1 Emergency Classification Rev 46
8 Page9of46 -.
Simulator Instructor 10203
- stuck rod 26-27 imf rd06r2627 imf rd04r2627 (e3 0)
- hwc malfunction imf ogo5a (e5 0) 85 br xa5553a[l0] (e5 0) alarm_on trg 16 nrc20110440 trg 16 = mmf og05a 100 36099
- rps A trip imf rp0la (elO 0) mrf rpO4 (eli 0) A irf rpO3 (eli 0) reset irf rpO9 (eli 0) reset
- FWP trip imf fwi5c (elS 0)
- LOOP LOCA ATWS/major bat atws95 mrf ed27a (ei9 0) close mrf edlO (e24 0) alt trg 21 = bat atws-l trg 22 = bat app0lf trg 23 = bat appO2 trg 25 = bat sdv imf th22 (e20 0) 100 imf th2l (e20 0) 0.1 imf edOl (e20 0) imf dgo3a (e20 0) imf edl5c (e20 0) imf dg06b (e20 0) imf edl5d (e20 0) imf hpO7 (e20 0) trg 26= bat eecw trg 27 = bat eecw-l trg 28 = bat ca trg 29= bat rpsreset
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8 Pal1 of 46 Simulator Event Guide:
Event 1 Normal: Warm RFPT B 3-01-3 5.6 Step 20 SRO Directs BOP to continue warming RFPT B per 3-01-3 Section 5.6, step 20 BOP Continues warming RFPT B per 3-01-3 Section 5.6, starting at step 20 5.6 Warming the Second and Third RFP/RFPT (continued)
[20] VERIFY RFPT Speed Control in MANUAL GOVERNOR.
[20.1] ADJUST RFPT 3B SPEED CONT RAISE/LOWER switch, 3-HS-46-9A, as necessary until RFPT speed is approximately 1000 rpm, as indicated on RFPT SPEED.
[21] PLACE RFPT 3B TURNING GEAR MOTOR, 3-HS-3-127A, in AUTO.
[22] DEPRESS RFPT 3B TRIP, 3-HS-3-1 51 A.
[22.1] VERIFY HP and LP Stop Valves close.
[23] VERIFY Turning Gear automatically engages or REP rolling on minimum flow.
NOTE Turning Gear Motor will lockout if Turning Gear does NOT engage within five seconds of reaching zero speed. Lockout can be reset by placing control switch to OFF and pulling out (at OFF).
[24] DEPRESS RFPT 3B TRIP RESET, 3-HS-150A.
[24.1] VERIFY the following:
. Blue light extinguishes.
CAUTION DO NOT RAISE RFP discharge pressure to greater than Reactor Pressure to prevent injection to the vessel.
. NOTE Normal operating range for RFP lube oil to bearings is I 10°F to 120°F. Illustration 7 has additional instruction for controlling Raw Cooling Water through RFP lube oil cooler.
8 Page42of46 Simulator Event Guide:
Event 1 Normal: Warm RFPT B 3-01-3 5.6 Step 20 (continued)
BOP [25] PLACE RFPT 3B START/LOCAL ENABLE, 3-HS-46-138A, in START.
[25.1] OBSERVE RFPT accelerates to approximately 600 rpm on RFPT 3B SPEED, 3-Sl-46-9A (Panel 3-9-6)
[26] IF the lube oil to bearings is NOT at least 110°F, THEN PERFORM the following:
[26.1] RAISE RFPT speed to 1100 rpm using RFPT 3A(3B)(3C) SPEED CONT RAISE/LOWER switch, 3-HS-46-9A.
. MAINTAIN this speed until lube oil to the bearings reaches 110°F.
[26.2] WHEN oil to the bearings reaches 110°F, THEN REDUCE RFPT speed using RFPT 3B SPEED CONT RAISE/LOWER switch, 3-HS-46-9A, until RFP 3A(3B)(3C) DISCH PRESS, 3-Pl-3-9A, is less than Reactor Pressure.
8 Page-l3 of 46 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 34 18-43 at 08 SRO Direct Power increase using Control Rods per 3-G0l-1 00-lA, Section 5.4 5.4 Withdrawal of Control Rods while in Mode 2
[67] CONTINUE to withdraw control rods to raise Reactor power to approximately 8%. (REFER TO 3-01-85 and 3-SR-3.1 .3.5(A))
ATC Raise Power with Control Rods per 3-01-85, Section 6.6 Group 34= 18-43, 42-43, 42-19, 18-19 from O8to 12 Group 35 = 26-35, 34-35, 34-27, 26-27 from 08 to 12 Group 36 = 02-35, 26-59, 34-59, 58-35, 58-27, 34-03, 26-03, 02-27 from 00 to 12 6.6.1 Initial Conditions Prior to Withdrawing Control Rods
[2] VERIFY the following prior to control rod movement:
- CRD POWER, 3-HS-85-46 in ON.
- Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).
6.6.2 Actions Required During and Following Control Rod Withdrawal
[4] OBSERVE the following during control rod repositioning:
- Control rod reed switch position indicators (four rod display) agree with the indication on the Full Core Display.
- Nuclear Instrumentation responds as control rods move through the core (This ensures control rod is following drive during Control Rod movement.)
[5] ATTEMPT to minimize automatic RBM Rod Block as follows:
- STOP Control Rod withdrawal (if possible) prior to reaching any RBM Rod Block using the RBM displays on Panel 3-9-5 and PERFORM Step 6.6.2[6].
[6] IF Control Rod movement was stopped to keep from exceeding a RBM setpoint or was caused by a RBM Rod Block, THEN PERFORM the following at the Unit Supervisors discretion to REINITIALIZE the RBM:
[6.1] PLACE CRD POWER, 3-HS-85-46 in the OFF position to deselect the Control Rod.
[6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position.
8 Pagel4of46 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 34 18-43 at 08 ATC 6.6.3 Control Rod Notch Withdrawal
[1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 3-XS-85-40.
[2] OBSERVE the following for the selected control rod:
- CRD ROD SELECT pushbutton is brightly ILLUMINATED
- White light on the Full Core Display ILLUMINATED
- Rod Out Permit light ILLUMINATED
[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.
[4] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.
[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.
[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:
[6.1] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.
[6.2] CHECK control rod coupled by observing the following:
- Four rod display digital readout and the full core Display, digital readout and background light remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator, 3-XA-55-5A, Window 14, does not alarm.
[6.3] CHECK the control rod settles into Position 48 and the ROD SETTLE light extinguishes.
[6.4] IF Control Rod Coupling Integrity Check fails, THEN REFER TO 3-AOI-85-2.
8 Page ISof 46 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 34 18-43 at 08 ATC 6.6.5 Return to Normal After Completion of Control Rod Withdrawal
[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:
[1.1] PLACE CRD POWER, 3-HS-85-46, in OFF.
[1.2] PLACE CRD POWER, 3-HS-85-46, in ON.
8
- Pagel6of46 Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27 DRIVER Verify malfunction (imf rd06r2627) to stick control rod 26-27 at 08 indicates TRUE ATC Reports Control Rod 26-47 failed to withdraw from position 08.
SRO Direct 3-01-85 Section 8.15 8.15 Control Rod Difficult to Withdraw
[1] VERIFY the control rod will not notch out. Refer to Section 6.6.
[2] REVIEW all Precautions and Limitations in Section 3.0 CAUTION
[NER!q Never pull control rods except n a deliberate, carefully controlled manner, while closely monitoring the Reactors response. ThJPG SOER6-OOIl
[3] [NRC/Cl IF RWM is enforcing, THEN VERIFY RWM is operable and LATCHED in to the correct ROD GROUP. [NRC-IR 84-02j NOTES I) Steps 8:1 5[4] through 8.1 5[6] should be used when the control rod is at Position 00 while Step 8.1 5[7] should be used when the control rod is at OR between Positions 02 and 46.
- 2) Double clutching of a control rod at Position 00 will place the rod at the overtravel in stop, independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet lingers from engaging the 00 notch.
- 3) Step 8.1 5[4j may be repeated as necessary until it is detemined that this method will not free the control rod.
8 Pai7of46 Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27
[7] IF the control rod is at or between Positions 02 and 46, THEN PERFORM the following to withdraw the control rod using elevated drive water pressure:
[7.1] RAISE CRD DRIVE WTR HDR DP, 3-PDI-85-17A, to 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
ATC [7.2] ATTEMPT to withdraw the control rod using CRD CONTROL SWITCH, 3-HS-85-48.
[7.3] IF the control rod successfully notches out, THEN LOWER CRD DRIVE WTR HDR DP, 3-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A, and PROCEED to Section 6.6.
CAUTION To prevent a drive from double notching in a high rod worth region and to reduce exposure of chive seals and directional control valves to excessive pressures, the CR0 DRIVE WTR HDR DP should be returned to between 250 psid and 270 psid as soon as possible.
[7.4] IF the control rod failed to notch out, THEN RAISE CRD DRIVE WTR HDR DP, 3-PDI-85-1 7A, to 350 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[7.5] ATTEMPT to withdraw the control rod using CRD CONTROL SWITCH, 3-HS-85-48.
[7.6] LOWER CRD DRIVE WTR HDR DP, 3-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
8 Pa13of46 -
Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27 drifts After drive water prissure i aéd, ad Oerator withdraws Control Rod 26-2 Driver Dnver INSERT trigger 3 for Control Rod Drift ATC Reports CONTROL ROD DRIFT alarm and Control Rod 26-27 is drifting out SRO Directs ATC to respond per ARP and enters 3-AOI-85-6 3-ARP-9-5A window 28 CONTROL ROD DRIFT ATC A. DETERMINES which rod is drifting from Full Core Display.
B. IF no control rod motion is observed, THEN RESETS rod drift as follows:
- 1. PLACE ROD DRIFT ALARM TEST switch, 3-HS-85-3A-S7, in RESET and RELEASE.
- 2. RESET the annunciator.
C. IF rod drifting in, THEN REFER TO 3-AOl-85-5 and 3-AOI-85-7.
D. IF rod drifting out, THEN REFER TO 3-AOI-85-6 and 3-AOI-85-7.
E. REFER TO Tech Spec Section 3.1.3 and 3.10.8.
ATC Responds per 3-AOI-85-6 Monitors Full Core Display for a second Control Rod Drift, per Immediate Actions of 3-AOI-85-6.
ATC 3-AOI-85-6 Control Rod Drift 4.1 Immediate Actions
[1] IF multiple control rod drifts are identified, THEN MANUALLY SCRAM the reactor and enter 3-AOl-i 00-1.
8 Page 9cf 46 Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27 drifts ATC 3-AOI-85-6 Control Rod Drift (continued) 4.2 Subsequent Actions
[1] IF a Control Rod is moving from its intended position without operator actions, THEN SELECT the drifting control rod and INSERT to the FULL IN position.
[2] IF control rod drive does NOT respond to INSERT signal, THEN
[3] NOTIFY the Reactor Engineer to Evaluate Core Thermal Limits and Preconditioning Limits for the current Control Rod pattern.
[4] IF another Control Rod Drift occurs before Reactor Engineering completes the evaluation, THEN MANUALLY SCRAM Reactor and enter 3-AOl-i 00-1.
[5] IF the control rod will not latch into position 00 and continues to demonstrate occurrences of inadvertent withdrawal, THEN
[6] IF the control rod is latched into position 00, THEN REMOVE associated HCU from service per 3-01-85.
[7] EVALUATE Tech Spec 3.1.3.
[8] INITIATE Service RequestNIork Order.
ATC 3-AOI-85-6 Control Rod Drift (continued) 4.2 Subsequent Actions(conti nued)
[9] NOTIFY Reactor Engineer to perform the following for current condition:
- EVALUATE condition of core to assure no resultant fuel damage has occurred.
- EVALUATION of impact on thermal limits and PCIOMOR restraints. (N/A if scram was initiated.)
- DETERMINE if other control rods need to be repositioned in order to safely restore core symmetry to prevent local fuel damage. (N/A it scram was initiated.)
[10] NOTIFY System Engineering to PERFORM 0-Tl-20, Control Rod Drive System Testing and Troubleshooting to determine problem with faulty control rod.
8 Page2Oof4&
Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27 drifts
[13] NOTIFY Reactor Engineer to EVALUATE impact on preconditioning envelope, prior to returning to normal power operation.
ATC Selects Control Rod 26-27 and inserts to position 00.
When Drive In is released control rod will start to drift out again, maintains drive in for control rod 26-27 until Operator dispatched to scram rod with SRI switch.
When requested to scram control rod 26-27 insert IRF RD1Or2627 SCRAM, oncd scrammed delete control rod drift on 26-27 and to return SRI to Normal insert MRF Driver Driver RD1Or2627 NORM.
As Reactor Engineer, inform that Core Thermal Lrrrnts and Preconditioning Limits for the current Control Rod pattern will bpevaluated.
SRO Evaluates Tech Spec 3.1.3, Condition C.
3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
ACTIONS NOTE Separate Condition entry is allowed for each control rod.
C. One or more control rods Ci inoperable for reasons RWM may be bypassed other than Condition A or as allowed by B. LCD 3.3,2.1w if required, to allow insertion of inoperable control rod and confinued operation.
Fully insert inoperable 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control rod.
AND 0.2 Disarm the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> CRD.
8 Page 21 of 46-Simulator Event Guide:
Event 3 Component: Control Rod stuck, 26-27 drifts Driver Driver The SRO will direct the associated HCU removed from service Acknowledge order to remove HCU from seniice, verify what steps in 3Ol-85 will be us to isolate th HCU.
Wait 20 minutes, then INSERT malfunction rdO8 tc bring in accumulator low.
pressure alarm and report HCU removed from servicer Notifies Reactor Engineer to perform the following for current condition:
. Evaluation of condition of core to assure no resultant fuel damage has occurred.
- Evaluation of impact on thermal limits and PCIOMOR restraints.
. Determination if other control rods need to be repositioned in order to safely restore core symmetry to prevent local fuel damage.
Notifies System Engineering to perform O-Tl-20, Control Rod Drive System Testing and Troubleshooting to determine problem with faulty control rod.
Enters 3-G0I-100-12, Power Maneuvering, for the power change that occurred.
Directs associated HCU removed from service, lAW 3-01-85.
ive Driv& Acknowledge all positions informed in step 2 of Subsequent Actions, IF contacted as Work Control, inform that you wfll getwqrking on a Work Order/service Request IF contacted as Reactor Engineers inform that you will evaluate all conditions listed above
8 Page-22 of 46 - -
Simulator Event Guide:
Event 4 Instrument: Hydrogen Water Chemistry Malfunction Driver When directed by NRC, insert malfunction OGO5a to cause a Hydrogen Water Chemistry MaLfunction.
BOP Respond to Off Gas Panel Alarms 9-53-10, 3, and 13 53-10, H2 Water Chemistry Abnormal A. Checks H2 concentration on H2 analyzer on 3-9-53.
B. Dispatches personnel.
53-3 and 13, High Offgas % H2 Train A and B A. CHECK H2 concentration on OFF-GAS HYDROGEN ANALYZER, at 3-H2R-66-96 (CH2), on Panel 3-9-53 to verify H2 concentration..
B. IF alarm is valid, THEN REFER TO 3-AOl-66-1.
SRO Announces entry into 3-AOI-66-1, Off Gas H2 High.
Driver When dspathéd to the Panel ii ONE rninüté reports H2injection rates abov (high) setpoint cannot adjusts If requested to Shutdown H2water chemistry loca1I inform control room yo cannot access the swich BOP 3-AOl-66-1, Off Gas H2 High
[2) IF HWC System injection is in service, THEN PERFORM the following (otherwise N/A):
[2.1] At HYDROGEN WATER CHEMISTRY CONTROL PANEL, 3-LPNL-925-0589, VERIFY that H2 and 02 injection rates are normal at Operator Interface Unit (OIU). (H2 injection rate should match the setpoint on the OIU. The 02 injection rate should match the setpoint on the OIU, which should be half of the H2 injection rate during normal steady state conditions.)
[2.2] IF H2 and 02 injection rates do NOT meet the above conditions, THEN NOTIFY the Unit Supervisor and INITIATE a HWC System shutdown using either:
. 3-HS-4-40A H2 WATER CHEMISTRY CONTROL
[Panel 3-9-53) or
. 3-HS-4-40B H2 WATER CHEMISTRY CONTROL
[Panel 3-9-5] or
. 3-HS-4-39 HWC SHUTDOWN SWITCH [3-LPNL-925-0588].
BOP Shutdown HWC System using either 3-HS-4-40A at panel 9-53 or 3-HS-4-40B at panel 9-5 SRO [4] IF hydrogen concentration is 4%, THEN REFER TO TRM 3.7.2 dC NRC Once HWC is Shutdown H2 Concentration will begin to lower slowI
8
- Page 23of46 Simulator Event Guide:
= Event 4 Instrument: Hydrogen Water Chemistry Malfunction SRO 3-AOI-66-1, Off Gas H2 High SRO NOTE Fuel failure is indicated by, but NOT limited to, rising activity on the following:
- OFF-GAS PRETREATMENT RADIATION recorder, 3-RR-90-157 (Panel 3-9-2)
- MAIN STEAM LINE RADIATION recorder, 3-RR-90-1 35 (Panel 3-9-2)
- OFFGAS POST-TREATMENT RADIATION recorder, 3-RR-90-265
- On MAIN CONDENSERS (MN COND) ICS display:
Offgas pretreatment, post treatment, and stack radiation
[5j IF high hydrogen concentration is a result of possible fuel failure, THEN REDUCE core flow to 50 60 % (otherwise N/A).
NR NB No indication of Fuel Failure Exists, step 5 should be N.
BOP Report H2 Concentration lowering slowly.
CAUTION When hydrogen concentration is suspected or being greater than 4% (by volume), Off-Gas System valves (other than nonspark producing SJAE inlet valves 3-FCV-066-OOl I and 3-FCV-0613-OO1 5) are NOT allowed to. be operated until the unit is shutdown and hydrogen concentration is confirmed to be < 4% (by volume).
8 Page24of 4&
Simulator Event Guide:
Event 4 Instrument: Hydrogen Water Chemistry Malfunction SRO
[7] WHEN any of the following conditions exist, THEN INITIATE actions to reduce hydrogen concentration within 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s:
. Hydrogen Analyzer on Panel 3-9-53 indicates 4%
. Chemistry Lab grab samples indicate 4% hydrogen SRO REFER TO TRM 3.7.2 TR 3J PLANT SYSTEMS TR 3.7,2 Airborne Efflunt LCO 3.7.2 Whenever the SJAE is in serdice, the noen1rs1ibn ci hydrogen in the offgas downstnssm of the reoornbiners shaU be limited to 4%
by )ume.
APPLICABILITY: During mn condenser offgas trement rsern cperstion NOTE TRM ICC 3.0.3 is rt ltoable.
AT1ONS CONDON REQUIRED ACTION COMPLETiON TIME A With the concentration AI Restore the 4 bours ci hydrogen >4% by concerticn o wftNn lune. the lirniL
8 Page 25 of 46 Simulator Event Guide:
Event 5 Component: Loss of RPS A DRIVER DRiVER When directed by the NRC, insert malfunction imf rpoia Trigger 10 to cause a loss of RPS A ATC Recognizes a loss of RPS A ATC REACTOR PROT 1 20V PWR SYS ABNORMAL 2-XA-55-5B, Window 10 Operator Action:
A. REFER TO 1 -AOl-99-1.
SRO Directs entry into 3-AOI-99-1, Loss of Power to One RPS Bus 4.1 Immediate Actions
[1) STOP all testing with potential RPS half-scrams or PCIS logic isolation signals.
NOTE The blanks to the side of steps contained in Section 4.0 Operator Actions are intended for place keeping only. Initials are NOT required. If necessary, place keeping marks may be made directly in the Control Room copy of this instruction. CONTACT Management Services for a replacement copy when time permits.
4.2 Subsequent Actions NOTES
- 1) If power cannot be restored promptly to a de-energized RPS Bus, plant operation may continue until repairs are made provided all plant operational limits are met.
- 2) With Reactor Building Ventilation isolated, Main Steam Line Tunnel Area temperature can reach PCIS Group I isolation trip setpoint in less than 10 minutes unless the Main Steam Tunnel Booster Fan is in service.
- 3) With Drywell Control Air isolated, MSIV accumulator air can bleed down and cause MSIVs to close.
8 Page 2&ot-4&
Simulator Event Guide:
Event 5 Component: Loss of RPS A Crew [1] VERIFY automatic actions occur.
[2] ATTEMPT to determine cause of loss of RPS Bus using indicating lights inside RPS Circuit Protector cabinets.
[3] NOTIFY Chemistry RWCU is isolated and no longer in-service and a sampling LCO per TRM 3.4.1 is to be entered.
[4] NOTIFY Electrical Maintenance to correct cause.
[5] RESTORE power to RPS Bus A(B) using alternate power supply. REFER TO 3-01-99 section for Immediate Restoration of Power to RPS Bus A(B) Using Alternate Power Supply.
[5.1] DISPATCH operator to Aux. Instrument Room to reset ATU GROSS FAILURES.
Driver After dispatched to RPS MG report trip of MG Set cause unknown at this time When regiested to restore RPS A using alternate insert trigger I t
[6] WHEN system restoration is desired, THEN RESTORE systems to normal. REFER TO 3-01-99 section for Restoration to Normal Following RPS Bus Power Loss.
8 Page27 of 46 Simulator Event Guide:
Event 5 Component: Loss of RPS A Crew Restoration of systems to normal per 3-01-99 8.3 Restoration to Normal Following RPS Bus Power Loss NOTES
- 1) This section provides instructions for resetting the various system isolations and reopening affected valves to allow those systems to be restored to normal operation in accordance with their respective operating instructions.
2> The following steps are performed at Panel 3-9-5 unless otherwise noted.
- 3) When RPS Bus power is lost to some scram discharge volume level switches, their RTD heater is de-energized. Following the restoration of power, a time delay, dependent on how long the level switch was de-energized, prevents resetting the half scram signal. This may take up to 37 seconds after RPS power is restored.
Precaution 30M can be referred to for more information on these level switches.
[1] OBTAIN Unit Supervisor/SROs permission to restore to normal.
[2] MOMENTARILY PLACE SCRAM RESET, 3-HS-99-5A1S5, as follows:
ATC A. RESET FIRST B. RESET SECOND C. NORMAL
[3] CHECK the following conditions:
A. All eight SCRAM SOLENOID GROUP A/B LOGIC RESET lights ILLUMINATED.
B. The following four lights ILLUMINATED:
- SYSTEM A BACKUP SCRAM VALVE, 3-I L-99-5AIAB
- SYSTEM B BACKUP SCRAM VALVE, ATC 3-I L-99-5A/CD C. Scram Discharge Volume vent and drain valves indicate OPEN.
D. Points SOEO33 and SOEO35 on ICS computer or on the First Out Printer reads NOT TRIP for RPS A.
E. Points SOEO34 and S0E036 on ICS computer or on the First Out Printer reads NOT TRIP for RPS B.
8 Page 28 of 46 Simulator Event Guide:
Event 5 Component: Loss of RPS A
[4] At Panel 3-9-4, RESET PCIS trip logic as follows:
[4.1] MOMENTARILY PLACE PCIS DIV I RESET, 3-HS-64-16A-S32, to left and right RESET positions.
[4.2] CHECK the following red lights ILLUMINATED:
. MSIV GROUP Bl, 3-lL-64-Bl
[4.3] MOMENTARILY PLACE PCIS DIV II RESET, BOP 3-HS-64-16A-S33, to left and right RESET positions.
[4.4] CHECK the following red lights ILLUMINATED:
- MSIV GROUP A2, 3-IL-64-A2 B P
- MSIV GROUP B2, 3-IL-64-B2 NOT!
Steps 8.3(5] through 8.3(21] can be performed in any order.
BOP [5] CHECK the green lights are ILLUMINATED on all 5 of the QLVPS located at bottom of Panel 9-14.
BOP [6] RESTORE Reactor and Refuel Zone Ventilation to normal operation. REFER TO 3-AOI-64-2D, Group 6 Ventilation System Isolation.
[7] RESTORE Standby Gas Treatment System to standby readiness. REFER TO 0-01-65.
8 Page29of46 Simulator Event Guide:
Event 5 Component: Loss of RPS A BOP 3-AOl-64-2D Group 6 Ventilation System Isolation
[1 3} WHEN initiating signal has been corrected AND necessary repairs are made, THEN
[13.1] VERIFY PCIS RESET:
- RESET PCIS DIV I RESET, 3-HS-64-16A-S32.
- RESET PCIS DIV II RESET, 3-HS-64-16A-S33.
[13.2] RESET Reactor/Refuel isolation logic, as required:
- PLACE REFUEL ZONE FANS AND DMPRS, 3-HS-64-3A, in OFF.
- PLACE REACTOR ZONE FANS AND DMPRS, 3-HS-64-1 1A, in OFF.
[13.3] START Reactor/Refuel zone ventilation, as required:
- PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-11 A, in SLOW A (SLOW B).
- PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS 3A, in SLOW 3A (SLOW 3B).
[13.4] For the fans started, VERIFY that the dampers open and fans start as indicated by illuminated red lights above the following switches:
- The two green lights A(B) above REACTOR ZONE FANS AND DAMPERS Switch 3-HS-64-1 1A, extinguish and the two red lights A(B) illuminate.
- The two green lights A(8) above REFUEL ZONE FANS AND DAMPERS Switch 3-HS-64-3A, extinguish and the two red lights A(B) illuminate.
- REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
- REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
- REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
- REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6
8 Page 30of 4&
Simulator Event Guide:
Event 5 Component: Loss of RPS A BOP [8] At Panel 3-9-3, PLACE PSC head tank pumps in service as follows:
. PLACE SUPPR POOL DRAIN INBD ISOL VALVE, 3-HS-75-57A, in AUTO After OPEN.
. PLACE SUPPR POOL DRAIN OUTBD ISOL VALVE, 3-HS-75-58A, in AUTO After OPEN.
[1 1] At Panel 3-9-4, RESTORE Drywell Floor and Equipment Drain Systems to normal operation as follows:
[11.1] NOTIFY Radwaste Operator that Drywell Equipment and Floor_Drain_Sump_isolation_valves_are_being_reopened.
[11.2] PLACE DRYWELL EQPT DR INBD ISOL VLV, 3-HS-77-15A, in AUTO After OPEN.
[11.3] PLACE DRYWELL EQPT DR OUTBD ISOL VLV, 3-HS-77-15B, in AUTO After OPEN.
[11.4] PLACE DRYWELL FLOOR DR INBD ISOL VLV, 3-HS-77-2A, in AUTO After OPEN.
[11.5] PLACE DRYWELL FLOOR DR OUTBD ISOL VLV, 3-HS-77-2B, in AUTO After OPEN.
BOP [12] At Panel 3-9-2, RESTORE Radiafion Monitoring System as follows:
[12.1] DEPRESS RESET pushbutton.
[12.2] VERIFY OPEN the associated valve.
[12.3] RELEASE pushbutton.
8 Page-3-1 of 46 Simulator Event Guide:
Event 6 Component : Trip of RFPT C DRIVER When directed by NRC, insert malfunction (imf fwl5c (e5 O)to trip the RFPTC ATC Recognizes Trip of RFPT C and reports to SRO.
ATC Verifies Feedwater is recovering RPV Water Level in Automatic.
Responds to the following alarms:
ATC RFPT TRIPPED (3-XA-55-6C, Window 29)
RFPT C ABNORMAL (3-XA-55-6C, Window 15)
RFPT TRIPPED (3-XA-55-6C, Window 29):
Operator Action:
A. VERIFY reactor power is within the capacity of operating REPs.
B. CHECK core limits.
ATC C. WHEN RFPT coasts down to zero speed, unless RFPT is rolling on minimum flow, THEN VERIFY turning gear motor starts and engages.
D. REFER TO 3-AOI-3-1 or 3-01-3, Section 8.1.
RFPT C ABNORMAL (3-XA-55-6C, Window 15)
Operator Action:
A. CHECK other RFP alarms on Panel 3-9-6 to determine problem area.
ATC B. REFER TO appropriate alarm response procedure.
C. IF no other annunciator on Panel 3-9-6 is in alarm, THEN PERFORM an alarm summary on alarm types.
8 Page 32 of 46 Simulator Event Guide:
Event 6 Component : Trip of RFPT C SRO Enters 3-AOl-3-1, Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-AOI-3-1, Loss Of Reactor Feedwater or Reactor Water Level High/Low CAUTION
£NRC!C1 Operations outside of the allowable regions shown on the Recirculation System ATC Operating Map could result in thermal-hydraulic power oscillations and subsequent fuel damage. REFER TO 3-GO1-100-12A for required actions and monitoring required during a power reduction. co 94Q245001]
[1] VERIFY appHcable automatic actions.
[2] IF level OR Feedwater flow is lowering due to loss of Condensate, Condensate Booster, or Feedwater Pump(s), THEN REDUCE Recirc flow as required to avoid scram on low level.
[3] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s).
[10] IF REPs are in manual control, THEN RAISE speed of operating RFPs.
ATC Takes manual control of RFPT B and restores water level.
When directed by the NRC Inseit Trigger 20 for loss of all offsite power and trigger Dnv er DrIVe 25 on the scrarli If an AutornatIQ Scram occurs or the crew scrarns the reactor insert trigger Driv Die 20and25
8 Page 33of Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS Report Reactor Scram and places Mode Switch to Shutdown ATC Scram Report:
. Rods out
. MSIVs closed
. Reactor Level lowering and no Feedwater Available
. Reactor Pressure increasing
. Initiated One Channel of ARI BOP Report loss of all offsite power SRO Directs entry into 0-AOl-57-1A SRO Enters EOI-1 on Low Reactor Water Level and High Reactor Pressure RCIQ Monitor and Control Reactor Power Verify the Mode Switch is on Shutdown Initiate second channel of ARI Verify Recirc Pump Runback. (Pump speed 480rpm or less)
Answers NO to: Is Reactor Power above 5% or Unknown?
Directs ATC to perform:
. Appendix 2, Defeat ARI Logic Trips CT#2
- Appendix 1 F, Defeat RPSA Logic Trips
- Appendix 1 D, Drive Control Rods Recognizes CRD 3A Pump is not available due to power loss. Starts 3B CRD ATC pump. He will have to coordinate this step if RHR 3A pump is already running on 4KV S/D Board EA (3EA diesel supplying).
BOP/A-rC Dispatch personnel to perform Appendix 2 and outside portions of Appendix 1 F.
Dispatch personnel to close, 3-FCV-85-586. (While awaiting completion of Appendix 1 F.)
Drive Rods per Appendix 1 D, while waiting for completion of Appendix 1 F.
Driver Drive When directed to pérforni Appendix 2 and outside portions of Appendix 1 F wait minutes. Insert triggers 22 (bat appOif) and 23 (bat appO2) then report corn p1etion.
8
- Pag&34of46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS DRIVER When scram is reset insert trigger 21 (bat atws.1)
ATC Appendix iF CT#2
- 3. VERIFY OPEN Scram Discharge Volume vent and drain valves.
- 4. DRAIN SDV UNTIL the following annunciators clear:
- WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 1)
- EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 29).
- 5. DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER ISOL.
- 6. WHEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.
- 7. CONTINUE to perform Steps 1 through 6 UNTIL ANY of the following exists:
- ALL control rods are fully inserted, OR
- NO inward movement of control rods is observed, OR
- SRO directs otherwise.
DRIVEf If directed to close 3-FGV85-586, wait 3 minutes, then insert mWrdO6 clse and report qompietior IfMlhen directed to re-open 3-FCV-85-586, waitS minutes, then insert rnrf rdO6 open and report qompietion
8
- Page3Sof46-Simulator Event Guide:
Event 7 Major: Loss of Otfsite Power, LOCA, ATWS ATC AppendixiD CT#2
- 1. VERIFY at least one CRD pump in service.
- 2. IF Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SOV
- 3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.
- 4. BYPASS Rod Worth Minimizer.
- 5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:
- a. SELECT control rod.
- b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inward.
- c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.
- 6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SOV (RB NE, El 565 ft).
RCIP Monitor and Control RPV Pressure.
Answers YES to: Is any MSRV cycling?
Directs BOP to maintain RPV Pressure 800-1000 psig using SRVs Appendix hA BOP Maintains RPV Pressure as directed with SRVs
8
- Page36of46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS SRO RC/L Monitor and Control RPV Water Level.
Verify as Required:
- PCIS Isolations (Groups 1, 2 and 3)
- RCIC Answers YES to: Can water level be Restored and Maintained above (+) 2 inches?
Direct level control with RCIC lAW 3-EOI Appendix-5C with a level band of
+2 to 51 inches BOP Control Reactor Water Level with RCIC lAW 3-EOl Appendix-5C
- 4. VERIFY 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 620 gpm.
- 5. OPEN the following valves:
- 3-FCV-71-39, RCIC PUMP INJECTION VALVE
- 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE
- 6. PLACE 3-HS-71 -31 A, RCIC VACUUM PUMP, handswitch in START.
- 8. CHECK proper RCIC operation by observing the following:
- a. RCIC Turbine speed accelerates above 2100 rpm.
- c. 3-FCV-71 -40, RCIC TESTABLE CHECK VLV, opens by observing 3-ZI-71-40A, DISC POSITION, red light illuminated.
- d. 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE, closes as flow rises above 120 gpm.
- 9. IF BOTH of the following exist:
- RCIC Initiation signal is NOT present, AND
- 0. ADJUST 3-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller as necessary to control injection.
8 Page3-7of-46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS SRQ Enters 0-AOl-57-1A, Loss of Offsite Power (161 and 500 KV)/Station Blackout 0-AOl-57-1A, Loss of Offsite Power (161 and 500 KV)/Station Blackout 4.1 Immediate Actions BOP
[1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators.
Recognizes 3EA Diesel Generator started but the breaker failure auto close and BOP CT#1 manually closes the breaker to energize the 4KV shutdown board 3EA.
BOP Recognized the failure of 3EB, 3EC, and 3ED diesels to start. Manually starts 3EC and 3ED. 3EB does not start.
[2] VERIFY two EECW Pumps (not using the same EECW strainer) are in BOP service supplying Diesel Generators.
[4] PERFORM the following to ensure at least one train of Diesel BOP Generator Room Fans are energized:
. VERIFY 480V DSL Aux Board 3EA or 3EB energized.
NRC DG 3EAt the only DG that will pick up Its respective $hutdowp Barct
8 Page 3&of 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS 4.2 Subsequent Actions
[1] IF ANY EOI entry condition is met, THEN REFER to the appropriate EOl(s). (Otherwise N/A)
CT#1 BOP [2] VERIFY automatic actions and PERFORM any that failed to occur.
Recognizes 3EA Diesel Generator started but the breaker failure auto close and BOP CT#1 manually closes the breaker to energize the 4KV shutdown board 3EA.
NOTES
- 1) If a Unit is in a Station Blackout condition, performance of this instruction will also require implementation of 1(2)(3)-AOl-30B-1, Reactor Building Ventilation Failure, on the Unit in Station Blackout.
- 2) EECW supply valves to the Control Air Compressors and RBCCW are air operated. If initial air pressure is low, air compressors may trip on high temperature, until cooling water flow is established.
- 3) The North header supply to Unit 1 RBCCW, the North header supply to Unit 2 RBCCW and the South header supply to Unit 3 RBCCW are normally isolated with a manual valve; therefore no flow will occur when either 1-FCV-67-60, 2-FCV-67-50 or 3-FCV-67-51 opens.
[3] WHEN EECW header pressure is restored above the reset pressure setpoint (psig) for the valves listed below, THEN Common Unit 1 Unit 2 Unit 3 0-FCV-67-53 106 - - -
FCV-67-50 - 90 91 92 BOP FCV-67-51 - 107 109 113 RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel 1-LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. Refer to the EECW to the ROW Crossties for Control Air & RBCCW section of 0-01-67.
DRIVEF When requested to restore EECW in the above step, insert trigger 26 (bat eecw) and then trigger 27_bat eecw$
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8 Page 4Gof 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS Executes all legs of EOl-2 concurrently EOl-2 DWJT Monitor and control Drywell Temperature below 160°F, using available Drywell Cooling.
Answers NO to: Can Drywell Temperature be maintained below 160°F?
PC/P SRO Monitor and control Primary Containment pressure below 2.4 psig using the Vent System (Appendix 12) as necessary Direct Appendix 12 Answers No to can Primary Containment Pressure be maintained below 2.4 psig Before Suppression Chamber Pressure rises to 12 psig Initiate Suppression Chamber Sprays using only those pumps not required for Adequate Core Cooling Directs Suppression Chamber Srravs EOl-2 PC/H Monitor and control Drywell and Suppression Chamber:
. Hydrogen at or below 2.4%
AND
. Oxygen at or below 3.3%
SRO Using the Nitrogen Makeup System (APPX 14A).
EOI-2 SPIT Monitor and control Suppression Pool temperature below 95°F, using available Suppression Pool Cooling (APPX 17A) as necessary.
SRO EOl-2SP/L Monitor and control Suppression Pool Level between -1 inch and -6 inches.
Can Suppression pool level be maintained above -6 inches YES Can Suppression pool level be maintained below -1 inch YES
8 Page 41 of 4&
Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS ATC/BOP Vent Containment lAW Appendix 12 VERIFY at least one SGTS train in service.
- 2. VERIFY CLOSED the following valves (Panel 2-9-3 or Panel 2-9-54):
2-FCV-64-31, DRYWELL INBOARD ISOLATION VLV, 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE.
Steps 3, 4, 5 and 6 are If / Then steps that do not apply
- 7. CONTINUE in this procedure at:
Step 8 to vent the Suppression Chamber through 2-FCV-84-1 9, OR Step 9 to vent the Suppression Chamber through 2-FCV-84-20.
- 8. VENT the Suppression Chamber using 2-FIC-84-1 9, PATH B VENT FLOW CONT, as follows:
- a. PLACE keylock switch 2-HS-84-35, DW/SUPPR CHBR VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).
- b. VERIFY OPEN 2-FCV-64-32, SUPPR CHBR VENT INBD ISOL VALVE (Panel 2-9-54).
- c. PLACE 2-FIC-84-1 9, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).
- d. PLACE keylock switch 2-HS-84-19, 2-FCV-84-19 CONTROL, in OPEN (Panel 2-9-55).
- e. VERIFY 2-FIC-84-1 9, PATH B VENT FLOW CONT, is indicating approximately 100 sctm.
- f. CONTINUE_in_this_procedure_at_step_12.
8 Page42*of 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS BOP Vents Primary Containment lAW Appendix 12
- 9. VENT the Suppression Chamber using 2-FIC-84-20, PATH A VENT FLOW CONT, as follows:
- a. VERIFY OPEN 2-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 2-9-3).
- b. PLACE keylock switch 2-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to SUPPR-CHBR position (Panel 2-9-54).
- c. VERIFY OPEN 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV (Panel 2-9-54).
- d. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 2-9-55).
- e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-55).
- f. VERIFY 2-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scfm.
- g. CONTINUE in this procedure at step 12.
- 12. ADJUST 2-FIC-84-19, PATH B VENT FLOW CONT, or 2-FIC-84-20, PATH A VENT FLOW CONT, as applicable, to maintain ALL of the following:
Stable flow as indicated on controller, AND 2-PA-84-21, VENT PRESS TO SGT HIGH, alarm light extinguished, AND Release rates as determined below:
iii. IF Venting for ANY other reason than items i or ii above, THEN MAINTAIN release rates below Stack release rate of 1.4 x 107 pCi/s AND 0-SI-4.8.Bi .a.1 release fraction of 1.
DRIVER Acknowledge Notification
8 Page 43 of 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS ATC/BOP Initiate Suppression Chamber Sprays per Appendix 17C
- 1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.
- 2. IF Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
ATC/BOP 5: INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN...PLACE keylock switch 2-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 2-FCV-74-53(67), RHR SYS 1(11) INBD INJECT VALVE, is OPEN, THEN...VERIFY CLOSED 2-FCV-74-52(66), RHR SYS 1(11) OUTBD INJECT VALVE.
- e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spray.
- g. OPEN 2-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
- h. IF RHR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN.. CONTINUE in this procedure at Step 5.k.
- i. VERIFY CLOSED 2-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
- j. RAISE System flow by placing the second RHR System 1(11) pump in service as necessary.
8 Page 44of 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS ATC/BOP I. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s).
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
SRO When Suppression Chamber Pressure exceeds 12 psig, determines that Drywell Sprays are required.
Directs Loop II of RHR to be placed in Drywell Sprays per EOl Appendix 17B.
ATC/BOP Drywell Sprays per appendix 17B
- 1. IF Adequate core cooling is assured OR Directed to spray the Drywell irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
. PLACE 1-HS-74-155A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.
. PLACE 1-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.
- 2. VERIFY Recirc Pumps and Drywell Blowers shutdown.
THEN CONTINUE in this procedure at Step 6 using RHR Loop 1(11).
8 Page 4&of4&
Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS ATC/BOP 6. INITIATE Drywell Sprays using RHR Loop I as follows:
- a. BEFORE drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 9.
- c. IF EITHER of the following exists:
. LPCI Initiation signal is NOT present, OR
- Directed by SRO, THEN PLACE keylock switch 1-XS-74-122(130), RHR SYS I LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- e. IF1-FCV-74-53, RHR SYS I LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 1-FCV-74-52, RHR SYS I LPCI OUTBD INJECT VALVE.
- f. VERIFY OPERATING the desired System 1(11) RHR pump(s) for Drywell Spray.
- g. OPEN the following valves:
1-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV 1-FCV-74-61, RHR SYS I DW SPRAY INBD VLV.
- h. VERIFY CLOSED 1-FCV-074-0007, RHR SYSTEM I MIN FLOW VALVE.
I. THROTTLE the following in-service RHRSW outlet valves to obtain between 1,350 and 4,500 gpm RHRSW flow:
.__1-FCV-23-34,_RHR_HX_1A_RHRSW_OUTLET VLV
8 Page 46 of 46 Simulator Event Guide:
Event 7 Major: Loss of Offsite Power, LOCA, ATWS
- 9. WHEN EITHER of the following exists:
- Before drywell pressure drops below 0 psig, OR
- Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:
- a. VERIFY CLOSED the following valves:
SRO REP Classification is an Alert. EAL 2.1-A Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.
All Control Rods are inserted Reactor Level is maintained Drywell or Suppression Chamber has been Sprayed
8 of Page47 46 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:
3% power. 3-G0l-100-1A Section 5.4 Step 63.3 Operations/Maintenance for the Shift:
Warm RFPT B lAW 3-01-3 Section 5.6 step 20. Continue to pull rods for Mode Change.
Units 1 and 2 are 100% Power Unusual Conditions/Problem Areas:
None
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 10
- Op-Test No.: 1108 Examiners: Operators: SRO:_
ATC:
BOP:
Initial Conditions: 95% power. DG 3A is OOS Turnover: Alternate Refuel and Reactor Zone Fans lAW 3-Ol-30A and 3-Ol-30B and raise power to 100%
with Recirc Flow Event Maif. No. Event Type* Event Description No.
N-BOP Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 3-1 N-SRO OI-30B R-ATC 2 Raise Power to 100% with flow R-SRO C-ATC 3 CUO4 RWCU Leak with failure to Auto isolate TS-SRO C-BOP Bus Duct Cooling Fan Trip with failure of standby fan to auto 4 EGJ3a C-SRO start C-ATC 5 TFJO3a R-ATC RR Pump A Trip with power oscillations TS-SRO C-BOP Level 2 instrument failures (58A and 58D) cause HPCI and RCIC 6 TH3Oa / d TS-SRO to Auto initiate 7 HPO8 M-ALL HPCI Steam Leak without Isolation 8 TCO2 C No bypass valves with ATWS 9 ED12a C Loss of 480V RMOV BD 3A (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or
Appendix D Scenario Outline - - Form ES-D-i Critical Tasks Three CT#1 -With reactor at power and with a primary system discharging into the secondary containment, manually scram the reactor before any area exceeds the maximum safe operating level.
- 1. Safety Significance:
Scram reduces to decay heat energy that the RPV may be discharging into the secondary containment.
- 2. Cues:
Procedural compliance Secondary containment area temperature, level, and radiation indication Field reports
- 3. Measured by:
Observation With a primary system discharging into secondary containment, a reactor scram is initiated before a maximum safe condition is reached.
OR Observation With a primary system discharging into secondary containment, US transitions to EOP-1 and RO initiates scram upon report that a maximum safe condition has been reached.
- 4. Feedback:
Control rod positions Reactor power decrease CT#2-With a primary system discharging into the secondary containment, when two or more areas are greater than their maximum safe operating values for the same parameter, RO initiates Emergency Depressurization as directed by US.
- 1. Safety Significance:
Places the primary system in the lowest possible energy state, rejects heat to the suppression pooi in preference to outside the containment, and reduces driving head and flow of system discharging into the secondary containment.
- 2. Cues:
Procedural compliance Secondary containment area temperatures, level, and radiation indication Field reports
- 3. Measured by:
Observation US transitions to C-2 and RO opens at least 6 SRVs when two or more areas are greater than their maximum safe operating values for the same parameter.
- 4. Feedback:
RPV pressure trend SRV status indications
Appendix D Scenario Outline Form ES-D-1 CT#3-With a primary system discharging into the secondary containment, take action to manually isolate the break.
- 1. Safety Significance:
Isolating high energy sources can preclude failure of secondary containment and subsequent radiation release to the public.
- 2. Cues:
Procedural compliance Area temperature indication
- 3. Measured by:
With the reactor at pressure and a primary system discharging into the secondary containment, operator takes action to manually isolate the break.
- 4. Feedback:
Valve position indication
AppendixD Scenario Outline Form ES-D-1 VENTS
- 1. BOP Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 3-OI-30B
- 2. ATC raises power to 100%
- 3. The ATC will respond to RWCU alarms indicating a leak and RWCU will fail to isolate. The ATC will isolate RWCU and take actions JAW 2-AOI-64-2A. The SRO will evaluate Technical Specification 3.6.1.3 Condition A is required. The SRO will evaluate TRM 3.4.1 and direct Chemistry to sample in order to satisfy TSR 3.4.1.1.
- 4. BOP will respond to Bus Duct Cooling 3A Fan trip and take action JAW with ARPs, start standby Bus Duct Cooling Fan 3B.
- 5. Reactor Recirculation 3A Pump will trip, ATC will respond JAW 3-AOI-68-1A. The ATC will close 3A RR Pump Discharge Valve. Small power oscillations will develop. The ATC will insert control rods to dampen oscillations and exit region 2. The SRO will evaluate Technical Specification 3.4.1 Condition A is required.
- 6. Level transmitter 58A and 58D will fail to less than -45 inches. This failure will result in a HPCI and RCIC auto initiation. The BOP Operator will respond JAW ARPs. BOP Operator will verify that level is in normal band and secure HPCI and RCIC. The SRO will evaluate Technical Specification 3.3.4.2 Condition A and B, 3.3.5.1 Condition A, B and F, 3.3.5.2 Condition A and B, and 3.8.1 Condition D and J.
- 7. After the HPCI initiation a steam leak will develop in the HPCI Room, HPCJ will fail to automatically and manually isolate. When attempting to manually isolate HPCI steam valve 73-2 the 3A 480V RMOV Board will be lost due to an electrical fault.
- 8. The crew will enter EOI-3 and scram the Reactor. A small ATWS will occur on the scram; power, level and pressure will be controlled LAW EOI 1. When the second MAX safe temperature is reached the crew will Emergency Depressurize.
- 9. Turbine Bypass Valves will not be available on the scram with an ATWS of 20 rods.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.
All but 6 Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained.
Appendix D Scenario Outline Form ES-D-1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 10 8 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
10
- Page7of3S Scenario Tasks TASK NUMBER K/A RO SRO Alternate Reactor and Refuel Zone Fans RO U-30A-NO-2 288000A4.01 3.1 2.9 Raise Power with Recirc Flow RO U-068--NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 RWCU Leak with Failure to Auto Isolate RO U-069-AL-10 223002A2.03 3.0 3.3 SRO S-000--EM-12 Bus Duct Cooling Fan Trip RO U-047-AL-13 245000A2.05 3.6 3.8 Reactor Recirculation Pump Trip RO U-068-AB-1 20200 1A2.03 3.6 3.7 SRO S-068-AB-1 Level 2 Instrument Failures RO U-073-NO-5 216000A3.01 3.4 3.4 RO U-071-NO-5 SRO S-000-AD-27 HPCI Steam Leak RO U-073-AL-06 295032EA2.03 3.8 4.0 SRO S-000-AB -03 SRO S-000-EM-12 SRO S-000-EM-15
10 Page 8 of 38 Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 3-0I-30A Refuel Zone Ventilation System Rev 26 3-Ol-30B Reactor Zone Ventilation System Rev 20 3-G0l-100-12 Power Maneuvering Rev 35 3-01-68 Reactor Recirculation System Rev 80 3-ARP-9-3D, W17 RWCU Leak Detection Temperature High Rev 28 3-AOl-64-2A Group 3 RWCU Isolation Rev 9 TS 3.6.1.3 Primary Containment Isolation Valves Amd 212 TRM 3.4.1 Coolant Chemistry Rev 21 3-EOl-3 Secondary Containment Control Rev 10 3-ARP-9-7A, W31 Generator Bus Duct Fan Failure Rev 22 Recirc Pump Trip/Core Flow Decrease OPRMs 3 AOl 68 1A R ev 6
- Operable 3-ARP-9-3F, W29 Reactor Water Level Low Low HPCI/RCIC Initiation Rev 28 Anticipated Transient Without Scram Recirculation TS 3.3.4.2 Amd 213 Pump Trip (ATWS-RPT) Instrumentation Emergency Core Cooling System (ECCS)
TS 3 3 5 1
. m Instrumentation Cooling (RCIC) System TS 3.3.5.2 oIaton1 0
- Amd 244 TS 3.5.1 ECCS Operating
- Amd 244 TS 3.5.3 RCIC System Amd 244 3-ARP-9-3F, W10 HPCI Leak Detection Temperature High Rev 28 3-EOI-1 RPV Control Rev 8 3-AOl-i 00-1 Reactor Scram Rev 53 3-EOI-App-3A SLC Injection Rev 1 3-E0I-3-C-5 Level/Power Control Rev 9
10 Page Procedures Used/Referenced:
Procedure Number Procedure Title Procedure Revision 3-EOI-App-4 Prevention of Injection Rev 4 3-EOl-3-C-2 Emergency RPV Depressurization Rev 8 3-EOl-App-6A Injection Subsystems Lineup Condensate Rev 2 Injection Subsystems Lineup RHR System II LPCI 3-EOI-App-6C Rev 3 3-EOI-App-2 Defeating ARI Logic Trips Rev 4 3-EOI-App-1 F Manual Scram Rev 2 Insert Control Rods using Reactor Manual Control 3-EOI-App-1 D Rev 2 3-EOl-2 Primary Containment Control Rev 8 3-EOI-App-1 7A RHR System Operation Suppression Pool Cooling Rev 5 EPIP-1 Emergency Classification Rev 46 EPIP-4 Site Area Emergency Rev 32
10 Rage 10 of 38 Simulator Instructor 10205
- 3A DG tagged out br ypobkrl 838 fail_ccoil mrf dgola open icr zIo3hs2l 1 3ea9a[1 j off
- RWCU seal leak no auto iso imf cuO6 imf cuO4 (el 0) 100 30050
- bus duct cooling fan trip imf egl3a (e5 0)
- Recirc pump A trip, with pwr oscillations imf th03a (elO 0) imf cr02a (elO 30) 10 120 imf cr02b (elO 30) 10 120
- RCIC/HPCI Initiate due to failed lnstr imfth30a (e15 0)40 imfth30d (e15 0)356084
- ATWS/major HPCI Leak (have to manually modify fpO2 to close) mrf fpO2 (e20 0) close imf hpO9 imfhpo8(e200)87004 trg2l nrc2011732 trg2l =imfedl2a ior ypovfcv733 (e20 0) fail_now bat nrcstick20 imf tcO2 (e20 0) 0 trg 26 = bat appOlf trg 27 = bat appo2 trg 28 = bat appo8ae trg 29 = bat nrcunstickl4 Scenario 10 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 205 Simulator Setup Load Batch bat nrcllO8-l0 Simulator Setup manual Clearance DG 3A Simulator Setup Verify file loaded Simulator Setup RCP required (95% 100% with flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12
10 Page 11of8 Simulator Event Guide:
Event 1 Normal: Alternate Refuel and Reactor Zone Fans SRO Direct alternating Refuel and Reactor Zone Supply and Exhaust Fans lAW 3-Cl-30A, section 6.1 and 3-OI-30B, section 6.1 BOP Alternates Refuel Zone Supply and Exhaust Fans lAW 3-Ol-30A, sections 6.1 6.1 Alternating Refueling Zone Supply and Exhaust Fans
[1] NOTIFY Unit 1 and Unit 2 Operators that the Refuel Zone fans are being alternated.
[4] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in OFF.
[5] CHECK that the two red lights A(B) extinguish and the two green lights A(B) illuminate above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A.
[6] CHECK the red (open) damper position indication lights extinguish and the green (closed) lights illuminate above the following control switches:
- REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
- REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6
[7] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B) to start alternate fans.
[8] CHECK that the two green lights A(B) extinguish and the two red lights A(B) illuminate above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A.
[9] CHECK the red (open) damper position indication lights illuminate and green (closed) lights extinguish above the following control switches:
- REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
- REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6
10
- Pagei2otS8 Simulator Event Guide:
Event 1 Normal: Alternate Refuel and Reactor Zone Fans 6.1 Alternating Refueling Zone Supply and Exhaust Fans (contd)
NOTE A five minute time delay should be observed following Refuel Zone Supply and Exhaust Fan SLOW Start. The time delay allows the discharge dampers to fully open after SLOW start.
[10] IF Refueling Zone Supply and Exhaust Fan FAST speed operation is necessary, THEN: PERFORM the following:
[10.1] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in FAST 3A(FAST3B).
[10.2] CHECK that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A.
[11] CHECK the following conditions:
- SUPPLY FANS FILTER DIFF PRESS Indicator, 3-PDI-064-0022, indicates less than 0.6 inches H2O at the Reactor Building/Refuel Floor Supply fan intake room at El 565.
- REFUELING ZONE STATIC PRESS INTLK, 1-PDS-064-0061 A/C, on refuel floor Panel 25-220 indicates between.(negative) -0.25 inches to -0.40 inches.
briver When contacted as RBAUO, wait 6 minutes and report Supply Fan Filter
. Differential Pressure indicates .25 inches H20 and the Refueling Zone Static Pressure on Panel 25-220 indicates -O3O inches BOP Alternates Reactor Zone Supply and Exhaust Fans lAW 3-OI-30B, sections 6.1 6.1 Alternating Reactor Zone Supply and Exhaust Fans
[5] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in OFF
[6] [lI/C] VERIFY dampers close and fans stop as indicated by illuminated green lights above the following switches:
- REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
- REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
- REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1 A
10 Page l3of-38 Simulator Event Guide:
Event 1 Normal: Alternate Refuel and Reactor Zone Fans 6.1 Alternating Reactor Zone Supply and Exhaust Fans (contd)
[7] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in SLOW A (SLOW B) to start alternate fans.
[8] VERIFY dampers open and fans start as indicated by illuminated red lights above the following switches:
- REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
- REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
- REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1A
[9] IF fast speed Reactor Zone Supply and Exhaust Fan operation is required, five minutes should be allowed after slow start for the discharge dampers to FULLY OPEN, THEN
[9.1] PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-1 1A, in FAST A (FAST B).
[9.2] VERIFY that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A.
[10] VERIFY the following conditions:
[10.1] VERIFY REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-064-0002, on 3-LPNL-925-0213, located at R17-P El 639, indicates between -0.25 inches and -0.40 inches H20.
[10.2] IF REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-64-2, is not between -0.25 inches and -0.40 inches H2O,THEN REFER TO 3-AOI-30B-1, Reactor Building Ventilation Failure.
river When contacted as RBAUO, wait 2 minutes and report Reactor Zone Pressur Differential indicates -0.30 inches
10
- Page-14of38-Simulator Event Guide:
Event 2 Reactivity: Raise Power to 100% with Recirc Flow SRO Notifies ODS of power increase.
Directs Power increase using Recirc Flow, per 3-GOt-i 00-12.
[21] WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes.
REFER TO 3-SR-3.3.5(A) and 3-01-68.
ATC Raise Power w/Recirc, lAW 3-01-68, Section 6.2
[1] IF desired to control Recirc Pumps 3A and/or 3B speed with Recirc Individual Control, THEN PERFORM the following;
- Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM),
3-HS-96-i 5A(1 5B).
AND/OR
- Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM),
3-HS-96-1 6A(1 6B).
[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A &
3B using the following push buttons as required:
RAISE SLOW, 3-HS-96-31 RAISE MEDIUM, 3-HS-96-32 Driver When directed by NRC, insert trigger I for RWCU Leak with failure to Auto isolate
10 Pag& 15-of 38 Simulator Event Guide:
Event 3 Component: RWCU leak with failure to auto isolate Report alarm RWCU LEAK DETECTION TEMP HIGH (3-9-3D Window 17)
ATC A. IF this alarm is received in conjunction with RWCU ISOL LOGIC CHANNEL A TEMP HIGH [3-XA-55-5B, window 32] and RWCU ISOL LOGIC CHANNEL B TEMP HIGH [3-XA-55-5B, window 33j, THEN EXIT this procedure and GO TO 3-ARP-9-5B. Otherwise, CONTINUE in this procedure.
Report alarms RWCU ISOL LOGIC CHANNEL A TEMP HIGH, RWCU ISOL LOGIC_CHANNEL_B_TEMP_HIGH A. VERIFY alarm by checking:
- 1. ATUs on Panel 3-9-83 and 3-9-85.
- 3. Area temperature indications on LEAK DETECTION SYSTEM TEMPERATURE, 3-TI-69-29, on Panel 3-9-21.
B. IF leak is suspected, THEN MANUALLY ISOLATE RWCU or if RWCU automatically isolates, REFER TO 3-AOI-64-2A.
C. IF TIS-69-835A(C) indicates greater than 131 °F, THEN ENTER 3-EOI-3.
ATC Reports RWCU Valve 69-1 failed to isolate CT #3
) Closes 69-1 to stop RWCU Leak CT #3 SRO Directs Penetration Isolated or concurs with the closure of 69-1 SRO Enter EOI-3 and 3-AOI-64-2A ATC 4.1 Immediate Actions
[1] VERIFY automatic actions occur.
[2] PERFORM any automatic actions which failed to occur.
Driver Acknowledge Notifications, when dispatched to AT Us report high t,mperatures ir RWCU_HX room_and temperature lowering
10 Page 6&S8 Simulator Event Guide:
Event 3 Component: RWCU leak with failure to auto isolate BOP 4.2 Subsequent Actions
[5] CHECK the following monitors for a rise in activity:
- AREA RADIATION, 3-RR-90-l, Points 9, 13, and 14 (Panel 3-9-2)
- AIR PARTICULATE MONITOR CONSOLE, 3-MON-90-50, 3-RM-90-55 and 57 (Panel 3-9-2)
- RB, TB, and Refuel Zone Exhaust Rad on CHEMISTRY CAM, MONITOR CONTROLLER, 0-MON-90-361 (Panel 1-9-2)
[6] IF it has been determined that leakage is the cause of the isolation, THEN NOTIFY RADCON of RWCU status.
[7] NOTIFY Chemistry that RWCU has been removed from service for the following evaluations:
- The need to begin sampling Reactor Water
- The need to remove the Durability Monitor from service
[8] IF the isolation cannot be reset, THEN
[9] EVALUATE Technical Requirements Manual Section 3.4.1, Coolant Chemistry, for limiting conditions for operation.
SRO Evaluate Technical Specification 3.6.1.3 and determine Condition A required and TRM 3.4.1. Notifies Chemistry that continuous monitoring is no longer available and to commence sampling per TRM Surveillance 3.4.1.1 Enters EOI-3 on High Secondary Containment Temperature.
Secondary Containment Temperature Monitor and Control Secondary Containment Temperature.
Operate available ventilation, per Appendix 8F.
Answers YES to: Is Any Area Temp Above Max Normal?
Isolate all systems that are discharging into the area Verifies RWCU Isolated Secondary Containment Radiation Monitor and Control Secondary Containment Radiation Levels.
Answers NO to: Is Any Area Radiation Level above Max Normal?
Secondary Containment Level Monitor and Control Secondary Containment Water Levels.
Answers NO to: Is Any Floor Drain Sump Above 66 inches?
AND Answers NO to: Is Any Area Water Level Above 2 inches?
When directed by NRC, insert trigger 5 for Bus Duct Cooling Fan Trip Driver,
10 Rage i7of 38 Simulator Event Guide:
Event 4 Component: Bus Duct Cooling Fan Trip with failure of standby fan to auto start BOP Responds to alarm GEN BUS DUCT FAN FAILURE 7A-31 A. VERIFY Main Bus Cooling Fans, 3-HS-262-1A or 1-HS-262-2A, indicates running on Panel 3-9-8 AND START GEN BUS DUCT HX FAN A(B) using 3-HS-262-1 A(2A), on panel 3-9-8 to start the standby fan.
B. IF no Fans are operating and the Generator is tied to the grid and loaded to greater than the self cooled bus rating of 16,500 amps THEN, IMMEDIATELY INSERT a manual reactor scram, AND TRIP the Main Generator.
C. IF while executing this procedure, the Bus Duct Temperature is at or above the Temperature Excursion limit of 120°C, THEN IMMEDIATELY INSERT a manual reactor scram, AND TRIP the Main Generator.
D. DISPATCH personnel as necessary to check the following:
- 1. Main Bus Cooling Fan on elevation 586 to check fan condition.
- 2. Monitor Bus Duct temperature by available means including using a portable temperature monitor device locally at the 14 in-service thermostats.
REFER to Window 32, Figure.1.
- 3. 480V Unit Board 3A on elevation 586 to check breaker 5C closed.
- 4. 480V Unit Board 3B on elevation 604 to check breaker SC closed.
E. VERIFY the system is operating in accordance with 3-01-47.
BOP Start Standby Bus Duct Cooling Fan B and dispatches personnel SRO Concurs with start or directs start of Bus Duct Cooling Fan B BOP Dispatch personnel to breaker and bus duct cooling fans Driver Breaker for bus duct coollng fan A is tripped, no abnormal indications apparent, if asked to reset breaker, breaker trips aqain, no problems noted at fans Driver When directed byNRC, insert trigger 1 GlorRH pump trip
10 Page 18-of-38 Simulator Event Guide:
Event 5 Component: RR Pump A Trip with power oscillations ATC Respond to numerous alarm and Report Trip of RR Pump A SRO Enter 3-AOl-68-1A Recirc Pump Trip/Core Flow Decrease OPRMs Operable ATC [1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (Otherwise N/A),
[2] IF a single Recirc Pump tripped, THEN CLOSE tripped Recirc Pump discharge valve.
Closes 3A Recirc Pump Discharge Valve ATC [3] IF Region I or II of the Power to Flow Map is entered, THEN IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline.
Report in Region 2 of Power to Flow Map SRO Directs Load Line reduction to <95%
ATC Insert Rods per Emergency shove sheets until <95% Load Line Driver When First Control rod is inserted d&ete Power Oscillations crO2a and crQ2b Driver When directed by NRC3 insert trigger 15 forLeve! 2intrurnent failures (58A and 580) cause HPCI and RCIC to Auto initiat
10 Page 1of3&
Simulator Event Guide:
Event 6 Instrument: Level 2 instrument failures (58A and 58D) cause HPCI and RCIC to Auto initiate BOP Report alarm 3F-29 RX WTR LVL LOW LOW HPCI/RCIC INIT ATC/BOP A. CHECK RPV water level using multiple indications.
Report indicated water level on B instrument is less than -45 inches but other indicators are normal.
ATC/BOP Trip RCIC and HPCI Crew Determine that Level instruments 58A and 58D have failed Driver If dispatched to investigate failures, report 58A is due. to a problem with the slave trip unit LS-3-58A arid still investigating the cause of the 580 instrument problem but it appears to be in the transmitter LT3-58D SRO Technical Specification 3.3.4.2 Condition A, B and C Required Action A Completion Time 14 days Required Action B Completion Time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Required Action C Completion Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 3.3.5.1 Condition A, B, and F Required Action A Completion Time Immediately Required Action B.1 Completion Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Required Action B.2 Completion Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Required Action B.3 Completion Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action F.1 Completion Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Required Action F.2 Completion Time 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> and 8 days 3.3.5.2 Condition A and B Required Action A Completion Time Immediately Required Action B.1 Completion Time 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Required Action B.2 Completion Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.8.1 Condition D and J Required Action D.1 Completion Time 7 Days Required Action J.1 Completion Time Enter 3.0.3 Immediately 3.5.1 Condition C and G Required Action C.1 Completion Time Immediately Required Action C.2 Completion Time 14 Days Required Action G.1 Completion Time 12 Hours 3.5.3 Condition A and B Required Action A.1 Completion Time Immediately Required Action A.2 Completion Time 14 Days Required_Action_B.1_Completion_Time_12_Hours Diver When directed by NRC, insert trigger 20 for HPCI steam leak without isoIation Manually modify fpO2 to Close
10 Page2O of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation Crew Recognize rising HPCI Room Temperatures and Radiation Levels.
HPCI LEAK DETECTION TEMP HIGH A. CHECK HPCI temperature switches on LEAK DETECTION SYSTEM TEMPERATURE, 3-TI-69-29 on Panel 3-9-21.
B. IF high temperature is confirmed, THEN ENTER 3-EOl-3 Flowchart.
C. CHECK following on Panel 3-9-1 1 and NOTIFY RADCON if rising radiation levels are observed:
- 1. HPCI ROOM EL 519 RX BLDG radiation indicator, 3-RI-90-24A.
- 2. RHR WEST ROOM EL 519 RX BLDG radiation indicator, 3-Rl-90-25A.
ATC/BOP VERIFIES HPCI STEAM LINE INBD ISOL VLV, 3-FCV-73-2 AND HPCI STEAM LINE OUTBD ISOL VLV, 3-FCV-73-3 CLOSE.
Attempts to isolate HPCI Steam Supply Valves.
Reports HPCI fails to isolate.
ATC/BOP During attempts to isolate HPCI Steam Supply Valves, report a loss of 3A RMOV
________ Board. (Loop 1 RHR and Loop 1 Core Spray unavailable.)
Crew Contacts personnel to investigate loss of 3A RMOV Board.
Crew Dispatches personnel to transfer RPS A to alternate.
Dr[vei When req uestedL waft 4 minutesánd pIe FP$ A on áIternate irf rO4nd mQ3.
Crew PA announcement to evacuate the HPCI quad or Reactor Building
10
- Page2I-of-38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation SRO Enters EOI-3 on Secondary Containment (Area Radiation or Temperature).
SRO IF Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels are below 72 mr/hr, THEN Restart Reactor Zone and Refuel Zone Ventilation, per Appendix 8F. Defeat isolation interlocks if necessary, Ajendix 8E.
If ventilation isolated and below 72 mr/hr, directs Operator to perform Appendix 8F.
Drh,er If requestedwait3 minutes and report Appendix 8Ecomplete, enter bat appO8é CT #1 SRO Enters EOl-1 RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe.
CT #2 SRO Stops at Stop sign When temperatures in two or more areas are above Max Safe, Then Emeraencv Derressurization is required.
10
-Page 22 otS8 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation CT #1 SRO Enters EOI-1 RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe.
CT #2 SRO Stops at Stop sign When temperatures in two or more areas are above Max Safe, Then Emeraencv Deressurization is reauired.
SRO EOI-3 Secondary Containment (Temperature)
Monitor and Control Secondary Containment Temperature.
Is Any Area Temp Above Max Normal? YES -
Isolate all systems that are discharging into the area except systems required to:
- Suooress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment?
-YES Proceeds to the STOP sign Before any area temp rises to Max Safe (table 5)
Continue:
Crew Monitors for Max Safe Temperatures, reports when two areas are above MAX Safe (H PCI Room greater than 270° F and RHR System II Pump Room greater than 21 5°F)
SRO EOI-3 Secondary Containment (Level)
Monitor and Control Secondary Containment Water Levels.
Is Any Floor Drain Sump Above 66 inches? NO Is Any Area Water Level Above 2 inches? NO -
10
-Page2SotS8 Simulator Event Guide:
Event 7 Major: HPC( Steam Leak without Isolation SRO EOl-3 Secondary Containment (Radiation)
Monitor and Control Secondary Containment Radiation Levels.
Is Any Area Radiation Level Max Normal? - YES Isolate all systems that are discharging into the area except systems required to:
. Be operated by EOls OR
.Suppressa Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES-Proceeds to the STOP sign Before any area radiation rises to Max Safe (table 4)
Continue
10
- Page24of38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation CT #1 SRO Enters EOl-1 RPV Control and directs Reactor Scram before any temperature exceeds MAX Safe based on EOI-3 step SCIT-6.
CT #1 ATC Inserts Reactor Scram, Initiates One Channel of ARI and reports rods out SRO Enters EOl-1 from EOl-3 step SCIT-6 Verify Reactor Scram EOI-1 RC/P Monitor and Control RPV pressure Exits RCIP and enters C-2, Emergency RPV Depressurization, based on Override step RCIP-4.
EOl-1 RC/L Monitor and Control RPV Water Level Verify as Required:
- PCIS Isolations (Groups 1,2 and 3)
- RCIC Exits RO/L and enters 0-5, Level/Power Control, based on override RCIL-3 EOI-1 RCIQ Monitor and Control Reactor Power
- Crew may determine Reactor Subcritical and exit RC/Q, as long as NQBoron has been injected, at any point during execution. If this is done Crew would enter AOl-i 00-1, Reactor Scram, based on override RC/Q-2.
(The following steps will be executed through AOI-100-1 if RCIQ exited)
- Verify Reactor Mode Switch is in Shutdown
- Initiate second channel of ARI
- Verify Recirc Pump Runback (Pump speed 480rpm or less)
- Answers No to is Reactor Power above 5% or Unknown (The Following steps N/A if RC/Q exited)
- Before Suppression Pool Temperature rises to 1 bE, determines Boron Injection is Required.
- Initiates SLC per Appendix 3A
10
- Page 25 of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation
= SRO EOI-1 RC/Q (cont)
Inhibit ADS Verify RWCU System Isolation Answers Yes to is SLC injecting into the RPV Stops at step RC/Q-1 8 until SLC has injected into the RPV to a tank level of 43%, then exits RC/Q and enters AOl-i 00-1 Trips the SLC pump when SLC tank level drops to 0%
ATC Initiates Second Channel of ARt and reports no rod movement.
Verifies Recirc Pump at 480 rpm or less.
Reports Reactor Power less than 5% during Scram Report Should insert IRMs to determine if Reactor is subcritical BOP/ATC Verify and Report PCIS Isolations, ECCS and RCIC If directed, Initiate SLC per Appendix 3A, Inhibit ADS, and Verify RWCU System Isolation (These steps N/A if RC/Q exited and AOl-i 00-1 entered)
10 Page 26 of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation BOP/ATC Appendix 3A
- 1. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A13B, control switch in START PUMP 3A or START PUMP 3B position.
- 2. CHECK SLC System for injection by observing the following:
- Selected pump starts, as indicated by red light illuminated above pump control switch.
- SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 20).
- System flow, as indicated by 3-IL-63-1 1, SLC FLOW, red light illuminated on Panel 3-9-5,
- SLC INJECTION FLOW TO REACTOR Annunciator in alarm on Panel 3-9-5 (3-XA-55-5B, Window 14).
- 3. IF Proper system operation CANNOT be verified, THEN RETURN to Step 1 and START other SLC pump.
- 4. VERIFY RWCU isolation by observing the following:
- RWCU Pumps 3A and 3B tripped
- 3-FCV-69-12, RWCU RETURN ISOLATION VALVE closed.
- 5. VERIFY ADS inhibited.
- 6. MONITOR reactor power for downward trend.
- 7. MONITOR 3-LI-63-1A, SLC STORAGE TANK LEVEL, and CHECK that level is droinci aroximateIv 1% ier minute.
10 Page 2-7 ot 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation SRO Enters C-5 from EOI-1 step RC/L-3 Override Step C5-1, states that IF Emergency Depressurization is required, THEN continue at step C5-19, however, if the SRO has not determined that ED is required at this time then he will continue at step C5-2 (below)
Inhibit ADS Answers Yes to is any Main Steam Line Open Bypass the following Isolation Interlocks:
. MSIV Low Low Low RPV Water Level (APPX (8A)
. RB Ventilation Low RPV Water Level (APPX 8E)
Crosstie CAD to DW Control Air, if necessary (APPX 8G) (Step N/A)
Driver. When requested for appendix 8A and 8E wait 4 minutes and insert trigger ?8 fot:
bat appO8ae and report complete SRO Answers No to is Reactor Power Above 5% or Unknown Establishes Reactor Water Level Band between -180 and +51 inches utilizing available injection sources listed on step 05-15.
Upon determination that Emergency Depressurization is required continues at step Cs-i 9 and enters C-2 by direction of EOI-3 step SC/T-9 and from EOI-i step RCIP 4 and directs Crew to Stop and Prevent all Injection Sources to the RPV Except from RCIC, CRD and SLC per step C5-20, in accordance with Appendix 4.
BOP/ATC Inhibits ADS (if not already done per Appendix 3A)
If directed, dispatches personnel to perform Appendices 8A and 8E.
Maintains Reactor Water Level until directed to Stop and Prevent per Appendix 4.
When directed performs Appendix 4 to Stop and Prevent all Injection Sources to the RPV Except from RCIC, CRD and SLC
10 Page 28otS8 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation BOP/ATC Appendix 4
- 1. PREVENT injection from HPCI by performing the following:
- a. IF HPCI Turbine is NOT at zero speed, THEN PRESS and HOLD 3-HS-73-18A, HPCI TURBINE TRIP push-button.
- b. WHEN HPCI Turbine is at zero speed, THEN PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP control switch in PULL TO LOCK and RELEASE 3-HS-73-18A, HPCI TURBINE TRIP push-button.
- 3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.
- 4. PREVENT injection from LPCI SYSTEM I by performing the following:
NOTE Injection may be prevented by performing EITHER step 4.a or step 4.b.
- a. Following automatic pump start, PLACE RHR SYSTEM I pump control switches in STOP.
- b. BEFORE RPV pressure drops below 450 psig,
AND
- 5. PREVENT injection from LPCI SYSTEM II by performing the following:
NOTE Injection may be prevented by performing EITHER step 5.a or step 5.b.
- a. Following automatic pump start, PLACE RHR SYSTEM II pump control switches in STOP.
10 Page29 of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation BOP/ATC Appendix 4 (continued)
- b. BEFORE RPV pressure drops below 450 psig,
AND
- 6. PREVENT injection from CONDENSATE and FEEDWATER by performing the following:
- a. IF Immediate injection termination from a reactor feedwater pump is required, THEN PERFORM step 6.d for the desired pump.
- b. LOWER RFPT 3A(3B)(3C) speed to minimum setting (approximately 600 rpm) using ANY of the following methods on Panel 3-9-5:
- Using 3-LIC-46-5, REACTOR WATER LEVEL CONTROL, in MANUAL AND individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C) SPEED CONTROL in AUTO, OR
- Using individual 3-SIC-46-8(9)(10), RFPT 3A(3B)(3C)
SPEED CONTROL in MANUAL, OR
- Using individual 3-HS-46-8A(9A)(1OA), RFPT 3A(3B)(3C)
SPEED CONT RAISE/LOWER switch in MANUAL GOVERNOR.
- c. CLOSE the following valves BEFORE RPV pressure drops below 450 psig:
- 3-FCV-3-1 9, REP 3A DISCHARGE VALVE
- 3-ECV-3-12, REP 3B DISCHARGE VALVE
- 3-FCV-3-5, REP 30 DISCHARGE VALVE
- 3-LCV-3-53, RFW START-UP LEVEL CONTROL
- d. TRIP RFPTs as necessary to prevent injection by DEPRESSING the following push-buttons:
- 3-HS-3-125A, REPT 3A TRIP
- 3-HS-3-1 51 A, REPT 3B TRIP
- 3-HS-3-176A, REPT 3C TRIP.
10 Page 3&-oF 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation CT #2 SRO Determines Emergency Depressurization is required and enters C-2 Answers No to will the reactor remain subcritical under all conditions. Waits until he receives the report that Appendix 4 is complete.
Answers Yes to is Suppression Pool Level above 5.5 ft Directs All ADS Valves opened Answers Yes to can Six ADS Valves be opened Stops execution of C-2 until:
- The Reactor is Subcritical and No Boron has been injected into the RPV Stops execution of execution of C-2 until Shutdown Cooling RPV Pressure Interlocks are clear Maintain RPV in Cold Shutdown Der Appendix 17D CT#2 BOP/ATC Reports when Appendix 4 is complete Reports Suppression Pool Level in Feet when Directed Opens and Verifies Open ALL ADS Valves when directed SRO Upon commencement of Emergency Depressurization Continues in C-5 at step C521 Answers Yes to are at least 2 MSRVs open per C-2, Emergency RPV Depressurization Stops until RPV Pressure is below MARFP (1 9opsig with 6 MSRVs open)
Then continues Directs crew to StarL and Slowly raise RPV Injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSH limits and Suppression Pool level per Appendix 6A or per Appendix 6C
10 Page3I of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation
= BOP/ATC Appendix 6A
- 1. VERIFY CLOSED the following Feedwater heater return valves:
- 3-FCV-3-71, HP HTR 3A1 LONG CYCLE TO CNDR
- 3-FCV-3-72, HP HTR 3B1 LONG CYCLE TO CNDR
- 3-FCV-3-73, HP HTR 3C1 LONG CYCLE TO CNDR
- 2. VERIFY CLOSED the following REP discharge valves:
- 3-FCV-3-19, RFP 3A DISCHARGE VALVE
- 3-FCV-3-12, RFP 3B DISCHARGE VALVE
- 3-FCV-3-5, RFP 3C DISCHARGE VALVE
- 3. VERIFY OPEN the following drain cooler inlet valves:
- 3-FCV-2-72, DRAIN COOLER 3A5 CNDS INLET ISOL VLV
- 3-FCV-2-84, DRAIN COOLER 3B5 CNDS INLET ISOL VLV
- 3-ECV-2-96, DRAIN COOLER 3C5 CNDS INLET ISOL VLV
- 4. VERIFY OPEN the following heater outlet valves:
- 5. VERIFY OPEN the following heater isolation valves:
- 6. VERIFY OPEN the following REP suction valves:
- 3-FCV-2-83, REP 3A SUCTION VALVE
- 3-FCV-2-95, REP 3B SUCTION VALVE
- 3-FCV-2-108, REP 3C SUCTION VALVE
- 7. VERIFY at least one condensate pump running.
- 8. VERIFY at least one condensate booster pump running.
- 9. ADJUST 3-LIC-3-53, RFW START-UP LEVEL CONTROL, to control injection (Panel 3-9-5).
10 Page32of3S Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation BOP/ATC Appendix 6C
- 1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 3-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.
- 4. VERIFY CLOSED the following valves:
- 3-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE
- 5. VERIFY RHR Pump 3B and/or 3D running.
- 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.
- 7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 3-FCV-68-3, RECIRC PUMP 3A DISCHARGE VALVE.
- 10. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers discharging to the RPV.
- 11. THROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:
10
-Page 33 of38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation
= BOP/ATC Starts and Slowly raises RPV Injection to Restore and Maintain RPV Water Level above -180 inches irrespective of pump NPSH limits and Suppression Pool level per Appendix 6A or per Appendix 6C SRO EOl-1 RC/Q steps RCIQ-20 and RCIQ-21 Reset ARI Defeat ARI Logic Trips if necessary (APPX 2) (This step is N/A, however, crew may choose to perform this step)
Insert Control Rods by performing Appendix 1 F and 1 D Appendix 1 F: Scram Valves Opened but SDV is Full
- 2) Drain SDV
- 3) Recharge Accumulators
- 4) Initiate Reactor Scram Appendix 10: Manual Control Rod Insertion Method
- 1) Drive Control Rods. Bvass RWM if necessary BOP/ATC Dispatch personnel to perform Appendix 2(N/A) and outside portions of Appendix iF.
Dispatch personnel to close 3-FCV-85-586 (while awaiting completion of Appendix 1 F)
Drive Rods per Appendix 1 D while waiting for completion of Appendix 1 F
10 Page 34 of 3S Simulator Event Guide:
Event 7 Major:HPCI Steam Leak without Isolation
= ATC Appendix iF
- 3. VERIFY OPEN Scram Discharge Volume vent and drain valves.
- 4. DRAIN SDV UNTIL the following annunciators clear:
- WEST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 1)
- EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 29).
- 5. DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER ISOL.
- 6. WHEN CRD Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.
- 7. CONTINUE to perform Steps 1 through 6 UNTIL ANY of the following exists:
- ALL control rods are fully inserted, OR
- NO inward movement of control rods is observed, OR
- SRO directs otherwise.
Drivet. When directed to perform Appendix 2 and outside portions of Appendix 1 F wait minutes. Insert Triggers 26, 27, and 29 then report completion.
If directed to close 3-FCV-85586wait3 minutes then insert nrt rdO6 close. Thed report compIetion If/When directed to re-open 3-FCV85-586 wait a minutes then insert mrf rdO6 open. Then report completion.
10 Page 35 of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation ATC Appendix 1 D
- 1. VERIFY at least one CRD pump in service.
- 2. IF Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGING WATER SOV
- 3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.
- 4. BYPASS Rod Worth Minimizer.
- 5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:
- a. SELECT control rod.
- b. PLACE CRD NOTCH OVERRIDE switch in EM ERG ROD IN position UNTIL control rod is NOT moving inward.
- c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.
- 6. WHEN NO further control rod movement is possible or desired, THEN DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGING WATER SOV (RB NE. El 565 ft.
ATC Continue performance of Appendix 1 F andl D until all rods inserted OR Until EOI-1 RC/Q is exited due to Reactor determined to be Subcritical at which point continue to insert rods per 3-AOI-100-1 and 3-01-85
10
- Page3&of38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation
= SRO Executes all legs of EOI-2 concurrently EOl-2 DW/T Monitor and control Drywell Temperature below 1 60F using available Drywell Cooling Answers Yes to can Drywell Temperature be maintained below 1 60F EOl-2 PC/P Monitor and control Primary Containment pressure below 2.4 psig using the vent system (APPX 12) as necessary Answers Yes to can Primary Containment pressure be maintained below 2.4 psig EOl-2 PC/H Monitor and control Drywell and Suppression Chamber
- Hydrogen at or below 2.4%
AND
- Oxygen at or below 3.3%
Using the Nitrogen Makeup System (APPX 14A)
EOI-2 SP/T Monitor and control Suppression Pool temperature below 95F using available Suppression Pool Cooling (APPX 17A) as necessary Answers No to can Suppression Pool temperature be maintained below 95 F (This is assuming Emergency Depressurization is complete and Reactor Water Level has been restored, it Emergency Depressurization has not been conducted yet, the answer will be Yes. It Reactor Water Level has not been restored yet, after Emergency Depressurization, this is not a priority.)
Directs Line up of all available Suppression Pool Cooling using only RHR pumps not required to assure adequate core cooling by continuous injection (APPX 1 7A) (After Emergency Depressurization complete and Reactor Water level restored)
BOP Performs Appendix 17A to place Suppression Pool cooling in service after Emergency Depressurization and restoration of Reactor Water level.
10 Page-37 of 38 Simulator Event Guide:
Event 7 Major: HPCI Steam Leak without Isolation BOP Appendix 17A
- 1. If Adequate core cooling is assured, OR Directed to cool the Suppression Pool irrespective of adequate core cooling, Then BYPASS LPCI injection valve auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD INJ VLV BYPASS SEL in BYPASS.
- 2. PLACE RHR SYSTEM 1(11) in Suppression Pool Cooling as follows:
- c. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- d. If Directed by SRO, Then PLACE 3-XS-74-1 22(1 30), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.
CTMT SPRAY/CLG VLV SELECT in SELECT.
- f. If 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, Then VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.
10 Page 38 of 38 Simulator Event Guide:
Event 7 Major:HPCI Steam Leak without Isolation
= BOP Appendix 17A (cont)
- h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.
I. THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, to maintain EITHER of the following as indicated on 3-Fl-74-50(64), RHR SYS 1(11) FLOW:
- Between 7000 and 10000 gpm for one-pump operation.
- At or below 13000 gpm for two-pump operation.
- j. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) MIN FLOW VALVE.
I. NOTIFY Chemistry that RHRSW is aligned to in-service RHR Heat Exchangers.
- m. If Additional Suppression Pool Cooling flow is necessary, Then PLACE additional RHR and RHRSW pumps in service using Steps 2.b through 2.1.
SRO Emergency Plan Classification is 3.1-S.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
All but six Control Rods are inserted Emergency Depressurization complete Reactor Level is restored and maintained
10 Page 39 of 38 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:
DG 3A is Out of Service Operations/Maintenance for the Shift:
Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 3-01-308 and raise power to 100% with Recirc Flow.
Unit 1 and 2 are 100% Power Unusual Conditions/Problem Areas:
None