ML112800671
| ML112800671 | |
| 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: ML112800671 (275) | |
Text
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 1 - Op-Test No.: 1108 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 Maif. 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 RHRISW pump C3 trip TS-SRO R-ATC 3 Commence a power increase with rods R-SRO fic-85-1 1 C-ATC CRD Controller Failure 0-100(L) C-SRO C-BOP Steam Packing Exhauster 2A trip and failure 213 discharge 5 Batch file C-SRO valve 6 CUO4 I-ATC RWCU Leak with failure to Auto isolate TS-SRO 7 PC14 M-ALL Non-isolable leak on torus 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
Appendix 0 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, JAW 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 JAW with ARPs.
- 6. 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-A0I-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.
- 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 EOI-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-E0I-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 Critical Tasks Three CT#1-When Suppression Pool level cannot be maintained above 11.5 feet the US directs the Reactor scrammed and either 1) anticipates Emergency Depressurization and depressurizes using bypass valves or 2) directs Emergency Depressurization before suppression pool level lowers to 11.5 feet.
- 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-Wben Suppression Pool Level cannot be maintained above 12.75 feet HPCI secured to prevent damage prior to reaching 12.75 feet.
- 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 0 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
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 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)
Appendix 0 Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER KJA RO SRO 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 271000A1.0l 3.3 3.2 SRO S-047-AB-3 RWCU Leak with Failure to Auto Isolate RO U-069-AL-10 223002A2.03 3.0 3.3 SRO S-000-EM-12 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
Appendix 0 Scenario Outline Form ES-D-1 Simulator Instructor
- 1. Setup IC-90 110801 Preference File F3 bat NRC/110801 F4 imf sw03j F5 mrf swO6 close F6 trg! Eli F7 bat NRC/110202-i F8 imfpci4 10036010 110801 Batch File
- RWCU seal leak no auto iso imf cuO6 imfcu04 (el 0)100 300 50 10 SRV overrides Trg e3 NRC/singleelemeut Trg e3 = bat NRC/ii0202-4 br zdihs7330a close Oir ypobkrrhrswpa3 fail_ccoil br zlohs2385a [1] off ior zlohs6635a{i] on ior ypomtrspea (eli 0) fail_controlpower ior ypovfcv6635 (eli 0) failpower_now trg 10 NRC/spe trg 10 = bat NRC/i 10801-1
- Steam packing blower trip 110801-i Dor ior ypovfcv6635 Dor ior zlohs6635a TRG SPE Zdihs6635a[3]. Eq. 1
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC-i Op-Test No.: ii 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 RHRJSW pump C3 trip TS-SRO R-ATC 3 Commence a power increase with rods R-SRO flc-85-1 1 C-ATC CRD Controller Failure 0-100(L) C-SRO C-BOP Steam Packing Exhauster 2A trip and failure 2B discharge 5 Batch file C-SRO valve 6 CUO4 I-ATC RWCU Leak with failure to Auto isolate TS-SRO 7 PC14 M-ALL Non-isolable leak on torus 8 IOR C HPCI minimum flow valve will not open C All SRVs except 3 fail to open for Emergency 9 JOR Depressurization (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or
Appendix D Scenario Outline Form ES-D-1 Critical Tasks - Three CT#1-When Suppression Pool level cannot be maintained above 11.5 feet the US directs the Reactor scrammed and either 1) anticipates Emergency Depressurization and depressurizes using bypass valves or 2) directs Emergency Depressurization before suppression pool level lowers to 11.5 feet.
- 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 prior to reaching 12.75 feet.
- 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*
Appendix B Scenario Outline Form ES-B-i 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
Appendix D Scenario Outline Form ES-fl-i 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 JAW with ARPs.
- 6. The ATC will respond to RWCU alarms indicating a leak and RWCU will fail to isolate. The ATC will isolate RWCU and take actions LAW 2-A0I-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.
- 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, HPCJ MN FLOW VALVE will fail to open. The crew will open the RCIC CST SUCTION VALVE and RCIC PUMP MN 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-E0I-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 EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 3 EOIs used: List (1-3)
I EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix B Scenario Outline Form ES-B-i Scenario Tasks TASK NUMBER KJA RO SRO Shutdown Suppression Pool Cooling RO U-92B-NO-05 219000A4.01 3.8 3.7 EECW Pump Trip RO U-067-NO-12 400000A2.O1 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 271000A1.0l 3.3 3.2 SRO S-047-AB-3 RWCU Leak with Failure to Auto Isolate RO U-069-AL-10 223002A2.03 3.0 3.3 SRO S-000-EM-12 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 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 1S 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-Ol-47C Seal Steam System Rev 24 2-ARP-9-3D, W17 RWCU Leak Detection Temperature High Rev 28 2-A0l-64-2A Group 3 RWCU Isolation Rev 25 TS 3.6.1.3 Primary Containment Isolation Valves Amd 212 TRM 3.4.1 Coolant Chemistry Rev 21 2-ARP-9-3B, W15 Suppression Chamber Water Level Abnormal Rev 28 2-EOl-3 Secondary Containment Control Rev 12 2-E0l-2 Primary Containment Control Rev 12 Suppression Pool Water Inventory Removal and 2-E0I-App-18 Rev 8 Makeup 2-E0l-1 RPV Control Flowchart Rev 12 2-AOl-i 00-1 Reactor Scram Rev 95 2-E0I-App-5A Injection Systems Lineup Condensate/Feedwater Rev 9 2-E0l-App-6A Injection Subsystems Lineup Condensate Rev 4 2-E0I-2-C-2 Emergency RPV Depressurization Revision 6 Alternate RPV Pressure Control Systems Main 2-E0I-App-1 H Rev 6 Condenser EPIP-1 Emergency Classification Procedure Revision 46 EPIP-4 Site Area Emergency Revision 32
1 Page7.cf3 Console Operator Instructions A. Scenario File Summary 110801 Preference File F3 bat NRC/110801 F4 imf swO3j F5 mrf swO6 close F6 trg! Eli F7 bat NRC/110202-l F8 imfpcl4 10036010 110801 Batch File
- RWCU seal leak no auto iso imf cuO6 imf cuO4 100 300 50 10 SRV overrides Trg e3 NRC/singleelement Trg e3 = bat NRC/110202-4 br zdihs7330a close Oir ypobkrrhrswpa3 fail_ccoil br zlohs2385ajij off ior zlohs6635a[i] on ior ypomtrspea (eli 0) fail_control_power ior ypovfcv6635 (eli 0) fail_power_now trg 10 NRC/spe trg i0=batNRC/110801-1
- Steam packing blower trip 110801-1 Dor iorypovfcv6635 Dor ior zlohs6635a TRG SPE Zdihs6635a[3j. Eq. 1 Scenario 1 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-100-12
1 Pageaaf 35 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 86 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, acknowledge removing Suppression Pool Cooling from 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 CLG/TEST 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 Pag 9 of 3 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.
D dhfleud BOP [5] CLOSE RHR SYS 1(11) SUPPR POOL CLG/TEST VLV, 2-FCV-74-59(73).
[6] WHEN RHR SYS 1(11) SUPPR POOL CLGJTEST 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 CHBRIPOOL 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.
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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.
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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 03,
- 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 I shall be addressed.
[1] To line up RHRSW Pump Cl for EECW System operation, PERFORM the following:
[1 .1] 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 Pel1Qf35.
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 & C2 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.
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- Page I2of35. -
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 runnin9 to be operable for EECW.
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hornl SRO Evaluate Technical Specification 3.7.2 before the Cl EECW Pump is aligned (D 4kiw (-.\
Condition A: Jw required EECW pumps inoperable. (A3 and C3)
Required ActionA -------%1-- )
R4-re. vi 7 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 Fagt3Qf35 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-1 00-1 2.
[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:
22-31...00 to 12 30-31...00 to 48 14-31...00 to 16 22-39...00 to 16 30-39...00 to 12 30-15...00 to 16 22-23...00 to 16 38-21...OOtol2 46.31...OOtol6 38.23...OOtol6 30-23... 00 to 12 30-47... 00 to 16 38-39. ..00 to 16 NOTES Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches.
When continuously withdrawing a control rod to a position other than position 48, the CRD Notch Override Switch is held in the Override position and then the CRD Control Switch is held in the Rod Out Notch position.
Both switches should be released when the control rod reaches two notches prior to its intended position.
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].
1 Page 14oL35 Simulator Event Guide:
Event 3 Reactivity: Raise Power with Control Rods
[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.
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 light extinguishes.
PI Eitjjnja enIi I ó-1dLi
1 Pag15of35 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.
1 Fe 1 6QL35 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.
1 Pagei7of35.
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.
Driyer
1 Pag18&35 Simulator Event Guide:
Event 4 Component: CRDH Controller Failure ATC Report Alarm 5A-1O CRD ACCUM CHG WTR HDR PRESS HIGH A. VERIFY pressure high on CRD ACCUM CHG WTR HDR 2-PI-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 P1v
1 Pt9Qf5.
Simulator Event Guide:
Event 5 Component: Steam Packing Exhauster 2A trip and failure 2B discharge valve 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.
NOTE: SPE.B BloWer indiation williave BRed and Green1iht. in order for Red!.
llghtn1y1fldióation thecrew wuldhavetostop the KSPE. lAW 2-01-470 Driver When dispatché4,wái5 minutes adreport iio obvious pbIms atSPE or Brker rieedtobedistiactedJ Dr17i When directd by NJQJ insert for RWCU,Leak with allure t Auto iolate (irnfvu6) imf cuO4 100 33J0 50)
1
- Pge2Q of 35 Simulator Event Guide:
Event 6 Instrument: RWCU leak with failure to auto isolate Report alarm RWCU LEAK DETECTION TEMP HIGH (2-9-3D Window 17)
A. IF this alarm is received in conjunction with RWCU ISOL LOGIC CHANNEL A TEMP HIGH [2-XA-55-5B, window 32] and RWCU ISOL LOGIC ATC CHANNEL B TEMP HIGH [2-XA-55-5B, window 33], THEN EXIT this procedure and GO TO 2-ARP-9-5B. Otherwise, CONTINUE in this procedure.
Report alarms RWCU ISOL LOGIC CHANNEL A TEMP HIGH, RWCU ISOL ATC LOGIC CHANNEL B TEMP HIGH A. VERIFY alarm by checking:
- 1. ATUs on Panel 2-9-83 and 2-9-85.
- 3. Area temperature indications on LEAK DETECTION SYSTEM TEMPERATURE, 2-Tl-69-29, on Panel 2-9-21.
B. IF leak is suspected, THEN MANUALLY ISOLATE RWCU or if RWCU automatically isolates, REFER TO 2-AOI-64-2A.
C. IF TIS-69-835A(C) indicates greater than 131°F, THEN ENTER 2-EOI-3.
ATC Reports RWCU Valve 69-1 failed to isolate T#3 ATC 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 2-AOI-64-2A 4.1 Immediate Actions
[1] VERIFY automatic actions occur.
[2] PERFORM any automatic actions which failed to occur.
Dye cknowIedgjrcons when cippatched toATlJs repojjiigh tenpératurs ir RWCUHX room and femperature lowerihgj 4.2 Subsequent Actions
[5] CHECK the following monitors for a rise in activity:
- AREA RADIATION, 3-RR-90-1, 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)
[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 I.
- The need to remove the Durability Monitor from service
1 Page al t35 -
Simulator Event Guide:
Event 6 Instrument: RWCU leak with failure to auto isolate
[8] IF the isolation cannot be reset, THEN ATC/BOP
[9] EVALUATE Technical Requirements Manual Section 3.4.1, Coolant Chemistry, for limiting conditions for operation.
ATC/BOP Inserts substitute data per 2-01-69 7.0 SYSTEM SHUTDOWN 7.1 ICS Temperature Point Subsitution for Heat Balance
[1] IF removing Reactor Water Cleanup System from service when operating at power, THEN PERFORM RWCU ICS Temperature Point Substitution for Heat Balance adjustments:
NOTE The following values are to be substituted for RWCU Inlet and Outlet temperatures so RWCU parameters provide conservative input to the Integrated Computer System (ICS) thermal power calculation.
a 625 degrees F for 69-6A, RWCU LOOP INLET TEMP.
a 420 degrees F for 69-6D, RWCU LOOP OUTLET TEMP.
Evaluate Technical Specification 3.6.1.3 and determine Condition A required and SRO TRM 3.4.1. Notifies Chemistry that continuous monitoring is no longer available and to commence sampling per TRM Surveillance 3.4.1.1 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
LCO 3.8.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.8.1, HPrimary Containment Isolation Instrumentation,
1
.P9,e22of35 Simulator Event Guide:
Event 6 Instrument: RWCU leak with failure to auto isolate CONDITION REQUIRED ACTION COMPLETION TIME A. --- NOTE-------- A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated AND automatic valve, closed SRO manual valve, blind 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured.
MSIV leakage not within limits.
AND (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 NOTE -
Isolation devices in high radiation areas may be verified by use of administrative means.
Verify the affected Once per3l days penetration flow path is for Isolation isolated. devices outside primary containment
1
- Page23.çf35 Simulator Event Guide:
Event 6 Instrument: RWCU leak with failure to auto isolate TR 34 REACTOR COOLANT SYSTEM TR 3.4.1 Coolant Chemistry SRO LcO 3.4.1 Reactor coolant chemistry shaft be maintained within the limits of Table 3.4.1-1.
APPLICABILITY: According to Table 3.4.1-1 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3,4,1.1 -----------------------NOTE------------------------- Continuously Not required when there is no fuel in the reactor vessel.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when Monitor reactor coolant conductivity, the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or S OR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the continuous conductivity monitor is inoperable and the reactor is in MODE 4 or 5
1 Page 24 of 35 - -
Simulator Event Guide:
Event 6 Instrument: RWCU leak with failure to auto isolate 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?
DR1yR When dircted PtNC leak from the torus.
1 PaQZQf3 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.
Øjy WJen. d pptcJdiait4 njnutes and reportfWater level i4 inches and rising in he $outeastQpa f1cJng in1rç1e ThrusAre& Unab1p trmjp source of(he4eak SRO Enters E0I-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
- - PageZ6of35 Simulator Event Guide:
Event 7 Major: Non-Isolable leak on Torus SRO Enters EOl-3 on Flood Alarms EOI-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 EOl-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 When Suppression Pool Level cannot be maintained above 12.75 feet, HPCI secured to prevent damage.
SRO Sets a Value for HPCI to place in Pull to Lock, prior to 12.75 feet.
ATC/BOP Places HPCI in Pull to Lock, before Suppression Level lowers to 12.75 feet.
1 Pg.27.of35 --
Simulator Event Guide:
Event 8 C: HPCI minimum flow valve will not open SRO Directs Appendix 18 BOP Appendix 18
- 6. IF Directed 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:
- 2) OPEN 2-FCV-71-34, RCIC PUMP MIN FLOW VALVE.
BOP Attempts to makeup water to the Suppression Pool using HPCI; 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 2-FCV-73-30 fails closed when the Torus leak is inserted, crew will dispatch personnel to investigate Acknowledge investjgation and provide no further informationi SRO Determines a trigger value for inserting a Reactor Scram on lowering Suppression CT #1 Pool Water Level and enters EOI-1, Scrams Reactor 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 P 28 ot3 Simulator Event Guide:
Event 7 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-O100, RHR SYS I U-i DISCH XTIE Breaker compartment is maintained in the OPEN position as an Appendix R requirement
- d. NOTIFY Unit I Operator to perform the following
- 2) OPEN 1-FCV-23-57, STANDBY COOLANT VALVE FROM RHRSW (Unit 1, Panel 1-9-3).
- 3) DISPATCH personnel to place 2-BKR-074-O100, RHR SYS I U-i DIXCH XTIE in ON (480V RMOV BD 1B, Compartment i9A).
rJier WJierersonI,e1 raft 6 mintehecIo btker an reporf, Ietepvemde for brealcer control power. Whenhquestç 1-FCy-23-52is closed When eqjp 9L-YJflmn9 f&tion só9 open and reporf 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).
1 Pae29Qfa5 -
Simulator Event Guide:
Event 7 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.
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 EOI-1 from EOI-2 step SPIL-7 Verify Reactor Scram EOI-1 RCIL 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 EOl-1 RCIQ Monitor and Control Reactor Power
. Crew will exit RC/Q and enter 2-AOl-i 00-1 based on RC/Q-2.
SRO SRO expected to lower pressure band to commence cooldown prior to anticipating Emergency Depressurization.
SRO May Anticipate Emergency Depressurization and Rapidly Depressurize using CT #1 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-i 00-1 SRO EOI-1 RCIP 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 P3Q_of Simulator Event Guide:
Event 7 Major: Non-Isolable leak on Torus ATC 2-AOl-I 00-1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AIS3A and 2-HS 5AIS3B, 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-Xl-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-EOI-App-5A to maintain RPV water level in band 2-EOI-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)(1O), RFPT 2A(2B)(2C) SPEED
[_____
CONTROL in AUTO.
1
_Pag 31 of 3 Simulator Event Guide:
Event 7 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-EOI-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 RFP discharge valves:
- 2-FCV-3-19, RFP 2A DISCHARGE VALVE
- 2-FCV-3-12, REP 2B DISCHARGE VALVE
- 2-FCV-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-FCV-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 32 of 35 Simulator Event Guide:
Event 9 Component: All SRVs except 3 fail to open for Emergency Depressurization When Suppression Pool level cannot be maintained above 1 1.5 feet the US determines that Emergency Depressurization is required; RO initiates Emergency CT #1 SRO Depressurization as directed by US.
When Emergency Depressurization is required exits RC/P and enters C-2, SRO 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 BOP Opens ADS Valves CT #1 Opens additional MSRVs as necessary in an attempt to establish 6 MSRVs open Identifies only 3 valves open and informs SRO
1 Page 33 of 35 Simulator Event Guide:
Event 7 Major: Non-Isolable leak on Torus SRO Directs BOP to Rapidly Depressurize the RPV to less than 80 psi above CT #1 Suppression Chamber pressure utilizing App 1 1 H BOP 2-EOI-App-1 I 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 n 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.
T #1 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 34 of 35 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
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
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1 .3.5(A)
UNIT 2 - - REV 0021 H
ATTACHMENT 2 (Page 1 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-31 12 00 N/A 30-39 12 00 N/A 38-31 12 00 N/A 30-23 12 00 N/A 30-3 1 48 00 N/A 30-15 16 00 N/A 46-3 1 16 00 N/A 30-47 16 00 N/A 14-3 1 16 00 N/A 22-23 16 00 N/A 38-23 16 00 N/A 38-39 16 00 N/A 22-39 16 00 N/A 14-23 48 00 N/A 14-39 48 00 N/A 46-39 48 00 N/A 46-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously tOO. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: / Issued by I Unit Supervisor Date Reactor Engineer Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1.3.5(A)
UNIT 2 REV 0021 ATTACHMENT 2 (Page 2 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-47 48 00 N/A 38-47 48 00 N/A 38-15 48 00 N/A 22-15 48 00 N/A 14-47 48 00 N/A 46-47 48 00 N/A 46-15 48 00 N/A 14-15 48 00 N/A 06-3 1 48 00 N/A 30-55 48 00 N/A 54-3 1 48 00 N/A 30-07 48 00 N/A 06-39 48 00 N/A 54-39 48 00 N/A 54-23 48 00 N/A 06-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously to 00. Insertion may stop after completion of any 2roun.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: / Issued by /
Unit Supervisor Date Reactor Engineer Date
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U2 NRC Exam I Unit 2 Rod Pattern Adjustment
I. BEN Reactivity Control Plan Attachment 7 (Page 1 of 2)
Reactivity Control Plan Form BEN Unit: 2 Valid Date(s): 8/7/11 8/19/11 Reactivity Control Plan #: U2 NRC Exam I Are Multiple Activations Allowed: No (If yes, US may make additional copies)
Prepared by I Reviewed by: /
Reactor Engineer Date Qualified Reactor Engineer Date Approved by: / Concurrence: I RE Supervisor Date WCC/Risk/US SRO Date Approved by: I Authorized by: I Ops Manager or Supt. Date Shift Manager Date RCP Activated: / RCP Terminated: I Unit Supervisor Date Unit Supervisor Date Title of Evolution: Unit 2 Rod Pattern Adjustment PurposelOverview of Evolution: Adjust Rod Pattern for 100% power operation Maneuver Steps
- 1. Withdraw Control rods lAWAttachment 2 provided by Reactor Engineer.
- 2. Increase flow to 100% power.(No Ramp Rate limits)
BFN Reactivity Control Plan Attachment 7 (Page 2 of 2)
Reactivity Control Plan Form Operating Experience and General Issues: U2 NRC Exam I Previously known control rod issues:
4 172292 05/28/2009 Control Rod 46-27 double notched during the performance of the Unit 2 sequence exchange, 00 to 04.
4 150002 08/10/2008 During power ascension activities, control rod 46-27 double notched from position 00 to 04.
4 149981 08/09/2008 Control Rod 38-35 double notched during control rod withdrawal from 00 to 04.
4 148263 07/12/2008 While pulling control rods during U2 startup, CR 38-03 double-notched twice. 10 to 14 and 14 to 18 Cautions/Error Likely Situations/Special Monitoring RequirementslContingencies:
- Rod Out Notch Override is authorized, for Rod Out Notch Override follow the guidance in 2-01-85 section 6.6.4.
- This plan is NOT valid if the unit is operating with a suspected or known fuel leaker and is not to be used. Contact Reactor Engineering if there are indications of a fuel leak.
L BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP I of 2 Reactivity Maneuver Plan # U2 NRC Exam I Description of Step: Withdraw Control rods lAWAttachment 2 provided by Reactor Engineer.
Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 850-1025 MFLCPR .80 - .85 MW Thermal 2600-31 00 MAPRAT .45 - .55 Core Flow 78-82mIbm/hr MFDLRX .65 - .70 Loadline 102-1 04 Core Power 85-90% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions Comments I Notes: Rod Out Notch Override is authorized, for Rod Out Notch Override follow the guidance in 2-01-85 section 6.6.4.
Step Complete AND Reviewed by: I Unit Supervisor I Date
BEN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 2 of 2 Reactivity Maneuver Plan # U2 NRC Exam I Description of Step: Raise reactor power to 100% using core flow. NO Ramp Rate Limits apply.
Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 1150 MFLCPR .85 .95 MW Thermal 3400-3450 MAPRAT .60 .70 Core Flow 8595 mlbm/hr MFDLRX .70 .75 Loadline 102-1 04 Core Power 100% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions N/A Comments I Notes:
- 1. Raise Reactor Power to 100% RTP
- 2. Document core flow changes on Attachment 10 Step Complete AND Reviewed by: I Unit Supervisor I Date
BEN Reactivity Control Plan Attachment 10 (Page 1 of 1)
Recirc Flow Maneuver Instructions Reactivity Control Plan # U2 NRC Exam I RCP Elow Time Target Delta Target Completed (RO)
Step # Step # Power Flow
(%RTP or +/-(MWe) (MLb/Hr)
MWe) 2 1 100%
Comments I Notes:
Reviewed by:
Unit Supervisor / Date
I BFN Reactivity Control Plan ATTACHMENT 4 ROD PATTERN STEP THROUGH MAPS Reactivity Maneuver Plan # U2 NRC Exam I 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 Prior to RCP I 59 59 55 55 51 51 47 00 47 43 43 39 00 00 00 39 35 35 31 00 00 00 00 00 31 27 27 23 00 00 08 23 19 19 15 00 15 11 11 07 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 After RCP I 59 59 55 55 51 51 47 16 47 43 43 39 16 12 16 39 35 35 31 16 12 12 16 31 27 27 23 16 12 16 23 19 19 15 16 15 11 11 07 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58
BEN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1 .3.5(A)
UNIT2 - REV.0021 .-
ATTACHMENT 2 (Page 1 of 1)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-31 00 12 N/A 30-39 00 12 N/A 38-31 00 12 N/A 30-23 00 12 N/A 30-3 1 00 48 N/A 30-15 00 16 N/A 46-3 1 00 16 N/A 30-47 00 16 N/A 14-3 1 00 16 N/A 22-23 00 16 N/A 3 8-23 00 16 N/A 38-39 00 16 N/A 22-39 00 16 4
REMARKS Control Rod Pattern Adjustment NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by:____________________________ Issued by /
Unit Supervisor Date Reactor Engineer Date
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.
C-BOP Returning to service Condenser Circulating Water Pump 3A, 1
C-SRO JAW 3-01-27 section 8.2, Pump Disch Valve Failure to Open R-ATC 2 Commence power increase with flow R-SRO C-BOP 3 RCO2 Inadvertent RCIC start w/ trip pushbutton failure 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 RCJC 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 dc-energize, ART inserts all Rods 9 HPO 15 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 Events
- 1. BOP returns the Condenser Circulating Water Pump 3A to service JAW 3-01-27 Condenser Circulating Water System, section 8.2. BOP Operator determines that 3A CCW Pump discharge valve fails to open. Orders a reset of breaker or secures 3A CCW Pump.
- 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 determine 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 insert Control Rod 46-19 until it becomes stuck at position 20. SRO will determine Control Rod 46-19 is Inoperable Technical Specification 3.1.3 Condition A.
- 6. A RCJC Steam Leak will result in high Room temperature with a failure of RCIC to Isolate.
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 EOI-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 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 MSL Tunnel temperature reaches 189°F OR any area radiation reaches 1000 mr/hr.
- 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. Within 5 minutes of the parameters exceeding max safe.
- 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 ART 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 ART 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 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2 9 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur afier EOI entry: List (1-4) 4 Abnomial Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 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 KIA 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-138 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-12 CRD Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 Control Rod Drift RO U-085-AL-12 201003A2.03 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 5-000-EM-b
Appendix D Scenario Outline Form ES-D-1 Simulator Instructor 199
- CCW 3A Disch Valve FTO br zaoei27l0a (el 0) 330 ior zlohs27l 3a[1] on
- RCIC inadvertent start imfrc02(e50) br zdihs7l 9a[1] null
- RCIC steam leak imf rclO ior zdihs7l 2a[2J auto imf rcO9 (e6 0)50 120 10 ior zdihs7l 9a[1] null
- CR 46-19 drift in imf rd0la (elO 0) imfrdO7r46l9 (e12 0) imf rd06r4619 (e13 0)
- MSL A break inside containment imfth35a (e15 0)36000 imfms06a imfms06b imf hpO3 (e15 0)10 ior xa557c[8] alarm_off imf rpO7
- FueI failure imfth23 (e20 120)4600 1 iorzdihso3l25[1](e20 10) trip br zdihs03l 51 [1] (e20 10) trip iorzdihso3l76[1] (e20 10) trip
Appendix D Scenario Outline Form ES-P-i 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.
C-BOP Returning to service Condenser Circulating Water Pump 3A, 1
C-SRO JAW 3-01-27 section 8.2, Pump Disch Valve Failure to Open R-ATC 2 Commence power increase with flow R-SRO C-BOP 3 RCO2 Inadvertent RCIC start w/ trip pushbutton failure 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 RCJC 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)onnal, (R)eactivity, (T)nstrument, (C)omponent, (M)ajor
Appendix B Scenario Outline Form ES-B-i Critical Tasks Four CT#1-With reactor at power and with a primary system discharging into the secondary containment, manually scram the reactor before MSL Tunnel temperature reaches 189°F OR any area radiation reaches 1000 mr/hr.
- 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. Within five minutes of the parameters exceeding max safe.
- 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 U 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 ART pushbuttons armed and depressed to cause control rod insertion.
- 4. Feedback:
Reactor power trend Rod status indication
Appendix U Scenario Outline Form ES-U-i Events
- 1. BOP returns the Condenser Circulating Water Pump 3A to service lAW 3-01-27 Condenser Circulating Water System, section 8.2. BOP Operator determines that 3A CCW Pump discharge valve fails to open. Orders a reset of breaker or secures 3A CCW Pump.
- 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 determine 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 JAW 3-AOI-85-5 Control Rod Drift In. ATC will insert Control Rod 46-19 until it becomes stuck at position 20. SRO will determine Control Rod 46-19 is Inoperable Technical Specification 3.1.3 Condition A.
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 EOI-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 ES-fl-i 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)
I Major Transients: List (1-2) 2 EOIs used: List (1-3) 1 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Appendix B Scenario Outline Form ES-B-i Scenario Tasks TASK NUMBER K/A RO SRO Condenser Circ Water Pump Start RO U-027-NO-5 400000A4. 01 3.1 3.0 Raise Power with Recirc Flow RO U-068-NO-1 7 SRO S-000-NO-1 38 2.1.23 4.3 4.4 RCIC Inadvertent Start RO U-071-NO-5 21 7000A2.01 3.8 3.7 RCIC Steam Leak RO U-071-AL-20 217000A2.15 3.8 3.8 SRO S-000-EM-12 CRD Pump Trip RO U-085-AL-07 201 001A2.01 3.2 3.3 SRO S-085-AB-03 Control Rod Drift RO U-085-AL-1 2 201 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 5-000-EM-b
2 PZ5 Procedures Used/Referenced:
[_Procedure_Number Procedure Title Procedure Revision 3-01-27 Condenser Circulating Water System Rev 58 3-GOl-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-AOl-85-3 CRD System Failure Rev 10 3-AOl-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-E0l-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-E0l-1 RPV Control Rev 8 Restoring Refuel Zone and Reactor Zone Ventilation 3-EOl-Appendix-8F Rev 2 Fans_Following_Group_6_lsoIaton Bypassing Group 6 Low RPV level and High Drywell 3-E0l-Appendix-8E Rev 1 Pressure_Isolation_Interlocks 3-EOl-Appendix-i 1A Alternate Pressure Control Systems MSRVs Rev 2 3-E0l-Appendix-5D Injection System Lineup HPCI Rev 5 3-EOI-3-C-2 Emergency RPV Depressurization Rev 8 3-E0l-Appendix-6A Injection Subsystems Lineup Condensate Rev 2 3-EOl-2 Primary Containment Control Rev 8 3-E0l-Appendix-17A RHR System Operation Suppression Pool Cooling RevS EPIP-1 Emergency Classification Rev 46
2 6ot S5 Simulator Instructor - IC-199
- CCW 3A Disch Valve FTO iorzaoei27l0a (el 0)330 ior zlohs27l 3a[1] on
- RCIC inadvertent start imf rcO2 (e5 0) br zdihs7l9a[1] null
- RCIC steam leak imf rclO br zdihs7l2a[2] auto imf rcO9 (e6 0)50 120 10 br zdihs7l9a[1] null
- CR 46-19 drift in imf rd0la (elO 0) imf rd07r4619 (e12 0) imf rd06r4619 (e13 0)
- MSL A break inside containment imfth35a (e15 0)36000 imfms06a imfms06b imf hpO3 (e15 0)10 br xa557c[81 alarm_off bmf rpO7
- Fuel failure imfth23(e20 120)46001 ior zdihsO3l25[1] (e20 10) trip iorzdihs03l5l[1] (e20 10) trip bor 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 Page 9 t 3 Simulator Event Guide:
Event 1 Component: Returning to service CCW Pump 3A, Disch Valve FTC SRO Directs CCW Pump 3A returned to service lAW 3-01-27, section 8.2 rive Insert trigger one whn operator starts 3A CCW Pump to epppdisphØrg ilve does riot fully operiandamps indicatethectly BOP 8.2 Returning a CCW Pump to Service
[4] VERIFY CLOSED the CCW PUMP 3A(3B)(3C) DISCH ISOL VALVE, 3-FCV-27-1 3(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 COW 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-27-1OA(18A)(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 CCW Pump 3A, recognizes CCW Pump 3A discharge valve does not fully open
2
-. PglQQf 35 -.
Simulator Event Guide:
Event I Component: Returning to service CCW Pump 3A, Disch Valve FTC When B ecognizesJj1ure of 90W Purçp 3 Dis4Valve to fully opeij carwildate may eilhej1) cal(AUO to reset bceaK&ORj5 Septlre the Pump
!iivej if cjpwsecçres3A CC jq3JJcontact U9it$upervipças theShiftManager ançi infatp91erpeed to coies first byOE)S req ues i$ver If crv contasAUQ ti ivestiaf gea)Øror sefyai3 irn es ajid repoit tlf 4
ofloai i i I( P w w r orderect o resej breaKeR report bleaker has bepneseraskopertor to tlçe carge valve íó Open wqa few seconds and DOR zlohs271 3[ikoh nd DORzaoøt27l Od raieturn 3A COW pump alps to normal BOP If crew ordered AUO to reset breaker, BOP shall verify that 3A CCW pump discharge valve travels full open or open the valve BOP IF breaker is NOT reset, BOP operator secures 3A CCW Pump
2 Page 1iQf35 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.
[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 D. Individual pump speeds should be mismatched by -60 RPM during dual pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short time periods for testing or maintenance).
[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 38 using, RAISE SLOW (MEDIUM), 3-HS 16A(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 Oiie puh qf Raip Medjpm on therrç pct ço JP peed bange at I RPM per second Crew may oHy be ihcreasTng pwerto26H Vacuum dependmng on outcome o wit pjn. inyrtjjpf RGIQ Jjiy Jiiehyj,
2 PagelZQt35 -
Simulator Event Guide:
Event 3 Component: Inadvertent RCIC start wI trip pushbutton failure BOP Responds to alarm 9-3B, Window 27, RCIC Gland Seal Vacuum Tank Pressure High A. VERIFY RCIC VACUUM PUMP, 3-HS-71-31A, 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 lAW 3-01-71, Section 8.4 8.4 RCIC Turbine Trip NOTES
- 1) The following signals cause a RCIC turbine trip. The RPV High Water Level Trip Signal closes the RCIC TURBINE STEAM SUPPLY VLV, 3-FCV-71-8, and RCIC PUMP MIN FLOW VALVE, 3-FCV-71-34. All other trip signals close the RCIC TURBINE TRIP/THROTTLE VLV, 3-FCV-71-9 and RCIC PUMP MIN FLOW -VALVE, 3-FCV-71 -34.
. High RPV Water Level (÷51 inches, Auto Reset at -45)
. Low Pump Suction Pressure (10 inches HG vacuum)
. High Turbine Exhaust Pressure (50 psig)
. Turbine Overspeed (Mechanical, 1223% of rated signal)
- Automatic Isolation
. Manual Pushbutton
- 2) All operations are performed at Panel 3-9-3 unless otherwise noted.
[1] IF RCIC Turbine did NOT trip from high RPV water level, THEN VERIFY the following automatic actions:
A. RCIC TURB TRIP/THROTTLE VLV, 3-FCV-71-9, closes B. RCIC PUMP MIN FLOW VALVE, 3-FCV-71-34, closes C. RCIC TURB SPEED, 3-SI-71-42A, indicates zero rpm 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
2 iQf3 Simulator Event Guide:
Event 3 Component: Inadvertent RCIC start w/ trip pushbutton failure 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 Priyei Ickoedge dtptGh5th breakeçwait3 miritjtes and report on station at 25P4 tIOf Bø B, jppt5D when directed)neit overe to opp *erfpr 74 vaive ioifypoyf cv7 j34 faiLpowerjiow cknowIedge disptas Instrument MecTIj 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 PW When directed bNRQ a11 the BOP operator toha,e candiciate checkwhich 9P umpare yutqngJncLwhep BO1 isay frop papej 7 1Ofq CRD Pump 3A triç4
2
-Pag 14 Qf 35 -
Simulator Event Guide:
Event 4 Component: CRD Pump 3A trip ATC Reports Trip of CRD Pump 3A.
SRO Announces entry into 3-AOI-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.
[1.2] START associated standby CRD Pump using one of the following:
- CRD PUMP 3B, 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.
bnver If DispIt9he to Rb Pdmp3A, pump Esexfremety ot totouq CR0 p3BoiJ jes tr brd ppp sta PRO 3A report breakertripped on over cirrent, Eectricaj Maint called Nac Wn ATQ Position 40 Dtjve, NRqJ tJer noi6d pyIete tp44ft
2 Page 1 of 3 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-AOI-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-AC 1-100-1.
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.
[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).
ATC Reports rod 46-19 stopped drifting at position 40 Drjver Whe ATbeipsto insert contioj od 4- 9 stickthe atppsitrn 20 insertjpg trigger I ATC Inserts Control Rod 46-19 to position 00.
ATC Reports Control Rod 46-19 is stuck at position 20 SRO May or May NOT direct actions for Control Rod difficult to insert ATC 8.16 Control Rod Difficult to Insert
[1] VERIFY the control rod will not notch in, in accordance with Section 6.7 or Section 8.19.
[2] REVIEW all Precautions and Limitations in Section 3.0.
[3] [NRC/C] IF RWM is enforcing, THEN VERIFY RWM operable and LATCHED in to the correct ROD GROUP. [NRC IR 84-02]
[4] CHECK CRD SYSTEM FLOW is between 40 gpm and 65 gpm, indicated by 3-FIC-85-i 1.
2 Pagel&at35 Simulator Event Guide:
Event 5 Component: Control Rod 46-19 drifts in to position 40 ATC 8.16 Control Rod Difficult to Insert (contd)
[5] CHECK CRD DRIVE WTR HDR DP, 3-PDI-85-17A is between 250 psid and 270 psid.
[6] IF CRD SYSTEM FLOW or CRD DRIVE WTR HDR DP had to be adjusted, THEN PROCEED to Section 6.7.
[7] IF control rod motion is observed, but the CRD fails to notch-in with normal operating drive water pressure, THEN:
[7.1] NOTIFY Reactor Engineer to determine what parameters should be recorded for further evaluation.
[7.2] RAISE CRD DRIVE WTR HDR DP, 3-PDI-85-17A, not to exceed 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[7.3] INSERT control rod as directed in Section 6.7.
[7.4] LOWER CRD DRIVE WTR HDR DP, 3-PDI-85-1 7A, to between 250 psid and 270 psid using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[12] IF the control rod still fails to notch in, THEN:
[12.1] NOTIFY the Unit Supervisor and Reactor Engineer to Refer to section Stuck Control Rod-Test to distinguish a Hydraulic Problem from Mechanical Binding, 0-Tl-20, and RETURN to Section 8.16.
[12.2] REQUEST the Unit Supervisor and Reactor Engineer to evaluate the control rod operability. Refer to Tech Spec 3.1
2
- Pge 17Qf 35 Simulator Event Guide:
Event 5 Component: Control Rod 46-19 drifts in to position 40 riy Reacoreçgiqr ackftQwlede rod Jrçdp ?djtnt, inform ,crew that &oU ar orkihg on ii w-v As AU -aftej dispatched reportscran3 vives are norma$
SRO Evaluate Tech Spec 3.1.3 Condition A One withdrawn control rod stuck Required Action A. 1 Verify stuck control rod separation criteria are met Completion Time Immediately AND Required Action A.2 Disarm the associated CRD Completion Time 2 Hours AND Required Action A.3 Perform SR3. 1.3.3 for each withdrawn OPERABLE control rod.
Completion Time 24 Hours from discovery of Condition A AND Required Action A.4 Perform 5R3.1.1.1 Completion Time 72 Hours jj rRM 3& may be p priatJFcQ position indicatiop is ayaiIabJ 1RC Wben ready tep teak in the RCIC room RC1CStearn line isolation vaf,es 3-FCV-71-Tand wi1i n& ait6Jsoiat riv
2
- - - FagQ Qf 3 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 3-9-11 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-Rl-90-26A on Panel 3-9-1 1 and NOTIFY RADCON if rising radiation levels are observed.
E. DISPATCH personnel to investigate.
2 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 that 71-2 and 71-3 failed to auto isolate._Based_on_amber_lights_on_Panel_3-9-3.
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 Driv djspaJ4 QRç rarpqrter5 mLputes that cahp cçssrea attflj tim&i 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:
. 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 PCIVs 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 Page2QQt35 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 EOI-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 EOls OR
. Suppress a Fire Ensures no systems are still discharging to Secondary Containment, remains in EOl-3 until entry conditions are cleared.
SRO Enters EOI-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 -
N1cI h ,?yJ4kicp ML yi1c oIo trye jjjjj
2 Pag2iot35 -
Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close BOP Responds to Rx Building Area Radiation High Alarm, 3A-22 and Rx BLDG, TURB BLDG,_RF ZONE_EXH_Radiation_High_3A-4_alarms.
A. DETERMINE area with high radiation level on Panel 2-9-11. (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 9-11 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 EOl-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 OR any area radiation reading reaches 1000 mr/hr.
SRO May or May NOT direct Core flow runback prior to Reactor Scram ATC If directed insert Core Flow Runback, by depressing the core flow runback pushbutton.
Driver When ATC arms ay depresses AFI, insrt tggrQ for RFPT trips a4 Fuel Failure
2 Pag22of5. -
Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close Orlyer After Açs ps ggpQqf Ip adf aUure CT #1 SRO Directs ATC to insert manual Rx Scram prior to MSIV isolation at a Steam Tunnel Temperature of 189F OR any area radiation reading reaches 1000 mr/hr.
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-99-5AJS3A and 3-HS 5A/S3B, 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-S1, in REFUEL.
[3.2] CHECK REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46, ilium mates.
[3.3] IF REFUEL MODE ONE ROD PERMISSIVE light, 3-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, 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) o ARI Manual Initiate, 3-HS-68-1 1 9A OR
- 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, eve 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 AR1 by Arming and Depressing one of the ARI Manual Initiate collars and pushbuttons then reports I have rod motion.
Verifies all rods insert and makes Scram Report to the SRO
2 Page.2a of 35 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 EOl-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 MSIVs failed to isolate manually or automatically SRO Enters EOl-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 BOP Dispatches personnel to investigate A MSIVs and manually close Outboard MSIVs Driver If requestedwait3 mifliLtes an report Appen8E complete, enter bat appO8e If dipatchidf& MSIVackededjatcn
2 P24f3 - -
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 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
2 Page 25 of Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO EOl-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:
. Be operated by EOls OR
. 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 EOI-1 (EOI-1 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 Page26ofi35 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO Enters EOI-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-EOI-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.
r T1nffct, FbLa
2 Page 27 of 35 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-IIA, if necessary
- 1. IF Drywell Control Air is NOT available, THEN EXECUTE EOl 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 -1 79 MN STM LINE A RELIEF VALVE.
- b. 2 3-PCV-1-180 MN STM LINED RELIEF VALVE.
- c. 3 3-PCV1-4 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.
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 H PCI, 3-EOI-Appendix-5D.
2 Page28Qf3 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 EOl Appendix 1 6E 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 EOI Appendix 16L 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 140F may result in equipment damage.
- 4. VERIFY 3-lL-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 start UNTIL 3-FCV-73-16, HPCI TURBINE STEAM SUPPLY VLV, starts to open.
2 Pae2aQfa5 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 injection.
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
_Pae30Qf35 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-AOl-i 00-1 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 within five minutes of exceeding the MAX safe values.
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 REP discharge valves:
. 3-FCV-3-19, REP 3A DISCHARGE VALVE
. 3-FCV-3-12, REP 3B DISCHARGE VALVE
. 3-ECV-3-5, REP 3C DISCHARGE VALVE
- 3. VERIFY OPEN the following drain cooler inlet valves:
. 3-ECV-2-72, DRAIN COOLER 3A5 CNDS INLET ISOL VLV
. 3-ECV-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-ECV-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 P ai ofa5 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 RFP suction valves:
- 3-FCV-2-83, RFP 3A SUCTION VALVE
- 3-FCV-2-95, RFP 3B SUCTION VALVE
- 3-FCV-2-108, RFP 30 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).
ATC Verifies REP discharge valves are closed prior to Reactor Pressure dropping below condensate system discharcie pressure to prevent overfeedinci the Reactor
2 Page 32 of 35 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 DryweN 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)
Monftor 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 EOl-2 Suppression Pool (Temperature)
Monitor and Control Suppression Pool Temperature Below 95°F, Using Available Suppression Pool Coolinq 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 Pane 33 of 35 Simulator Event Guide:
Event 7 Major: MSL A Break in Reactor BLDG with MSL A valves failing to close SRO EOI-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-74-1 22(1 30), 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 Pg34of 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.
- i. THROTTLE 3-FCV-74-59(73), RHR SYS 1(11) SUPPR POOL CLGITEST 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.
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 throuah 2.1.
SRO Emergency Plan Classification 3.2-S
2 Page35Qf35 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
None Operations/Maintenance for the Shift:
Return Condenser Circulating Water Pump 3A to service, lAW 3-01-27, section 8.2[4]
Section 8.2 of 3-0I-27has been completed thru [3.3j. 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
BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3.1.3.5(A)
UNIT 3 REV 0023 ATTACHMENT 2 (Page 1 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-23 10 00 N/A 22-39 10 00 N/A 38-23 10 00 N/A 38-39 10 00 N/A 30-3 1 14 00 N/A 14-15 16 00 N/A 14-47 16 00 N/A 46-47 16 00 N/A 46-15 16 00 N/A 14-3 1 12 00 N/A 30-47 12 00 N/A 46-3 1 12 00 N/A 30-15 12 00 N/A 22-3 1 48 00 N/A 30-39 48 00 N/A 38-31 48 00 N/A 30-23 48 00 4
REMARKS Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods ilytTiO. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 3-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I Issued by I Unit Supervisor Date Reactor Engineer Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3.1.3.5(A)
UNIT 3 - - REV 0023 ATTACHMENT 2 (Page 2 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-47 48 00 N/A 38-47 48 00 N/A 38-15 48 00 N/A 22-15 48 00 N/A 14-39 48 00 N/A 46-39 48 00 N/A 46-23 48 00 N/A 14-23 48 00 N/A 06-3 1 48 00 N/A 30-55 48 00 N/A 54-3 1 48 00 N/A 30-07 48 00 N/A 06-39 48 00 N/A 54-39 48 00 N/A 54-23 48 00 N/A 06-23 48 00 4
REMARKS Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously to 00. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 3-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I Issued by /
Unit Supervisor Date Reactor Engineer Date
TENNESSEE VALLEYAUTHORJTY -
BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U3 NRC Exam 2 Raise Reactor Power to 100%
BFN Reactivity Control Plan Attachment 7 (Page 1 of 2)
Reactivity Control Plan Form BEN Unit: 3 Valid Date(s): 8/8/Il 8/19/11 Reactivity Control Plan #: U3 NRC Exam 2 Are Multiple Activations Allowed: No (If yes, US may make additional copies)
Prepared by: / Reviewed by: I Reactor Engineer Date Qualified Reactor Engineer Date Approved by: I Concurrence: I RE Supervisor Date WCC/Risk/US SRO Date Approved by: I Authorized by: I Ops Manager or Supt. Date Shift Manager Date RCP Activated: I RCP Terminated: /
Unit Supervisor Date Unit Supervisor Date Title of Evolution: Raise Power to 100%
Purpose/Overview of Evolution: Raise Power to 100%
Maneuver Steps
- 1. Raise reactor power to 100% using core flow. (NO Ramp Rate Limits Apply)
BFN Reactivity Control Plan Attachment 7 (Page 2 of 2)
Reactivity Control Plan Form Operating Experience and General Issues: U3 NRC Exam 2 This plan is NOT valid if the unit is operating with a suspected or known fuel leaker and is not to be used. Contact Reactor Engineering if there are indications of a fuel leak.
Known Issues:
CautionslError Likely Situations/Special Monitoring Requirements/Contingencies:
NONE
BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP I of I Reactivity Maneuver Plan # U3 NRC Exam 2 Description of Step: Raise reactor power to 100% using core flow. No Ramp Rate Limits apply.
Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No X (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 900-1150 MFLCPR .80 .85 MW Thermal 2850-3450 MAPRAT .55 .65 Core Flow 65-96 mlbm/hr MFDLRX .65 .75 Loadline 105-1 08 Core Power 80% 100%
- Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions MWth Limit 3458 Comments I Notes:
- 1. Raise Reactor Power to 100% RTP
- 2. Document core flow changes on Attachment 10 Step Complete AND Reviewed by: I Unit Supervisor I Date
BFN Reactivity Control Plan Attachment 10 (Page 1 of 1)
Recirc Flow Maneuver Instructions Reactivity Control Plan # U3 NRC Exam 2 RCP Flow Time Target Delta Target Completed (RO)
Step # Step # Power Flow
(%RTP or +/-(MWe) (MLbIHr)
MWe) 1 100%
Comments I Notes:
Reviewed by:
Unit Supervisor I Date
Appendix 0 Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 4 - Op-Test No.: 1108 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 starting at step [4]. 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-OI-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 EGO3 Turbine Generator Voltage Regulator Failure C-SRO C-ATC LOCA Recirculation Pump B Inboard and Outboard seal R-ATC TH1O/1 lb failure TS-SRO C-BOP 6 EGO2 Stator Water Cooling Pump Trip C-SRO C-ATC 7 TCJOb EHC 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 Events
- 1. BOP Transfers 4KV Unit Board 3A from USST 3B to Start Bus 1A JAW 0-OI-57A section 8.15.1 starting at step [4].
- 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. 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.
- 5. #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-68-1A. 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.
- 6. Stator Water Cooling Pump trip, BOP operator starts standby pump and restores stator water cooling prior to a turbine trip.
- 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 JAW 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 RCIC 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 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 JAW EOI-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeep5,ing 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 EOI 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 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-NO-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-2O 262001A2.09 3.1 3.4 SRO S-57A-AB-4 RR Pump Seal Failure RO U-068-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
Appendix D Scenario Outline Form ES-D-1 Simulator Instructor IC-204
- HPCI tagout bat nrchpcito
- Tech Spec call SRO Core Spray System #1 ior ypovfcv7525 (el 0) fail_now ior xa553c[27] (el 0) crywolf
- B stator water pump trip irf egO2 (e5 0) off ior ypobkrscwpa (e5 0) fail_ccoil ior zdihs3535a[2] (e5 0) stop ior zlohs3535a[1] (e5 0) off
- Turbine Generator Voltage Regulator failure imfego3 (elO 0)
- B Recirc pump seal failures imfthl2b (e15 0) imfthl0b (e15 0)100 imfthllb (e15 180) 100600
- B EHC Pressure transducer failure bat atws70 ior zdihs0l 16{1] (e20 0) select ior zdihs472O4[1] (e20 0) null iorzlohs0ll6[1] off ior z1ohs47204[ 1] on imftcl0b (e20 0) 86 1200 79
- RCIC steam supply valve fails to auto open imf rcO8 trg25 =batsdv trg 26 bat atws-l trg 27 = bat appolf trg 28 = bat app02 trg29=batapp08ae
- After Scram manually insert under DI Overide
- 3-FIC-85-1 1 0-100(L)
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.: 1108 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 starting at step [4]. 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-OI-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 4 EGO3 Turbine Generator Voltage Regulator Failure LOCA Recirculation Pump B Inboard and Outboard seal 5
TH1O/llb failure TS-SRO C-BOP 6 EGO2 Stator Water Cooling Pump Trip CSRO 7 TC1Ob EHC Pressure Transducer Failure 8 M-ALL ATWS, without MSJVs 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 P Scenario Outline Form ES-P-i 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 BITT) 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-l 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 / B Pump control switch in ON, when directed by US.
AND Control Rod insertion commenced in accordance EOI 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 Temiinated and Prevented.
- 4. Feedback:
RPV pressure trend RPV level trend ADS annunciator status
Appendix D Scenario Outline Form ES-D-1 Events
- 1. BOP Transfers 4KV Unit Board 3A from USST 3B to Start Bus 1A JAW 0-OI-57A section 8.15.1 starting at step [4].
- 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. 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.
- 5. #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-68-1A. 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.
- 6. Stator Water Cooling Pump trip, BOP operator starts standby pump and restores stator water cooling prior to a turbine trip.
- 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 JAW 3-AOI-47-2.
- 8. ATWS exists on the scram the crew will enter EOJ-1, EOI-2 and EOJ-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 RCJC 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 B Scenario Outline Form ES-B-i 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 EQIs used: List (1-3) 1 EOl 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 KJA RO SRO Transfer 4KV Unit Board RO U-57A-NO-1 26200 1A4.05 3.3 3.3 SRO S-57A-NO-4 Lower Power with Recirc Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 Stator Water Cooling Pump Trip RO U35A-AL-2 245000A4.03 2.7 2.8 SRO S-070-AB-1 Turbine Generator Voltage Regulator Failure RO U47-AL-20 262001A2.09 3.1 3.4 SRO S-57A-AB-4 RR Pump Seal Failure RO U-068-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 295015AA2.01 4.1 4.3 SRO S-000-EM-1 SRO S-000-EM-2 SRO S-000-EM-3
4 PageeLcif3 -
Procedures Used/Referenced:
Procedure Number Procedure Title ]_Procedure_Revision 0-OI-57A Switchyard and 4160V AC Electrical System Rev.141 3-G0I-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 BFN-UNIT 3 Tech Spec 3.5-1 Amend No. 244 December 1, 2003 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-A01641 Rev. 3 Excessive Leakage Into Drywell Recirc Pump Trip/Core Flow Decrease OPRMs 3-AOI-68-1A Rev. 6 Operable 3-AOI-47-2 Turbine EHC Control System Malfunctions Rev. 6 3-EOI-1 RPV CONTROL FLOWCHART Rev. 8 INSERT CONTROL RODS USING REACTOR 3-EOI APPENDIX-iD Rev. 2 MANUAL_CONTROL SYSTEM 3-EOI-2 PRIMARY CONTAINMENT CONTROL FLOWCHART Rev.7 3-EOI APPENDIX-SC INJECTION SYSTEM LINEUP RCIC Rev. 3 3-EOI 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 Pae9Qt35 Simulator Instructor IC-204
- HPCI tagout bat nrchpcito
- Tech Spec call SRO Core Spray System #1 ior ypovfcv7525 (el 0) fail_now ior xa553c[27] (el 0) crywolf
- B stator water pump trip irfego2(e5 0)off br ypobkrscwpa (e5 0) fail_ccoil bor zdihs3535a[2] (e5 0) stop ior zIohs3535a[1] (e5 0) off
- Turbine Generator Voltage Regulator failure imf egO3 (el 0 0)
- B Recirc pump seal failures imfthl2b (e15 0) imfthl0b (e15 0)100 imfthllb(e15 180)100600
- B EHC Pressure transducer failure bat atws70 iorzdihs0ll6[1] (e20 0) select ior zdihs472O4[1] (e20 0) null or zlohs0l 16[1] off ior z1ohs47204[1] on imftcl0b (e20 0)86 120079
- RCIC steam supply valve fails to auto open imf rcO8 trg 25 = bat sdv trg 26 = bat atws-1 trg 27 = bat appOif trg 28 = bat appo2 trg 29 = bat app08ae
- After Scram manually insert under DI Overide
- 3-FIC-85-1 1 0-100(L)
Scenario 4 DESCRIPTIONIACTION 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
_PagA0of35 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-Ol-57A section SRO 8.15.1 8.15.1 Transfer 4kV Unit Board 3A from USST to Start Bus BOP
[1] REVIEW all Precautions and Limitations in Section 3.0.
CAUTIONS I) This board transfer can cause a power interruption causing a loss of Computer Rooms and Communication Battery Board ACU, Computer UPS AOL), 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 COW pumps.
- 3) If 4kV Unit Board 3A is fed from the Alternate Power Supply (Start Bus), then Auto transfer must be blocked for:
. 4kV UNIT BD 1A, IB, IC, 2A, 28, and 20. (Ref. 3-45E721 OPL) e 4kV COM BD A and B. (3-45E721 OPL)
- 4) If either 4kV UNIT BD IA, 18, 2A or 28 is aligned to a Start Bus, prior to aligning UNIT BD 3A to the Start Bus, check Technical Specifications 3.8.1 .a and 3.8.2.a to determine operability of qualified AC circuits between the offsite transmission network and the onsite Class I 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.
[2.21 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 Page1i of 35 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.
[3.5] 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.
When rquestdjo RE-ALIGN 4kv UNtT BD Auto transfer Schem. Reprt switches fqr 4 KV Unaf Boards IA, B2G AND Cmçjj Brd A nd B have been p1ce in MANUAL.
[4] TRANSFER 4kv UNIT BD 3A to the ALT ED
[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-El-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 Page12Df35 -
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.
[4.111 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.
Weft requesédackowiedge that a autin Orde- will need to be pIed on ondensate cBPpS, 1R or 0GW Pupp jIg çe4tQ pLc bRIVER b1VE When requested, acnowIedge that the Computer Rooms,, Gomthuniatiri Bar oard Computer QPS and CoWmunicatiop rpoms ACUs are to breturped tq wrc < ryicç pr cocppJetehsp.
4 Page 13 of 35 Simulator Event Guide:
Event 2 Reactivity: Lower Reactor Power with Recirc Flow SRO Notify ODS of power decrease Directs Power Reduction using Recirc Flow per 3-G0I-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-GOI-100-12 to recover from Recirc Pump Trip) 3-01-68 Precaution and Limitations 3.5.3 Dual Pump Operation D. Individual pump speeds should be mismatched by 60 RPM during dual ATC pump operation between 1200 and 1300 RPM to minimize harmonic vibration (this requirement may be waived for short time periods for testing or maintenance).
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-96-15A(15B). (Otherwise N/A)
- Lower Recirc Pump 3A 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-96-16A(16B). (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-H S-96-33 LOWER MEDIUM, 3-HS-96-34 LOWER FAST, 3-HS-96-35
4
.Paget4of35. - -
Simulator Event Guide:
Event 3: Core Spray Loop 1 Inoperable failed FCV-75-25 NRC c When satisfied-with Reactivi manipulation, nove on to5re Spray [.oop lhóprable failed FCV-75-25 15-25 DRIVER DRIVER IiserttRtGGE TT ause a loss SPJAY s7 i INBD INJEcT 1ALE1 7-25 W -
r As eactort5Idg AUO, cafl te ojitrol. roop and report thaVyou discoverec beáker 48O/ R4OV Bd 3A, Cornp 148, frF GORE SPRAY lN8D INJEGT yA1YEttripped AND ilI NOT reset.
Relays field report to US. And recognizes 3-FCV-&T, CORE SPRAY SYS I BOP INBD INJECT VALVE, does not have indication. 725 SRO References Tech Spec 3.5.1 and enters Conditions A and D.
3.5 EMERGENCY CORE COO UNG SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 35.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray 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 vtth reactor steam dome pressure 150 psig.
ACTIONS LCO 3.0.4.b is not applicable to HPCI.
CONDIT1ON REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A. I 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. 0.1 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.
0.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem lo OPERABLE status.
4 Page 15 of 35 Simulator Event Guide:
Event 4 Component: Turbine Generator Voltage Regulator Failure 1nse R1GR ethTr nrage Rpgo fiigfl P1Y P&ER cmi This faiIu N 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-El-57-39, between 20,900V and 23,1 OOV.
- 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.
Takes Voltage Regulator to Manual and adjusts MVARs to comply with 0-GOI-300-4, Switchyard Manual P&L I:
A 300 MVAR maximum outgoing limit applies to all units for both the 500kV and 161kV offsite power source qualification. If the outgoing MVAR limit is exceeded for BOP a unit and is not corrected within 15 minutes, the TOp must immediately inform BFN that both offsite power sources are disqualified for the unit that is exceeding the limit.
Offsite power qualification is not impaired for the unit(s) whose outgoing MVARs are under the limit.
Crew Make notifications. Must notify Load Dispatch when voltage regulator not in Auto iik J nqn tL1 thej
4 Pe16of3. -
Simulator Event Guide:
Event 5 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure PWc 11 gNW ETRJER1USeIiRecc pump iaisjii ATC Reports failure of the #1 Reactor Recirc Pump B Seal RECIRC PUMP B NO. I 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 conjunction with this window.
s Rising drywall pressure and/or temperature.
s Increased leakage into the drywall sr.imp.
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 Pag17ot3 Simulator Event Guide:
Event 5 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-AOI-64-1, Drywell Pressure and/or Temperature High, or Excessive Leakage Into Drywell.
3-AOI-68-IA, 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-Tl-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-FI-68-48).
4 Page t8Qf35 Simulator Event Guide:
Event 5 Component: LOCA Recirculation Pump B Inboard and Outboard seal failure lnsrts Rods per Emergency shove sheet to get below 95% loadline.
- 2. Rod 14-15 12 to 00 9. Rod 30-1512 to 00
- 4. Rod 46-47 12 to 00 11. Rod 30-3948 to 00.
ATC 5. Rod 46-15 12 to 00 12. Rod 38-31 48 to 00.
- 7. Rod3O-4712to00 AT When less than 95% load line, raises core flow to greater than 45%.
SRO AOI-64-1 Directs BOP to Vent the Drywell 3-AOI-64-lDryweII Pressure andlor 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, 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 NIA)
W
4 Pagei9of35 Simulator Event Guide:
Event 5 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
- - Pag2Q of 35 -
Simulator Event Guide:
Event 6 Component: Stator Water Cooling Pump Trip DRIVER TV 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.
Responds to annunciator ARP 3-XA-55-9-8A Window 1, TURBINE TRIP TIMER INITIATED:
Automatic Action: Main Turbine Trip 60 seconds after the alarm is received (Total of 70 seconds)
Operator Action:
A. CHECK Stator Cooling Water Flow and Temperature and Generator Stator temperatures using 105.
BOP B. VERIFY all available Stator Cooling Water Pumps running.
C. IF all of the following conditions exist
. Alarm fails to reset,
. Low Stator Cooling Water flow OR High Generator or Stator
. Cooling temperatures are observed on ICS,
. Reactor Power is above turbine bypass valve capability, THEN, SCRAM the reactor.
Operator starts the standby Stator Water Cooling Pump and restores Stator Water BOP Cooling.
H trr J$i
4 Page ?1of 35 Simulator Event Guide:
Event 7 Component: EHC Pressure Transducer Failure At NFC cWrection, 1nSer TRIGGER 20 fo cause the B EFiC Pressure transducer to fail Venfy tcb jtppr toiisertjigjriggçr Qnap iij1 fail to isolatej Responds to annunciator, 3-9-7B Window 4, HEADER PRESS SETPOINT OUT OF RANGE:
A. CHECK header pressure setpoint >960 on EHC SETPOINT, 3-Pl-47-204.
B. STOP any power ascensions until alarm can be reset (power reductions may be performed).
C. IF it is desired to be in Reactor Pressure Control, THEN TRANSFER control to Reactor Pressure Control. REFER TO Transferring EHC Pressure ATC/BOP Control from Header Pressure To Reactor Pressure section in 3-01-47. (N/A if Step D will be performed).
D. IF Reactor Pressure Control is NOT available, THEN LOWER setpoint to below 960 psig by using LOWER pushbutton, 3-HS-47-1 62A (N/A if Step C was performed).
E. VERIFY alarm will reset.
F._RECORD_events_in_narrative_log.
ATC Recognizes lowering Reactor Pressure and generator megawatts.
SRO Directs entry into 3-AOl-47-2.
3-AOI-47-2 Turbine EHC Control System Malfunctions
[1] 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.
Jt Scram manuall9 tnsefl Under DI Oyer4 FlC-85.110jj(Lj a4djnset D11VE DRP TRIGGERv25 to enter bt SDV BOP Recognizes one main steam line D failed to isolate. Closes the MSIVs.
SRO Enter 3-EOl-1, RPV Control.
SRO EOI-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 ofCurve3?-NO
4 Page22of35 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 1A.
ATC/BOP Maintain directed pressure band, lAW Appendix hA.
EOl-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 1 1 F, Main Steam System Drains SRO Appendix 11 D, Steam Seals Appendix 1 1 G, SJAEs Appendix 1 1 G, Off Gas Preheater Appendix 1 1 G, RWCU Appendix 1 1 E.
ATC/BOP Pressure Control lAW Appendixi 1A, RPV Pressure Control SRVs
- 1. IF Drywell Control Air is NOT available, THEN:
EXECUTE EOl 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. 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 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs ATC/BOP Pressure Control lAW Appendixi 1A, 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 Page 24 of 35 Simulator Event Guide:
Event 8 Major: A1WS, 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 BOPIATC 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 Page 25 of 35 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 26of 35 Simulator Event Guide:
Event 8 Major: A1WS, 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, REP 2C 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, RFPT 3A TRIP
- 3-HS-3-151A, REPT 3B TRIP
- 3-HS-3-176A, RFPT 30 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 ZT#2 Reactor Level reaches -162 inches, THEN Continue:
Directs a Level Band with RCIC.
4 Page 7 of3 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs (continued)
SRO EOl-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 ARI.
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 1 F
. AppendixiD
ØRr7gT i&h1Appendix 1 Ap ai5if inufeünirt TRiGGER 27 tpenM WHEN. theScram hasbeenrest THEN insert TRIGGER 26 toenter .bat ATWS1.
CT#1 ATC Inserts Control Rods, lAW Appendix 1 D and 1 F.
4 Page 28 of 35 - -
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 29 of 35 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).
. 3-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.
. System flow, as indicated by 3-lL-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-69-1, RWCU INBD SUCT ISOLATION VALVE closed.
. 3-FCV-69-2, RWCU OUTBD SUCT ISOLATION VALVE closed.
. 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 dropping approximately 1 % per minute.
4 Page 30 of 35 Simulator Event Guide:
Event 8 Major: ATWS, without MSIVs SRO ENTER 3-EOI-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 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.
ATC/BOP Place an RHR System in Pool Cooling, when directed lAW Appendix 17A.
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 Page 31 of 35 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 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 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 35 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 Page 33 qf35 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-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? 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).
pv EIJspaced IG IosC) ari butofiJj 2)iiute ançepprt SfO5-O86 cIqped/(mropcte) edo oeft Chping OS-O586 open. {mrf4dO6 oph)
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 Page35 of 35 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
None Operations/Maintenance for the Shift:
100% power. HPCI is out of service.
Complete transfer of 4kV Unit board 3Afrom USSTto Start Bus IA 0-Ol-57A section 8.15.1 starting at step [4]. Steps [1], [2], and [3] of 0-OI-57A section 8.15.1 are complete.
When 4kV Unit board 3A transfer is complete, lower reactor power to 90% using recirc for surveillance testing.
Unit I and 2 at 100% Power Unusual Conditions/Problem Areas:
None
BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3.1.3.5(A)
UNIT 3 REV 0023 ATTACHMENT 2 (Page 1 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 30-31 48 00 N/A 14-15 12 00 N/A 14-47 12 00 N/A 46-47 12 00 N/A 46-15 12 00 N/A 14-3 1 12 00 N/A 30-47 12 00 N/A 46-3 1 12 00 N/A 30-15 12 00 N/A 22-3 1 48 00 N/A 30-39 48 00 N/A 38-3 1 48 00 N/A 30-23 48 00 4
REMARKS Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously to 00. Insertion may ston after completion of any 2roun.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 3-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I issued by I Unit Supervisor Date Reactor Engineer Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 3-SR-3.1.3.5(A)
UNIT 3. - REV OO2 ATIACHMENT 2 (Page 2 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-47 48 00 N/A 38-47 48 00 N/A 38-15 48 00 N/A 22-15 48 00 N/A 14-39 48 00 N/A 46-39 48 00 N/A 46-23 48 00 N/A 14-23 48 00 N/A 06-3 1 48 00 N/A 30-55 48 00 N/A 54-3 1 48 00 N/A 30-07 48 00 N/A 06-39 48 00 N/A 54-39 48 00 N/A 54-23 48 00 N/A 06-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously to 00. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 3-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: / Issued by /
Unit Supervisor Date Reactor Engineer Date
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U3 NRC Exam 4 Lower Reactor Power to 90%
BFN Reactivity Control Plan Attachment 7 (Page 1 of 2)
Reactivity Control Plan Form BEN Unit: 3 Valid Date(s): 8/8/11 8/19/11 Reactivity Control Plan #: U3 NRC Exam 4 Are Multiple Activations Allowed: No (If yes, US may make additional copies)
Prepared by: / Reviewed by:
Reactor Engineer Date Qualified Reactor Engineer Date Approved by: I Concurrence:
RE Supervisor Date WCC/RiskIUS SRO Date Approved by: / Authorized by: I Ops Manager or Supt. Date Shift Manager Date RCP Activated: I RCP Terminated: /
Unit Supervisor Date Unit Supervisor Date Title of Evolution: Lower Power to 90%
PurposelOverview of Evolution: Lower Power to 90%
Maneuver Steps
- 1. Lower reactor power to 90% using core flow. NO Ramp Rate Limits Apply
BFN Reactivity Control Plan Attachment 7 (Page 2 of 2)
Reactivity Control Plan Form Operating Experience and General Issues: U3 NRC Exam 4 This plan is NOT valid if the unit is operating with a suspected or known fuel leaker and is not to be used. Contact Reactor Engineering if there are indications of a fuel leak.
Known Issues:
Cautions/Error Likely Situations!Special Monitoring Requirements/Contingencies:
NONE
BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP I of I Reactivity Maneuver Plan # U3 NRC Exam 4 Description of Step: Lower reactor power to 90% using core flow. No Ramp Rate Limits apply Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No X (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 950-1050 MFLCPR .80 - .85 MW Thermal 2950-3150 MAPRAT .55 - .65 Core Flow 75-80 mlbm/hr MFDLRX .65 - .75 Loadline 105-1 08 Core Power 88% 90%
- Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments / Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions Comments I Notes:
- 1. Lower Reactor Power to 90% RTP
- 2. Document core flow changes on Attachment 10 Step Complete AND Reviewed by: I Unit Supervisor I Date
BFN Reactivity Control Plan L
Attachment 10 (Page 1 of 1)
Recirc Flow Maneuver Instructions Reactivity Control Plan # U3 NRC Exam 4 RCP Flow Time Target Delta Target Completed (RO)
Step # Step # Power Flow
(%RTP or +/-(MWe) (MLbIHr)
MWe) 1 90%
Comments I Notes:
Reviewed by: I Unit Supervisor I Date
Appendix D Scenario Outline Form ES-D-1 Facility: Browns Ferry NPP Scenario No.: NRC 6-Op-Test No.: 1108 Examiners:____________________ Operators: SRO:____________________
ATC:______________
BOP:______________
Initial Conditions: 80% power. RCIC is out of service and Breake r 1624 Alternate Feed to SD BD C.
Turnover: Place RFPT A in service from 600 RPM in accordance with 2-01-3 section 5.7 and then raise power to 100%
Event Maif. No. Event Type* Event Description No.
N-BOP Place RFPT A in service from 600 RPM in accordance with 1 2-N-SRO 0I-3section 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 stai C-ATC RBCCW pump B trips, RBCCW sectionalizing valve fails to 6 Batch file TS-SRO auto close OGO5a M-ALL Explosion in Off-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 TSSRO (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Events
- 2. ATC increases power with Control Rods
- 3. Control Rod 30-15 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 30-15.
- 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 JAW 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. Directs spraying the drywell before exceeding the PSP curve or reaching 280°F 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 0 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 JAW with EOI 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 JAW with EOI Appendix 17B AND Observation RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment
AppendixD Scenario Outline Form ES-D-1 CT#2- Teminate 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 JAW with EOJ Appendix 17B AND Observation RO secures Drywell Sprays
- 4. Feedback:
RHR flow to containment lowering RHR Sprays Valves closed
Appendix 0 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)
Appendix 0 Scenario Outline Form ES-D-1 Scenario Tasks TASK NUMBER KIA 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 ROU-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 SOOO-EM-5
Appendix 0 Scenario Outline Form ES-D-1 SIMULATOR Instructor IC92 Batch File 1108-6 Pref File 110806 Imfdg0lc F3 bat NRC/i l08rcicto Trg e4 NRC/dgstart F4 Trg e4 = drnf ed09c F5 bat NRC/i 10806 br zdi0hs2i iOcO2a[1] trip F6 imfrdO6r3Oi5 br zlo0hs2 11 OcO2a[ 1] off F7 dmfrdO6r3Oi5 br zlo0hs2 ii OcO2a[2] on F8 imfogo4a br zlohs7O48a[2] on F9 imfed09c br zlohs7O48a{l] off FlO imfsw02b br xa554c19 alarm off Fl 1 mrfsw02 align Trg el 7048-1 F12 imfogOl Trgei =batNRC/110806-1 Si imfogo5a 80 1200 100 br zlohs66ia[i] on S2 ior zdihs66la open br zlohs6óla{2] off br zdixs74i2i[1] reset Trg e2 modesw Imfth2l (e2 180) 0.5 600 0.1 bmf dgo3c (e2 0)
Appendix D Scenario Outline Form ES-B-i a7cility: Browns Ferry NPP Scenario No.: NRC 6 -
Op-Test No.: 1108 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 fiom 600 RPM in accordance with 2-01-3 section 5.7 and then raise power to 100%
Event Maif. No. Event Type* Event Description No.
NBOP 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 M-ALL Explosion in Off-gas system, Loss of condenser vacuum OGO 8 TH21 C LOCA, Loss of SD BD C I
9 IOR RHR Sys 1 Containment Spray Valve select switch failure TSSRO (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix B Scenario Outline Form ES-B-i 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 EQI Appendix 1 7B AND Observation RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment OR CT#i 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 17B 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 JAW 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 30-15 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 30-15.
- 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 Nonnal 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 JAW 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 E0J-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. Directs spraying the drywell before exceeding the PSP curve or reaching 280°F 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 B Scenario Outline Form ES-B-i 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 44 Scenario Tasks TASK NUMBER KJA RO SRO P lace RFPT A in Service RO U-003-NO-4 259002A4.03 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 ROU-085-NO-19 201003A2.O1 3.4 3.6 Loss of SJAE A ROU-066-NO-7 295002AA2.Ol 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 ROU-070-AL-3 206000A2.17 3.9 4.3 SRO S-070-AB-1 LOCA RO U-000-EM-1 295024EA1.1l 4.2 4.2 RO UOOO-EM-5 SRO S-000-EM-1 SRO S-000-EM-2 SRO S-000-EM-5
Page7of Procedures UsedlReferenced:
Procedure Number Procedure Procedure Title Revision 2-01-3 Reactor Feedwater System Revision 136 2-G0l-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-AOl-66-1 Off-Gas H2 High Revision 19 2-AOl-i 00-1 Reactor Scram Revision 95 2-EOI-1 RPV Control Flowchart Revision 12 2-EOl-2 Primary Containment Control Flowchart Revision 12 2-E0l-2-C-1 Alternate Level Control Flowchart Revision 9 2-EOI-2-C-2 Emergency RPV Depressurization Revision 6 2-EOl Appendix-6D Injection Subsystems Lineup Core Spray System I Revision 7 2-EOl-APPENDIX-17A RHR System Operation Suppression Pool Cooling Revision 12 2-EOI Appendix-5C Injection System Lineup RCIC Revision 5 2-E0l Appendix-7B Alternate RPV Injection System Lineup SLC System Revision 6
Page &
Procedures Used/Referenced Continued:
Procedure Number Procedure Procedure Title Rev is 10 fl 2-EOl Appendix-i 1A Alternate RPV Pressure Control Systems MSRVs Revision 4 2-EOI Appendix-12 Primary Containment Venting Revision 4 2-EOI Appendix-5B Injection System Lineup CRD Revision 3 2-EQ I Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode Revision 8 2-EOl Appendix-17C RHR System Operation Suppression Chamber Sprays Revision ii EPIP-1 Emergency Classification Procedure Revision 46 EPIP-5 General Emergency Revision 41
Pag of 44 Console Operator Instructions A. Scenario File Summary Batch File 1108-6 Pref File 110806 Irnfdg0lc Irnf dgO3 c F3 bat NRC/i l08rcicto Trg e4 NRC/dgstart F4 Trg e4 = dmf ed09c F5 bat NRC/i 10806 br zdi0hs2 11 OcO2a[ 1] trip F6 imfrdO6r3Oi5 br zloOhs2ilOcO2a[i] off F7 dmfrdO6r3Ol5 br zloOhs2l lOcO2a[2j on F8 imfog04a br zlohs7O48a[2] on F9 irnfed09c br zlohs7O48a[ij off FlO irnfsw02b br xa554c19 alarm off Fl 1 rnrfsw02 align Trg el 7048-1 F12 imfogOl Trg el bat NRC/i 10806-1 Si imfogo5a 80 1200 100 br zlohs66la[1] on S2 ior zdihs66ia open br zlohs66la[2] off br zdixs74l2i[l] reset Trg e2 modesw Imfth2i (e2 180) 0.5 600 0.1 Irnfdg03c (e2 0)
Scenario 6 DESCRIPTION/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 10 of 44 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 l) FAILURE to monitor SJAE/OG CNDR ONDS FLOW, 2-Fl-2-42. on Panel 2-9-6 for proper flow (between 2 x 10 and 3 x 1O 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 low 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[2.2.8} 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):
[7] 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- 1f 44 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[10] 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[1 3] 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.L1
[12] SLOWLY RAISE RFP speed.
- CHECK discharge flow and discharge pressure rise.
[13] WHEN REP speed is approximately equal to operating RFP(s) speed, THEN on RFPT 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:
- REPT A HP STEAM SHUTOFF ABOVE SEAT DRAIN, 2-ECV-6-1 53 (local control)
RFPTA(B)(C) LP STEAM SHUTOFF ABOVE SEAT DRAIN, 2-ECV-6-154 (local control)
Driver When called report 2-FCV-6-163 and 2-FCY-6-154 closed
Page 12-cyF44 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:
RFPT (B)(C) LP STEAM SHUTOFF ABOVE SEAT DR, 2-FCV-6-(156)(158) (local control)
[16] VERIFY both RFPT Main Oil Pumps running.
Driver When called report 2-EVC-6-1 56/1 58 are closed
[17] IF desired to stop Turning Gear for in service RFPT, THEN PLACE appropriate handswitch in STOP and RETURN to AUTO:
- RFPT 2A TURNING GEAR MOTOR, 2-HS-3-1O1A
[18] GO TO Section 6.0.
[18.1] CONTROL and MONITOR RFW System operation.
Page 13 of 44 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods SRO Direct Power Increase lAW RCP SRO Notify 0DS of power increase ATC Raise Power with Control Rods per 2-01-85, section 6.6. Control Rods 22-31, 30-39, 38-31, and 30-23 from 00 to 08, 30-3 1 from 00 to 48, 30-15 00 to 24 ATC Withdraw control rods lAW 2-01-85 NOTES
- 1) Continuous control rod withdrawal may be used when a control rod is to be withdrawn greater than three notches.
- 2) When in areas of high notch worth, single notch withdrawal should be used instead of continuous rod withdrawal. Information concerning high notch worth is identified by Reactor Engineering in Control Rod Coupling Integrity Check, 2-SR-3. I .3.5A.
- 3) When continuously withdrawing a control rod to a position other than position 48, the CRD Notch Override Switch is held in the Override position and then the CRD Control Switch is held in the Rod Out Notch position.
. Both switches should be released when the control rod reaches two notches prior to its intended position.
(Example: If a control rod is to be withdrawn from position 00 to position 12, the CRD Notch Override Switch and the CRD Control Switch would be used to move the control rod until reaching position 08, then both switches would be released.)
. If the rod settles in a notch prior to the intended position, the CRD Control Switch should be used to withdraw the rod to the intended position.
(using the above example; If the control rod settles at a notch prior to the intended position of 12, the CRD Control Switch would be used to withdraw the control rod to position 12.)
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.
Page 14 of 44 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods ATC 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 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 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 in to correct ROD GROUP when Rod Worth Minimizer is enforcing.
[4] PLACE CRD CONTROL SWITCH, 2-.HS-85-48, in ROD OUT NOTCH and RELEASE.
[5] OBSERVE control rod settles into desired position AND ROD SETTLE light extinguishes.
[6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check
Pag&i5of 44 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods
[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 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 the CRD Power, 2-HS-85-46 to the OFF position to deselect the control Rod.
[6.2] PLACE the CRD Power, 2-HS-85-46 to the ON position.
[6.3] IF desired, THEN CONTINUE to withdraw Control Rods and PERFORM applicable section for Control Rod withdraw ATC 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)
- 1. VERIFY correct control rod selected. U Z VERIFY Rod Out Permit light is not illuminated to ensure selected rod withdrawal is inhibited. U
- 3. CHECK annunciator LPRM HIGH (1-xa-55-5a. Window 12) and matrix light, Panel [-9-5 to determine if the alarm is due to high flux. U
- 4. DESELECT then RESELECT the desired Control Rod to reset the alarm and reinitialize the REM back to normalized 1OD%. U Event 3 DRIVER Insert Malfunction to F6 (imf RDO6r3OI 5) to stick rod 30-15
Page-16 of 44 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.
[2] REVIEW all Precautions and Limitations in Section 3.0 CAUTION ERC Never pull control rods except in a dellberate, carefully controlled manner, while closely monitoring the Reactors response. :PDsoc--a:i
[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 54] through 8.15161 should he used when the control rod is at Position 00 while Step 8.15[7J 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 oveilravel 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.i54] may be repeated as necessary until it is determined that this method will not free the control rod.
Pag& 17 of 44 Simulator Event Guide:
Event 3 Component: Control Rod Difficult to Withdraw (continued)
DRIVER Delete malfunction F7 (dmf rdO6r3Ol 5) when double clutch is used.
NRC NOTE The procedure first directs double clutching be used ATC
[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].
Page 18 of 44 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 AOl-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.
BOP 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(S) less than 60 psig, 2-PI-2-34(0).
Panel 26-105.
. SJAE 2A(25) CONDENSATE INLET \JALVE closed at 2-HS-2-31A(36).
Panel 2-9-S.
. SJAE 2A(2B) CONDENSATE OUTLET \IALVE closed at 2-HS-235A(z1A).
Panel 2-9-S.
. STEAM TO SJAE A(S) STAGE I & II, 2-Pl-1-150(152). Panel 25-105 is less than 155 psig. (disabled for the SJAE selected by 2-HS-001-0375)
. Loss of l&C bus A(S): power is required to be restored to return the SJAE to service.
- 3) 2-HS-O01-0375, 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.
Page 1of 44-Simulator Event Guide:
Event 4 Component: Loss of SJAE A (continued)
[11.1] 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 925-1 05 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-90-1 55, 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-66-14(18), AUTO/OPEN (Panel 2-9-8)
[11.5] PLACE SJAE 2A(2B) PRESS CONTROLLER 2-HS-1 -1 50(152) 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-1-1 51/1 66 (1 53/1 67).
- 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) OG OUTLET VALVE, 2-HS-66-18(14) (Panel 2-9-8).
- VERIFY CLOSED SJAE 2B(2A) PRESSURE CONTROLLER, 2-HS-1 -152(150) (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)
- Page2Qof44 Simulator Event Guide:
Event 4 Component: Loss of SJAE A (continued)
BOP [1 1 .10] IF the HWC System had previously been in service, (otherwise N/A)
AND WHEN stable SJAE operation is confirmed, THEN REFER TO 2-01-4, HWC System,_for_shut_down_and_restart_guidance.
BOP Notifies Chemistry of HWC Shutdown
Page 21 of44 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, and immediately delete the malfunction.
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 C to SRO.
DRIVER When requested to investigate the shutdown board, report that the NORMAL Feeder Breaker to Shutdown Board C is tripped and the smell of smoke at the compartment.
SRO Evaluates Tech Specs 3.8.7 (condition A) and 3.8.1 (condition B)
NRC NOTE SRO may identify 3.8.1 G (Off site Circuit mop) and this action is met with the 3.8.7 action.
3.8 ELECTRiCAL POVVER SYSTEMS 3.8,7 Distribution Systems - Operahng LCD 3.8.7 The fobowin AC and DC electric! power dstribuuon subsystei,s shaI be OPERABLE:
- a. Unit 1 and 2 4.16 kV Siutdcwn Boards:
- b. Urut 2 480 V Shutdown Boards:
- c. Unft 2 480 V RMO! Boards 2A, 28. 2D, and 28:
ci. Unit I and 2 DG Auxtlior Boards:
t.. Unit 1 and 2 Shutdown Board DC Dstriution Panels; and cj. Unit 1 and 3 AC and DC Boards needed to suppon equpment requtred to be OPERABLE by LCD 3.6.4.3, Standay Gas Tceatmenr (SOT) System, and LCD 3.7.3, Control Room Emercienc Veniflation CREV) System.
T APPLICABT Y : LI MODES 1, 2, and 3.
Page 22 of 44 Simulator Event Guide:
Event 5 Component: DG C Auto Start Failure SRO ACTIONS CONDiTION REQUIRED ACTION COMPLETION TIME A. One Unit I and 2 4.16 kV NOTE Shutdown Board Enter applicable Conditions and inoperable. Required Actions of Condition 8, C, D and G when Condition A.
results in no power source to a required 480 volt board.
Al Restore the Unit 1 and 2 5 days 4.16 k\f 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) 3.8 ELECTRICAL POVER SYSTEMS 3.8.1 AC Sources - Operating LCO 3,8.1 The following AC eIectrica power sources shall be OPERABLE:
- a. Two cualified circuits between the offsite tmnsmission netor and the onsite Class IE AC Eiec:rica! Power Distribution System:
- b. Unit 1 and 2 diesel generators (DGS witn two divisions of 480 V ioad shed Ioaic arid common accident signal logic OPERABLE: and
- c. Unit 3 OGIs) capable of supplying the Unit 34.16 kV shutdown boards) required by LCO 3.8.7, Distribution Systems -
Operating.
APPLICABILITY: r.1O.DES 1, 2, and 3.
Page 2o44 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 /> DO inoperable, from the offsite transmission network. AND 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 1 and 2 DO(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 DO(s).
AND B.4 Restore Unit 1 and 2 DO 7 days to OPERABLE status.
AND 14 days from discovery of failure to meet LCO
-Pag&24 o44 Simulator Event Guide:
Event 6 Component: Loss of RBCCW DRIVER Insert malfunction FlO (sw02b) 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 CLOSING/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 following:
- 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-AOl-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 EOI entry condition is met, THEN ENTER appropriate EOl(s) (Otherwise N/A).
Page 25 of 44 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 tn pped RBCCW pump(s).
DRIVER When dispatched, report RBCCW Pump 2B breaker is tripped. There is also a smell of burnt wiring and 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 service, wait 3 minutes (IRE SWO2 align). THEN inform Unit 2 Operator that spare RBCCW Pump is in service.
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 Responds to alarm RBCCW 2-ECV-70-48 Closed B. OPEN 2-FCV-70-48, RBCCW Sectionalizing Valve, when conditions permit.
C. IF unable to reopen 2-FCV-70-48, THEN if desired, REMOVE RWCU from service. REFER TO 2-01-69.
Crew Notifies Chemistry of TRM, TSR 3.4.1.1 for Reactor Coolant Conductivity monitoring ATC Opens Sectionalizing Valve, 2-FCV-70-48.
Page 26 of 44 Simulator Event Guide:
Event 6 Component: Loss of RBCCW (continued)
TR 3.4 REACTOR COOLANT SYSTEM TR 3.4.1 Coolant Chemistry LCO 3.4.1 Reactor coolant chemistry shall be maintained within the limits of Table 3.4.1-1.
APPLICABILITY: According to Table 3.4.1-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Conductivity greater A.1 Verify by administrative Immediately than the limit of means that conductivity Table 3.4.1-1 has not been > 1.0 Column B but 10 Jrnho/cn, at 25 C for> 2 3
!.Lmholcm at 25CC. weeks in the past year.
B. Chloride concentration B.1 Verify by administrative Immediately greater than the limit means that chloride of Table 3.4.1-1 concentration has not Column B or E but been> 0.2 ppm for> 2
= 0.5 ppm. weeks in the past year.
C. pH not within limits of C.1 Restore pH to within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Table 3.4.1-1 limits.
Column A. B, and E.
(continued)
Page 27 of 44 Simulator Event Guide:
Event 6 Component: Loss of RBCCW (continued)
TSR 3.4.1.1 Monitor reactor coolant conductivity.
Continuously OR 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the continuous conductivity monitor is inoperable and the reactor is not in MODE 4 or 5 OR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the continuous conductivity monitor is inoperable and the reactor is in MODE 4 or 5
Page 2& of 44 Simulator Event Guide:
Event 6 Component: Loss of RBCCW (continued)
ATC!BOP 2-01-69 RWCU 3.10 Nuclear Heat Balance (NHB)
A. When a RWCU demin is removed from service it is required to ensure proper heat balance auto subsitution for the RWCU Demin removed from service per Section 8.16.
ATC!BOP 7.0 SYSTEM SHUTDOWN 7.1 ICS Temperature Point Subsitution for Heat Balance
[1] IF removing Reactor Water Cleanup System from service when operating at power, THEN PERFORM RWCU lOS Temperature Point Substitution for Heat Balance adjustments:
NOTE The following values are to be substituted for RWCU Inlet and Outlet temperatures so RWCU parameters provide conservative input to the Integrated Computer System (lOS) thermal power calculation.
420 degrees_F_for_69-6D,_RWCU_LOOP_OUTLET TEMP.
A. TYPE SV in the yellow block at the top of the ICS display and depress Enter key.
B. At the prompt ENTER POINT ID, TYPE 69-6A and DEPRESS Return key.
C. At the prompt ENTER SUBSTITUTE VALUE, TYPE 525 and DEPRESS Return key.
D. At the prompt ENTER POINT ID, TYPE 69-6D and DEPRESS Return key.
E. At the prompt ENTER SUBSTITUTE VALUE, TYPE 420 and DEPRESS Return key.
F. At the prompt ENTER SUBSTITUTE VALUE, DEPRESS the CANCEL key.
Page-29 o 44 Simulator Event Guide:
Event 7 Major: Explosion in Off Gas due to high hydrogen Loss of condenser Vacuum.
DRIVER Insert F12 (imf ogOl) and Shift Fl, to cause High Offgas Hydrogen 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 (CH 1-Analyzer 2A, CH 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 (iorzdihs66la 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 EOl-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 RC/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 1 1A)
IF Steam cooling is required? - NO IF Suppression Pool level and temperature cannot be maintained in the safe area of Curve 3?- NO
Page 30 of 44 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 100°F/hr cooldown.
ATC Control Reactor Pressure in assigned band, lAW Appendix hA.
ATC/BOP Pressure Control lAW AppendixilA, RPV Pressure Control SRVs.
NA 1. IF Drywell Control Air is NOT available, THEN EXECUTE EOl 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-1-179 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-31 of 4+
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:
EQ 1-1 RPV Pressure Augment RPV Pressure control, as necessary; with one or more of the following depressurization systems:
. HPCIAppendixllC
. RCICAppendixllB
. RFPTs on minimum flow Appendix 1 1 F SRO
. Main Steam System Drains Appendix 1 1 D
. Steam Seals Appendix 11 G
. SJAEs Appendix 1 1 G
. Qif Gas Preheater Appendix 11 G
. RWCU Appendix 1 1 E.
ATC/BOP Augments RPV Pressure Control, if directed by SRQ.
Monitor and Control Reactor Water Level.
Directs_Verification_of_PCIS_isolations.
ATC/BQP Verifies PCIS isolations.
SRO Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with one or more of the following injection sources. (HPCI, Appendix 5D)
ATC Maintains the prescribed level band, lAW Appendix SD.
- 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 CANNQT be maintained below 4.25 in., THEN EXECUTE EQI Appendix 16E concurrently with this procedure to bypass HPCI High Suppression Pool Water Level Suction Transfer Interlock.
Rage 32 of44 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 16L 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 140F may result in equipment damage.
- 4. VERIFY 2-lL-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 NQI 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.
Page 33 of 44 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 34 of 44 Simulator Event Guide:
Event 8 Component: LOCA, Loss of SD BD C SRO Enters EOl-2, all legs.
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? - NO SRO Directs H202 Analyzers placed in service, lAW Appendix 19.
BOP Places H202 analyzers in service, lAW Appendix 19.
SRO EOI-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! 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.
Page 35 of 44 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 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 2-9-55).
- e. VERIFY 2-FIC-84-19, 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 17C NOTE Sprays are unavailable on Loop I of RHR due to failed 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 SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN ... CONTINUE in this procedure At Step 7 using RHR Loop I OR At Step 8 using RHR Loop II.
Page 36 o44 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:
- [PCI Initiation signal is NOT present, OR
- 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.
- 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 37 of 44 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 SRO 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-74-10i, RHR SYS II U-3 DISCH XTIE.
- d. STOP RHR Pumps 2A and 2C (2B and 2D).
Pag38of44 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, H2/02 ANALYZER ISOLATION BYPASS switch in BYPASS (Panel 2-9-54).
- 2. DEPRESS 2-HS-76-91, H2/02 ANALYZER ISOLATION RESET.
- 3. IF H21O2 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-76-55156 OPEN using 2-IL-76-49-1.
- c. VERIFY OPEN SMPL RTN VLVS 2-FSV-76-57!58 using 2-lL-76-49-3.
- 4. IF H2/02 Analyzer is to sample the Drywell, THEN ALIGN Analyzer as follows (Panel 2-9-54):
- a. PLACE 2-HS-76-1 10, H2/02 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.
Page 39-cf 44 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-76-110 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!02 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/O2 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 17A)
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.
Page 4Qof44 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-FI-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.
Page 41 of 44 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
CT #1 SRO When Suppression Chamber Pressure exceeds 12 psig, determines that Drywell Sprays are required.
Directs Spraying the Drywell before exceeding the PSP curve or reaching 2800 with Loop II of RHR to be placed in Drywell Sprays per EOI Appendix 17B.
CT #1 ATC/BOP Drywell Sprays per appendix 178
- 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:
e 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).
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 I(II) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
Page 42 of 44 Simulator Event Guide:
Event 9 Instrument: RHR Sys 1 Containment Spray Valve select switch failure (continued)
- d. MOMENTARILY PLACE 1-XS-74-121(129), 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:
- 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-23-46, RHR HX lB RHRSW OUTLET VLV
. 1-FCV-23-52, RHR HX 1D RHRSW OUTLET VLV.
Page 43 of 44 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 psig, OR
- 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-44of44 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 is out of service Operations/Maintenance for the Shift:
RFPT A is operating at 600 RPM.
Place RFPT A in service from 600 RPM in accordance with 2-Ol-3section 5.7.
Once RFPT A is in service perform Rod Pattern adjustment and then raise power to 100% with flow in accordance with the RCP.
Units 1 and 3 are at 100% Power Unusual Conditions/Problem Areas:
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1.3.5(A)
IJNIT2 - -
REV 0021 ATTACHMENT 2 (Page 1 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 30-15 24 00 N/A 46-3 1 24 00 N/A 30-47 24 00 N/A 14-31 24 00 N/A 30-3 1 48 00 N/A 22-3 1 08 00 N/A 30-39 08 00 N/A 38-3 1 08 00 N/A 30-23 08 00 N/A 22-23 16 00 N/A 38-23 16 00 N/A 38-39 16 00 N/A 22-39 16 00 N/A 14-23 48 00 N/A 14-39 48 00 N/A 46-39 48 00 N/A 46-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously tOO. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I Issued by Unit Supervisor Date Reactor Engineer Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1.3.5(A)
UNIT 2 REV 0021 ATTACHMENT 2 (Page 2 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-47 48 00 N/A 38-47 48 00 N/A 38-15 48 00 N/A 22-15 48 00 N/A 14-47 48 00 N/A 46-47 48 00 N/A 46-15 48 00 N/A 14-15 48 00 N/A 06-3 1 48 00 N/A 30-55 48 00 N/A 54-3 1 48 00 N/A 30-07 48 00 N/A 06-39 48 00 N/A 54-39 48 00 N/A 54-23 48 00 N/A 06-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously tOO. Insertion may stop after completion of group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I Issued by I Unit Supervisor Date Reactor Engineer Date
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U2 NRC Exam 6 Unit 2 Rod Pattern Adjustment
BEN Reactivity Control Plan Attachment 7 (Page 1 of 2)
Reactivity Control Plan Form BEN Unit: 2 Valid Date(s): 8/7/11 8/19/11 Reactivity Control Plan #: U2 NRC Exam 6 Are Multiple Activations Allowed: No (If yes, US may make additional copies)
Prepared by: / Reviewed by:
Reactor Engineer Date Qualified Reactor Engineer Date Approved by: I Concurrence: /
RE Supervisor Date WCC/RiskIUS SRO Date Approved by: I Authorized by: I Ops Manager or Supt. Date Shift Manager Date RCP Activated: / RCP Terminated: /
Unit Supervisor Date Unit Supervisor Date Title of Evolution: Unit 2 Rod Pattern Adjustment PurposelOverview of Evolution: Adjust Rod Pattern for 100% power operation after RFPT A returned to service.
Maneuver Steps
- 1. Withdraw Control rods lAWAttachment 2 provided by Reactor Engineer.
- 2. Increase flow to 100% power. No Ramp Rate Limits apply
BEN Reactivity Control Plan Attachment 7 (Page 2 of 2)
Reactivity Control Plan Form Operating Experience and General Issues: U2 NRC Exam 6 Previously known control rod issues:
4 172292 05/28/2009 Control Rod 46-27 double notched during the performance of the Unit 2 sequence exchange, 00 to 04.
4 150002 08/10/2008 During power ascension activities, control rod 46-27 double notched from position 00 to 04.
4 149981 08/09/2008 Control Rod 38-35 double notched during control rod withdrawal from 00 to 04.
4 148263 07/12/2008 While pulling control rods during U2 startup, CR 38-03 double-notched twice. 10 to 14 and 14 to 18 Cautions/Error Likely SituationslSpecial Monitoring Requirements/Contingencies:
- Rod Out Notch Override is authorized, for Rod Out Notch Override follow the guidance in 2-01-85 section 6.6.4.
- This plan is NOT valid if the unit is operating with a suspected or known fuel leaker and is not to be used. Contact Reactor Engineering if there are indications of a fuel leak.
BEN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP I of 2 Reactivity Maneuver Plan # U2 NRC Exam 6 Description of Step: Withdraw Control rods lAWAttachment 2 provided by Reactor Engineer.
Conditions : To be recorded at the Completion of Step Recorded:
(byRO) (Date)
QRE presence required in the Control Room? Yes No (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 875-1100 MFLCPR .80 - .85 MW Thermal 2650-3200 MAPRAT .45 - .55 Core Flow 80-84mlbm/hr MFDLRX .65 - .70 Loadline 103-106 Core Power 88-92% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions Comments I Notes: Rod Out Notch Override is authorized, for Rod Out Notch Override follow the guidance in 2-01-85 section 6.6.4.
Step Complete AND Reviewed by: I Unit Supervisor I Date
BEN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP 2 of 2 Reactivity Maneuver Plan # U2 NRC Exam 6 Description of Step: Raise reactor power to 100% using core flow. No Ramp Rate Limits apply Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No X (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 1150 MFLCPR .85 - .95 MW Thermal 3400-3450 MAPRAT .60 - .70 Core Flow 8595 mlbm/hr MFDLRX .70 - .75 Loadline 103-1 06 Core Power 100% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments / Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions N/A Comments / Notes:
- 1. Raise Reactor Power to 100% RTP
- 2. Document core flow changes on Attachment 10 Step Complete AND Reviewed by: /
Unit Supervisor / Date
BEN Reactivity Control Plan Attachment 10 (Page 1 ofl)
Recirc Flow Maneuver Instructions Reactivity Control Plan # U2 NRC Exam 6 RCP Flow Time Target Delta Target Completed (RO)
Step # Step # Power Flow
(%RTP or +/-(MWe) (MLbIHr)
MWe) 2 1 100%
Comments / Notes:
Reviewed by:
Unit Supervisor I Date
BEN Reactivity Control Plan ATTACHMENT 4 ROD PATTERN STEP THROUGH MAPS Reactivity Maneuver Plan # U2 NRC Exam 6 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 Prior to RCP 6 59 59 55 55 51 51 47 00 47 43 43 39 16 00 16 39 35 35 31 00 00 00 00 00 31 27 27 23 16 00 16 23 19 19 15 00 15 11 11 07 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 AfterRCP6 59 59 55 55 51 51 47 24 47 43 43 39 16 08 16 39 35 35 31 24 08 08 24 31 27 27 23 16 08 16 23 19 19 15 24 15 11 11 07 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58
BEN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1.3.5(A)
..UMT2 REVOO21 ATTACHMENT 2 (Page 1 ofi)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-3 1 00 08 N/A 30-39 00 08 N/A 38-31 00 08 N/A 30-23 00 08 N/A 30-31 00 48 N/A 30-15 00 24 N/A 46-3 1 00 24 N/A 30-47 00 24 N/A 14-3 1 00 24 4
REMARKS Control Rod Pattern Adjustment NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second U0, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thennal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. I?N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by:____________________________ Issued by Unit Supervisor Date Reactor Engineer Date
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 JAW 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 TS-SRO section 5.2, Relative Humidity heater fails for TS action R-ATC 2 Raise Power with Flow RSRO R-ATC 3 ADO1a TS-SRO ADS SRV 1-5 fails open C-BOP CATC VFD Cooling Water Pump 2B trips with failure of the standby 4 THJ8d C-SRO pump to auto start RATC 5 FWO5b C-BOP B2 Feedwater Heater Leak C-SRO 6 FW3Oa Feedwater Pump 2A Governor Drifts Up 7 Batch File M-ALL Earthquake, Loss of All High Pressure injection 8 Loss of LPCI MG sets, loss of ALL Level Control Systems C
Steam Cooling 9 ALL Emergency Depressurization (N)onal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or
Appendix D Scenario Outline Form ES-D-1 EVENTS
- 1. BOP starts SBGT Fan C and aligns to Reactor Bldg lAW 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 detennine 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-i.
- 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-1, EOI-2 and EOJ-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-l, 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.
Loss of All injection sources When crew enters steam cooling, one LPCJ 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 Outilne 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 -195 inches, direct Emergency Depressurization prior to -215 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 when RPV level lowers to -200 inches.
- 4. Feedback:
RPV pressure trend.
SRV status indications.
CT#2 With RPV pressure below the Shutoff I-lead 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.
Appendix D Scenario Outline Form ES-D-1 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 AIB INHIBITED annunciator status.
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 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 ROU-O65NO-O2 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 ROU-068-AL-33 202001A2.22 3.1 3.2 SRO S-068-AB-O1 Loss of Feedwater Heating RO U-006-AB-O1 2.1.43 4.1 4.3 SRO 5-006-AB-Ol Reactor Feed Pump Turbine Governor Failure RO U-003-AL-9 259002A4.O1 3.8 3.6 SRO S-003-AB-1 Steam Cooling ROU-000-EM-15 295031EA2.04 4.6 4.8 SRO S-000-EM-15 SRO T-000-EM-16
Appendix 0 Scenario Outline Form ES-.D-1 Simulator Instructor 1C93 Batch 1108-7 Preference File 110807 Trg 11 NRC/msrvinhibit F3 bat NRC/110807 Trg 11 =dmfad0la F4 bat csloop2to br zlohs682b2a[i] on F5 imfadola70 br zlohs682b2a{2] off F6 ior zdihs682bla[1J off Mrfthl8d trip F7 imffw05b 100 300 75 Trg 15 NRC/bvfd F8 Trg 15 =batNRC/110807-1 F9 th22 100 5:00 50 Trg 1 rnodesw FlO dmfedl2a Trg 1 = bat NRC/i 10807-4 Fil mrf edO9 norm br zdihs858a[1] close F12 mrfrp02 reset Trg 17 NRC/rcic Si mrfsl0i align Imfrc09 (e17 1:00) 100 1:00 S2 imfadOla 100 Trg 10 NRC/rfptarnaual Trg 10 = dmffw30a br ypovfcvo52l fail_power now br zlohsO52la[2] on Trg 5NRC/fwheating zdihso52la{1].eq. I Trg 5 = bat NRC/i 10807-2 br ypomtrsbgtrrh fail_control_power Batch 110807-4 Batch 110807-2 Imfsl0ia and slOib br ypovfcv052l fail_power_now Imfedlia and edlib (el 4:00) br zlohsO52la[2] on Irnfedl2a and edl3a (ei 1:00)
Imfedlic and edlid (el 5:00) Batch 110807-2 Irnffwl9 (el 0)100 3:00 mrfthi8d close Imfth2l (el 8:00) .25 1200 dor zlohs682b2a[ij Imfrd0ia (el 3:00) dor zlohs682b2a[2]
Manually Enter FW3OA
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 JAW 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 5 FWO5b C-BOP B2 Feedwater Heater Leak C-SRO C-ATC 6 FW3Oa Feedwater Pump 2A Governor Drifts Up CSRO 7 Batch File M-ALL Earthquake, Loss of All High Pressure injection Loss of LPCJ MG sets, loss of ALL Level Control Systems 8 C Steam Cooling 9 ALL Emergency Depressurization (N)ormal, (R)eactivity, (J)nstrument, (C)omponent, (M)aj or
Appendix B Scenario Outline Form ES-B-i Critical Tasks - Four CT#1 With NO injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -195 inches, direct Emergency Depressurization prior to -215 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 when RPV level lowers to -200 inches.
- 4. Feedback:
RPV pressure trend.
SRV status indications.
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 or 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.
Appendix B Scenario Outline Form ES-B-i 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 AIB IN}IIBITED annunciator status.
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 B Scenario Outline Form ES-B-i 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 iC 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-1, 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.
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 B Scenario Outline Form ES-B-i 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 KIA RO SRO Manual bitiation of SBGT Fai 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-001-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 S-068-AB-01 Loss of Feedwater Heating RO U-006-AB-O1 2.1.43 4.1 4.3 SRO S006-AB-O1 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 Pge7of 34 Procedures UsedlReferenced:
Procedure Number [
0-01-24 Procedure Title Standby Gas Treatment System Procedure Revision Revision 53 1
TS 3.6.4.3 Containment Systems Amendment 290 2-G0I-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-A0I-6-IC 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-l .
Revision 20 HighlLow 0-AOI-lOO-5 Earthquake Revision 33 2-AOI-lOO-l Reactor Scram Revision 95 2-E0I-1 RPV Control Flowchart Revision 12 2-EOI-2 Primary Containment Control Flowchart Revision 12 2-EOI-2-C-i Alternate Level Control Flowchart Revision 9 2-EOI-2-C-2 Emergency RPV Depressurization Revision 6 2-E0I Appendix-6D Injection Subsystems Lineup Core Spray System I Revision 7 2-E0I-.APPENDIX-l7A RHR System Operation Suppression Pool Cooling Revision 12 2-EOI Appendix-5C Injection System Lineup RCIC Revision 5
Scenario 7 Pag&gyf34 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 Confrol Systems MSRVs Revision 4 2-EOI Appendix-12 Primary Containment Venting Revision 4 2-EQI Appendix-i 7C RHR System Operation Suppression Chamber Sprays Revision 0 EPIP-1 Emergency Classification Procedure Revision 46 EPIP-5 General Emergency Revision 41
Scenario 7 Page of 34 Console Operator Instructions Batch 1108-7 Preference File 110807 Trg 11 NRC/msrvinhibit F3 bat NRC/110807 Trg 11 =dmfad0la F4 bat csloop2to br zlohs682b2a[l] on F5 imfadola 70 br zlohs682b2a[2] off F6 ior zdihs682bla[i] off Mrfthl8d trip F7 imffw05b 100 300 75 Trg 15 NRC/bvfd F8 Trg 15 =batNRC/110807-1 F9 th22 100 5:00 50 Trg 1 rnodesw FlO dmfedl2a Trg 1 =batNRC/110807-4 Fil mrfed09nonn br zdihs858a[1] close F12 mrfrpO2 reset Trg 17 NRC/rcic Si mrf slOb align brnfrc09 (e17 1:00) 100 1:00 S2 imfadOla 100 Trg 10 NRC/rfptarnaual Trg 10 = dmffw30a br ypovfcvo52i fail_power now br zlohsO52la[2] on Trg 5NRC/fwheating zdihs052l a[ 1].eq. 1 Trg 5 bat NRC/i 10807-2 br ypomtrsbgtrrh fail control_power Batch 110807-4 Batch 110807-2 ImfslOla and slOib br ypovfcvo52l fail_power now Imfedllaandedllb (el 4:00) br zlohs052la[2] on bmfedi2a and edl3a (ci 1:00)
Imf cdi ic and edi ld (el 5:00) Batch 110807-2 Imffwl9 (el 0)100 3:00 mrfthi8d close Imfth2l (ci 8:00) .25 1200 dor zlohs682b2a[i]
Imfrd0ba(ei 3:00) dor zlohs682b2a[2]
Manually Enter FW3OA Scenario 7 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 93 Simulator Setup Load Batch RestorePrefNRC/110807 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 lOof 34 Simulator Event Guide:
Event I Normal: Start SBGT Fan C and align to Reactor Bldg JAW 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 JAW 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-65-51A 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-llof34 Simulator Event Guide:
Event 1 Normal: Start SBGT Fan C and align to Reactor Bldg JAW 0-01-65 section 5.2 BOP should identify failure of the RH during procedure execution, step 6.2 on previous page. If BOP turns the RH control switch out of the AUTO position, 2-9-3B, window 5 (SGT TRAiN C SWITCHES MISALIGNED), will alarm, however, the RH will not work with switch in either position (AUTO or ON)
NRC NRC IF the BOP fails to inform the SRO that the relative humidity heater failed to energize, THEN the Chief examiner will notify the booth driver to call the SRO (as UO) and inform
/DRIVER /DRIVER him of the problem.
2-ARP-9-3B, Window 5 - SGT TRAIN C SWITCHES MISALIGNED A. CHECK each hand switch in normal operating position in accordance with 0 65, Attachment 2.
B. If possible, CLEAR initiating signal. Otherwise REFER TO Tech Spec 3.6.4.3.
C. NOTIFY UNIT SUPERVISOR/SRO and Unit 1 and Unit 3.
SRO SRO Evaluate Technical Specification 3.6.4.3 LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
Condition A One SGT subsystem inoperable Required Action A. 1 Restore SGT subsystem to OPERABLE status Completion Time 7 Days NOTE: This LCO applies to ALL 3 UNITs
Scenario 7 1age 12 of 34 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-GOT- 1 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 2-SR-3.3.5(A) and 2-01-68.
ATC Raise Power w/Recirc JAW 2-01-68, section 6.2
[1] IF desired to control Recirc Pumps 2A and/or 2B 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-96-1 6A(1 6B).
[2] 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 Manipulation ADS SRV Fails Open requiring power to be lowered to less thaii 90%
Driver Driver At lead floor instructor direction F5, for failure of ADS SRV 1-5.
Scenario 7 P-age 13 of34 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open BOP Report alarm MAIN STEAM RELIEF VALVE OPEN (2-9-3 C 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. REFERTO2-AOI-1-l.
SRO Enters 2-AOl-i-i BOP 4.1 Immediate Action
[1] IDENTIFY stuck open relief valve by OBSERVING the following:
SRV TAILPIPE FLOW MONITOR, 2-FMT-1-4, on Panel 2-9-3, OR
- MSRV DISCHARGE TAILPIPE TEMPERATURE recorder, 2-TR- 1 -1 on Panel 2-9-47.
BOP Identifies ADS SRV 1-5 open ATC [2] IF relief valve transient occurred while operating above 90% power, THEN PERFOR11 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] WIULE 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:
[1] 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-i-202 in IN}{IBIT:
CT#4 Observe and report when 2-XS-l-202 is placed in Inhibit, ADS SRV 1-5 closes.
Scenario 7 Page 14 of34 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open BOP [5] IF relief valve closes, THEN OPEN breaker or PULL fuses as necessary using 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 Driver Driver If MSRV AUTO Actuation Logic Inhibit Switch is returned to Auto prior to pulling fuses insert imf adO 1 a (shift F2)
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-i-S 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 Fuse 2E-F6B (Block AA, F14)
Scenario 7 Page 15 of 34 Simulator Event Guide:
Event 3 Component: ADS SRVs Fail Open Driver Driver When directed to remove power from SRV 1-5, insert mrf adOla OUT in two minutes 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 R}IRSW Pump is operating on each EECW Header.
[7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows:
[7.1] START an RHRSW Pump to supply RHR Heat Exchanger A(C).
[7.2] ESTABLISH RHRSW flow by perfonning one the following:
[7.2.2] 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.
[7.3] VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 2-FCV-74-53.
[7.4] VERIFY CLOSED RHR SYS I SUPPR POOL CLG/TEST VLV, 2-FCV-74-59.
[7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-58.
[7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60.
[7.7] VERIFY OPEN RHR SYS I SUPPR CHBR/POOL ISOL VLV, 2-FCV-74-57.
[7.9] START RHR PUMP A(C) using 2-HS-74-5A(16A).
[7.10] ThROTTLE RHR SYS I SUPPR 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 hi 2 Operatn Loop 6ow 7000 to <1 3,000 pni &
10.000 gpm & 6[ue ue Firjht IiQht iliumthated ruminatd
[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 REPERFOR1eVIING_Step_8.5[7]_for the_second pump.
Scenario 7 Page 16-of 34 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) andlor 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 INBD 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.
[10.7] VERIFY OPEN RHR SYS II STJPPR CHBR/POOL ISOL VLV, 2-FCV-74-71.
[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.
nnps in 1 2 on Loop Finw 7000 to <1 3,000 pm &
10.000 opni & 5[ue 3Iue Fight icih .ilinrnnated illun1natd
[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 Driver At lead floor instructor direction , for trip of 2-B-i VFD Cooling Pump
Scenario 7 Page 17 of34 Simulator Event Guide:
Event 4 Component: VFD Cooling Water Pump 2-B-i Failure ATC Reports the following annunciators 4B-i2, 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 DRIVE 2B COOLANT FLOW LOW STARTS RECIRC DRIVE cooling water pump 2-B-2 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.
DRIVER Wait 4 minutes after dispatched, THEN report tripped VFD Pump 2-B-i is chot to the touch,_internal_bkr_closed,_480_volt bkr_tripped_(480_V_SD_BD_2A-5D).
DRIVER Upon Lead examiner direction 7 for Loss of Feedwater Heating
Scenario 7 Page 18of 34 Simulator Event Guide:
Event 5 Component: B2 Feedwater Heater Leak DRIVER When directed by NRC insertF7 for Loss of Feedwater Heating and 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV Fail to isolate.
ATC/BOP 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.
ATC/BOP 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-6-1A or 2-AOI-6-1C.
ATC/BOP 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 Page49 of-34 Simulator Event Guide:
Event 5 Component: B2 Feedwater Heater Leak ATC/BOP Announces B1 and B2 High Pressure Heater Extraction Isolation SRO Directs crew to enter 2-AOI-6-1A or 2-AOI-6-IC ATC/BOP 2-AOI-6-1A High Pressure Feedwater Heater String/Extraction 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 TTJRB 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 20 of 34 Simulator Event Guide:
Event 5 Component: B2 Feedwater Heater Leak ATC/BOP 2-AOI-6-1A High Pressure Feedwater Heater String/Extraction Steam Isolation (continued) 4.2 Subsequent Actions (continued)
[9] 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 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 Directs power reduction to 920 MWe (79%) power (Power Reduction with RCP flow or Control Rods) per 2-01-6, Illustration 1 SRO 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 21 of 34 Simulator Event Guide:
Event 5 Component: B2 Feedwater Heater Leak 2-AOI-6-1A Attachment 1 Closes the following Feedwater Valves Manually 2-FCV-3-3 1, 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 HEATER 2B1 EXTR ISOL VLV BOP 2-FCV-5-21, HP HEATER 2B2 EXTR ISOL VLV 2-FCV-6-74, MOISTURE SEP LC RES Bi 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 213 HP HTR string (i.e. tube leak) and reports to crew DRIVER After HS for 2-FCV-5-21 taken to closed, verif 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 Pag22 of 34 Simulator Event Guide:
Event 6: Feedwater Pump 2A Governor Drifts Up DRIVER When NRC directs, insert imf fw3 0a check current setting of fw3 Oa 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 must be manually reinserted and deleted when the operator returns the Governor control knob back down to force the operator to control level manually. For Example (irnf fw30a 100 1200 67.05)
NOTE NRC Annunciator 2-9-6C Window 32, RFP DISCH FLOW LOW, will alarm at approximately 83% of malfunction severity if the crew does not notice the failure before the alarm.
ATC Report Rising Reactor Water Level and RFPT is not responding.
SRO Direct manual control of operating RFPT and Enter 2-AOI-3 1. -
NOTE NRC The crew may decide to trip the 2A RFPT per 2-AOI-3-1 step 4.2 [6].
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:
[2.1] PLACE individual RFPT Speed Control Raise/Lower switches in MANUAL GOVERNOR (depressed position with amber light illuminated).
[2.2] 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.
[8] 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 inserted or at NRC direction initiate for LOCA and make Earthquake calls
Scenario 7 Page 23-of 34 Simulator Event Guide:
Event 7 Major: Earthquake Driver Driver Report confirmed earthquake Unit 1 is handling 0-AOl-i 00-5, 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-AOl-lOU-i, EOI-1 and EOI-2 on High Drywell Pressure ATC 2-AOl-lOU-i
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AJS3A and 2-HS-99-5AJS3B, on Panel 2-9-5
[2] IF scram is due to a loss of RPS, THEN (Otherwise N/A)
[3] REFUEL MODE ONE ROD PERMISSIVE light check:
[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-Sl, 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-Sl, in SHUTDOWN position.
Driver Driver Ensure trigger 1 goes active on MODESWITCH
Scenario 7 Page 24-of 34 Simulator Event Guide:
Event 7 Major: Earthquake Feedwater Line Break Driver Driver Report confirmed earthquake Unit 1 is handling 0-AOl- [00-5, 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-l 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 llA..
RC/L Monitor and Control RPV Water Level.
Verify as Required:
. PCIS Isolations (Groups 1, 2 and 3)
. ECCS
. RCIC Directs level band of +2 to +51 inches, with Appendix 5C, SB and/or 7B.
Scenario 7 Pag25 of 34 Simulator Event Guide:
Event 7 Maj or: Earthquake Feedwater Line Break ATC/BOP Pressure Control lAW Appendix 1 1A, RPV Pressure Control SRVs 1 IF Drywell Control Air is NOT available, THEN:
EXECUTE EO1 Appendix SG, 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. 2-PCV-1-179 MN STM LINE A RELIEF VALVE
- b. 2-PCV-1-180 MN STM LiNE D RELIEF VALVE.
- c. 2-PCV-i-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-i-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
- Page26 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems Steam Cooling NOTE NOTE When RCIC is started, a break will occur on the RCIC 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 27 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems Steam Cooling SRO Enters C-i, Alternate Level Control Verifies ADS Inhibited Directs lineup of Injection Systems frrespective 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 Driver Driver Once steam cooling is entered insert FlU (dmf edl2a), Then close normal feeder breaker to RMOV Bd 2A insert JJ (mrf edO9 norm). Notify crew that RMOV Bd 2A is restored. Then insert (mrfrpo2 reset) to reset RPS B.
NOTE NOTE Restoration of RMOV Bd2A makes Core Spray Loop I available.
SRO C-i, Alternate Level Control 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 Pag28 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems Steam Cooling SRO C-i, Alternate Level Control (Cont.)
CT#i If RPV Water Level drops to 495 inches continue Emergency Depressurization is required Enters C-2 Directs maximizing RPV Injection from all available sources irrespective of pump NPSH and Vortex Limits Directs Emergency Depressurization before RPV Level reaches -215 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#1 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#i 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 Pag29 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems 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 SIJPPR 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-i, RPV Control at step RC/L-1 SRO Enters EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below I 60F 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 30 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems 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 1NBD ISOL VALVE, 2-FCV-64-34, SUPPR CHBR INBOARD ISOLATION VLV, 2-FCV-64-32, SUPPR CHBR VENT LNBD 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, SIJPPR CHBR VENT 1NBD 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 2-9-5 5).
- d. PLACE keylock switch 2-HS-84-19, 2-FCV-84-19 CONTROL, in OPEN (Panel 2-9-55).
- e. VERIFY 2-FJC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scfm.
- f. CONTINUE in this procedure at step 12.
Scenario 7 Page31 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCJ MG sets, loss of ALL Level Control Systems 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 CHBRDW VENT ISOL BYP SELECT, to SIJPPR-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 2-9-5 5).
- e. PLACE keylock switch 2-HS-84-20, 2-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 2-9-5 5).
- 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 iCi/s AND 0-SI-4.8.B.1.a.1 release fraction of 1.
DRIVER Acknowledge Notification
Scenario 7
- Page32of34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems 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 1 7C
- 1. BEFORE Suppression Chamber pressure drops below 0 psig, CONTINUE in this procedure at Step 6.
- 2. IF Adequate core cooling is assured OR Directed to spray the Suppression Chamber irrespective of adequate core cooling, THEN BYPASS LPCI injection valve open interlock as necessary:
PLACE 2-HS-74-155A, LPCI SYS I OUTBD NJ VLV BYPASS SEL in BYPASS.
EOI Program Manual 0-Vu-A 3.9 I. Use of Containment Cooling Modes While Executing C-l Alternate Level Control.
During execution of C-I Alternate Level Control, if less thin two Condensate. LPCI, CS Injection Subsystems can be aligned with pumps running per step C1-5, then available RF[R injection subsystems must be aligned until the two subsystem requirement is met. Containment Cooling NRC must be secured from those RHR subsystems that are aligned for injection.
Step C1-5 does not count the number of pumps running. It counts the number of independent injection subsystem paths aligned that have at least one pump running, from the following five subsystems: Condensate, LPCI System I, LPCI System II, CS System I, CS System II.
If at least two Condensate, LPCI, CS Injection Subsystems are aligned with pumps running in each per step C1-5, any RHR loop that is excess to the two required injection subsystems may be aligned for Containment Cooling mode.
Scenario 7
-Page 33 of 34 Simulator Event Guide:
Event 8 Major: Loss of LPCI MG sets, loss of ALL Level Control Systems 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), R}{R 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.
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 The Emergency classification is 1.1-Gi
Scenario 7 Page-34 of 34 SHIFT TURNOVER SHEET Equipment Out of Service/LCOs:
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 lAW 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
BFN - Reactivity Control Plan TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT Reactivity Maneuver Plan U2 NRC Exam 7 Raise Reactor Power to 100%
BFN Reactivity Control Plan Attachment 7 (Page 1 of 2)
Reactivity Control Plan Form BFN Unit: 2 Valid Date(s): 8/7/1 1 8/19/11 Reactivity Control Plan #: U2 NRC Exam 7 Are Multiple Activations Allowed: No (If yes, US may make additional copies)
Prepared by: / Reviewed by: /
Reactor Engineer Date Qualified Reactor Engineer Date Approved by: / Concurrence: /
RE Supervisor Date WOO/Risk/US SRO Date Approved by: / Authorized by: /
Ops Manager or Supt. Date Shift Manager Date RCP Activated: / RCP Terminated:
Unit Supervisor Date Unit Supervisor Date Title of Evolution: Power increase with flow to 100%
Purpose/Overview of Evolution: Raise Power to 100%
Maneuver Steps
- 1. Increase flow to 100% power. No Ramp Rate Limits apply
BFN Reactivity Control Plan Attachment 7 (Page 2 of 2)
Reactivity Control Plan Form Operating Experience and General Issues: U2 NRC Exam 7
- This plan is NOT valid if the unit is operating with a suspected or known fuel leaker and is not to be used. Contact Reactor Engineering if there are indications of a fuel leak.
Cautions/Error Likely Situations/Special Monitoring Requirements/Contingencies:
NONE
BFN Reactivity Control Plan Attachment 8 Reactivity Maneuver Instructions STEP I of I Reactivity Maneuver Plan # U2 NRC Exam 7 Description of Step: Power increase with flow to 100%. No Ramp Rate Limits apply Conditions : To be recorded at the Completion of Step Recorded:
(by RO) (Date)
QRE presence required in the Control Room? Yes No X (check)
Predicted Actual Predicted Actual (may be ranges) (may be ranges)
MW Electric 1 050-1 150 MFLCPR .85 - .95 MW Thermal 3300-3450 MAPRAT .60 - .70 Core Flow 85-94mlbm/hr MFDLRX .70 - .75 Loadline 104-108 Core Power 95-100% Other Critical Parameters: To be recorded DURING Step. IF parameters are outside of the predictions, THEN discuss with the RE AND record conclusions in the Comments I Notes section.
Description including frequency, method of monitoring, AND High Low contingency actions Maximum MWth 3458 Comments I Notes:
- 1. Raise Reactor Power to 100% RTP
- 2. Document core flow changes on Attachment 10 Step Complete AND Reviewed by: I________
Unit Supervisor I Date
BFN Reactivity Control Plan Attach ment 10 (Page 1 of 1)
Recirc Flow Maneuver Instructions Reactivity Control Plan # U2 NRC Exam 7 RCP Flow Time Target Delta Target Completed (RO)
Step # Step # Power Flow
(%RTP or ÷(MWe) (MLb/Hr)
MWe)
I 1 100%
Comments / Notes:
Reviewed by:
Unit Supervisor / Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.1.3.5(A)
UNIT 2 -.
REV 0021 ATTACHMENT 2 (Page 1 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM 1 ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 30-31 22 00 N/A 22-23 08 00 N/A 3 8-23 08 00 N/A )O
.)O-.) 08 00 N/A 22-39 08 00 N/A 30- 15 48 00 N/A 46-3 1 48 00 N/A 30-47 48 00 N/A 14-3 1 48 00 N/A 14-23 48 00 N/A 14-39 48 00 N/A 46-39 48 00 N/A AF l
-fO-L.) 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Jisert Rods Continuously to 00. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: L Issued by I Unit Supervisor Date Reactor Engineer Date
BFN CONTROL ROD COUPLING INTEGRITY CHECK 2-SR-3.l.3.5(A)
UNIT 2 REV 0021 -
ATTACHMENT 2 (Page 2 of 2)
Date: Today CONTROL ROD MOVEMENT DATA SHEET RWM ROD Rod Movement Completed GP NUMBER FROM TO INITIALS 2
N/A 22-47 48 00 N/A 38-47 48 00 N/A 38-15 48 00 N/A 22-15 48 00 N/A 14-47 48 00 N/A 46-47 48 00 N/A 46-15 48 00 N/A 14-15 48 00 N/A 06-3 1 48 00 N/A 30-55 48 00 N/A 54-31 48 00 N/A 30-07 48 00 N/A 06-39 48 00 N/A 54-39 48 00 rA N/A )
.)-+-L.D 48 00 N/A 06-23 48 00 REMARKS 4 Emergency Shove Sheet Loadline reduction or Unit Shutdown Insert Rods Continuously to 00. Insertion may stop after completion of any group.
NOTES:
(1) RWM Group may be marked N/A if not applicable (i.e., when above the LPSP).
(2) For all rod moves to position 48, this signoff verifies coupling integrity was checked in accordance with 2-01-85.
(3) Second-party verification by a second UO, RE, or STA is required ONLY when the RWM is inoperable or bypassed with core thermal power < 10%. A Peer Checker (not required in emergencies) may initial when second party is not required. N/A if not applicable.
(4) Record the rod number and any problems encountered, as applicable.
(5) Peer check by RE or SRO. The SRO should be checking the FROM and TO control rod positions as a minimum. The RE or SRO should be checking the positions identified for agreement with the predictor cases. Anytime the SRO feels the Peer check is beyond his knowledge level, then call in a second RE to perform the required Peer check.
Reviewed by: I Issued by /
Unit Supervisor Date Reactor Engineer Date