ML13214A269
| ML13214A269 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/01/2013 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| Download: ML13214A269 (266) | |
Text
NRC Scenario 1 acthty Browns Ferry NPP Scenario No NRC 1- Op-Test No j.
Examiners:_____________________ Operators: SRO:______________________
ATC:_______________
BOP:_______________
Initial Conditions: Reactor Power is 70%. HPCI is tagged out. Condensate Booster Pump 3A has been returned to service.
Turnover: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3. Commence a power increase in accordance with Reactivity Control Plan to 90%.
Event Maif. No. Event Type* Event Description No.
N-BOP 1 Place 3A RFPT in service from 600 RPM JAW 3-01-3 N-SRO R-ATC 2 Power increase with Control Rods R-SRO C-ATC 3 RDOJA CRD Pump Trip C-SRO C-ATC Control Rod 3 0-23 High CR1) Temperature 4 sev file TS-SRO C-BOP 5 ed07b Loss of 480V Unit Board 3B C-SRO 1-BOP 6 0ff Gas Recombiner Catalyst failure, with Hydrogen explosion ogO5aIogOl TS-SRO Batch M-ALL Loss of Feedwater / Fuel Failure / RCIC Steam Leak / ED on High 7
Radiation 8 RHO1A/C C 3A and 3C RHR Pumps trip 9 IOR I ADS SRV 1-34, Acoustic Monitor indication failure
- (N)ormal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor 1
NRC Scenario 1 Critical Tasks Three 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 without delay.
- 1. Safety Significance:
Places the primary system in the lowest possible energy state, rejects heat to the suppression pooi in preference to outside the containment, and reduces driving head and flow of system discharging into the secondary containment.
- 2. Cues:
Procedural compliance.
Secondary containment area temperatures, level, and radiation indication.
Field reports.
- 3. Measured by:
Observation US transitions to C-2 and RO opens at least 6 SRVs when two or more areas are greater than their maximum safe operating values for the same parameter.
- 4. Feedback:
RPV pressure trend.
SRV status indications.
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 1 7B AND Observation RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment 2
NRC Scenario 1 Critical Tasks - Three When MAIN STEAM LINE RADIATION HIGH-HIGH is received and SLC injection is NOT required in accordance with RC/Q of EOI-1, MSIVs are closed. SRO will direct MSIV closure without delay and the RO will close MSIVs.
- 1. Safety Significance:
Isolate potential discharge path to environment.
- 2. Cues:
Procedural compliance.
Reactor Pressure trend.
Pressure Control on SRVs
- 3. Measured by:
When ARP-9-3A window 27 annunciator is received MSIVs are closed if alarm is verified valid and SEC injection lAW EOI-l RC/Q is NOT required.
- 4. Feedback:
Reactor Pressure trend.
MSIVs indicate closed.
3
NRC Scenario 1 Events
- 2. ATC will commence to raise power with control rods to a load line of 105 to 106, by withdrawing four control rods from 00 to 24.
- 3. CRDH pump 3A trips while the ATC is withdrawing the fourth control rod. ATC will perform 3-AOI-85-3 actions to start the Standby CRD Pump and restore CRD parameters.
- 4. Once Control Rod 3 0-23 is withdrawn to position 24, this control rod will experience a high CRD mechanism temperature. The ATC will respond JAW ARPs and 3-01-85, the operator will perform CRD flushing for a control rod at a position other than 48. The SRO will declare control rod 30-23 slow JAW Technical Specification 3.1.4-1 Note 1. Enter Tech Spec 3.1.4 or 3.1.3.
- 5. Crew will respond to a loss of 480V Unit Board 3B, the BOP will restore from a trip of Bus Duct Cooling 3A Fan and take action lAW with ARPs, start standby Bus Duct Cooling Fan 3B.
- 6. The Hydrogen Water Injection system will malfunction resulting in high hydrogen concentration in Off Gas. The BOP Operator will respond JAW with ARPs and 3 -AOI 1 an Off Gas System explosion will occur once off gas hydrogen concentration exceeds 5%. The SRO will evaluate TRM 3.7.2 and enter Condition A.
- 7. On the SCRAM a loss of Unit Board 3C will occur with a trip of Condensate Booster Pump 3A and 3B, resulting in a loss of feedwater. RCIC will auto start and slowly start to restore Reactor Level. Shortly after the scram a LOCA will occur requiring Suppression Chamber and Drywell Sprays. Reactor Level will begin to lower. An unisolable RCIC steam leak will develop and Emergency Depressurization will be required due to High Secondary Containment Radiation Levels
- 8. RHR Pumps on Division I will trip and be unavailable for Containment Cooling. The crew will use Division 2 RHR Pumps for Containment Cooling functions.
- 9. When the Emergency Depressurization is required the Acoustic Monitor on ADS SRV 1-34 will fail, requiring the crew to open an additional SRV.
- 10. Once Emergency Depressurization is complete the crew will restore Reactor Level with Core Spray Loops 1 or 2, and the Condensate Pumps.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Drywell Sprays initiated and secured Emergency Depressurization complete Reactor Level is restored and maintained 4
NRC Scenario 1 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 1 8 Total Malfunctions Inserted: List (4-8) 3 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) 65 Validation Time (minutes) 3 Crew Critical Tasks (2-5)
YES Technical Specifications Exercised (Yes/No) 5
NRC Scenario I Scenario Tasks TASK NUMBER Q iQ Place a RFPT in Service RO U-003-NO-04 259001A4.0l 3.6 3.5 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-NO-31 2.2.2 4.6 4.1 CR1) Pump Trip RO U-085-AL-07 201001A2.01 3.2 3.3 SRO S-085-AB-03 CRD Mechanism Temperature High RO U-085-AL-08 201003A2.10 3.0 3.4 SRO S-085-AB-03 Loss of 480V Unit Board 3B RO U-57A-AL-15 26200 1A2.04 3.8 4.2 SRO S-57B-NO-07 Off Gas Explosion RO U-066-AB-O1 271000 A2.06 3.5 3.9 SRO S-066-AB-01 Fuel Failure RO U-090-AL-07 272000A2.01 3.7 4.1 SRO S-000-EM-02 RCIC Steam Leak RO U-000-EM-10 295033EA2.01 3.8 3.9 SRO 5-000-EM-b LOCA RO U-000-EM-01 295009AA2.01 4.2 4.2 SRO S-000-EM-01 Secondary Containment High Radiation RO U-090-AL-04 295033EA2.01 3.8 3.9 SRO S-000-EM-15 6
NRC Scenario 1 Procedures Used/Referenced:
Procedure Number ] Procedure Titk 3-01-3 Reactor Feedwater System 3-G0I-100-12 Power Maneuvering 3-01-85 Control Rod Drive System 3-A0I-85-3 CRD System Failure 3 -TI-3 93 Evaluation of CRD Temperature Alarms Technical Specifications 3-ARP-9-7A Panel 9-7 3-XA-55-7A 3-ARP-9-8A Panel 9-8 3-XA-55-8A 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-8C Panel 9-8 3-XA-55-8C 3-ARP-9-53 Panel 9-53 3-XA-55-53 3-A0I-66-i Off Gas H2 High 3-A0I-iOO-i Reactor Scram 3-E0I-i RPV Control 3-E0I-2 Primary Containment Control 3 -E0I-3 Secondary Containment Control 3-E0I Appendix-5C Injection System Lineup RCIC
. Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 3-E0I Appendix-8F Isolation 3-E0I Appendix-17C RHR System Operation Suppression Chamber Sprays 3 -E0I Appendix-i 7B RHR System Operation Drywell Sprays 3-ARP-9-3A Panel 9-3 3-XA-55-3A 3-EOI Appendix-i iA Alternate RPV Pressure Control Systems MSRVs 3-EOI Appendix-6A Injection Subsystems Lineup Condensate 3-E0I Appendix-6D Injection Subystems Lineup Core Spray System I 3-E0I Appendix-6E Injection Subystems Lineup Core Spray System II 3 -EOI-2-C-2 Emergency RPV Depressurization EPIP-i Emergency Classification Procedure EPIP-4 Site Area Emergency 7
NRC Scenario 1 Console Operator Instructions A. Scenario File Summary Batch File
- HPCI tagout bat nrchpcito
- bus duct cooling fan trip imfed07b (el 0)
- Rod 3 0-23 high temperature sev rdkxleak(63) -1.01 sev yua85795 240 trg9=sevyua85795 345 trg 10 = sev rdkxleak(63) -.9 Do NOT use this trigger trg 11 = sev rdkxleak(63) -.96 Do NOT use this trigger trg 12 = sev rdkxleak(63) .768
- hwc malfunction imfogo5a(e3 0)85 1200 100 imfogOl (e5 0) mrfogo6 byp
- CRD A Pumptrip imfrd0la (e7 0)
- Major ior zaoxi0l34 0 Acoustic monitor failure ADS SRV 1-34 imf ed08c (e20 0) Loss of Unit Board 3C imf fw02a (e20 0) Condensate Booster Pump 2A Trip imf fw02b (e20 0) Condensate Booster Pump 2B Trip trg e20 MODESW imfrc09 (e20 12:20) 25 120 RCIC Steam Leak imfth2l (e20 5:00) .2 LOCA imfpcl6a (e20 5:10) Vacuum Breaker failure imfpcl6b (e20 5:10) imfth23 (e20 0)10 1000 Fuel Failure imf rc 10 RCIC Steam Valves fail to Auto Close ior zdihs7l2a open RCIC Steam Valve fails to close ior zdihs7l3a open RCIC Steam Valve fails to close trg e30 nrcRHR3C imfrh0lc (e30 :30) RHR3C Trip imf rhO 1 a RHR3A Trip 8
NRC Scenario 1 Trigger Files MODESW ZDIHS465(4) .NE. 1 nrcRHR3C 3C RHR Pump to start zdihs74l6a[3] .eq. 1 Scenario 1 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 201 Simulator Setup close HPCI Valves and open drain valves manual so steam line pressure decays to zero Simulator Setup Load Batch bat nrcl3O6-l Simulator Setup Tag HPCI, Shift Bus Duct Fans to B on manual and A in Standby, swap lens covers on Bus_duct,_Adjust_CRD_flow to_60 Simulator Setup Verify file loaded, leave simulator in freeze until crew is walking down.
Monitor control rod 30-23 temperature and adjust file sev rdkxleak(63) as necessary to ensure CR1) temp alarm does not come in until 3 0-23 is being withdrawn. Less than 1.0 temp down, greater than_1.0_temp_up.
RCP required (70% 90% with control rods and flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12 9
NRC Scenario 1 Simulator Event Guide:
Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3 5.7 Placing the Second and Third RFPIRFPT In Service
[2] IF RFPIRFPT is NOT warmed, reset and rolling, THEN PERFORM the following: (Otherwise N/A)
[3] VERIFY RFP 3A MIN FLOW VALVE, 3-HS-3-20, in OPEN position.
- CHECK OPEN M1N FLOW VALVE, 3-FCV-3-20.
[4] SLOWLY RAISE speed of RFPT using RFPT 3A SPEED CONT RAISE/LOWER, 3-HS-46-8A, to establish flow and maintain level in vessel.
[5] WHEN RFPT discharge pressure is within 250 psig of reactor pressure, THEN VERIFY OPEN RFP 3A DISCHARGE VALVE, 3-FCV-3-19.
[6] SLOWLY RAISE RFPT speed using RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A, to slowly raise RFP discharge pressure and flow on the following indications (Panel 3-9-6):
[7] WHEN sufficient flow is established to maintain RFP 3A MIN FLOW VALVE, 3-FCV-3-20, in CLOSED position (approximately 2 x 106 lbm!hr), THEN PLACE RFP 3A MIN FLOW VALVE, 3-HS-3-20, in AUTO.
[8] OBSERVE lowering in speed and discharge flows of other operating RFPs.
10
NRC Scenario 1 Simulator Event Guide:
Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3
[9] IF transferring RFPT from MANUAL GOVERNOR to individual RFPT Speed Control PDS, THEN PERFORM the following: (Otherwise N/A)
[9.1] PULL RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A, to FEEDWATER CONTROL position.
[9.2] VERIFY amber light at switch extinguished above RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A.
[9.3] PERFORM the following on RFPT 3A SPEED CONTROL(PDS),
3-SIC-46-8 (Panel 3-9-5):
[9.3.1] SELECT Column 3.
[9.3.2] VERIFY PDS in MANUAL.
[10] IF transferring control of RFPT from individual RFPT Speed Control PDS to AUTO control using REACTOR WATER LEVEL CONTROL PDS, 3-LIC-46-5, THEN PERFORM the following: (Otherwise N/A)
[10.1] VERIFY REACTOR WATER LEVEL CONTROL (PDS),
3-LIC-46-5 is functioning properly and ready to control second or third RFP.
[10.2] SLOWLY RAISE RFP discharge flow and pressure by raising RFP speed.
[10.3] WHEN RFP speed is approximately equal to operating RFP(s) speed, THEN PERFORM the following on RFPT 3A SPEED CONTROL (PDS), 3-SIC-46-8:
[10.3.1] PLACE PDS in AUTO.
[10.3.2] VERIFY Colunm 3 selected.
11
NRC Scenario 1 Simulator Event Guide:
Event 1 Normal: Place 3A RFPT in service from 600 RPM in accordance with 3-01-3
[11] WHEN RFP in automatic mode on REACTOR WATER LEVEL CONTROL, (PDS) 3-LIC-46-5, THEN CLOSE the following valves:
[12] VERIFY CLOSED the following valves on first RFP started in Section 5.5:
3 -FCV-006-0 156(0158) (local control)
[13] VERIFY both RFPT Main Oil Pumps running.
[14] IF desired to stop Turning Gear for in service RJFPT, THEN PLACE appropriate handswitch in STOP and RETURN to AUTO:
- RFPT 3A TURNING GEAR MOTOR, 3-HS-3-1O1A
[15] REFER TO Section 6.0.
- CONTROL and MONITOR RFW system operation.
12
NRC Scenario 1 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods SRO Notify ODS of power increase.
Direct Power increase using control rods JAW 3 -GOl- 100-12.
[21] WhEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes. REFER TO 3-SR-3.3.5(A) and 3-01-68.
- MONITOR Core thermal limits using ICS, and/or 0-TI-248 ATC Raise Power with Control Rods JAW 3-01-85, section 6.6. Control Rods:
22-3 1 from 0 to 24, 30-39 from 0 to 24, 38-3 1 from 0 to 24 and 30-23 from 0 to 24 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, 3-HS-85-46 in ON.
- Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing.
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 S_________ Panel 3-9-5 and PERFORM_Step_6.6.2[6].
13
NRC Scenario 1 Simulator Event Guide:
Event 2 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, 3-HS-85-46 in the OFF position to deselect the Control Rod.
[6.2] PLACE CRD POWER, 3-HS-85-46, in the ON position.
ATC 6.6.3 Control Rod Notch Withdrawal
[ 1] SELECT the desired control rod by depressing the appropriate CRD ROD SELECT pushbutton, 3-XS-85-40.
[2] OBSERVE the following for the selected control rod:
e CRD ROD SELECT pushbutton is brightly ILLUMiNATED.
o White light on the Full Core Display ILLUMiNATED.
. Rod Out Permit light ILLUMINATED.
[3] VERIFY Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when the Rod Worth Minimizer is enforcing.
[4] PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH, and RELEASE.
[5] OBSERVE the control rod settles into the desired position and the ROD SETTLE light extinguishes.
Driver At NRC direction, trip CR1) Pump 3A trigger 7 14
NRC Scenario 1 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods ATC 6.6.4 Continuous Rod Withdrawal
[1] SELECT desired Control Rod by depressing appropriate CRD ROD SELECT, 3-XS-85-40.
[2] OBSERVE the following for the selected control rod:
- CRD ROD SELECT pushbutton is brightly ILLUMiNATED.
- White light on the Full Core Display ILLUMiNATED.
- Rod Out Permit light ILLUMINATED.
[3] VERIFY Rod Worth Minimizer 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:
[5.1] PLACE AND HOLD CRD NOTCH OVERRIDE, 3-HS-85-47, in NOTCH OVERRRIDE.
[5.2] PLACE AN]) HOLD CR1) CONTROL SWITCH, 3-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, 3-HS-85-47 and CR1) CONTROL SWITCH, 3-HS-85-48.
[5.4] IF control rod settles at notch before intended notch, THEN PLACE CRD CONTROL SWITCH, 3-HS-85-48, in ROD OUT NOTCH and RELEASE.
Driver At NRC direction, trip CR1), Pump 3A trigger 7 15
NRC Scenario 1 Simulator Event Guide:
Event 2 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, 3-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)
ATC 6.6.5 Return to Normal After Completion of Control Rod Withdrawal
[1] WHEN control rod movement is no longer desired AND deselecting control rods is desired, THEN:
[1.1] PLACE CRD POWER, 3-HS-85-46, in OFF.
[1.2] PLACE CRD POWER, 3-HS-85-46, in ON.
16
NRC Scenario 1 Simulator Event Guide:
Event 2 Reactivity: Raise Power with Control Rods ATC Responds to annunciator 9-5A window 24: RBM HIGH/INOP A. IF moving control rods for start-up or power maneuvering THEN PERFORM the ATC following:
- 1. VERIFY correct control rod selected.
- 2. VERIFY Rod Out Permit light is NOT illuminated to ensure selected rod withdrawal is inhibited.
- 3. CHECK annunciator LPRM HIGH (3-XA-55-5A, Window 12) and matrix light, Panel 3-9-5 to determine if the alarm is due to high flux.
- 4. DESELECT then RESELECT the desired Control Rod to reset the alarm and Reinitialize the RBM back to Normalized 100%.
B. IF NOT moving control rods but a rod is selected THEN VERIFY Rod Out Permit light is NOT illuminated to ensure selected rod withdrawal is inhibited.
(Receiving a rod block when NOT moving a rod, may be an indication of a failure of the RBM or an indication of a Reactor power reduction with a rod selected.)
17
NRC Scenario 1 Simulator Event Guide:
Event 3 Component: CRD Pump 3A trip ATC Reports Trip of CRD Pump 3A.
SRO Announces entry into 3-AOI-85-3, CR1) System Failure.
4.1 Immediate Actions
[11 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-l 1, in MAN at minimum setting.
[1.2] START associated standby CRD Pump using one of the following:
- CR1) PUMP 3B, using 3-HS-85-2A
[1.3] ADJUST CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, to establish the following conditions:
- CR1) CLG WTR HDR DP, 3-PDI-85-18A, approximately 20 psid
- CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, between 40 and 65 gpm.
[1.4] BALANCE CR1) SYSTEM FLOW CONTROL, 3-FIC-85-l 1, and PLACE in AUTO or BALANCE.
Driver If Dispatched to CRD Pump 3A, pump is extremely hot to touch.
CRD Pump 3B oil levels in band, pump ready for start, conditions normal after the start.
CRD 3A report breaker tripped on over current, Electrical Maint called.
While control rod 3 0-23 is being withdrawn initiate severity files for High CRD Temperature on Control Rod 30-23. Initiate trigger 9 CRD Temp for 3 0-23 will rise to 345 and climb to 350 from there and continue to climb.
Driver Once the crew is no longer monitoring CRD Temps insert trigger 12 to return simulator to normal otherwise the last sev entered will stay in, resetting does not clear.
18
NRC Scenario 1 Simulator Event Guide:
Event 4 Component: Control Rod 30-23 High Temperature ATC Responds to annunciator 9-5A window 17: CONTROL ROD DRIVE UNIT TEMP NIGH A. CHECK high temp of CRD on recorder 3-TR-85-7A, & 3-TR-85-7B (Panel 3-9-47) or on ICS.
B. IF alarm is VALID, THEN PERFORM the following, as directed by the Unit Supervisor.
- CHECK cooling water pressure and flow normal on Panel 3-9-5.
- DISPATCH personnel to check for HCU scram discharge valve leaking as indicated by elevated discharge piping temperatures for associated CRD.
- PERFORM 3-TI-393 for control rods with high temperatures or failed thermocouples.
- REFER TO OPDP-4, 3-01-85, 3-AOI-85-3.
- FLUSH CRD to unblock restricted cooling water flow. REFER TO 3-01-85.
- DECLARE the control rod, which is in alarm, SLOW as directed by 3-TI-393 per Tech Spec Table 3.1.4-1 Note 1.
- RAISE CRD Flow, as directed by Unit Supervisor, if required to keep the drives cool per CRD Pump Operation At Elevated Flow section of 3-01-85.
When called report CR]) Temperature from ICS, report no elevated pipmg temperature Driver 19
NRC Scenario 1 Simulator Event Guide:
Event 4 Component: Control Rod 30-23 High Temperature ATC Flush of Control Rod 30-23 lAW 3-01-85 Section 8.29 CAUTION This section may only be used at the direction of the CRD System engineer to flush a CRD that has the characteristics of the cooling water flow being restricted by debris within the CRD. This flushing section should not be performed repeatedly (Greater than one time per week)on the same CRD to prevent thermal cycling on the CRD. GESIL 173.
[1] IF flushing rods at Position 48, THEN:
NOTE The following steps (performed only on rods at positions other than 48) perform an insert flush at reduced drive water DP to prevent the CRD from being inserted.
[2] II? flushing rods at positions other than 48, THEN:
[2.1] LOWER the CRD DRIVE WTRHDRDP, 3-PDI-85-17A, to <75 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[2.2] VERIFY CRD POWER, 3-HS-85-46, in ON.
[2.3] SELECT the control rod to be flushed by depressing the appropriate CRD ROD SELECT pushbutton, 3-XS-85-40.
[2.4] IF while performing this section Rod Motion is observed, THEN
[2.4.1] IMMEDIATELY RELEASE CRD CONTROL SWITCH 3-HS-85-48 and NOTIFY System Engineer.
[2.4.2] RAISE the CRD DRIVE WTR HDR DP, 3-PDI-85-17A to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[2.4.3] STOP further flushing until authorized by the Unit Supervisor.
NOTE It could take up to two minutes for CRD DRIVE WTR HDR DP to stabilize.
[2.5] CHECK CRD DRIVE WTR HDR DP, 3-PDI-85-17A, is stable and <75 psid.
[2.6] PLACE AND HOLD CRIJ CONTROL SWITCH 3-HS-85-48, in ROD IN for 30 seconds.
Driver When called CRD System Engineer recommends flushing 20
NRC Scenario 1 Simulator Event Guide:
Event 4 Component: Control Rod 30-23 High Temperature
[2.7] ADJUST the CR1) DRIVE WTR FIDR DP, 3-PDI-85-17A to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 3-HS-85-23A.
[2.8] PESELECT the control rod after flushing by PLACING CR1) POWER, 3-HS-85-46, in OFF, THEN PLACE CRD POWER, 3-HS-85-46, in ON.
Evaluate 3-T1-393 and Tech Spec 3.1.4 EVALUATION OF CRD TEMPERATURE ALARMS SCOPE This procedure is utilized by Operations personnel to evaluate CR1) high temperature alarms as directed by 3-ARP-9-5A.
FREQUENCY This instruction will be performed for the following conditions:
As required to evaluate new CR]) high temperature alarms, 4.0 PRECAUTIONS AND LIMITATIONS 4.1 GE SIL 173 discusses the potential effects on scram times for CRDs whose operating temperature is above 350 °F. Scram times can be increased by 0.150 to 0.500 seconds for a CRD operating between 350 °F and 525 °F. CRDs with temperatures above 350 °F are declared slow in accordance with Technical Specification Section 3.1.4 and are added to the population of rods scram time tested per 3-SR-3.1.4.1.
SRO 7.1 Evaluation of New Alarms 7.1.1 Record the CR1) number on Appendix B, Evaluation of New CRD Temperature Alarms. Record the CR1) temperature or indicate that the alarm is due to an open thermocouple input (upscale trip of 3-TA-85-7 and unknown temperature data displayed on ICS).
7.1.2 Determine if the scram outlet valve is leaking by comparing the scram outlet piping temperature for the affected CR1) to the scram outlet piping temperatures of adjacent HCUs.
7.1.4 If the CR]) temperature is greater than 350 °F and it has been determined that the scram outlet valve is not leaking, perform Section 8.28 of 3-01-85 to flush the CRD as directed by the System Engineer. If the CR]) temperature returns to normal after flushing, document in the Comments field of Appendix B.
7.1.5 If the CR1) temperature remains greater than 350 °F, 7.1.5.1 Declare the affected rod SLOW per Note 1 of Tech Spec Table 3.1.4-1, and contact Reactor Engineering to update the scram time data base.
Driver At direction of NRC initiate trigger 1 for loss of 480V Unit B]) 3B 21
NRC Scenario 1 Simulator Event Guide:
Event 4 Component: Control Rod 30-23 High Temperature SRO Evaluate 3-TI-393 and Tech Spec 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1 A Control Rod Scram Times LCO 3.1.4 a No more than 13 OPERABLE control rods shall be slow, in accordance with Table 3.1.4-1; and
- b. No more than 2 OPERABLE control rods that are slow shall occupy adjacent locations.
APPLICABILITY: MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCD Ai Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> not met.
Table 3.1.4-1 (page 1 of 1)
Control Rod Scram Times
.NOTES
- 1. OPERABLE control rods with scram times not within the limits of this Table are considered slow.
- 2. Enter applicable Conditions and Required Actions of LCD 11 3, Control Rod OPERABILITY, for control rods with scram times 7 seconds to notch position 06.
These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered slow.
SRO Contacts Reactor Engineering and Operations Management Entries Tech Spec 3.1.4 Condition A, once flushing is started and control rod temperature SRO lowers below 3500 F can exit TS 3.1.4 condition A Driver When Reactor Engmeer called report that will evaluate Driver At direction of NRC initiate trigger 1 for loss of 480V Unit BD 3B 22
NRC Scenario 1 Simulator Event Guide:
Event 4 Component: Control Rod 3 0-23 High Temperature May enter Tech Spec 3.1.3 and declare control rod inoperable and fully insert to exit TS SRO 3.1.4 Condition A action statement.
3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
Condition C. One or more control rods inoperable for reasons other than Condition A or B
Required Action C.1 Fully insert inoperable control rod Required Action C.2 Disarm the associated CR1)
Completion Time C.1 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Completion Time C.2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SRO Contacts Reactor Engineering and Operations Management Entries Tech Spec 3.1.4 Condition A, once flushing is started and control rod temperature SRO
. lowers below 350° F can exit TS 3.1.4 condition A
. Driver At dfrection of NRC itiate trigger 1 for loss of 480V UnitBD 3B 23
NRC Scenario 1 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, Crew 9-8C Window 22, 23, and 30 and 9-7A Window 31 Announce loss of 480V Unit Board 3B SRO Prioritize annunciators 9-7A Window 31, Highest Priority BOP 9-7A Window 31 GEN BUS DUCT FAN FAILURE A. VERIFY Main Bus Cooling Fans, 3-HS-262-1A or 1-HS-262-2A, indicates running on Panel 3-9-8 AND START GEN BUS DUCT HX FAN A using 3-HS-262-1A, on panel 3-9-8 to start the standby fan.
B. IF no Fans are operating and the Generator is tied to the grid and loaded to greater than the self cooled bus rating of 16,500 amps THEN, IMMEDIATELY INSERT a manual reactor scram, AN]) TRIP the Main Generator.
C. II? while executing this procedure, the Bus Duct Temperature is at or above the Temperature Excursion limit of 120°C, THEN IMMEDIATELY INSERT a manual reactor and TRIP the Main Generator.
D. DISPATCH personnel as necessary to check the following:
- 1. Main Bus Cooling Fan on elevation 586 to check fan condition.
- 2. Monitor Bus Duct temperature by available means including using a portable temperature monitor device locally at the 14 in-service thermostats. REFER to Window 32, Figure 1.
- 3. 480V Unit Board 3A on elevation 586 to check breaker 5C closed.
- 4. 480V Unit Board 3B on elevation 604 to check breaker SC closed.
E. VERIFY the system is operating in accordance with 3-01-47.
When sent to mvestigate mform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (51U) 24
NRC Scenario 1 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8A Window 4 and 35 Window 4 EXCITATION SYSTEM ABNORMAL A. CHECK the following displays on ICS to identify the cause:
- AVR ALARM CODES (AVRALM)
- AVR FAULT CODES (AVRFLT)
B. ADJUST, as required, VOLTAGE REGULATOR LOWER/RAISE ADJUST, 3-HS-57-26, to maintain the following:
- GENERATOR VOLTS, 3-EI-57-39, between 20,900V and 23,100V.
- GENERATOR MVARS, 3-EI-57-51 Window 35 COMMUNICATIONS ROOM COMMON ALARM A. Unit 3 ICS display:
1.SYSTEMMIMICS
- 2. ANNUNCIATOR MIMICS
- 3. ANNUNCIATOR 8A
- 4. COMM ROOM COMMON ALARM
- 5. Determine alarm point in alarm by EOR point number with RED ALM for point QUAL.
When sent to investigate inform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (51U) 25
NRC Scenario 1 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8B Window 17 Window 17 480V UNITBD 3B UV OR XFR Automatic action A. Undervoltage Trip (loss of following):
- EHC hyd fluid system Pump 3B
- Stator cooling water Pump 3B Motor suction Pump
- RFPT 3B 3B1 Main Oil Pump
- Cond vac pump 3B
- Lose one of two 480V AC power supplies to Generator Voltage Regulator.
Voltage Regulator remains in service.
COMPLETE listing on reference drawings.
B. Auto Transfer:
- Alternate breaker closes
- No loss of equipment A. VERIFY automatic action has occurred.
B. INSPECT 480V Unit Bd 3B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
C. REFER TO 0-OI-57B to re-energize board.
D. REFER TO appropriate 01 for recovery or realignment of equipment.
When sent to investigate inform crew that an ACTUATION OF NORMAL SUPPLY Driver BREAKER OVERCURRENT RELAY (5 1U). Unit Board 3B failed to transfer 26
NRC Scenario 1 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8C Window 22, 23, and 30 Window 22 480V TB VENT BD 3B UV OR XFR A. VERIFY alarm by checking associated annunciator, TURBINE BLDG VENTILATION ABNORMAL, (3-XA-55-3D, Window 4) in alarm. D B. DISPATCH personnel to TB VENT Bd 3B to check equipment, board status, and abnormal conditions.
- Turbine spaces supply fans 3A and 3B
- Turbine room supply fans 3C, 3D, 3E
- Turbine room exhaust fans 3E, 3F, 3G, 3H, 3J
- Electrical space supply fan 3A, 3B, and exhaust fan C. REFER TO 0-OI-57B to re-energize or transfer the board.
D. REFER TO appropriate 01 for recovery or realignment of equipment.
Window 23 480V TURB MOV BD 3B UV OR XFR A. VERIFY alarm by checking light indication in Control Room for the following equipment:
- Steam packing exhauster blower 3B
- RFPT injection water pump 3B
- RFPT oil tanks vapor extractor
- RFPT 3A 3A2 Main Oil Pump
- RFPT 3C 3C2 Main Oil Pump B. NOTIFY Radwaste of loss of the following:
- Station sump pump 3A
- Turb Bldg equipment and floor drain sump pump 3B
- Turb Bldg cnds pump pit equipment and floor drain sump pump 3B C. CHECK board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
D. REFER TO 0-OI-57B to re-energize or transfer the board.
B. REFER TO appropriate 01 for recovery or realignment of equipment.
When sent to mvestigate mform crew that 480V TB Vent BD 3B transferred and 480V Driver TURB MOV BD 3B transferred DO NOT re-energize 480V Umt BD 3B, electrical Maintenance will have to investigate actuation of overcurrent relay 27
NRC Scenario I Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 3B Respond the following Annunciators: 9-8A Window 4 and 35, 9-8B Window 17, C rew 9-8C Window 22, 23, and 30 and 9-7A Window 31 BOP Responds to remaining annunciators 9-8C Window 22, 23, and 30 Window 30 480V TURB MOV BD 3C UV OR XFR A. VERIFY alarm by checking light indication in Control Room for the following equipment:
- H2 main seal oil pump
- H2 seal oil vacuum pump
- H2 recirculating seal oil pump
- Steam packing exhauster blower 3A B. ChECK MOV board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
C. REFER TO O-OI-57B to re-energize or transfer the board.
D. REFER TO appropriate 01 for recovery or realignment of equipment.
SRO Will call outside US and determine if board can be re-energized
. When sent to investigate inform crew that 480V TURB MOV BD 3C transferred. DO NOT Driver re-energize 480V Unit BD At NRC direction insert trigger 3 for Off Gas Malfunction when hydrogen is above 5% and D river at NRC direction initiate trigger 5 for Hydrogen Explosion 28
NRC Scenario 1 Simulator Event Guide:
Event 6 Component: OFF-GAS H2 RECOMBINER CATALYST FAILURE BOP Responds to alarm the following alarms:
HIGH OFFGAS % 112 TRAIN A (9-53, Window 3)
HIGH OFFGAS % H2 TRAiN B (9-53, Window 13)
H2 WATER CHEMISTRY ABNORMAL, (9-53, window 10)
BOP Reports a rise in hydrogen concentration on OFF GAS HYDROGEN ANALYZERs (CH 1-Analyzer 3A, CH 2-Analyzer 3B) recorder, 3-XR-66-103/1 and 2, Panel 9-53.
SRO Enters 3-AOI-66-1, Off-Gas H2 High.
4.1 Immediate Actions None 4.2 Subsequent Actions
[1] PLACE both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK / BYPASS control switches, 3-HS-066-1007 and 1008, on OFFGAS SAMPLE PANEL, 3-LPNL-925-0588, in BYPASS to assure continuous availability of hydrogen monitoring.
[2] IF HWC System injection is in service, THEN PERFORM the following:
[2.1] At HYDROGEN WATER CHEMISTRY CONTROL PANEL, 3-LPNL-925-0589, VERIFY that H2 and 02 injection rates are normal at Operator Interface Unit (OJU). (H2 injection rate should match the setpoint on the OIU and the 02 injection rate should match the setpoint on the OIU which should be half of the H2 injection rate during normal steady state conditions.)
[2.2] IF H2 and 02 injection rates do NOT meet the above conditions, THEN
[2.3] IF oxygen concentration as indicated on OFFGAS TRAiN A or TRAIN B SAMPLE 02 INDICATOR, 3-021-066-105 1 or 1052, on 3 -LPNL-925-05 88 indicates less than 5% oxygen AND an automatic HWC shutdown has NOT occurred, THEN INITIATE a HWC System shut down using either, 3-HS-4-40A H2 WATER CHEMISTRY CONTROL
[Panel 3-9-53], 3-HS-4-40B 112 WATER CHEMISTRY CONTROL [Panel 3-9-5] or 3-HS-4-39 HWC SHUTDOWN SWITCH [3-LPNL-925-0588].
When directed wait 2 minutes and report that both OFFGAS TRAIN A(B) AUTO CHANNEL CHECK / BYPASS control switches, 3-HS-066-1007 and 1008 are m bypass Driver Report that 112 and 02 mjection rates are normal When called to report 02 concentration for step 2 3 above report a concentration of 4 5%
29
NRC Scenario 1
/ Simulator Event Guide:
Event 6 Component: OFF-GAS H2 RECOMI31NER CATALYST FAILURE
[3] VERIFY proper operation of in service SJAE. (Steam jet may have failed to isolate on low supply steam pressure.)
[3.11 IF a failure of the in service SJAE is indicated AND hydrogen concentration is less than 4%, THEN PLACE standby SJAE in service.
REFER TO 3-01-66 (otherwise N/A)
[4] IF hydrogen concentration is? 4%, THEN REFER TO TRM 3.7.2 SRO TR 3.7.2 Airborne Effluents LCO 3.7.2 Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to 4% by volume.
APPLICABILITY: During main condenser offgas treatment system operation Condition A: With the concentration of hydrogen >4% by volume.
Required Action A. 1: Restore the concentration to within the limit.
Completion Time: 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> When directed wait 2 minutes and report that both OFFGAS TRAIN A(B) AUTO Driver CHANNEL CI{ECK / BYPASS control switches, 3-HS-066-1007 and 1008 are in bypass.
I________ Report that H2 and 02 injection rates are normal.
30
NRC Scenario 1 Simulator Event Guide:
Event 6 Component: OFF-GAS H2 RECOMBINER CATALYST FAILURE NOTE Fuel failure is indicated by, but NOT limited to, rising activity on the following:
- OFF-GAS PRETREATMENT RADIATION recorder, 3-RR-90-157 (Panel 3-9-2)
- MAiN STEAM LINE RADIATION recorder, 3-RR-90-135 (Panel 3-9-2)
- OFFGAS POST-TREATMENT RADIATION recorder, 3-RR-90-265
- On MAIN CONDENSERS (MN COND) ICS display: Offgas pretreatment, post treatment, and stack radiation
[5] IF high hydrogen concentration is a result of possible fuel failure, THEN REDUCE core flow to 50 60 % (otherwise N/A).
[6] WHEN any of the following conditions exist OBTAIN US approval, ThEN SCRAM the Reactor (REFER TO 3-AOI-100-1):
- Hydrogen ignition/explosion as indicated by rising temperature, and/or pressure, and/or flow in the Off-Gas System which may cause some or all of the following annunciators to alarm:
- OFF-GAS HOLDUP VOLUME PRESS HIGH, (3-XA-55-7A, Window 1)
- OFF-GAS HOLDUP VOLUME TEMP HIGH, (3-XA-55-7A, Window 2)
- OG CHARCOAL BED VESSEL TEMP HIGH, (3-XA-55-53, Window 9)
- HOLDUP LINE INLET FLOW HIGH, (3-XA-55-53, Window 14)
- CHARCOAL BED VAULT TEMP HIGH, (3-XA-55-53, Window 19)
CHARCOAL BED GAS REHTR OUTL DEW PT TEMP HIGH, (3-XA-55-53, Window 27)
- CHARCOAL TRAIN INLET/OUTLET PRESS HIGH, (3-XA-55-53, Window 28)
- Charcoal combustion in Adsorber beds as indicated by rising temperature and/or radiation in the Off-Gas System which may cause some or all of the following windows to alarm:
- CHARCOAL BED VESSEL TEMP HIGH, (3-XA-55-53, Window 9)
- CHARCOAL BED VAULT TEMP NIGH, (3-XA-55-53, Window 19)
- OG POST TRTMT RADIATION HIGH, (3-XA-55-4C, Window 33)
- Unit Supervisor direction The Following alarms will come in when Off Gas Explosion occurs: Panel 7A window NRC 1,4,9, Panel 4C window 33,34,3 5, Panel 3A window 6,13, Panel 53 window 4,14,8,9,18,27,28 SRO May Direct Core Flow runback prior to Reactor Scram SRO Directs_Reactor_SCRAM_and_enters_3-AOl- 100-1 Driver Just before scram or as scrarnis directed insert trigger 15 for loss of feedwater 31
NRC Scenario 1 Simulator Event Guide:
Event 6 Component: OFF-GAS H2 RECOMB]NER CATALYST FAILURE Respond to Annunciator 4C window 35: OG POST TRTMT RAD MONITOR HI HI BOP HJI1NOP A. VERIFY alarm condition on the following:
- OFFGAS POST-TREATMENT RADIATION recorder, 3-RR-90-265 on Panel 3-9-2
- OG POST-TREATMENT CHAN A RAD MON RTMR radiation monitor, 3-RM-90-266A on Panel 3-9-10
- OG POST-TREATMENT CHAN B RAD MON RTMR. radiation monitor, 3-RM-90-265A on Panel 3-9-10 B. CHECK the following is in alarm:
OFF-GAS ISOLATION VALVE CLOSED, 3-XA-55-7A Window 4.
C. VERIFY OFF-GAS SYSTEM ISOLATION VALVE, 3-FCV-66-28 has the mechanical restraint DISENGAGED_and_3-FCV-66-28_is_CLOSED.
BOP Closes OFF-GAS SYSTEM ISOLATION VALVE, 3-FCV-66-28 SRO May Direct Core Flow runback prior to Reactor Scram SRO Directs Reactor SCRAM and enters 3-AOI-100-l l[ Driver Just before scram óras scram is dected inserttrigger 15 for loss of feedwater 32
NRC Scenario 1 Simulator Event Guide:
Event 6 Component: OFF-GAS H2 RECOMB1NER CATALYST FAILURE BOP Responds to annunciator 9-7A window 1, OFF-GAS HOLDUP VOLUME PRESS HIGH A. CHECK off-gas flow recorder, 3-FR-66-20 on Panel 3-9-8 and pressure to off-gas preheater 3-PI-66-71 on Panel 3-9-53.
B. CHECK for indications of H2 ignition:
1.0FF GAS HOLDUP VOLUME TEMP HIGH alarm, 3-XA-55-7A, Window 2.
2.0FF GAS SAMPLE FLOW ABNORMAL alarm, 3-XA-55-3A, Window 33.
- 3. Various abnormal indications or alarms NOT limited to, but may include one or more of the following on Panel 3-9-53:
- a. GAS RHTR OUTLET DEW POINT HIGH TEMP alarm, 3-XA-55-53, Window 27.
- b. Upscale reading on Recorder 3-TRS-66-108.
- c. Adsorber Vault (3-TRS-66-120) or Adsorber Vessel (3-TRS-66-1 15) high temperature.
- d. Lowering inlet recombiner temperature 3-TI-66-75A(B).
C. IF indication of H2 ignition is present, ThEN VERIFY the following valves are closed:
- 3-FCV-66-14, 18 - SJAE discharge valves.
- 3-FCV-66-1 1, 15 SJAE Air inlet valves.
3-PCV-1-15 1,153,166,167 SJAE steam pressure control valves.
- 3-FCV-1-155,156,172,173 SJAE steam isolation valves.
- 3-FSV-1-150,152 SJAE inter-condenser drain valves, AND REFER TO 3-AOI-66-1.
Verify the following Valves Closed
- 3-FCV-66-14, 18 SJAE discharge valves.
- 3-FCV-66-1 1, 15 SJAE Air inlet valves.
- 3-PCV-1-15 1,153,166,167 - SJAE steam pressure control valves.
- 3-FCV-1-155,156,172,173 SJAE steam isolation valves.
- 3-FSV-1-150,152 SJAE inter-condenser drain valves SRO May Direct Core Flow runback prior to Reactor Scram SRO Directs_Reactor_SCRAM_and_enters_3-AOl- 100-1 Driver Just before scram or as scram is directed insert trigger 15 for loss of feedwater 33
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater SRO Directs Reactor SCRAM and enters 3 -AOl- 100-1 4.1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5A/S3B, on Panel 3-9-5.
[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A)
[3] Refuel Mode One Rod Permissive Light check
[3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL.
[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46.
[3.3] II? 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 SHUTDOWN.
[5] REPORT the following status to the US:
- Reactor Scram Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Water Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
Power level 4.2 Subsequent Actions
[3] DRiVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit.
[3.1] DOWNRANGE IRMs as necessary to follow power as it lowers.
[4] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5.
[5] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, as follows:
Crew Reports loss of Unit Board 3C ATC Report Loss of all Condensate Booster Pumps 34
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater SRO Enters EOI- 1 on Reactor Level 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 or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO -
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 with MSIVs SRO EOI- 1 (Reactor Level)
Monitor and Control Reactor Water Level.
Directs_Verification_of PCIS_isolations.
ATC/BOP Verifies PCIS isolations.
Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with SRO the following injection source. (RCIC, App 5C)
ATC Initiates RCIC JAW App 5C 35
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C.
- 1. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRIP/THROT VALVE RESET.
- 2. 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-71-3 1A, 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 3ZI-7l-40A, DISC POSITION, red light illuminated.
- d. 3-FCV-71-34, RCIC PUMP M1N 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.
36
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater Crew Respond to Turbine Building Radiation High A. DETERMINE area with high radiation level on Panel 3-9-11. Report Numerous Turbine Building areas B. IF the TSC is NOT manned, THEN USE public address system to evacuate area where high airborne conditions exist.
Respond to MAIN STEAM LINE RADIATION HIGH A. CHECK following radiation recorders:
- 1. MAIN STEAM L1NE RADIATION, 3-RR-90-135 on Panel 3-9-2.
Respond to RX BLDG AREA RADIATION HIGH A. DETERMINE area with high radiation level on Panel 3-9-11. RX BLDG 565 Area 3-RJ-90-21A D. IF the TSC is NOT manned and a VALID radiological condition exists., THEN USE public address system to evacuate area where high radiological conditions exist I. ENTER 3-EOI-3 Flowchart.
Enters EOI-3 on Secondary Containment Radiation SRO SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO If Reactor Zone or Refuel Zone Exhaust Ventilation isolated and ventilation radiation levels SRO are below 72 mr/hr Then Restart Reactor Zone and Refuel Zone Ventilation.
Directs Restart of Reactor Zone and Refuel Zone Exhaust Ventilation (App 8F)
SRO Directs Defeat of isolation interlocks (App 8E)
Calls for Appendix 8E and when complete Restarts Reactor Zone and Refuel Zone Exhaust BOP/ATC Ventilation Is Any Area Radiation Level Above Max Normal? - YES Isolate all systems that are discharging into the area except systems required to:
- Suppress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment? NO, NO System Discharging Before any area radiation rises to Max Safe (table 4) Continue and enter EOI-1 (EOI- 1 has already been entered after Reactor Scram)
Crew Monitors for Max Safe Radiation and reports DRIVER If Called for Appendix 8E, Wait 4 mmutes, enter bat appO8e and report complete 37
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater ATC/BOP Appendix 8F RESTORiNG REFUEL ZONE AND REACTOR ZONE VENTILATION FANS FOLLOWiNG GROUP 6 ISOLATION 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. PLACE Reactor Zone Ventilation in service as follows (Panel 3-9-2 5):
- a. VERIFY 3-HS-64-l 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.
- b. PLACE 3-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch in SLOW A (SLOW B).
- c. CHECK two SPLY/EXH A(B) green lights above 3-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch extinguish and two SPLYIEXFI A(B) red lights illuminate.
- d. VERIFY OPEN the following dampers:
- 3-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR
- 3-FCO-64-14, REACTOR ZONE SPLY 1NBD ISOL DMPR
- 3-FCO-64-42, REACTOR ZONE EXH 1NBD ISOL DMPR
38
NRC Scenario 1 Simulator Event Guide:
(
Event 7 Major: SCRAM and Loss of Feedwater Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure DWIT Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 160F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B) 39
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Jnitiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)
When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 40
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-1),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)
When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 41
NRC Scenario 1 Simulator Event Guide:
Event 7 Major: SCRAM and Loss of Feedwater Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19)
When H2 is detected in PC (2.4% on control room indicators continue, does not continue Enters EOI-2 on High Drywell Pressure SP/T MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18)
Can suppr p1 lvl be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES 42
NRC Scenario 1 Simulator Event Guide:
Event 8 Component: RHR Pumps 3A and 3C Trip ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays 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 auto open signal as necessary by PLACING 3-HS-74-155A(B), LPCI SYS 1(11) OUTBD NJ VLV BYPASS SEE in BYPASS.
- 3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN CONTINUE in this procedure at Step 7.
- 4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.
- 5. INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 3-FCV-74-53(67), RHR SYS 1(11) 1NBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RUR SYS 1(11) OUTBD iNJECT VALVE.
- e. VERIFY OPERATING the desired RHR System 1(11) pump(s) for Suppression Chamber Spray.
- g. OPEN 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
ATC/BOP When RHR Pumps 3A and 3C are started they will trip, operator will report and align Loop 2 ofRHR 43
NRC Scenario 1 Simulator Event Guide:
Event 8 Component: RHR Pumps 3A and 3C Trip ATC/BOP 3-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays
- h. IF RRR System 1(11) is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5.k.
- i. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) M1N FLOW VALVE.
- j. RAISE system flow by placing the second RHR System 1(11) pump in service as necessary.
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
- 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 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
- b. VERIFY CLOSED 3-FCV-74-l00, RHR SYS I U-2 DISCH XTIE
- c. IF RI{R operation is desired in ANY other mode, THEN EXIT this EOI Appendix.
- d. STOP RHR Pumps 3A and 3C (3B and 3D).
44
NRC Scenario 1 Simulator Event Guide:
Event 8 Component: RHR Pumps 3A and 3C Trip ATCIBOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays
- 1. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.
- 2. IF Adequate core cooling is assured, OR Directed to spray the Drywell 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 1NJ VLV BYPASS SEL in BYPASS.
- 3. VERIFY Recirc Pumps and Drywell Blowers shutdown.
- 4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.
- 5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.
- 6. INITIATE Drywell Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRI), in MANUAL OVERRIDE.
- d. 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.
- e. VERIFY OPERATING the desired System 1(11) RHR pump(s) for Drywell Spray.
- f. OPEN the following valves:
ATC/BOP When RHR Pumps 3A and 3C are started they will trip, operator will report and align Loop 2 of RHR 45
NRC Scenario 1 Simulator Event Guide:
Event 8 Component: RHR Pumps 3A and 3C Trip ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays
- g. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) Mll4 FLOW VALVE.
- h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System 1(11) RHR Pump in service.
- j. VERIFY RHRSW pump supplying desired R}IR Heat Exchanger(s).
- k. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- 7. WHEN EITHER of the following exists:
- Before drywell pressure drops below 0 psig, OR
- Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:
- a. VERIFY CLOSED the following valves:
- 3-FCV-74-100, R}IR SYS I U-2 DISCH XTIE
- b. VERIFY OPEN 3-FCV-74-7(30), RHR SYSTEM 1(11) MN FLOW VALVE.
- c. IF R}IR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.
- d. STOP RHR Pumps 3A and 3C (3B and 3D).
46
NRC Scenario 1 Simulator Event Guide:
Event 9 Major: Fuel Failure ATC/BOP Report MAIN STEAM LINE RADIATION HIGH-HIGH A. VERIFY alarm on 3-RM-90-136 thru 137 on Panel 3-9-10.
B. CONFIRM main steam line radiation level on recorder 3-RR-90-135, Panel 3-9-2.
C. IF alarm is VALID and scram has NOT occurred, THEN PERFORM the following:
IF core flow is above 60%, THEN
- 1. LOWER core flow to between 50-60%.
- 2. MANUALLY SCRAM the Reactor.
- 3. REFER TO 3-AOI-100-1.
D. IF SLC injection per RC/Q of EOI-1 is NOT required, THEN VERIFY the MSIVs closed.
SRO Directs MSIVs Closed and transitions pressure control to Appendix 1 l ATC/BOP Close MSWs and commences pressure control with Appendix hA 47
NRC Scenario 1 Simulator Event Guide:
Event 9 Major: Fuel Failure ATC/BOP Commence pressure control with Appendix hA, Alternate RPV Pressure Control Systems MSRVs
- 1. IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.
- 2. IF Suppression Pool level is at or below 5.5 fi, THEN CLOSE MSRVs and CONTROL RPV pressure using other options.
- a. 1 3-PCV-1-179 MN STM LINE A RELIEF VALVE.
- b. 2 3-PCV-1-180 MN STM LINED RELIEF VALVE.
- c. 3 3-PCV-1-4 MN STM LINE A RELiEF VALVE.
- d. 4 3-PCV-1-31 MN STM LiNE C RELIEF VALVE.
- e. 53-PCV-1-23 MN STM LINE B RELIEF VALVE.
- f. 6 3-PCV-1-42 MN STM LINE D RELIEF VALVE.
- g. 73-PCV-1-30 MN STM LINE C RELIEF VALVE.
- h. 83-PCV-1-19 MN STM LINE B RELIEF VALVE.
- i. 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.
- 1. 12 3-PCV-1-18 MN STM LiNE B RELIEF VALVE.
- m. 13 3-PCV-l-34 MN STM LINE C RELIEF VALVE.
48
NRC Scenario I Simulator Event Guide:
Event 9 Major: Fuel Failure
= ATCIBOP Report RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH AND RCIC STEAM LiNE LEAK DETECTION TEMP HIGH RCIC STEAM LINE LEAK DETECTION TEMP HIGH A. CHECK following instrumentation:
I. RCJC temperature elements on LEAK DETECTION SYSTEM TEMPERATURE Recorder, 1 -TR-69-29 (Points 9-12) on 1-9-21.
- 3. RX REFUEL ZONE EXHAUST RADIATION, 1-RR-90-144, on Panel 1-9-2 B. IF RCIC is NOT in service AND 1-FI-71-1A(B), RCIC STEAM FLOW indicates flow, THEN ISOLATE RCIC AND verify temperatures lowering.
C. IF high temperature is confirmed, THEN ENTER 1-EOI-3 Flowchart.
ATCIBOP Reports rising temperature in RCIC Area, If temperature continues to rise it will cause a RCIC Isolation @ 165°F in Torus area or RCIC Pump Room temp.
SRO Re-Enters EOI-3 49
NRC Scenario 1 Simulator Event Guide:
Event 9 Major: Fuel Failure RE Enters EOI-3 on Secondary Containment Temperature SRO SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO 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.
SRO Directs Restart of Reactor Zone and Refuel Zone Exhaust Ventilation (App 8F)
Directs Defeat of isolation interlocks (App 8E)
BOP/ATC Calls for Appendix 8E and when complete Restarts Reactor Zone and Refuel Zone Exhaust Ventilation SRO Monitors and Controls Secondary Containment Temps Operate Available ventilation (APPX 8F)
Defeat_Isolation_interlocks_if necessary_(APPX_8E)
Is any Area Temp Above MAX Normal, Yes RCIC Room Isolate all systems that are discharging into the area except systems required to:
. Suppress a Fire ATC/BOP Attempt to Isolate RCIC, RCIC fails to isolate SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before ANY Area Temp Rises to MAX Safe Continue, Completed When Temps in 2 or more areas are above MAX Safe Then Continue, Does Not Continue in the Temperature leg, NO area at or above MAX Safe Temperature Enters EOI-3 on Secondary Containment Radiation SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before any area radiation rises to Max Safe (table 4) Continue When Radiation Levels in 2 or more Areas are Above MAX Safe Then Continue, Report Two areas are above MAX SAFE Radiation Levels 3-RI-90-26A CS/RCIC Room ATC/BOP and_3-RI-90-29A_Supp Pool Area_519_Elev.
SRO Continues to Emergency RPV Depressurization 3 -C-2 50
NRC Scenario 1 Simulator Event Guide:
Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure SRO Emergency RPV_Depressurization_3-C-2 RPV Water Level CANNOT be determined, NO Containment Water Level CANNOT be Maintained Below 44 Feet, NO DW Control Air Becomes Unavailable, NO Will the Reactor remain subcritical without boron under all conditions, YES Is DW pressure above 2.4 psig, YES Prevent Injection from ONLY those CS and LPCI Pumps NOT required to assure adequate core cooling (APPX 4), NOT required currently no injection systems aligned for injection Is Suppression Pool Level above 5.5 feet, YES Directs all ADS Valves Open ATC/BOP .
Opens six ADS Valves, reports indication on that ADS Valve 1-34 failed to open
SRO Can 6 ADS Valves be Opened, NO Open Additional MSRVs as necessary to establish 6 MSRVs open ATC/BOP Open an additional MSRV, reports 6 MSRVs are open 51
NRC Scenario I Simulator Event Guide:
Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure SRO EOI-1 RPV Control SRO Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with the following injection source. Condensate APPX 6A, Core Spray APPX 6D, 6E SRO Can RPV Water Level be maintained above 2 inches, Yes ATCIBOP Align injection sources directed by the SRO and Restores Level to 2 to 51 inches Injection Subsystems Lineup Condensate
- 1. VERII?Y CLOSED the following Feedwater heater return valves:
3-FCV-3-71, HP HTR 3A1 LONG CYCLE TO CNDR 3-FCV-3-72, HP HTR 3B1 LONG CYCLE TO CNDR 3-FCV-3-73, HP HTR 3C1 LONG CYCLE TO CNDR.
- 2. VERIFY CLOSED the following RFP discharge valves:
3-FCV-3-19, RFP 3A DISCHARGE VALVE 3-FCV-3-12, RFP 3B DISCHARGE VALVE 3-FCV-3-5, REP 3C DISCHARGE VALVE.
- 3. VERIFY OPEN the following drain cooler inlet valves:
3-FCV-2-72, DRAIN COOLER 3A5 CNDS INLET ISOL VLV 3-FCV-2-84, DRAIN COOLER 3B5 CNDS INLET ISOL VLV 3-FCV-2-96, DRAIN COOLER 3C5 CNDS INLET ISOL VLV.
- 4. VERIFY OPEN the following heater outlet valves:
3-FCV-2-124, LP NEATER 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.
- 5. VERIFY OPEN the following heater isolation valves:
3-FCV-3-38, HP HTR 3A2 FW INLET ISOL VLV 3-FCV-3-3 1, HP HTR 3B2 FW INLET ISOL VLV 3-FCV-3-24, HP HTR 3C2 FW INLET ISOL VLV 3-FCV-3-75, HP HTR 3A1 FW OUTLET ISOL VLV 3-FCV-3-76, HP HTR 3B1 FW OUTLET ISOL VLV 3-FCV-3-77, HP HTR 3C1 FW OUTLET ISOL VLV.
- 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 3C SUCTION VALVE.
52
NRC Scenario I Simulator Event Guide:
Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure ATC/BOP Injection Subsystems Lineup Condensate
- 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/BOP Injection Subsystems Lineup Core Spray System I
- 1. VERIFY OPEN the following valves:
3-FCV-75-2, CORE SPRAY PUMP 3A SUPPR POOL SUCT VLV 3-FCV-75-1 1, CORE SPRAY PUMP 3C SUPPR POOL SUCT VLV 3-FCV-75-23, CORE SPRAY SYS I OUTBD INJECT VALVE.
- 2. VERIFY CLOSED 3-FCV-75-22, CORE SPRAY SYS I TEST VALVE.
- 3. VERIFY CS Pump 3A anchor 3C RUNNING.
- 4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 3-FCV-75-25, CORE SPRAY SYS I INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.
- 5. MONITOR Core Spray Pump NPSH using Attachment 1.
53
NRC Scenario 1 Simulator Event Guide:
Event 10 Instrument: ADS SRV 1-34, Acoustic Monitor indication failure ATC/BOP Injection_Subsystems_Lineup_Core_Spray_System_II
- 1. VERIFY OPEN the following valves:
3-FCV-75-30, CORE SPRAY PUMP 3B SUPPR POOL SUCT VLV 3-FCV-75-39, CORE SPRAY PUMP 3D SUPPR POOL SUCT VLV 3-FCV-75-51, CORE SPRAY SYS II OUTBD iNJECT VALVE.
- 2. VERIFY CLOSED 3-FCV-75-50, CORE SPRAY SYS II TEST VALVE.
- 3. VERIFY Cs Pump 3B andlor 3D RUM4ING.
- 4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 3-FCV-75-53, CORE SPRAY SYS II INBD INJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.
- 5. MONITOR Core Spray Pump NPSH using Attachment 1.
SRO Emergency Classification EPIP-1 3.2-S An unisolable Primary System leak is discharging into Secondary Containment AND Any area radiation level at or above the Maximum Safe Operating Area radiation limit listed in Table 3.2.
TABLE 3.2 MAXIMUM SAFE OPERATING AREA RADIATION LIMITS AREA RAD MONITOR MAX S1E VALUE MRII-IR UNITI UNIT2 UNIT3 RHR West Room 90-25A 1000 1000 1000 RHR East Room 90-28A 1000 1000 1000 HPCI Room 90-24A 1000 1000 1000 CS!RCIC Room 90-26A 1000 1000 1000 Core Spray Room 90-27A 1000 1000 1000 Suppr Pool Area 90-29A 1000 1000 1000 CRDHCU West Area 90-20A 1000 1000 1000 CRD-HCU East Area 90-21A 1000 1000 1000 TIP Dre Area 90-23A 1000 1000 1000 North RWCU SystemArea 90-13A 1000 1000 1000 South RWCU System Area 90-14A 1000 1000 1000 RWCU System Area 90-9A 1000 1000 1000 MG Set Area 90-4A 1000 1000 1000 Fuel PooiArea 90-lA 1000 1000 1000 Service Fir Area 90-2A 1000 1000 1000 New Fuel Storage 90-3A 1000 NA NA 54
NRC Scenario 1 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
HPCI is Tagged Out Operations/Maintenance for the Shift:
Unit 3 is at 70% Power. Condensate Booster Pump 3A has been placed back in service. RFPT 3A is warmed and operating at 600 rpm, place RFPT 3A in service in accordance with 3-01-3 Reactor Feedwater System section 5.7. An RPHP is effect for placing the RFPT in service, all data and signatures are recorded on Appendix A.
Then commence power ascension with Control Rods in accordance with the reactivity control plan.
Units I and 2 are at 100% power Unusual Conditions/Problem Areas:
The following Control Rods are identified as SLOW: 30-19, 34-23, 14-5 1, 02-19, 46-5 1, and 06-43.
55
NRC Scenario 2 Facility: Browns Ferry NPP Scenario No.: NRC 2
- Op-Test No.:
Examiners: Operators: SRO:_
ATC:
BOP:
Initial Conditions: 4.5% power. 2-GOI-100-1A Section 5.4 Step 64 and 66 Turnover: Continue to pull rods to 8% power, verify IRM/APRM overlap at 5% and continue to 8% power and hold for Mode Change.
Event Maif. No. Event Type* Event Description No.
N-BOP Start SBGT Fan C and align to Reactor Zone Ventilation 1
TS-SRO lAW 0-01-65 section 5.2 2 Raise power with Control Rods; IRM B fails as is C-ATC 3 imfrd06 Control Rod 50-11 and 58-19 difficult to withdraw C-SRO 1-BOP 4 imf fw20a Condenser Hotwell Level Automatic Makeup controller failure I-SRO C-BOP 5 imfed07a Loss of 480V Unit Board 2A, failure of EHC Pump 2B to auto start C-SRO 6 imfth3lb TS-SRO Pressure Transmitter PT-3-22BB, falls high C-ATC 7 ior RFPT C Trip C-SRO 8 th23 M-ALL ATWS and Fuel Failure RWCU Leak, failure of auto isolation only, high Secondary 9 cu06alb C Containment Radiation, ED not required 10 imf rd0 1 C CRD Pump Trip (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or 1
NRC Scenario 2 Critical Tasks Three With a reactor scram required and the reactor not shutdown, initiate action to reduce power by inserting control rods.
- 1. Safety Significance:
Shutting down reactor can preclude failure of containment or equipment necessary for the safe shutdown of the plant.
- 2. Cues:
Procedural compliance CRD Pump B operating
- 3. Measured by:
Observation Control Rod insertion commenced in accordance EOI Appendixes.
- 4. Feedback:
Reactor Power trend.
Control Rod indications.
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 before two areas reach the MAX Safe radiation levels.
- 4. Feedback:
Valve position indication 2
NRC Scenario 2 Critical Tasks Three When MAiN STEAM LINE RADIATION HIGH-HIGH is received and SLC injection is NOT required in accordance with RC/Q of EOI- I, MSIVs are closed.
- 1. Safety Significance:
Isolate potential discharge path to environment through Off-Gas System which is currently bypassed and will not isolate or go to treat.
- 2. Cues:
Procedural compliance.
Reactor Pressure trend.
- 3. Measured by:
When ARP-9-3A window 27 annunciator is received MSIVs are closed if alarm is verified valid and SLC injection lAW EOI-l RC/Q is NOT required. MSIV are closed without delay when the SRO has been notified of the annunciator and has evaluated that SLC Injection is not required lAW EOI-1 RC/Q.
- 4. Feedback:
Reactor Pressure trend.
MSIVs indicate closed.
3
NRC Scenario 2 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 Continues Power ascension with control rods to 8%. IRM/APRM overlap required to be verified at 5% power, prior to continued ascension to 8%. IRM B fails as is and during power ascension the ATC will identify that IRM B is not responding to increasing power.
- 3. Control Rod 50-11 will not withdraw with normal methods from position 00, the ATC will take action JAW 2-01-8 5 for control difficult to withdraw. The ATC will double clutch control rod 50-11, the control rod will withdraw. Control Rod 58-19 will not withdraw with normal methods from position 12, the ATC will take action JAW 2-01-85 for control difficult to withdraw. The ATC will raise drive water pressure and the control rod will withdraw.
- 4. Condenser Hotwell Level Automatic Makeup controller will fail closed. The BOP operator will responds to alarms on Panel 9-6A-5, 6, 7 OR notice Hotwell Level on 2-LR-2-9 trending down or CST FLOW to Hotwell at 0 on 2-FT-2-48. The operator will take manual control and begin to restore hotwell level.
- 5. A loss of 480V Unit Board 2A will occur. EHC Pump 2A will trip with failure of standby pump to auto start, BOP operator will start EHC Pump 2B to prevent a loss of EHC pressure and closure of Turbine Bypass Valves.
- 6. Pressure Transmitter PT-3-22BB, fails high. This will cause a Half Scram and a high Reactor Pressure annunciator. The SRO will evaluate Tech Specs 3.3.1.1 and determine condition A is entered.
SRO will enter 2-AOJ-3 -1. While restoring Reactor Level if the restoration is too rapid, power oscillations will occur and the SRO may direct a Reactor Scram at this point.
- 8. After Reactor Level is restored a fuel failure will occur. Rising radiation levels will occur throughout the Unit, the SRO will determine when to scram. Eventually the Main Steam Line High-High will alarm requiring a Reactor Scram and MSJVs to be isolated if SLC is not required lAW the RC/Q leg of EOI- 1. An ATWS will result on the scram and the SRO will evaluate conditions.
- 9. Shortly after the scram RWCU will develop a small leak, with the fuel failure two areas in Secondary Containment will be at or above max safe radiation levels. The SRO will enter EOI-3 and evaluate the need to Emergency Depressurize. The crew will isolate RWCU and with no primary system discharging to Secondary Containment ED will not be required.
- 10. The operating CRD Pump will trip while rods are being driven requiring the RO to start the standby CRD Pump in order to continue rod movement.
4
NRC Scenario 2 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner.
Control Rods are being inserted Reactor Level is maintained RWCU has been isolated SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 2 10 Total Malfunctions Inserted: List (4-8) 3 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 1 EOIsused: List(1-3) 0 EOI Contingencies used: List (0-3) 80 Validation Time (minutes) 3 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No) 5
NRC Scenario 2 Scenario Tasks TASK NUMBER K/A RO SRO Manual Initiation of SBGT Fan C RO U-065-NO-02 261000A4.07 3.1 3.2 SRO S-000-AD-27 Raise Power with Control Rods RO U-085-NO-07 SRO S-000-AD-31 2.2.2 4.6 4.1 Control Rod difficult to Withdraw RO U-085-AB-07 201002A2.02 3.2 3.3 SRO S-085-AB-07 Hotwell Level Automatic Makeup controller failure RO U-002-AL-O1 201002A2.02 3.2 3.3 SRO S-002-AL-O1 EHC Pump Trip RO U-47A-AL-07 256000A2.06 3.2 3.2 SRO S-57B-NO-07 RFPT C Trip ROU-003-AB-O1 259001A2.O1 3.7 3.7 SRO 5-003-AB-Ol RWCU Leak RO U-069-AL-09 295033EA1.05 3.9 4.0 SRO S-000-EM-10 6
NRC Scenario 2 Scenario Tasks TASK NUMBER RQ SiQ Fuel Failure RO U-099-AL-05 272000A2.O1 3.7 4.1 SRO S-066-AB-02 ATWS RO UOOO-EM-O3 295015AA2.O1 4.1 4.3 RO U-000-EM-22 RO U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM-18 7
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NRC Scenario 2 Simulator Instructor IC 104 Batch File NRC/l3O6nrc-2 imf cu06a imf cu06b trg e5 NRC/ehc trg e5= bat NRC/ehcpumptrip-1 trg elO NRC/modesw imfcu04 (dO 2:00) 70 180 0 RWCU Leak ior zdihs74l55a normal imfrd06r50l 1 bat atws80 ior ypomtrsbgthtrrh fail_control_power ior ypobkrrhfpmb fail_ccoil imfnni05b 21 imf rd06r58 19 imffwO8c 15 Batch File NRC/l3O6nrc-2a dmf fw08c ior zdihs03 176 trip Preference File NRC/l3O6nrc-2 pflcOl tog pfk 02 ann silence pfk 03 bat NRC/l3O6nrc-2 p1k 04 imf fw2O 0 Condenser hotwell level auto makeup failure pflc 05 imfed07a 480V unit bd 2A loss p1k 06 imfth3 lb Pressure Transmitter PT-3-22BB fails high pflcO7 pflc 08 bat NRC/l3O6nrc-2a RFPT 2C Trip p1k 09 imfth23 15 180 0 Fuel Failure p1k 10 bat sdv p1k 11 bat appOlf p1k 12 bat app02 plksl p1k s2 p1k s3 bat app08ae p1k s4 mrfrd06 close p1k s5 mrf rdO6 open plks6 imfrd0la CRD Pump 2A trip 9
NRC Scenario 2 Scenario 2 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 104 restorep ref NRC/i 3O6nrc-2 Simulator Setup Load Batch F3 NRC/i306nrc-2 Simulator Setup Verify file loaded Verify Rod Worth Minimizer Working Verify IRM B on range 8, when the batch file is loaded range all IRMs upscale and then back to mask IRM B Failure, Dial 2-LC-2-3 Hotwell High Level Dumpback to an indication on 2-FT-2-47 of Manual about 151. Set 2-LC-2-3 at 30 in manual and 2-LC-2-6 at 30.5 in auto Swap lens covers on Bus Duct Cooling Fans Advance chart recorders in fast K Insert pref key f4 as soon as the crew assumes the tvianuai1 shift. imf fw2O 0 RCP required (4% - 8%), Marked up copy of 2-GOI-100-1A Unit Startup 10
NRC Scenario 2 Simulator Event Guide:
Event 1 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-65-69A/2 in START.
[5] CHECK SOT 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 SOT 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 11
NRC Scenario 2 Simulator Event Guide:
Event 1 Normal: Start SBGT Fan C and align to Reactor Bldg lAW 0-01-65 section 5.2 BOP should identify failure of the RI-I 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 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 SUPERVISORJSRO and Unit 1 and Unit 3.
SRO SRO Evaluate Technical Specification 3.6.4.3 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Three SGT subsystems shall be OPERABLE.
APPLICABILITY: MODES 1,2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs).
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 SRO Determines that cannot change MODE with SBGT Train C inoperable 12
NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 SRO Direct Power increase using Control Rods per 2-GOI-100-1A, Section 5.4 5.4 Withdrawal of Control Rods while in Mode 2
[64] VERIFY IRMJAPRM overlap by operator visual observation before exceeding 5% power.
[66] CONTINUE to withdraw control rods to raise Reactor power to approximately 8% per 2-01-85 and 2-SR-3.1.3.5(A).
ATC Raise Power with Control Rods JAW 2-01-85, Section 6.6 Group 40 = 18-03 and 02-19 from 00 to 12 Group 41 = 10-51, 50-51, 50-11, and 10-11 from 00 to 12 Group 42 = 02-43, 18-59, 42-59, 58-43, 58-19, 42-03, 18-03, 02-19 from 12 to 48 6.6.1 Initial Conditions Prior to Withdrawing Control Rods
[2] VERIFY the following prior to control rod movement:
- CRD POWER, 3-HS-85-46 in ON.
- Rod Worth Minimizer is operable and LATCHED into the correct ROD GROUP when Rod Worth Minimizer is enforcing (not required with no fuel in RPV).
6.6.2 Actions Required During and Following Control Rod Withdrawal
[1] IF control rod fails to withdraw, THEN Refer to Section 8.15 for additional methods to reposition control rod.
[2] IF control rod double notches, or withdraws past its correct/desired position, THEN Refer to Section 6.7 for inserting control rod to its correct/desired position.
[3] IF at any time while driving a selected rod during the performance of this section, the Control Rod moves more than one notch from its intended position, THEN Refer to 2-AOI-85-7, MISPOSITIONED CONTROL ROD.
[4] OBSERVE the following during control rod repositioning:
- Control rod reed switch position indicators (four rod display) agree with indication on 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 9-5 and perform step 6.6.2[6].
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NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42
[6] IF Control Rod movement was stopped to keep from exceeding a RBM Setpoint or ATC 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.
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.
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NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [6] IF control rod is notch withdrawn to rod notch Position 48, THEN PERFORM control rod coupling integrity check as follows:
[6.1] PLACE CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.
[6.2] CHECK control rod coupled by observing the following:
- Four rod display digital readout AND full core display digital readout AND background light remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window
- 14) does not alarm.
[6.3] CHECK control rod settles into Position 48 and ROD SETTLE light extinguishes.
[6.4] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2.
6.6.4 Continuous Rod Withdrawal 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-3.1.3 .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.)
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NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [1] SELECT the desired control rod by depressing the appropriate CR1) ROD SELECT pushbutton, 2-XS-85-40.
[2] OBSERVE the following for selected control rod:
- CR1) 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] 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 OVERRIDE.
[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 both CR1) 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-HS85-48, in ROD OUT NOTCH and RELEASE.
[5.5] WHEN control rod settles into the intended notch, THEN CHECK the following:
- Four rod display digital readout and full core display digital readout and background light will remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window
- 14) does not alarm.
[5.6] CHECK control rod settles at intended position and ROD SETTLE light extinguishes.
ATC During power ascension IRM B fails to respond to continuous steady counts, ATC reports to Unit Supervisor SRO Directs Startup to continue, have all needed IRIvI instruments. Need 6 of 8 IRMs. Direct IRM B bypassed lAW OI-92A, IRMs 16
NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC NOTE When continuously withdrawing a control rod to position 48, the control rod coupling integrity check can be performed by one of the two following methods:
- 1) Coupling integrity check while maintaining the CRD Notch Override Switch in the Override position and the CRD Control Switch in the Rod Out Notch position. If this method is selected, perform Step 6.6.4[6] and N/A Step 6.6A[7].
- 2) Coupling integrity check after releasing the CRD Notch Override Switch and the CRD Control Switch. If this method is selected, perform Step 6.6.4[7] and N/A Step 6.6.4[6].
[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 CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE.
[6.2] PLACE and HOLD CRD CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.
[6.3] MAINTAIN the CR1) 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 full core display digital readout and background light will remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window
- 14) does not alarm.
[6.5] RELEASE both CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1)
CONTROL SWITCH, 2-HS-85-48.
[6.6] ChECK control rod settles into position 48 and ROD SETTLE light extinguishes.
[6.7] fi? control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2.
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NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 ATC [7] IF continuously withdrawing the control rod to position 48 control rod coupling integrity check is to be performed after the CR1) NOTCH OVERRIDE, 2-HS-85-47, and CRD CONTROL SWITCH, 2-HS-85-48 are to be released, THEN PERFORM control rod coupling integrity check as follows (otherwise N/A):
[7.1] PLACE and HOLD CR1) NOTCH OVERRIDE, 2-HS-85-47, in OVERRIDE.
[7.2] PLACE and HOLD CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH.
[7.3] WHEN position 48 is reached, THEN RELEASE CR1) NOTCH OVERRIDE, 2-HS-85-47, and CR1) CONTROL SWITCH, 2-HS-85-48.
[7.4] VERIFY control rod settles into position 48.
[7.5] PLACE CR1) CONTROL SWITCH, 2-HS-85-48, in ROD OUT NOTCH and RELEASE.
[7.6] CHECK control rod coupled by observing the following:
- Four rod display digital readout and full core display digital readout and background light will remain illuminated.
- CONTROL ROD OVERTRAVEL annunciator (2-XA-55-5A, Window
- 14) does not alarm.
[7.7] CHECK control rod settles into position 48 and ROD SETTLE light extinguishes.
[7.8] IF control rod coupling integrity check fails, THEN Refer to 2-AOI-85-2.
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 CR1) POWER, 2-HS-85-46, in OFF.
[1.2] PLACE CRD POWER, 2-HS-85-46, in ON.
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NRC Scenario 2 Simulator Event Guide:
Event 2 Reactivity: Raise power with Control Rods, Group 40, 41 and 42 SRO Directs Startup to continue, have all needed IRM instruments. Need 6 of 8 lRMs. Direct IRM B bypassed lAW OI-92A, IRMs 6.1 Bypassing an 1PM Channel CAUTION NPG-SPP 10.4 requires approval of the Plant Manager or his designee prior to any planned operation with IRMs bypassed unless bypassing is specifically allowed within approved ATC procedures.
NOTES
- 1) It is not necessary for a bypassed IRM channel to have its detector inserted into the core.
- 2) Only one IRM in each trip system can be bypassed at a time.
- 3) All operations are performed on Panel 2-9-5 unless specifically stated otherwise.
[1] REVIEW all precautions and limitations in Section 3.0.
[2] PLACE the appropriate IRM Bypass selector switch to the BYPASS position:
- IRM BYPASS, 2-HS-92-7A/S4A
- IRM BYPASS, 2-HS-92-7A/S4B
[3] CHECK that the_Bypassed_light_is_illuminated.
NRC If Reactor Power is approximately 8%, SRO will contmue m 2-GOI-100-1A SRO When Reactor power is approximately 8%, continues in 2-GOI- 100-1 A Directs
[70] IF primary containment purge and/or Primary Containment Ventilation is in service, THEN PLACE the following switches in the BYPASS position (Panel 2-9-3):
- PC PURGE DIV I RUN MODE BYPASS, 2-HS-64-24.
- PC PURGE DIV II RUN MODE BYPASS, 2-HS-64-25.
BOP Bypasses PC PURGE DIV I RUN MODE BYPASS, 2-HS-64-24 and PC PURGE DIV II RUN MODE BYPASS, 2-HS-64-25.
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NRC Scenario 2 Simulator Event Guide:
Event 3 Component: Control Rod 50-1 land 58-19 difficult to withdraw 8.15 Control Rod Difficult to Withdraw ATC [1] VERIFY the control rod will not notch out. Refer to Section 6.6.
[2] REVIEW all Precautions and Limitations in Section 3.0
[3] IF RWM is enforcing, THEN VERIFY RWM is operable and LATCHED in to the correct ROD GROUP.
NOTES
- 1) Steps 8.15[4] through 8.15[6] should be used when the control rod is at Position 00 while Step 8.15[7] should be used when the control rod is at OR between Positions 02 and 46.
- 2) Double clutching of a control rod at Position 00 will place the rod at the overtravel in stop, independent of the RMCS timer, allowing maximum available time to establish over-piston pressure required to maintain the collet open and prevent the collet fmgers from engaging the 00 notch.
- 3) Step 8.15[4] may be repeated as necessary until it is determined that this method will not free the control rod.
[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.
ATC Will Double clutch Control Rod 50-11 from position 00 Driver when ATC goes to double clutch 50-Il, delete stuck rod 50-11 20
NRC Scenario 2 Simulator Event Guide:
Event 3 Component: Control Rod 50-1 land 58-19 difficult to withdraw
[7] IF the control rod is at or between Positions 02 and 46, ThEN PERFORM the following to withdraw the control rod using elevated drive water pressure:
[7.1] RAISE the CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to 300 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.
ATC [7.2] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48.
[7.3] IF the control rod successfully notches out, THEN LOWER CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL
[7.4] IF the control rod still fails to NOTCH OUT, THEN RAISE CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to 350 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.
[7.5] ATTEMPT to withdraw the Control Rod using CRD CONTROL SWITCH, 2-HS-85-48.
[7.6] LOWER CRD DRIVE WTR HDR DP, 2-PDI-85-17A, to between 250 psid and 270 psid, using CRD DRIVE WATER PRESS CONTROL VLV, 2-HS-85-23A.
[7.7] IF the control rod still fails to NOTCH OUT using elevated CRD DRIVE WTR HUR DP, THEN CONTACT Reactor Engineer and NOTIFY Unit Supervisor for further instructions.
ATC Will raise drive water pressure to 350 psid to successfully move Control Rod 58-19 from position 12 Driver when ATC raises drive water pressure to 350 psid, delete stuck rod 5 8-19 21
NRC Scenario 2 Simulator Event Guide:
Event 4 Instrument: Condenser Hotwell Level Automatic Makeup controller failure This malfunction will be in from the beginning of the scenario, hotwell level starts at 30 NRC inches and this alarm is at 24 inches. May notice before alarm and take action at that time Responds to Alarms on Panel 9-6A-5, 6, 7 OR notices Hotwell Level on 2-LR-2-9 trending BOP down or CST FLOW to Hotwell at 0 on 2-FT-2-48 6A-5, HOTWELL A LEVEL ABNORMAL 6A-6, HOT WELL B LEVEL ABNORMAL 6A-7, HOTWELL C LEVEL ABNORMAL A. VERIFY abnormal level on 2-LIC-2-3, Panel 2-9-6.
L IF level is high, THEN RAISE hotwell reject.
- 2. IF level is low, THEN RAISE makeup.
B. CHECK condenser vacuum normal.
C. CHECK:
- 1. Hotwell level recorder 2-LR-2-9C and 2-LIC-2-6.
- 2. Hotwell makeup flow 2-FR-2-48, Panel 2-9-6.
- 3. Hotwell dump flow 2-FR-2-47, Panel 2-9-6.
- 4. Conductivity rising recorder 2-CR-43 1A, Panel 2-9-6; recorder 2-CR-43-1 1A/12A, Panel 2-9-4.
- a. IF conductivity is rising, THEN REFER TO 2-AOI-2-l.
- 5. Bypass valves 2-LCV-2-4 and -7 closed.
- 6. Locally verify level in Hotwell level sightglass 2-LG-2-26 1.
D. IF cause is malfunction of automatic level control, ThEN REFER TO 2-01-2, Section 8.3, Manual HW Level Control.
BOP OPENS CONDENSATE MAKEUP BYPASS VLV, 2-LCV-2-7, using 2-HS-2-7A 22
NRC Scenario 2 Simulator Event Guide:
Event 4 Instrument: Condenser Hotwell Level Automatic Makeup controller failure BOP Refers to 2-01-2 Section 8.3 8.3 Manual Ilotwell Level Control
[3] IF desired to raise Hotwell Level, THEN PERFORM the following, as required:
[3.1] THROTTLE OPEN HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LCV-2-6, using 2-LC-2-6 in MAN to maintain desired level.
[3.2] OPEN CONDENSATE MAKEUP BYPASS VLV, 2-LCV-2-7, using 2-HS-2-7A, if necessary.
[3.3] VERU?Y CLOSED CONDENSATE DUMPBACK BYPASS VLV, 2-LCV-2-4, using 2-HS-2-4A.
[3.4] THROTTLE CLOSED HOTWELL HIGH LEVEL DUMPBACK CONTROL, 2-LC-2-3, as necessary to raise Hotwell level.
[5] IF desired to maintain Hotwell Level, THEN:
[5.1] PLACE HOTWELL HIGH LEVEL DUMPBACK CONTROL, 2-LC-2-3, in MAN and ESTABLISH approximately equal to or greater than 60 x 103 lbmlhr flow from the Hotwell to the CST as indicated on 2-FR-2-47.
[5.2] PLACE HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, in MAN and ESTABLISH the required amount of flow from the CST to the Hotwell as necessary to stabilize desired HOTWELL LEVEL.
[5.3] II? auto control of HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, is desired, THEN PLACE HOTWELL LOW LEVEL MAKEUP CONTROL, 2-LC-2-6, in AUTO with controller set at desired HOTWELL LEVEL.
BOP When hotwell level is restored CLOSES CONDENSATE MAKEUP BYPASS VLV, 2-LCV-2-7, using 2-HS-2-7A 23
NRC Scenario 2 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.
BOP Responds to the following alarms; 7A-22, 8C-3, 8C-lO, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.
7A-22 GEN STATOR COOLANT SYS ABNORMAL A. IF while performing the action of this ARP 2-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, 2-TR-57-59, Panel 2-9-8.
Generator NOT on line, verifies Stator water cooling pump running 8B-16 480V UNIT BD 2A UV OR XFR A. VERIFY automatic action has occurred.
B. INSPECT 480V Unit Bd A for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
C. REFER TO O-OI-57B to re-energize board.
D. REFER TO appropriate 01 for recovery or realignment of equipment.
Reports trip of 480V Unit Bd 2A, dispatches operators 8C-3 480V RX BLDG VENT BD 2A UV OR XFR A. VERIFY automatic action has occurred.
B. ChECK or START refuel floor and reactor zone exhaust fans 2A or 2B.
C. CHECK board for abnormal condition: relay targets, smoke, burned paint, breaker position, etc.
D. REFER TO O-OI-57B to re-energize or transfer board.
E. REFER TO appropriate 01 for recovery or realignment of equipment.
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NRC Scenario 2 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.
BOP Responds to the following alarms; 7A-22, 8C-3, 8C-lO, 8C-l5, 8C-16, 8C-25, 8B-16, 7B..l and 7B-15.
8C-lO 480V RX BLDG VENT BD 2B UV OR XFR A. VERIFY automatic action has occurred.
B. CHECK or START refuel floor and reactor zone fans 2A or 2B.
C. CHECK 480V Reactor Bldg Vent Bd 2B for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
D. REFER TO O-OI-57B to re-energize board.
E. REFER TO appropriate 01 for recovery or realignment of equipment.
8C-15 480V TB VENT BD 2A UV OR XFR A. VERIFY alarm by checking the following:
- Associated annunciator, TURBiNE BLDG VENTILATION ABNORMAL, (2-XA-55-3D, Window 4) in alarm.
- Power light on MTOT vapor extractor and EHC fluid heaters, Panel 2-9-7.
B. DISPATCH Personnel to 480V Turb Bldg Vent Bd 2A to CHECK equipment and board status for abnormal conditions.
- Mechanical spaces supply fan 2A, 2B and exhaust fan.
- Turb room supply fan 2A, 2B and exhaust fans 2A, 2B, 2C, 2D.
- EHC fluid transfer and filtering pump.
C. REFER TO O-OI-57B to re-energize or transfer the board.
D. REFER TO appropriate 01 for recovery or realignment of equipment.
If called to verify if the following boards transferred report that these boards have auto transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 25
NRC Scenario 2 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf edO7a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.
BOP Responds to the following alarms; 7A-22, 8C-3, 8C-1O, 8C-15, 8C-16, 8C-25, 8B-16, 7B-l and 7B-15.
8C-16 480V TURB MOV BD 2A UV OR XFR A. VERIFY alarm by checking light indication to the following equipment:
- RFW heaters (2B1,2B2,2C1,2C2) extraction isolation valves.
- RFPT 2B 2B2 Main Oil Pump.
- RFPT 2C 2C1 Main Oil Pump.
B. ChECK board inspected for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
C. REFER TO ICS screen VFDAAL or VFDBAL and verify PROCESS ALARM Internal HX Fan Power status is OK.
D. REFER TO O-OI-57B to re-energize or transfer the board.
E. REFER TO appropriate 01 for recovery or realignment of equipment.
8C-25 480V CNDS DEMIN BD 2 UV OR XFR A. VERIFY automatic transfer by dispatching personnel to 480V Cnds Demin Bd 2 to check for the following:
- Power available lights illuminated.
- Normal disconnect switch 1 A open and alternate disconnect 2A closed.
- Any abnormal conditions such as breaker trips.
B. NOTIFY Radwaste Operator.
C. IF power NOT available, THEN OBTAIN status of the following from Radwaste:
- Condensate demins precoat operation.
- Condensate backwash transfer operation.
- Backwash receiver pit floor drains.
D. REFER TO O-OI-57B for power restoration or transfer instructions.
If called to verify if the following boards transferred report that these boards have auto
.* transferred to their alternate supply. 480V RX BLDG VENT BD 2A, 480V RX BLDG Driver VENT BD 2B, 480V TB VENT BD 2A, 480V TURB MOV BD 2A, and 480V CNDS DEMIN BD 2 26
NRC Scenario 2 Simulator Event Guide:
Event 5 Component: Loss of 480V Unit Board 2A Driver When requested by the NRC insert preference key F5 for imf ed07a, when sent to investigate wait 3 minutes and inform the Normal Supply Breaker for Unit Board 2A has an actuation of the overcurrent relay.
BOP Responds to the following alarms; 7A-22, 8C-3, 8C-10, 8C-15, 8C-16, 8C-25, 8B-16, 7B-1 and 7B-15.
7B-1 EHC HYD FLUID HDR PRESS LOW A. VERIFY Standby EHC PUMP 2B(2A), 2-HS-47-2A(1A) running.
B. ChECK EHC HEADER PRESSURE indicator, 2-PI-47-7 between 1550 and 1650 psig.
C. DISPATCH personnel to inspect EHC pump unit.
D. IF EHC Hydraulic system fails, THEN VERIFY turbine trips at or below 1100 psig.
7B-15 STANDBY EHC PUMP FAILED A. On Panel 2-9-7:
NOTE Lights extinguish at 1300 psig lowering and illuminate at 1500 psig rising.
- 4. CHECK lights above EHC PUMP 2A TEST pushbutton 2-HS-47-4A and EHC PUMP 2B TEST pushbutton 2-HS-47-5A.
B. DISPATCH personnel to pumping unit to check for abnormal conditions.
C. IF EHC Hydraulic System fails, THEN VERIFY turbine trips at or below 1100 psig.
BOP Starts Standby EHC Pump 2B and verifies EHC pressure returns to normal Driver When directed by NRC insert preference key F6, and when asked to investigate in aux instrument room report 2-PIS-3-22BB is failed high 27
NRC Scenario 2 Simulator Event Guide:
Event 6 Instrument: Pressure Transmitter PT-3-22BB, fails high ATC Responds to alarms 5B-2 and 4A-9, reports half scram, power, pressure and level stable 5B-2 REACTOR CHANNEL B AUTO SCRAM A. VERIFY channel B relays dropped out by checking scram solenoid and backup scram valve lights extinguished.
B. IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s).
C. IF alarm due to inadvertent criticality during incore fuel movements, ThEN REFER TO 2-AOI-79-2.
D. IF alarm is from a control rod drop, THEN REFER TO 2-AOI-85-1.
E. With SRO permission, RESET half-scram. REFER TO 2-01-99.
4A-9 RX VESSEL PRESSURE HIGH HALF SCRAM A. VERIFY alarm by multiple indications.
B. IF the alarm is valid AND reactor has NOT scrammed, THEN MANUALLY SCRAM the reactor. ENTER 2-EOI-1.
C. DISPATCH personnel to the sensors to check for abnormal conditions.
D. IF alarm is invalid, THEN with SRO permission, RESET Half Scram.
28
NRC Scenario 2 Simulator Event Guide:
Event 6 Instrument: Pressure Transmitter PT-3-22BB, fails high SRO 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
APPLICABILITY: According to Table 3.3.1.1-1.
Table 3.3.1.1-i (page2of3)
Resctcr Protection System Ins8ijmentalion APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.i
- 3. RtorVessel&eamDome 12 0 SR 3.3.1.1.1 1090 psig Presre Hlghcd)
SR 3.3.1.1.8 SR 3.11.1.10 SR 3.3.1.1.14 SRO Condition A: One or more required channels inoperable Required Action A. 1: Place channel in trip.
Completion Time: 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR Required Action A.2: NOTE Not applicable for Functions 2.a, 2.b, 2.c, 2.d or 2.f.
Place associated trip system in trip.
Completion Time: 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRO May direct the following fuse pulled lAW 2-01-99, 2-FU 1-3 -22BA, (5AF5B) 2-RLY-099-O5AKO5B at Panel 9-17, ALARMS AND 1/2 SCRAM IN B CHANNEL Driver When directed by NRC insert preference key F8 for a trip of KEPT 2C. After RFPT 2C is tripped wait 3 minutes after dispatched and report a scaffold crew removing scaffold from the area believes they may have tripped RFPT.
When RFPT 3C is tripped verify imf fwO8c is deleted 29
NRC Scenario 2 Simulator Event Guide:
Event 7 Component: RFPT 2C Trip ATC Responds to a trip of RFPT 2C, reports Reactor level lowering, Reactor power lowering and reactor pressure stable.
Responds to the following alarms 6C-15, 6C-23, 6C-29, 6C-32 and 5A-8 6C-15 RFPT C ABNORMAL A. CHECK other RFP alarms on Panel 2-9-6 to determine problem area.
B. REFER TO appropriate alarm response procedure.
6C-23 RFPT TRIP CIRCUIT ABNORMAL A. VERIFY alarm and RFPT trip by checking Panel 2-9-6, RFPT speed, governor valve position and discharge flow.
B. VERIFY reactor power is within the capacity of operating RFPs.
C. IF BKR TRIPOUT PNL 2-9-9 DC DIST (2-XA-55-8C, alarm window 20) is illuminated, THEN Step is NA fl IF RFP is tripped, TIIEN REFER TO 2-01-3, Section 8.1 or 2-AOI-3-1.
6C-29 RFPT TRIPPED A. VERIFY reactor power is within the capacity of operating RFPs.
B. CHECK core limits.
C. WHEN RFPT coasts down to zero speed, unless RFPT is rolling on minimum flow, THEN VERIFY turning gear motor starts and engages.
D. REFER TO 2-AOI-3-l_or 2-01-3,_Section_8.1.
6C-32 RFP DISCH FLOW LOW A. VERIFY reactor Feedpump flow on Panel 2-9-6.
B. REFER TO 2-01-3, Section 8.7.
30
NRC Scenario 2 Simulator Event Guide:
Event 7 Component: RFPT 2C Trip ATC Responds to a trip of RFPT 2C, reports Reactor level lowering, Reactor power lowering and reactor pressure stable.
Responds to the following alarms 6C-15, 6C-23, 6C-29, 6C-32 and 5A-8 5A-8 REACTOR WATER LEVEL ABNORMAL A. VALIDATE Reactor water level hi/low using multiple indications including Average Narrow Range Level on 2-XR-3-53 recorder, 2-LI-3-53, 2-LI-3-60, 2-3-206 and 2-LI-3-253 on Panel 2-9-5.
B. IF alarm is valid, THEN REFER TO 2-AOI-3-1 or 2-01-3.
SRO Directs entry to 2-A0I 1, Loss of Reactor Feedwater or Reactor Water Level ATC 2-A0I-3-l 5.0 LOW REACTOR WATER LEVEL OR LOSS OF FEEDWATER
[7] IF RFPs in manual control, ThEN RAISE speed of operating RFPs.
[9] II? RFPT has tripped and will not be required to maintain level, THEN REFER TO 2-01-3 and SHUT DOWN tripped RFPT.
[11] IF unit remains on-line, THEN RETURN Reactor water level to 33?l (normal range).
ATC Takes control of RFPT 2B and raises RFPT speed and discharge pressure until RFPT 2B injects_and raises and maintains RPV Level.
NRC vill drop to about 3% as level lowers, depending on how aggressive level is restored o- may peak and cause and auto scram at 14% reactor power, or the SRO may direct a then he sees power oscillating depending on the actions of the ATC operator on IAvI control.
Driver irected by NRC or if the crew scrams the reactor insert preference key F9 for fuel Ti A K 1... T1 Ti A T 31
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Respond to the following radiation alarms 3A-29, 3A-5, 3A-22 and 3A-7 Alarm 3A-29, TURBiNE BLDG AREA RADIATION HIGH 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.)
B. IF the TSC is NOT manned, THEN USE public address system to evacuate area where high airborne conditions exist.
D. NOTWY RAD PRO.
Alarm 3A-5, OG PRETREATMENT RADIATION HIGH A. VERIFY high radiation on following:
- 1. OFFGAS RADIATION recorder, 2-RR-90-266 on Panel 2-9-2.
- 2. OG PRETREATMENT RAE) MON RTMR, 2-RM-90-157 on Panel 2-9-10.
- 3. OFFGAS RAD MON RTMR, 2-RM-90-160 on Panel 2-9-10.
B. CHECK off-gas flow normal.
C. CHECK following radiation recorders and associated radiation monitors:
- 1. MAiN STEAM LINE RADIATION, 2-RR-90-135 on Panel 2-9-2.
- 2. OFFGAS POST-TREATMENT RADIATION, 2-RR-90-265 on Panel 2-9-2.
- 3. STACK GAS/CONT RM RADIATION, 0-RR-90-147 on Panel 1-9-2.
D. NOTWY RAD PRO.
E. NOTWY Chemistry to perform radiochemical analysis to determine source 32
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Respond to the following radiation alarms 3A-29, 3A-5, 3A-22 and 3A-7 Alarm 3A-22, RX BLDG AREA RADIATION HIGH 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.)
B. IF Dry Cask storage activities are in progress, THEN NOTIFY CASK Supervisor.
C. IF alarm is from the HPCI Room while Flow testing is performed, NOTIFY personnel at the HPCI Quad to validate conditions.
D. NOTIFY RAE PRO.
E. IF the TSC is NOT manned and a VALID radiological condition exists, THEN USE public address system to evacuate area where high radiological conditions exist.
J. For all radiation indicators except FUEL STORAGE POOL radiation indicator, 2-RI-90-30, ENTER 2-EOI-3 Flowchart.
BOP Reports RWCU South Area 90-14a and RWCU 621 Elev 90-9a in alarm and each are EOI 3 entry conditions SRO Announce Entry to EOI-3 BOP Alarm 3A-7, MAIN STEAM LINE RADIATION HIGH A. CHECK following radiation recorders:
- 1. MAIN STEAM LINE RADIATION, 2-RR-90-135 on Panel 2-9-2.
- 2. OFFGAS RADIATION, 2-RR-90-266 on Panel 2-9-2.
- 3. STACK GAS/CONT RM RADIATION, O-RR-90-147 on Panel 1-9-2.
B. NOTIFY RAD PRO.
C. NOTIFY Chemistry to perform radiochemical analysis of primary coolant.
33
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS SRO Enters EOI-3 on Secondary Containment Radiation SRO If Reactor Zone or Refuel Zone Exhaust Radiation Level is above 72 mr/hr. NO 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. NO SRO WREN Any Area Radiation Level Above Max Normal? - YES, proceeds Isolate all systems that are discharging into the area except systems required to:
- Suppress a Fire Will Emergency Depressurization Reduce Discharge Into Secondary Containment? NO, NO System Discharging Before any area radiation rises to Max Safe (table 4) Continue and enter EOI- 1 (EOI- 1 has afready been entered after Reactor Scram)
SRO If Reactor not SCRAMMED, Direct Scram Crew Monitors for Max Safe Radiation and reports BOP Respond to the following radiation alarm 3A-27, MAIN STEAM LiNE RADIATION HIGH-HIGH A. VERIFY the alarm on 2-RM-90-136 and 2-RM-90-137 on Panel 2-9-10.
B. CONFIRM main steam line radiation level on recorder 2-RR 135, Panel 2-9-2.
C. IF alarm is valid and Reactor Scram has not occurred, THEN PERFORM the following:
- 1. IF core flow is above 60% THEN LOWER core flow to between 5 0-60%..
- 2. MANUALLY SCRAM the Reactor.
- 3. REFER to 2-AOl- 100-1.
D. IF SLC injection per RC/Q of EOI-1 is NOT required, THEN VERIFY the MSIVs closed.
E. NOTIFY RAD PRO.
SRO Directs Manual SCRAM, once scrammed evaluates conditions and with NO EOI entry conditions_for_EOI- 1,_directs_MSIVs_Closed BOP Closes MSIVs 34
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS SRO If have NOT directed SCRAM, directs SCRAM and entry to 2-AOl- 100-1, Reactor Scram ATC 4.1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5A/S3A and 2-HS-99-5A/S3B, on Panel 2-9-5.
[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA
[3] REFUEL MODE ONE ROD PERMISSIVE light check:
[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL.
[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46.
[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) Step is NA
[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN.
[5] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC 4.2 Subsequent Actions NOTES
- 1) Steps in this section are written in general order of importance for most anticipated events; however, they are NOT required to be performed in order, but as required to maintain stable conditions. When a step is entered, all associated sub steps are required to be completed in order except those in Step 4.2[33] (Return to Service).
Steps which are NOT applicable for this scram should be marked N/A.
[1] ANNOUNCE Reactor SCRAM over PA system.
[2] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following:
[2.1] INITIATE ARI by Arming and Depressing BOTH of the following:
- ARI Manual Initiate, 2-HS-68-1 19A
- ARI Manual Initiate, 2-HS-68-1 19B
[2.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed.
[2.3] REPORT ATWS Actions Complete and power level.
[3] DRiVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.
[3.1] DOWNRANGE IRMs as necessary to follow power as it lowers.
[5] MONITOR and CONTROL Reactor Water Level between +2 and +5 1, or as directed by US, as follows:
[5.2] II? required to maintain reactor water level, THEN TRIP reactor feed pumps as necessary to prevent exceeding High Reactor Water Level Trip setpoint.
[5.5] IF Reactor Water Level exceeds +51 inches, ThEN VERIFY TRIPPED the following turbines:
[5.6] IF Reactor Water Level exceeds +55 inches, THEN VERIFY TRIPPED the following turbines:
- Reactor Feed Pump Turbines 36
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC [5] MONITOR and CONTROL Reactor Water Level between +2 and +51, or as directed by US, as follows:
[5.1] CONTROL RFPT/RFPs with any of the following controls on Panel 2-9-5:
REACTOR WATER LEVEL CONTROL (PDS), 2-LIC-46-5, in MANUAL with Column 3 selected OR with Programmed Scram Response inhibited, either in MANUAL (Column 3 selected) or in AUTO (Column 2 selected).
- Individual RFPT 2A(2B)(2C) SPEED CONT RAISE/LOWER switch, 2-HS-46-8A(9A)(1OA), in MANUAL GOVERNOR.
[5.2] IF required to maintain reactor water level, THEN TRIP reactor feed pumps as necessary to prevent exceeding High Reactor Water Level Trip setpoint.
[5.3] IF required to maintain reactor water level, ThEN START RCIC andJor HPCI as required. REFER TO 2-01-7 1 and/or 2-01-73.
[5.4] IF HPCI and/or RCIC are in service and injecting to the vessel, THEN
- MONITOR and CONTROL Reactor Water Level as necessary.
o TRIP HPCI and/or RCIC as necessary to prevent exceeding High Reactor Water Level setpoint.
[5.5] IF Reactor Water Level exceeds +51 inches, THEN VERIFY TRIPPED the following turbines:
- }IPCI
[5.6] IF Reactor Water Level exceeds +55 inches, ThEN VERIFY TRIPPED the following turbines:
- Reactor Feed Pump Turbines SRO When MSIVs are closed direct level control on RCIC or HPCI ATC/BOP Transition level control as directed by US to either HPCI or RCIC 37
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Control Level as directed with RCIC or HPCI Reactor Core Isolation Cooling 5.2 Manual Startup
[1] VERIFY the RCIC System is in Standby Readiness. REFER TO Section 4.0. D
[2] NOTIFY Radiation Protection of the impending action to manually start the RCIC System.
[3] REVIEW all Precautions and Limitations in Section 3.0.
[4] OBTAIN 2-SR-3.6.2.1.l to check Suppression Pool level and temperature every 5 mm.
[5] ESTABLISH communication with the AUO locally at the RCIC turbine.
[6] ENSURE all unnecessary personnel have exited the general area of the RCIC turbine and rupture discs prior to rolling the RCIC turbine.
[7] ANNOUNCE on the plant PA system, Unit Two is starting RCIC, all unnecessary personnel stay clear of the NW RX BLDG. QUAD.
[8] OPEN RCIC LUBE OIL COOLING WTR VLV, 2-FCV-71-25.
[9] START RCIC VACUUM PUMP, 2-HS-7l-31A.
[10] OPEN RCIC PUMP INJECTION VALVE, 2-FCV-71-39.
[11] OPEN RCIC PUMP MTh4 FLOW VALVE, 2-FCV-71-34.
[12] START RCIC Turbine by opening RCIC TURBINE STEAM SUPPLY VLV, 2-FCV-71-8, OBSERVE the following:
- Flow to the RPV stabilizes and is controlled automatically at 620 gpm.
6B, close.
[13] REFER TO Section 6.0 to control and monitor RCIC turbine operation.
38
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Control Level as directed with RCIC or HPCI High Pressure Coolant Injection System 5.2 Manual Startup
[1] VERIFY HPCI is in Standby Readiness. REFER TO Section 4.0.
[2] NOTIFY Radiation Protection of the impending action to manually start the HPCI System. RECORD time Radiation Protection notified in the Narrative Log.
[3] REQUEST Fire Ops to disable the HPCI pump area fire detection system and initiate required impairment.
[4] REVIEW Precautions and Limitations in Section 3.0.
[5] OBTAIN 2-SR-3 .6.2.1.1 to check Suppression Pool level and temperature every 5 minutes.
[6] VERIFY HWC has been set to the desired setpoint (if required) to lower radiation levels in the area.
[7] PLACE SGTS in operation. REFER TO 0-01-65.
[8] ESTABLISH communication with the AUO locally in the }JPCI room.
[9] DEPRESS and HOLD HPCI AUX OIL PUMP, 2-HS-73-47B, START push button (local) for approximately 2 minutes to prime the oil system.
[10] NOTIFY Radiation Protection that an RPHP exists for the impending action to manually start the HPCI turbine. RECORD time Radiation Protection notified in the NOMS Narrative Log.
[11] ENSURE all unnecessary personnel have exited the HPCI room prior to rolling the HPCI turbine.
[12] ANNOUNCE on the plant PA system, Unit Two is starting HPCI, all unnecessary personnel stay clear of the HPCI Room.
[13] CHECK HPCI SYSTEM FLOW/CONTROL, 2-FIC-73-33, is in AUTO and SET at 530 (5,300 gpm). IF required, DEPRESS AUTO operation mode transfer switch and ADJUST setpoint using Setpoint up/down keys.
39
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Control Level as directed with RCIC or HPCI
[14] PLACE HPCI AUXILIARY OIL PUMP handswitch, 2-HS-73-47A, in START.
[15] START HPCI STEAM PACKING EXHAUSTER using 2-HS-73-1OA.
[16] OPEN HPCI PUMP MIN FLOW VALVE, 2-FCV.-73..30.
[17] OPEN HPCI PUMP INJECTION VALVE, 2-FCV-73-44.
[18] OPEN HPCI TURBINE STEAM SUPPLY VLV, 2-FCV-73-16.
[19] VERIFY the following automatic actions occur:
- HPCI TURBINE STOP VALVE, 2-FCV-73-18, opens.
- HPCI TURBINE CONTROL VALVE, 2-FCV-73-19, opens.
- HPCI TURBINE SPEED, 2-SI-73-51, raises.
- HPCI STEAM LINE INBD DRAiN VLV, 2-FCV-73-6A, and }{PCI STEAM LINE OUTBOARD DRAIN VLV, 2-FCV-73-6B, close.
- HPCI HOTWELL PUMP INBD ISOL VLV, 2.FCV-73-17A, and HPCI HOTWELL PUMP OUTBD ISOL VLV, 2-FCV-73-17B, close.
[20] CHECK HPCI System Check Vlv DISC POSITION, 2-ZI-73-45A, indicates open.
[21] VERIFY HPCI PUMP MN FLOW VALVE, 2-FCV-73-30, closes as flow rises above 1255 gpm.
[22] VERIFY HPCI Auxiliary Oil pump stops and the Shaft Driven Oil pump is operating properly, THEN PLACE HPCI AUXILIARY OIL PUMP handswitch, 2-HS-73-47A, in AUTO.
[23] REFER TO Section 6.0 to control and monitor HPCI Turbine operation.
40
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC/BOP [5.7] THROTTLE COND SPE BYPASS FLOW CONTROL VALVE, 2-FCV 1 90A, using 2-HS 1 90A as necessary, to maintain the following indications (Panel 2-9-6):
- SJAE/OG CNDR CNDS FLOW, 2-FI-2-42, between 2 x 106 lbmlhr and 3xl06lbmlhr CNDS PRESS AFTER DEM1N, 2-PI-2-46 greater than 5 psig
- CNDS to RFP PRESS, 2-PI-2-105 greater than 150 psig
[6] IF Programmed Scram Response has initiated (or was previously inhibited), ThEN RESET logic by depressing SCRAM RESPONSE INHIBIT/RESET, 2-HS-46-5 on Panel 2-9-5.
[7] CHECK RFW Control System in SiNGLE ELEMENT (Green backlight illuminated for pushbutton 2-HS-46-6/1 on Panel 2-9-5).
[8] PLACE H2 WATER CHEMISTRY CONTROL switch, 2-HS-4-40/B, in SHUTDOWN (Panel 2-9-5).
BOP [9] At 50 MWe, or as directed by the Unit Supervisor, VERIFY TRIPPED the Main Turbine as follows:
[9.1] DEPRESS the TRIP pushbutton, 2-HS-47-67D on Panel 2-9-7.
[9.2] PERFORM the following as required to VERIFY OPEN GENERATOR PCB224:
[9.2.1] CHECK green light illuminated and red light not illuminated above handswitch GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A.
[9.2.2] IF Generator Breaker is CLOSED OR to RESET white BOP disagreement light at GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A, THEN (Otherwise NA)
A. CONFIRM AIN 2-XA-55-8A window 7, GEN REVERSE PWR FIRST RELAY OPERATION 2-EA-57-136 is received, B. PLACE GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A, in TRIP and release.
C. VERIFY the following at GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A:
- GREEN open target at 2-HS-242-224A
- GREEN open light illuminated
- RED closed light extinguished 41
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS
[9.2.3] IF Generator Breaker remains CLOSED following step 4.2[9.2.2],
BOP THEN (Otherwise NA)
A. PLACE GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A, in BYPASS and release (will feel additional resistance at handswitch as it passes through TRIP to BYPASS position).
B. VERIFY the following at GENERATOR PCB 224 CNTR W/REV BYPASS, 2-HS-242-224A:
- GREEN open target at 2-HS-242-224A
- GREEN open light illuminated
- RED closed light extinguished
[9.3] IMMEDIATELY PLACE VOLTAGE REGULATOR START/STOP SEL, 2-HS-57-24, to STOP and release.
[9.4] ChECK the following at 2-HS-57-24:
- GREEN light illuminated
- RED light extinguished
[10] MONITOR Main Turbine Vibration on TURBiNE GENERATOR VIBRATION, BOP 2-XR-47-15, during coast down.
[11] ADJUST TURBiNE OIL TEMPERATURE CONT, 2-TIC-24-75, setpoint to 85°F.
[12] WHEN turbine speed is less than 900 RPM, THEN START the following:
- TURBINE BEARThTG LIFT OIL PUMPS
- MOTOR SUCTION PUMP
- AC TURNING GEAR OIL PUMP NOTE Due to low power scenario Reactor pressure should be between 700 psig and 800 psig and not rising no actions required to control pressure rise 42
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS NOTE Suppression Pool cooling should be placed in service as soon as practicable following MSRV, HPCI and / or RCIC operation regardless of indicated Suppression Pool temperatures to verify that thermal stratification does not exist.
[13] MONITOR and CONTROL RPV pressure to keep below 1073 psig and stable, or as directed by US.
[13.1] IF RPV pressure is lowering rapidly, THEN CLOSE MSIVs. (Otherwise ATC/BOP N/A)
[13.2] IF MSRVs are cycling and bypass valves are available, THEN MANUALLY OPEN MSRVs on Panel 2-9-3 until RPV pressure is below 965 psig. (Otherwise N/A)
[13.3] IF MSRVs are cycling and bypass valves are NOT available, ThEN MANUALLY OPEN MSRVs on Panel 2-9-3 until RPV pressure is controlled between 800 and 1000 psig. (Otherwise N/A)
[14] IF any PCIS isolation signal is received, THEN VERIFY PCIS isolations using ATC/BOP any of the following: (Otherwise N/A)
. Containment Isolation Status System on Panel 2-9-4
- . PCIS Mimic and individual control switch indications
. .ICS
. 2-01-64 BOP
[15] VERIFY all available drywell cooling coils and blowers in service.
ATC [16] CHECK all control rods fully inserted as indicated on full core display or on ICS NSSS FULL CORE DISPLAY and REQUEST PRINT ROD POSITION LOG on ICS NSSS menu.
ATC [17] IF all rods are NOT inserted to Position 02 or beyond, THEN DIRECT Reactor Engineer to commence determination that reactor will remain subcritical under all conditions without boron. (Otherwise N/A)
Dnver When called for Appendix 2 wait 3 minutes and preference key F12, Appendix-iF wait 5 minutes and preference key Fl 1, If requested to close 85-586 wait 3 minutes and preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 43
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC [18] IF any control rod fails to fully insert and it is required to Re-scram, THEN PERFORM the following, as required:
[18.1] RESET the scram per Steps 4.2[25] thru 4.2[25.12].
[18.2] CHECK WEST and EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM annunciators (2-XA-55-4A-land 4A-29) reset.
[18.3] INITIATE manual scram.
[18.4] REPEAT Step 4.2[18], as necessary, as long as rod motion is observed.
[19] IF any control rod fails to fully insert and it is required to Drive Control Rods, THEN REFER TO 2-01-85.
ATC [20] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 2-9-4.
[21] IF Reactor Recirculation pump(s) have tripped, THEN PERFORM the following before exceeding differential temperature limits for start. (Otherwise N/A)
- VERIFY RWCU in service. REFER TO 2-01-69.
- RESTART affected Reactor Recirculation pumps prior to exceeding differential temperature limits for start. REFER TO 2-01-68.
Driver When called for Appendix 2 wait 3 minutes and preference key F 12, Appendix-i F wait 5 minutes and preference key Fl 1, If requested to close 85-586 wait 3 minutes and preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 44
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC Insert Control Rods JAW 2-01-8 5 8.19 Control Rods Which Fail to FULLY INSERT After Scram NOTE The operator should determine the most effective method to insert rods from the following sections:
- Individually Scram Control Rods (Section 8.19[3]).
- Insert Control Rods using Reactor Manual Control System (Section 8. 19[4]).
- Manual Insertion of Control Rods by Venting the Over Piston Area (Section 8.19[5]).
- Control Rod Insertion using Raised Cooling Water Differential Pressure (Section 8. 19[6])
[4] IF desired to Insert Control Rods Using Reactor Manual Control System, THEN:
[4.11 VERIFY reactor scram is reset. Refer to 2-AOl- 100-1.
[4.2] IF scram CANNOT be reset, THEN CLOSE CHARGiNG WATER SHUTOFF, 2-SHV-085-0586 (RB, EL 565, NE Corner).
[4.3] REVIEW all Precautions and Limitations in Section 3.0.
[4.4] DEMAND, Print Rod Position Log, to edit control rod positions.
[4.5] BYPASS Rod Worth Minimizer. Refer to Section 8.17.
[4.6] Refer to Illustration 4 and DEPRESS the appropriate CR1) Rod Select pushbutton on 2-XS-85-40.
[4.7] CHECK backlit CR1) ROD SELECT pushbutton is brightly illuminated and white indicating light on Full Core Display illuminated.
[4.8] CONTINUOUSLY INSERT control rod to Position 00, by holding CR1)
CONTROL SWITCH, 2-HS-85-48, in ROD iN OR CRD NOTCH OVERRIDE SWITCH, 2-HS-85-47, in EMERG ROD N.
[4.9] IF control rod is difficult to insert, THEN Refer to Section 8.16.
[4.10] REPEAT Steps 8.19[4.6] through 8.19[4.8] for each Control Rod to be inserted.
[4.11] PLACE Rod Worth Minimizer Normal Bypass Switch in NORMAL in accordance with Section 8.18.
[4.12] PLACE the Reactor Mode Switch in SHUTDOWN.
[4.13] VERIFY OPEN CHARGiNG WATER SHUTOFF, 2-SHV-085-0586 (RB, EL 565 NE Corner).
45
NRC Scenario 2 Simulator Event Guide:
Event 10 Component: CRD Pump Trip Driver When ATC has driven 5 or 6 control rods insert Shift F6 for a trip of CR1) Pump 2A ATC Report Trip of CRD Pump SRO Direct entry to 2-AOI-85-3, CRD Failure ATC Restore CRD lAW 2-AOI-85-3 4.1 Immediate Actions
[1] IF operating CRD pump has failed AND standby CRD pump is available, ThEN PERFORM the following at Panel 2-9-5: (Otherwise N/A)
[1.1] PLACE CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, in MAN at minimum setting.
[1.2] START associated standby CR1) Pump using one of the following:
- CR1) PUMP 1B, using 2-HS-85-2A.
- CR1) Pump 2A, using 2-HS-85-1A.
[1.3] IF CR1) Pump lB was started, THEN OPEN CRD PUMP lB DISCH TO U2, using 2-HS-85-8A
[1.4] ADJUST CRD SYSTEM FLOW CONTROL, 2-FIC-85-1 1, to establish the following conditions:
- CR1) SYSTEM FLOW CONTROL, 2-FIC-85-1 1, between 40 and 65 gpm.
[1.5] BALANCE CRD SYSTEM FLOW CONTROL, 2-FIC-85-11, AND PLACE in AUTO or BALANCE.
If called as Unit 1 CR1) Pump lB is not being used If dispatched to CR1) Pump 2A Pump Driver is extremely hot to touch 46
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS ATC [22] IF AT between Rx vessel bottom head temperature and moderator temperature precludes restart of Recirc pump OR forced Recirculation flow CANNOT be established for any reason, ThEN PERFORM the following: (Otherwise N/A)
[22.1] INITIATE plant cooldown to prevent exceeding the pressure limit for Rx vessel bottom head temperature indicated on 2-PNL-9-47, 2-TR-56-4 Pt. 10 and based on Tech Specs Figure 3.4.9-1.
[22.2] MONITOR pressure and temperature to ensure pressure limit vs. Rx vessel bottom head temperature curve from Tech Specs Figure 3.4.9-1 is NOT exceeded using ICS RPV PRESS/TEMP COMPARISON graph (type in RPVCOM or the graph may be selected from Ops Support touch screen).
[23] VERIFY Mode Switch in Shutdown and PERFORM the following:
ATC [23.1] REMOVE Mode Switch key.
[23.2] PLACE Mode Switch key under Shift Manager control.
[23.3] REQUEST Caution Order on Mode Switch to verif SRMs and IRMs operable.
ATC/BOP [24] With Unit Supervisors permission and conditions allow:
[24.1] VERIFY RESET PCIS Logic on Panel 2-9-4.
[24.2] DEPRESS TIP ISOLATION RESET, 2-HS-94-7D/S pushbutton on Panel 2-9-13.
Driver When called for Appendix 2 wait 3 minutes and preference key Fl 2, Appendix- iF wait 5 minutes and preference key Fl 1, If requested to close 85-586 wait 3 minutes and preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 47
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS
[25] WHEN scram signal(s) are reset
[25.1] OBTAIN Unit Supervisors permission to reset the scram.
[25.2]DEPRESS HS-85-36A CRD SCRAM DISCH VOL ISOL TEST VALVE pushbutton on Panel 2-9-5 to isolate the SCRAM DISCH VOL Vent & DR VLVS.
[25.3] IF either 2-XA-5 5-4A- 10, ATWS AUTO INITIATE, or 2-XA-5 5-4A-24, ART MANUAL INITIATE is in alarm, THEN VERIFY RESET ATWS/RPT/ARI as follows: (Otherwise N/A)
[25.3.1] VERIFY 30 seconds have elapsed from ATWS initiation as indicated by READY TO RESET white lights, 2-XI-68-121A and B, illuminated.
[25.3.2] DEPRESS both RESET pushbuttons 2-HS-68-121A and 2-HS-68-121B.
[25.3.3] VERIFY ARI MANUAL INITIATE switches, 2-HS-68-l 19A and 2-HS-68-l 19B, disarmed.
[25.3.4] CHECK ARMED white lights, 2-XI-68-l 19A and B, extinguished.
[25.3.5] ChECK the following annunciators are reset:
- 2-XA-55-4A-l0, ATWS AUTO INITIATE
- 2-XA-55-4A-24, ART MANUAL INITIATE ATC [25.4] PLACE SCRAM DISCH HI LEVEL BYPASS 2-HS-99-5A1S4 switch in BYPASS.
[25.5] RESET Scram by placing SCRAM RESET Switch in RESET FIRST and then to RESET SECOND position.
Driver When called for Appendix 2 wait 3 mmutes and preference key F12, Appendix-iF wait 5 minutes and preference key Fl 1, If requested to close 85-586 preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 48
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS
= ATC [25.6] CHECK SCRAM SOLENOID GROUP A and B LOGIC RESET, lights illuminate.
[25.7] WhEN scram is reset, THEN CHECK all blue lights on Full Core Display are extinguished and control rods have settled back to position 00 by either observation of Full Core Display, or ICS NSSS FULL CORE DISPLAY.
[25.8] IF any control rod(s) fail to settle back to 00, THEN RESET control rod(s) to 00 by applying rod insert signal. (Otherwise N/A)
[25.9] IF control rod cannot be seated in notch position 00, THEN ISOLATE affected HCU per 2-01-8 5, Removing a Hydraulic Control Unit from Service. (Otherwise N/A)
[25.10] RESET Control Rod Drift Lights.
[25.11] IF there are any indications that a scram outlet valve has NOT isolated, THEN ISOLATE the affected HCU per 2-01-85. (Otherwise N/A)
[25.12] WHEN there are NO indications of an open scram outlet valve and any HCU identified in step 4.2[25.11] have been isolated, THEN
[25.12.1] DEPRESS HS-85-36A CRD SCRAM DISCH VOL ISOL TEST VALVE pushbutton on Panel 2-9-5
[25.12.2] CHECK SCRAM DISCH VOL VENT & DR VLVS open by red indicating lights/illuminating on SDV display.
Driver When called for Appendix 2 wait 3 minutes and preference key F 12, Appendix- iF wait 5 minutes and preference key Fl 1, If requested to close 85-586 wait 3 minutes and preference key shift F4 to close and shift F5 to open. Do NOT delete ATWS change ATWS severity to 90 on both sides 49
NRC Scenario 2 Simulator Event Guide:
Event 9 Component: RWCU Leak, failure of Auto Isolation Reports alarm 3D-17, RWCU LEAK DETECTION TEMP HIGH ATC/BOP RE Enters EOI-3 on Secondary Containment Temperature SRO SRO Monitors and Controls Secondary Containment Temps Operate Available ventilation (APPX 8F)
Defeat_Isolation_interlocks_if necessary_(APPX_8E)
Is any Area Temp Above MAX Normal, Yes RWCU Room Isolate all systems that are discharging into the area except systems required to:
- Suppress a Fire Reports High Temperature RWCU Outboard Isolation Valve Area 2-TS-69-29F BOP Reports RWCU Radiation Monitors 90-1 4A and 90-1 3A at MAX Safe SRO Before ANY Area Temp Rises to MAX Safe Continue, Completed When Temps in 2 or more areas are above MAX Safe Then Continue, Does Not Continue in the Temperature leg, NO area at or above MAX Safe Temperature Enters EOI-3 on Secondary Containment Radiation SRO Will Emergency Depressurization Reduce Discharge Into Secondary Containment? YES Before any area radiation rises to Max Safe (table 4) Continue When Radiation Levels in 2 or more Areas are Above MAX Safe Then Continue, ATC Reports RWCU failed to Auto Isolate, RWCU has been manually isolated with the Hand Switches Will Emergency Depressurization Reduce Discharge Into Secondary Containment? NO SRO Does NOT direct Emergency Depressurization, NO systems Discharging into Secondary Containment.
50
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Places Suppression Pool Cooling in service as directed JAW 2-01-74 8.5 Initiation of Loop 1(11) Suppression Pool Cooling
[1] VERIFY RHR Loop 1(11) is in Standby Readiness. REFER TO Section 4.0
[2] REVIEW the precautions and limitations in Section 3.0.
[3] NOTIFY other units of placing Loop 1(11) of RUR in suppression pool cooling, the subsequent start of common equipment (i.e., RHRSW pumps) and associated alarms are to be expected.
[4] NOTIFY Radiation Protection for impending action to initiate Suppression Pool Cooling. RECORD name and time of Radiation Protection representative notified in NOMS narrative log
[5] IF possible, THEN BEFORE placing R}IRSW in service, NOTIFY Chemistry that RHRSW sampling is to be initiated (RHRSW sampling requirements).
[6] VERIFY at least one RHRSW Pump is operating on each EECW Header.
[7] PLACE RHR Pump and Heat Exchanger A(C) in service as follows:
[7.1] START an R}IRSW Pump to supply RHR Heat Exchanger A(C).
[7.2] ESTABLISH RHRSW flow by performing one the following:
[7.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, (RIHIRSW Pump A(C) and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23. OR
[7.2.2] THROTTLE OPEN RHR FIX 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 1NBD INJECT VALVE, 2-FCV- 74-53.
51
NRC Scenario 2 Simulator Event Guide:
Event 8 Major: Fuel Failure and ATWS BOP Places Suppression Pool Coolmg in service as directed JAW 2-01-74
[7.4] IF NO RI-IR PUMP (1A OR 1C) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RER SYS I SUPPR POOL CLG/TEST VALVE, 2-FCV-74-59.
[7.5] VERIFY CLOSED RHR SYS I SUPPR CHBR SPRAY VALVE, 2-FCV-74-5 8.
[7.6] VERIFY CLOSED RHR SYS I DW SPRAY OUTBD VLV, 2-FCV-74-60.
[7.7] VERIFY OPEN RHR SYS I SUPPR CHI3RIPOOL ISOL VLV, 2-FCV-74-5 7.
[7.8] IF desired to operate without the Drywell DP Compressor, THEN:
[7.8.1] OBTAIN permission from the Unit Supervisor
[7.8.2] OPERATE without the Drywell DP Compressor.
REFER TO 2-01-64.
[7.9] START RHR PUMP A(C) using 2-HS-74-5A(16A).
[7.10] THROTTLE RRR 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 in 1 2 Operation Loop Flow 7,000 to <13,000 gpm &
10,000 gpm & Blue Blue light light illuminated illuminated
[7.11] fi? desired to raise Suppression Pool Cooling flow and only one Loop I pump is in service, THEN PLACE the second Loop I RHR Pump and Heat Exchanger in service by REPERFORMING Step 8.5 [7]
for the second pump.
[8] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:
- Amber breaker spring charged light on,
- Closing spring target indicates charged.
52
NRC Scenario 2 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
None Operations/Maintenance for the Shift:
4.5% power. 2-GOI-lOO-1A Section 5.4 Step 64 and 66. Perform IRM!APRM overlap before exceeding 5% power.
RFPT 2B is warmed lAW 2-01-3 Section 5.6.
Engineering is ready in the field need to start SBGT Fan C with alignment to Reactor Zone Ventilation lAW 0-01-65 section 5.2.
Continue to pull rods to 8% lAW RCP.
Units 1 and 3 are 100% Power Unusual Conditions/Problem Areas:
Severe Thunderstorm Warnings are in affect for the entire area for the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
53
NRC Scenario 3 acuity: Browns Ferry NPP Scenario No.: NRC 3 Op-Test No.: 1306 Pvrninarc Operators: SRO:_
ATC:
BOP:
Initial Conditions: 95% power, CRD Pump 3A, EHC Pump 3A and B3 RHRSW Pump are tagged out.
Turnover: Remove DG 3B from Parallel operation in accordance with 3-01-82. Raise power to 100% with flow.
Event Maif. No. Event Type* Event Description No.
R-ATC 1 Commence power increase with flow to 100%
R-SRO N-BOP 2 IOR Remove DG 3B from Parallel operation lAW 3-01-82 TS-SRO C-BOP 3 sw03m D3 EECW Pump Trip TS-SRO I-ATC 4 RD22 CRD Flow Transmitter FT-85-1 1 fails high I-SRO C-ATC 5 Batch File RFPT 3C Trips, Reactor Recirc fails to runback C-SRO C-BOP Loss of 4KV Unit BD 3C, Loss of Circ Water Pump 3C and the 6 Batch File C-SRO discharge valve fails to close automatically 7 Batch File M-ALL Station Blackout 8 RCO4 I RCIC Controller fails in Auto, Manual available 9 sw03k C RHRSW Pump Dl Trips
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor 1
NRC Scenario 3 Critical Tasks Two With a loss of all Off Site Power and NO 480V SD BD energized, energizes 4KV SD BD 3EC from 4KVSDBD3ED.
- 1. Safety Significance:
Provides Power for ECCS Systems
- 2. Cues:
Procedural compliance 480V Shutdown Board 3B Energized, along with 480V RMOV Boards 3B, 3C, 3D, and 3E
- 3. Measured by:
Observation RO crossties to 4 KV SD BD 3ED to 3EC AND Observation 4 KV SD BD 3EC indicates energized
- 4. Feedback:
Power to 4 KV SD BD 3EC Power to 480V SD BD 3B With a Station Blackout on Unit 3 verify two EECW Pumps not using the same EECW strainer are in service supplying Diesel Generators. Cooling water is required to be established within 8 minutes.
I. Safety Significance:
Provides Cooling Water for only operating DG on Unit 3
- 2. Cues:
Procedural compliance RHRSW Pump Al or Cl in service aligned to EECW
- 3. Measured by:
Observation RO aligns RHRSW Pump Cl and verifies Cl RHRSW Pump starts AND Observation RO aligns RHRSW Pump Al and start Al RHRSW Pumps
- 4. Feedback:
No High Temperature alarms on Panel 9-23D for D Diesel Generator Al or Cl RHRSW Pumps operating 2
NRC Scenario 3 Events
- 1. ATC commences power increase 100% using recirculation flow.
- 2. BOP remove DG 3B from Parallel operation lAW 3-01-82 section 8.1 step 17. A low lube oil pressure condition will occur on DO 3B, requiring an Emergency Shutdown of the DO 3B. SRO will evaluate TS 3.8.1 and Enter Condition B.
- 3. D3 EECW pump will trip, with B3 EECW pump tagged, the crew will respond lAW the ARP and manually start the Dl EECW pump. The SRO will refer to Tech Specs and initially determine TS 3.7.2 Condition A. Once the Dl EECW pump has been aligned the SRO will determine TS 3.7.1 Condition A now applies.
- 4. The ATC will respond to the CRD flow element failing high causing 3-FIC-85-1 1 CRD flow control valve to close. The ATC will take action to take manual control of the CRD flow controller and restore CRD system parameters. Depending on how long the crew takes to restore CRD system parameters Tech Spec actions may be required if CRD temperatures increase high enough.
- 5. RFPT 3 C will trip, with the RFPT trip and when level reaches 27 inches a Reactor Recirc runback should have occurred. The runback will fail to occur, RFPT 3A and 3B speed will increase to maintain level but much higher than the prescribed procedure limit of 5050 rpm. The ATC will have to manually initiate the runback or lower power manually.
- 6. The crew will respond to a Loss of 4KV Unit Board 3C. The 3C CCW Pump will trip, the Discharge valve will fail to close and the crew will close the discharge valve, to prevent a loss of vacuum. If the crew fails to close the discharge valve vacuum will quickly degrade requiring a SCRAM.
- 7. On the SCRAM or just before a Loss of Offsite Power will occur, due to equipment failures Unit 3 will enter a Station Blackout. No 4KV Shutdown Boards will be energized. The crew will take action lAW 0-AOI-57-1A and restore 4 KV SD BD B. RCIC and HPCI will be available for Level and Pressure control.
- 8. The RCIC flow controller will fail in automatic, the crew will take manual control to maintain Reactor level.
- 9. RHRSW Pump Dl trips, No EECW pumps operating on EECW. Need to align Al or Cl RHRSW Pumps to EECW within 8 minutes of Dl tripping.
3
NRC Scenario 3 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are inserted Reactor Level is maintained Station Blackout exited when power restored to 4KV SD BD 3EC SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 3 9 Total Malfunctions Inserted: List (4-8) 3 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 Validation Time (minutes) 2 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No) 4
NRC Scenario 3 Scenario Tasks TASK NUMBER K/A RO SRO Shutdown DG 3B RO U-082-NO-05 264000 A4.04 3.7 3.7 SRO S-082-NO-02 Raise Power with Recire Flow RO U-068-NO-17 SRO S-000-NO-138 2.1.23 4.3 4.4 CRD Controller Failure RO U-085-AB-03 201001 A2.07 3.2 3.1 SRO S-085-AB-03 Loss Unit Board 3C/CCW Pump Trip RO U-047-AB-03 295002 AA1.07 3.1 2.9 SRO S-047-AB-03 EECW Pump Trip RO U-067-NO-12 400000 A2.01 3.3 3.4 SRO S-000-AD-27 RFPT Trip / Failure of Core Flow Runback ROU-003-AL-16 259001 A2.01 3.7 3.7 RO U-068-AL-20 SRO S-003-AB-01 Loss of Offsite Power/Station Blackout RO U-57A-AB-01 295003 AA1.03 4.4 4.4 SRO S-57A-AB-01 5
NRC Scenario 3 Procedures Used/Referenced:
Procedure Number Procedure Title 3-01-82 Standby Diesel Generator System 3-ARP-9-23B Panel 9-23 3-XA-55-23B 3-GOT-i 00-12 Power Maneuvering 3-01-68 Reactor Recirculation System Technical Specifications 3-ARP-9-20A Panel 9-20 3-XA-55-20A 0-01-67 Emergency Equipment Cooling Water System 3-A0I-85-3 CRD System Failure 3-01-85 Control Rod Drive System 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3-ARP-9-6A Panel 9-6 3-XA-55-6C 3-A0I-3 -1 Loss of Reactor Feedwater or Reactor Water Level High/Low 3-01-3 Reactor Feedwater System 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-AOI-47-3 Loss of Condenser Vacuum 3-A0I-100-i Reactor Scram 0-A0I-57-1A Loss of Offsite Power (161 and 500 KV)/Station Blackout 3-AOI-30B-i Reactor Building Ventilation Failure 3-A0I-78-i Fuel Pool Cleanup System Failure 3-E0I-i RPV Control 3-EOI Appendix-hA Alternate Pressure Control Systems MSRVs 3-EOI Appendix-5D Injection System Lineup HPCI 3-E0I Appendix-SC Injection System Lineup RCIC 3-EOI Appendix-i 7C RFIR System Operation Suppression Chamber Sprays 3-EOI Appendix-i 7B RHR System Operation Drywell Sprays 3-E0I Appendix-i 7A RHR System Operation Suppression Pool Cooling 3-EOI Appendix-il C Alternate RPV Pressure Control Systems }IPCI Test Mode 3-EOI-2 Primary Containment Control EPIP-3 Alert EPIP-4 Site Area Emergency EPIP- 1 Emergency Classification 6
NRC Scenario 3 Batch File
- tagout CRD 3A, EHC 3A, and B3 EECW ior ypobkrpumpa fail_ccoil ior zlohs85la[l] off br zlohs47la[1j off ior zdihs47la[1j ptl ior ypobkrrhrswpb3 fail_ccoil ior zdihs2388a[2j stop ior zlohs2388a[1] off
- DG 3B Loss of Lube Oil ior z1o3i18213b (e30 0) on ior 0xa5523b[2] (e30 0) alarm_on ior 0xa5523b[4] (e30 0) alarm on
- EECW D3 Trip imf sw03m (el 0) mrfsw05 (e2 0) close
- CRD flow controller fails high imfrd22 (e4 0)100
- RFPT 3C Trip! RR Fails to runback ior zdihs03 176 [lj (e8 0) trip imffw26e (e8 0) 60
- Circ water 3C pump trip and discharge valve fails to close, Loss of Unit BD 3C imfed08c (e12 0) ior ypovfcv2729 fail_power_now ior zlohs2729a[2] on trg e13 nrcccwpumpC trg e13= bat nrcccwpumpC
- Major imfedOl (e18 0) Loss of Off Site Power imfed09a (e18 0) Loss of 4KV SD BD 3EA imf dgo6c DG 3C fails to start trg e20 MODESW imfrcO4 (e20 0)10 RCIC Flow controller Failure imf sw03k (e20 240) trg e23 = bat eecw trg e24 = bat eecw- 1 trg e25 = bat rpsreset trg e26 = bat ca 7
NRC Scenario 3 Trigger Files nrcccwpumpC zdihs2729a[1] .eq. 1 MODESW ZDIHS465(4) .NE. 1 Scenario 3 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 202 Verify CRD Pump 3B in service, EHC manual Pump 3B in service and EHC Pump 3A in Pull-to-Lock Simulator Setup Load Batch bat nrcl3O6-3 Simulator Setup Tag CRD 3A, EHC 3A and EECW Pump manual B3 Simulator Setup Verify file loaded RCP required (95% 100% with flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12 Provide marked up copy of 3-01-82, section 8.1 up to step 17 Provide marked up Illustration 2 8
NRC Scenario 3 Simulator Event Guide:
Event 1 Reactivity: Power increase with Recirc Flow SRO Notifies ODS of power increase.
Directs Power increase using Recirc Flow, per 3-GOl- 100- 12.
[211 WHEN desired to restore Reactor power to 100%, THEN PERFORM the following as directed by Unit Supervisor and recommended by the Reactor Engineer:
- RAISE power using control rods or core flow changes.
REFER TO 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 andJor 3B speed with Recirc Individual Control, THEN PERFORM the following;
- Raise Recirc Pump 3A using, RAISE SLOW (MEDIUM), 3-HS , 15A(15B).
AND/OR
- Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM), 3-HS 16A(16B).
[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A & 3B using the following push buttons as required:
RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-HS-96-32 Driver When directedby NRC, Trigger ito trip EECW PumpD3 9
NRC Scenario 3 Simulator Event Guide:
Event 2 Normal/Component: Remove DG 3B from Parallel operation lAW 3-01-82 SRO Directs BOP to remove DG 3B from parallel operation JAW 3-01-82 section 8.1 BOP 8.1 Parallel with System Operation at Panel 9-23
[17] WHEN Parallel With System operation is no longer desired, THEN UNLOAD the Diesel Generator as follows:
CAUTION When unloading the Diesel Generator, failure to slowly approach the 300 kW/250 kVAR limit may result in a reverse power trip of the Diesel Generator output breaker.
[17.11 USE the associated Diesel Generators governor control switch and voltage regulator control switch to reduce generator load to approximately 300 kW and 250 kVAR.
DG 38 GOVERNOR CONTROL 3-HS-82-3B13A DG 38 VOLT REGULATOR CONT 3-HS-82-3B/2A 38 3-9-23 DG 38 KILOWATTS 3-JI-82-3BfA DG 38 KILOVARS 3-VAR-82-38/A
[17.2] PLACE the associated Diesel Generator breaker control switch in TRIP.
DG3B BKR 1842 3-HS-211-3EA/9A 3-9-23
[17.3] PULL and PLACE the associated Diesel Generator mode selector switch in SINGLE UNIT.
DG 3B Mode Select 3-HS-82-3B/5A 3-9-23
[17.41 RELEASE the Diesel Generator mode selector switch and OBSERVE the SINGLE UNIT light illuminated
[17.5] RECORD the time/date unloaded on Illustration 2.
[17.6] DISPATCH personnel to visually inspect the Diesel Generator output breaker to verify the closing springs are fully charged. Both the amber light and mechanical flag should be checked to indicate a charged spring.
[18] IF operation of the Diesel Generator is no longer required, THEN REFER TO Section 7.0 and SHUT DOWN the Diesel Generator.
DRIVER Prior to the operator shutting down the DG insert trigger 30, to cause a low lube oil condition and report significant oil leak from a damaged oil line. Unable to obtain oil pressure reading, cannot access panel. Delete override on amber light after Emergency Shutdown depressed.
10
NRC Scenario 3 Simulator Event Guide:
Event 2 Normal/Component: Remove DG 3B from Parallel operation lAW 3-01-82 Responds to annunciators DIESEL GEN 3B TROUBLE and DIESEL GEN 3B LUBE OIL BOP ABNORMAL DIESEL GEN 3B TROUBLE A. CHECK panel 9-23 alarm panel for other alarms.
B. DISPATCH personnel to diesel generator room to check alarm.
DIESEL GEN 3B LUBE OIL ABNORMAL A. CHECK panel 9-23 to see if Low Low Lube Oil pressure light is illuminated.
B. DISPATCH personnel to diesel generator room to check:
- 1. Oil pressure on local gauges for lube oil engine, lube oil filter inlet, and turbocharger compressor lube oil.
- 2. Scavenging, main, and piston cooling pumps running.
- 3. Oil level visible on dipstick.
- 4. Oil filters and strainers.
- 5. Any leakage.
C. SHUT DOWN the diesel generator by EMERGENCY SHUTDOWN Pushbutton, if necessary.
7.4 Emergency Shutdown at Panel 9-23
[1] DEPRESS the associated Diesel Generator Emergency SHUTDOWN (STOP) push-button.
DG 3B EMERGENCY SHUTDOWN 3-HS-82-3B/4 3-9-23 BOP [2] VERIFY OPEN DG 3B Output Bkr 1842.
[3] DISPATCH personnel to locally VERIFY Diesel Generator stops.
[6] INITIATE corrective action to return the Diesel Generator to an operable status.
[7] RECORD time/date stopped on Illustration 2.
Driver ReportflG is Stoppmg 11
NRC Scenario 3 Simulator Event Guide:
Event 2 NormallComponent: Remove DG 3B from Parallel operation lAW 3-01-82 SRO Evaluate Tech Spec 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:
- a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System;
- b. Unit 3 diesel generators (DGs) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and
- c. Unit 1 and 2 DG(s) capable of supplying the Unit 1 and 2 4.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems Operating.
APPLICABILITY: MODES 1,2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 8 (continued) B2 Evaluate availability of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> both temporary diesel generators (TDG5).
AND AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter B.3 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 3 DG, Condition B inoperable when the concurrent with redundant required inoperability of feature(s) are inoperable, redundant required feature(s)
B. One required Unit 3 DG 8.1 Verify power availability 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable. from the oftsite 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 12
NRC Scenario 3 Simulator Event Guide:
Event 2 NonnallComponent: Remove DG 3B from Parallel operation lAW 3-01-82 SRO Evaluate Tech Spec 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources Operating ACTIONS CONDITION REQU I RED ACTI ON COMPLETION TIME AND B4i Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Unit 3 DG(s) are not inoperable due to common cause failure OR B.42 Pertorm SR 38.li for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE Unit 3 DG(s).
AND B (continued) 85 Restore Unit 3 DG to 7 days from OPERABLE status. discovery of unavailabillty of TDG(s)
AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition B entry
.6days concurrent with unavailability of TDG(s)
AND 14 days 13
NRC Scenario 3 Simulator Event Guide:
Event 3 Component: EECW pump D3 trip BOP Respond to Motor Trip Out annunciator.
Report Trip of D3 EECW Pump, No EECW Pumps on the South Header C. The EECW System is aligned as follows:
- 1. At least one RHRSW pump, assigned to the EECW System, should be running on each header to maintain the header charged at all times. If no pumps are running on a header and header pressure lowers to S 0 psig, the header shall be declared inoperable and appropriate actions_taken,_as_required_by_Technical_Specifications.
DRIVER If contacted as Unit I operator, you did not secure the D3 EECW Pump SRO Direct operator to place Dl R}IRSW Pump in service to the South Header BOP 8A Operation of RHRSW Pump Dl (for EECW in place of D3)
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 Dl 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 Dl must be aligned to EECW, the pump started, and should remain running. RHRSW Pump Dl does NOT have the same auto start signals as RHRSW Pump D3.
- 3) Prior to aligning Dl RHRSW Pump to EECW, Technical Requirements Manual 3.5.2 must be reviewed to ensure Standby Coolant requirements are met.
- 4) When RHRSW Pump Dl is aligned for EECW, its RHRSW function required by the Safe Shutdown Program (Appendix R) is inoperable. Appendix R program equipment operability requirements of FPR-Volume 1 shall be addressed.
- 5) The RF1RSW pump control switches and amp meters are located on Panel 9-3, Unit 1, 2, and 3.
14
NRC Scenario 3 Simulator Event Guide:
Event 3 Component: EECW pump C3 trip 8.4 Operation of RIIRSW Pump Dl (for EECW in place of D3)
[1] To line up RHRSW Pump Dl for EECW System operation, PERFORM the following:
[1.11 VERIFY EECW System is in prestartup/standby readiness alignment in accordance with Section 4.0.
[1.2] REVIEW all precautions and limitations in Section 3.0.
[1.3] VERIFY RI{RSW Pump Dl is in standby readiness in accordance with 0-01-23.
[1.4] VERIFY RI{RSW Pump Dl upper and lower motor bearing oil level is in the normal operating range.
[1.5] UNLOCK and CLOSE RFIRSW PMP Dl & D2 CROSSTIE, 0-23-563 at RRRSW D Room.
[1.6] OPEN RIIRSW PMP Dl CROSSTIE TO EECW, 0-FCV-67-48 using one of the following:
- RFIRSW PUMP Dl SUPPLY TO EECW, 0-HS-67-48A13 on Unit 3
[1.7] REQUEST a caution order be issued to tag RHRSW Pump Dl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Dl pump should remain running to be operable for EECW.
[2] To start RI{RSW (EECW) Pump Dl, PERFORM the following:
[2.1] START RHRSW Pump Dl using one of the following:
- RHRSW PUMP Dl, 0-HS-23-23A/3 on Unit 3
[2.2] VERIFY RHRSW Pump Dl running current is less than 53 amps using one the following:
- RHRSW PUMP Dl AMPS, 0-EI-23-23/3 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.
tehed to check D3 EECW pump breaker, report breaker tnpped on overload and smells burnt but no visible smokeorflarnes(4kvSD;BDD) 15
NRC Scenario 3 Simulator Event Guide:
Event 3 Component: EECW pump C3 trip 8.4 Operation of RIIRSW Pump Dl (for EECW in place of D3)
[2.5] NOTIFY Chemistry of rurming RE[RSW (EECW) pump(s).
[2.6] VERIFY a caution order has been issued to tag R}IRSW Pump Dl and its associated crosstie valves to inform Operations personnel that it is aligned for EECW system operation and that the Dl pump should remain running to be operable for EECW.
Driver When chemistry contacted, acknowledge report When contacted as Work Control for Caution Order, acknowledge direction and inform will begin working on a Caution Order When dispatched as intake AUO to check Oil Levels and close 0-23-563 valve wait 2 minutes and insert trigger 2 (mrf swOS close), then report oil levels are normal and the 0-23-563 valve is closed When contacted to check breaker charging spring recharged for the Dl EECW pump, wait 2 minutes and inform amber breaker spring charged light is on and closing spring target indicates charged (BKR 4kv SD BD 3EB).
When contacted as Intake AUO for second Oil Level check, report Oil Levels are normal Evaluate Technical Specification 3.7.2 before the Dl EECW Pump is aligned 3.7.2 Emergency Equipment Cooling Water (EECW) System and Ultimate Heat Sink (UHS)
SRO LCO 3.7.2 The EECW System with three pumps and UHS shall be OPERABLE.
APPLICABILITY: MODES 1,2, and 3.
Condition A: One required EECW pumps inoperable. (D3 and B3)
Required Action A. 1: Restore the required EECW pump to OPERABLE status.
Completion Time: 7 days In addition due to DG 3B Inoperable, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from DG 3B inoperability C3 EECW Pump will have to be declared inoperable per Tech Spec 3.8.1 Condition B.3 and at that time Tech Spec 3.7.2 Condition B will be required to be entered.
Condition B: Two or more required EECW pumps inoperable. (D3, B3, and C3)
SRO Required Action B. 1: Be in MODE 3.
Completion Time: 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action B.2: Be in MODE 4.
Completion Time: 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 16
NRC Scenario 3 Simulator Event Guide:
Event 3 Component: EECW pump C3 trip Evaluate Technical Specification 3.7.1 after the Dl EECW Pump is aligned 3.7.1 Residual Heat Removal Service Water (RHRSW) System and Ultimate Heat Sink (UHS)
LCO 3.7.1 NOTE The number of required RHRSW pumps may be reduced by one for each fueled unit that has been in MODE 4 or 5 for LI 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SRO Four RHRSW subsystems and UHS shall be OPERABLE with the number of OPERABLE pumps as listed below:
- 3. 3 units fueled eight OPERABLE R}{RSW pumps.
APPLICABILITY: MODES 1,2, and 3.
Condition A: One required RHRSW pump inoperable Required Action A. 1: Only applicable for the 2 units fueled condition. (NA)
Required Action A.2: Restore required RHRSW pump to OPERABLE status.
Completion Time: 30 days In addition due to DG 3B Inoperable, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from DG 3B inoperability C3 EECW Pump SRO will have to be declared inoperable per Tech Spec 3.8.1 Condition B.3 and at that time Tech Spec_3.7.2 Condition A will be required_to be entered.
17
NRC Scenario 3 Simulator Event Guide:
/
Event 4 Component: CR1) flow element fails high causing 3-FIC-85-1 1 CRD flow control valve to close.
Driver When directed by the NRC trigger 4 (imf rd22 100) to fail the CRD flow element ATC Report Alarm 5A-10 CRD ACCUM CHG WTR HDR PRESS HIGH A. VERIFY pressure high on CRD ACCUM CHG WTR HDR 3-PI-85-13A, B. CHECK 3-FCV-85-1 1A (B) in service.
C. IF in-service controller has failed, THEN REFER TO 3-01-85.
D. IF pressure is still greater than 1510 psig after verifying proper controller operation, THEN THROTTLE PUMP DISCH THROTTLING, 3-THV-085-0527, to maintain between 1475 and 1500 psig.
ATC Report CRD controller has failed in Automatic, takes manual control and restores CR1)
Parameters including Drive Water DP between 250 and 270 psid SRO Directs ATC to take manual control of 3-FIC-85-1 1 and restore CRD parameters Ihe crew may use UI-IJI-I guidance listed below, or 3-U1-.5 Section 32 to take manual control of 3-FIC-85-1 1.
OPDP-1 Conduct of Operations Examiner 3.5 Manual Control of Automatic Systems Note A. If an automatic control or an automatic action is confirmed to have malfunctioned, take prompt actions to place that control in manual or to accomplish the desired function. (e.g.
Establishment of manual level control following automatic FCV failure to control level or manual start of an EDG that failed to auto start.)
8.32 AUTOMATIC/MANUAL Operation of 3-FIC-85-1 1
[1] REVIEW all Precautions and Limitations in Section 3.6.
[2] IF transferring 3-FIC-85-.1 1 from AUTO to MANUAL, THEN:
[2.1] PLACE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1 in BAL.
[2.2] BALANCE CRD SYSTEM FLOW CONTROL, 3-FIC-85-1 1, by turning ATC Manual Control Pot inside Control Selector Wheel until red deviation pointer is in the Green Band.
[2.3] PLACE CR1) SYSTEM FLOW CONTROL, 3-FIC-85-1 1, in MAN.
[2.4] ADJUST CR1) SYSTEM FLOW CONTROL, 3-FIC-85-1 1, manual potentiometer to establish the desired system flow. Refer to Section 5.1 or Section 6.11.
18
NRC Scenario 3 Simulator Event Guide:
Event 5 Instrument: RFPT C Flow instrument fails as is on a trip of RFPT C, Reactor Recirculation System will fail to runback.
Driver When directed by the NRC trigger 8 to trip RFPT C ior zdihs03 1 76[ I] trip and to fail FW flow element imffw26e 60 ATC Reports a Trip of RFPT 3C, and the following annunciators Panel 6C window 15, 23, and 29 RFPT C ABNORMAL, RFPT TRIP CIRCUIT ABNORMAL, RFPT TRIPPED RFPT C ABNORMAL A. CHECK other RFP alarms on Panel 3-9-6 to determine problem area.
B. REFER TO appropriate alarm response procedure.
RFPT TRIP CIRCUIT ABNORMAL A. VERIFY alarm and RFPT trip by checking Panel 3-9-6, RFPT speed, governor valve position and discharge flow.
B. VERIFY reactor power is within the capacity of operating RFPs.
D. IF RFP is tripped, ThEN REFER TO 3-01-3, Section 8.1 or 3-AOI-3-1.
RFPT TRIPPED A. VERIFY reactor power is within the capacity of operating RFPs.
B. CHECK core limits.
C. WHEN RFPT coasts down to zero speed, unless RFPT is rolling on minimum flow, THEN VERIFY turning gear motor starts and engages.
D. REFER TO 3-AOI-3-1 or 3-01-3, Section 8.1.
SRO Directs Entry to 3-AOI 1 Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-AOI-3 -1 Loss Of Reactor Feedwater or Reactor Water Level High/Low AUTOMATIC ACTIONS ATC Recirc Pumps receive run back signal to 75% speed at 27 (normal range) if the discharge flow of a RFP is less than 889,000 lb/hr 19% (rated flow).
[1] VERIFY applicable automatic actions.
[2] IF level OR Feedwater flow is lowering due to loss of Condensate, Condensate Booster, or Feedwater Pump(s), THEN REDUCE Recirc flow as required to avoid scram on low level.
19
NRC Scenario 3
/ Simulator Event Guide:
Event 5 Instrument: RFPT C Flow instrument fails as is on a trip of RFPT C, Reactor Recirculation System will fail to runback.
ATC Reports annunciator REACTOR WATER LEVEL ABNORMAL and level is below 27 inches and slowly lowering RFPTs 3B and 3A speed is increasing to recover level.
REACTOR WATER LEVEL ABNORMAL A. VERIFY Reactor water level hillow using multiple indications including Average Narrow Range Level on 3-XR-3-53 recorder, 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-LI-3-253 on Panel 3-9-5.
B. IF alarm is valid, THEN REFER TO 3-AOI-3-l or 3-01-3.
ATC Reports Power remained stable, pressure remained stable and level is recovering For operating Feed Pumps, monitor and maintain the following parameters within ranges described below.
RFPT Speed: 5050 rpm maximum (3-9-6).
ATC_reports_RFPT_3A and 3B_speed_at greater than_5400_RPM SRO If ATC has NOT initiated a power reduction Directs a Core flow runback ATC Initiates a Mid Power Core Flow Runback, to lower RFPT speeds to less than 5050 RPM When called to report reason for RFPT 3C trip, wa minutes and report laborers in area, D river they believe the tripped the RFPT. Delete override o- zdihsO3l76[l] trip At direction of NRC trigger 12 Loss of 4kv Unit Board 3 C, when called wait 4 minutes and
. report cause of Unit Board trip is actuation of Unit Board Feeder 86-XXX Lockout Driver . .
Relay. When operator closes Circ Water Pump Discharge Valve ensure trigger 13 goes active 20
NRC Scenario 3 Simulator Event Guide:
Event 6 Instrument: Loss of 4KV Unit BD 3C, Loss of Circ Water Pump 3C and the discharge valve fails to close automatically Crew Respond to 8B-14 4KV UNIT BD 3C UNDERVOLTAGE A. VERIFY Unit in stable condition by checking:
- Condensate Pump 3C
- Condensate Booster Pump 3C
- RCWPump3C
- CCW Pump 3C
- CRDPump3A B. IF undervoltage has occurred, THEN
- 1. CLEAR disagreement lights on breakers.
- 2. REDUCE load as necessary to maintain stable operating conditions.
- 3. Condenser discharge may need to be throttled for two CCW pump operation.
REFER TO 3-01-27.
C. ChECK Unit Bd 3C for abnormal conditions: relay targets, smoke, burned paint, etc.
D. REFER TO O-0I-57A to re-energize board.
Report loss of 4KV Unit Board 3C ATC Reports loss of Condensate Pump 3C and Condensate Booster Pump 3C, Power Pressure Level stable BOP Will respond to a loss of Circ Water Pump 3C, report failure of discharge valve to close and close discharge valve ATC/BOP If discharge Valve is NOT closed, report degrading condenser Vacuum SRO Enters 3-AOI-47-3 Loss of Condenser Vacuum IF Operator fails to close CCW Pump 3C Discharge Valve Vacuum will quickly degrade causing a Turbine Trip and Reactor SCRAM. If the Turbine Trips insert trigger 18 loss of DRIVER Offsite Power. If the operator closes the discharge valve insert trigger 18 at the direction of the NRC 21
NRC Scenario 3 Simulator Event Guide:
Event 6 Instrument: Loss of 4KV Unit BD 3C, Loss of Circ Water Pump 3C and the discharge valve fails to close automatically BOP/ATC 3-AOI-47-3 Loss of Condenser Vacuum
[2] IF unable to maintain hotwell pressure below -25 inches Hg as indicated on 3-XR-2-2, with_Reactor power less than 30%,_THEN_TRIP the main turbine.
[4] REDUCE reactor power in an attempt to maintain condenser vacuum.
[5] VERIFY automatic actions.
[6] CHECK CCW pumps for proper operation. If available, START additional CCW PUMPS.
[7] VERIFY CLOSED CONDENSER VAC BREAKERS 1AAND lB. 3-HS-66-1A, Panel 9-8.
[8] CHECK OFF-GAS FLOW TO 6-HOUR HOLDUP VOLUME, 3-FR-66-20, Panel 9-8, between 20 and 180 scfm.
[9] VERIFY OPEN, 3-FCV-66-28, OFF-GAS SYSTEM ISOLATION VALVE.
[10] IF SJAE 3A is in service THEN VERIFY the following:
- 3-PCV-001-0151 and 3-PCV-001-0166 OPEN using STEAM TO SJAE 3A STAGES 1,2, AND 3, 3-ZI-1-1511166 on Panel 3-9-7.
- SJAE 3A 1NTMD CONDENSER DRAIN, 3-ZI-l-150, on Panel 3-9-7, indicates OPEN.
- 3-FCV-066-00l 1 OPEN using SJAE 3A INLET VALVE, 3-HS-66-1 1 on Panel 3-9-8.
- Main Steam supply pressure at SJAE 3A STAGE I & II STEAM PRESS, 3-PT-001-0150, on 3-LPNL-925-0105, is being maintained between 190 and 225 psig.
(TB EL 586 T12-C)
IF Operator fails to close CCW Pump 3C Discharge Valve Vacuum will quickly degrade causing a Turbine Trip and Reactor SCRAM. If the Turbine Trips insert trigger 18 loss of DRIVER Offsite Power. If the operator closes the discharge valve insert trigger 18 at the direction of the NRC 22
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout SRO Directs_Reactor_SCRAM_and_enters_3-AOl- 100-1 4.1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5A/S3A and 3-HS-99-5AJS3B, on Panel 3-9-5.
[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (Otherwise N/A)
[3] Refuel Mode One Rod Permissive Light check
[3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL.
[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46.
[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-Sl, in SHUTDOWN.
[5] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Water Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
- Power level 4.2 Subsequent Actions
[3] DRIVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit.
[3.1] DOWNRANGE IRMs as necessary to follow power as it lowers.
[4] VERIFY SCRAM DISCH VOL VENT & DR VLVS closed by green indicating lights at SDV Display on Panel 3-9-5.
[5] MONITOR and CONTROL Reactor Water Level between +2 and +51 , or as directed by US, as follows:
Recognize Loss of Off Site Power, Report failure of 4KV SD BD 3EA, DG 3EB, DG 3EC, Crew Only DG operating is 3ED SRO Declares a Station Blackout (SBO) is defined as a loss of 161 and 500kV systems and a failure of the two diesel generators which supply normal power to the two 480V Shutdown Boards on a unit. Currently NO 480V Shutdown Boards are energized.
23
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A 4.1 Immediate Actions
[1] VERIFY Diesel Generators have started and tied to respective 4kV Shutdown Boards, THEN DISPATCH personnel to Diesel Generators.
[2] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators.
[3] IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes).
BOP Report only DG 3D is carry 4KV SD BD D, dispatches personnel to determine problems with DGs and SD Boards.
Crew Reports trip of RHRSW Pump Dl. iF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes).
BOP Performs Attachment 9 of O-AOI-57-1A SRO Declares a Station Blackout (SBO) is defined as a loss of 161 and 500kV systems and a failure of the two diesel generators which supply normal power to the two 480V Shutdown Boards on a unit. Currently NO 480V Shutdown Boards are energized.
When sent to investigate 4KV SD BD 3EA has an actuation of overcurrent relay 51. DG 3 A is operating normally. Unknown why DG 3C will not start contacting maintenance. If called Driver as Unit 1 or 2 report all 4 DGs on Unit 1 and 2 have started and are tied to their respective board. IF Unit 1 or is requested to start RHRSW Pumps report UNAVAILABLE to perform.
24
NRC Scenario 3 Simulator Event Guide:
Event 9 Component: Trip of R}IRSW Dl Crew Reports trip of RHRSW Pump Dl. IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes).
BOP Performs Attachment 9 of O-AOI-57-1A 1.0 EECW PUMP ELECTRICAL RESTORATION NOTES
- 1) EECW Pumps may be restored by using one or both of the methods listed in this attachment.
- 2) Actions in this attachment should be continued until two EECW Pumps from different strainers are in service.
[1] SECURE any Diesel Generator prior to 8 minutes of operation without cooling water.
[2] IF no EECW Pumps are in service supplying cooling water to the Diesel Generators, THEN PERFORM EITHER of the following to restore an EECW Pump electrically:
- RE-ENERGIZE 4KV Shutdown Board C per Attachment 8, Energizing a Unit 1/2 4KV Shutdown Board using Another Unit 1/2 DG, THEN VERIFY EECW Pump B3 in service.
- RE-ENERGIZE 4KV Shutdown Board D per Attachment 8, Energizing a Unit 1/2 4KV Shutdown Board using Another Unit 1/2 DG, THEN VERIFY EECW Pump D3 in service..
BOP B3 and D3 EECW Pumps are not available
[3] SECURE any Diesel Generator prior to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of operation with only one EECW Pump supplying cooling water.
[4] IF one EECW Pump is in service supplying cooling water to the Diesel Generators, THEN PERFORM one of the following:
BOP Step is4isNA If Cl is the RI{RSW pump when called wait one mmute and mrf swO6 open, call and report RHRSWPMP Cl & C2 CROSSTIE,O-SHV..23-544 CLOSED.
Driver If Al is the RHRSW Pump when called wait one mmute and mrf swO7 align and report RHRSW PMP Al CROSSTIE TO EECW, O-SHV-067-0088 is OPEN and RHRSW PMP Al_& A2 CROSSTIE, O-SHV-23-.504 is CLOSED At direction of NRC report problems from the trip of EECW Pump D3 or Dl has been iver repaired and EECW Pump D3 or Dl is available Delete imf swO3m or swO3k 25
NRC Scenario 3 Simulator Event Guide:
Event 9 Component: Trip of RETRSW Dl Crew Reports trip of RHRSW Pump Dl. IF two EECW Pumps (not using the same EECW strainer) are not in service supplying Diesel Generators, THEN PERFORM Attachment 9 (Cooling water is required to be established within 8 minutes).
BOP Performs Attachment 9 of O-AOI-57-1A BOP 2.0 EECW PUMP CROSSTIE RESTORATION
[1] CROSSTIE desired in-service Number One EECW Pump:
- OPEN RHRSW PMP Cl CROSSTIE TO EECW, O-FCV-067-0049 in Ul/213 MCR using one of the handswitches (if DG Aux Board A is energized), or locally in RHRSW C Room using the handwheel.
- OPEN RHRSW PMP Dl CROSSTIE TO EECW, O-FCV-067-0048 in U 1/2/3 MCR using one of the handswitches (if DG Aux Board B is energized), or locally in RHRSW D Room using the handwheel.
[2] UNLOCK and CLOSE desired in-service Number One EECW Pump crosstie valve to Number Two EECW Pump:
- RFIRSW PMP Dl & D2 CROSSTIE, O-SHV-23-563 in RNRSW D Room.
BOP 3.0 DIESEL GENERATOR RETURN TO SERVICE NOTE Diesel Generators must be in service before cooling water can be restored.
[1] VERIFY EECW is in service to the Diesel Generator to be placed in service.
[2] PLACE Diesel Generators in service.
[3] VERIFY Diesel Generators have started and tie to respective 4kV Shutdown Boards.
[4] VERIFY two EECW Pumps (not using the same EECW strainer) are in service supplying Diesel Generators.
At direction of NRC report problems from the trip of EECW Pump D3 or Dl has been Driver repaired and EECW Pump D3 or Dl is available Delete imf swO3m or swO3k 26
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power lAW O-AO1-57-1A Crew 4.2 Subsequent Actions
[1] IF ANY EOI entry condition is met, THEN REFER to the appropriate EOI(s).
[2] VERIFY automatic actions and PERFORM any that failed to occur.
NOTES
- 1) If a Unit is in a Station Blackout condition, performance of this instruction will also require implementation of 1 (2)(3 )-A01-3 OB- 1, Reactor Building Ventilation Failure, on the Unit in Station Blackout.
- 2) EECW supply valves to the Control Air Compressors and RBCCW are air operated. If initial air pressure is low, air compressors may trip on high temperature, until cooling water flow is established.
- 3) At US discretion, the O-FCV-67-53 valve can be placed in the open position with hand switch. The valve will automatically come open once EECW pressure is above setpoint. REFER to 01-67 for valve operation.
- 4) The North header supply to Unit 1 RBCCW, the North header supply to Unit 2 RBCCW and the South header supply to Unit 3 RBCCW are normally isolated with a manual valve; therefore no flow will occur when either l-FCV-67-50, 2-FCV-67-50 or 3-FCV-67-5 1 opens.
BOP/ATC 3-A0I-30B- 1, Reactor Building Ventilation Failure
[2] IF this procedure is entered due to Station Blackout (O-A0I 1A) THEN GO TO Step 4.2[14].
NOTE Steps 4.2[141 ,4.2[15] and 4.2[16] may be entered directly from Step 4.2[2] if this procedure is entered due to Station Blackout (O-A0I-57-1A)
[14] IF reactor building pressure CANNOT be maintained more negative than -0.25 inch H20, THEN START Standby Gas Treatment. REFER TO 0-01-65.
[15] IF the unit is in Mode 4 or Mode 5 THEN, Step 15 and 16 NA Unit not in Mode 4 or 5 BOP/ATC Verifies SGT Fans started 27
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A Crew 4.2 Subsequent Actions
[3] REFER to 1(2)(3)-AO1 1, FPC System Failure for a complete Loss of AC POWER, as necessary. NOT NECESSARY
[4] WHEN EECW header pressure is restored above the reset pressure setpoint (psig) for the valves listed below, THEN Common Unit 1 Unit 2 Unit 3 0-FCV-67-53 106 FCV-67-50 - 90 91 92 FCV-67-51 - 107 109 113 RESET EECW supplies to Control Air Compressors and RBCCW, at Unit 1 Panel l-LPNL-925-0032 and Unit 2,3 Panels 2(3)-25-32. Refer to the EECW to the RCW Crossties for Control Air & RBCCW section of 0-01-67.
[5] START Control Air Compressors A, D and G as required and MONITOR system pressure. Refer to 0-AOI-32-1.
[5.1] IF an air compressor trips on high temperature, THEN (Otherwise N/A)
NOTIFY Unit Supervisor for instructions.
[6] REFER to 3-AOI-32-2, Loss of Control Air, as necessary
[7] PLACE RPS MG Sets A and B in service. Refer to 3-01-99.
Crew Calls for Control Air and EECW reset, calls for RPS Reset once 4KV SD BD 3EB energized or can request now but should inform Outside US power is not available for reset of RPS.
Driver Only after EECW Pump is started, 3 minutes after called for EECW trigger 23 bat eecw and trigger 24 bat eecw- 1 and 4 minutes for Control Air trigger 26 bat ca Cannot reset RPS until operator has energized 4KV SD BD 3EB, when 4KV SD BD B is D river energized and 480 V SD BD 3B is energized wait 3minutes and trigger 25 bat rpsreset.
28
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A NOTES
- 1) Station Blackout (SBO) is defmed as a loss of 161 and 500kV systems and a failure of the two diesel generators which supply normal power to the two 480 V Shutdown Boards on a unit. Exiting the SBO can occur through Cross-connect capabilities as long as it does not place the Non-SBO unit in jeopardy. Analysis takes credit for only one unit being in an SBO Event.
- 2) This section is to be performed if at any time during the loss of 161 and 500 kV Offsite power, the required Diesel Generators (for the Units 480 V Shutdown Bds) become inoperable thereby placing the unit in a SBO event. All times start with the recognition of an SBO Event, except for the time since shutdown.
- 3) The purpose of the alternate curves are to replace the normal curves (especially the Crew PSP curve) which would force an Emergency Depressurization (thus losing RCIC level control) before the end of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping period of the SBO analysis. Cooldown must be commenced as soon as possible at near maximum allowable rates to ensure that reactor pressure on the SBO unit is below 235 psig before 155 minutes have elapsed.
- 4) To support one unit in a LOOP/LOCA and two units in a LOOP, 6 RIJR pumps, 2 Core Spray pumps, 6 RHRSW pumps and 2 EECW pumps are required long term (greater than 10 minutes). The units in the LOOP each require 2 RHR pumps and 2 RHRSW pumps in suppression pool cooling for long term cooling requirements until shutdown cooling can be placed in service where only 1 RHR pump and 1 RHRSW pump per non-accident unit is required. DG load management will ensure the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> de-rated DG limit is not exceeded by manually removing non-required loads.
29
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power lAW O-AOI-57-1A
[8] IF the Unit(s) are under a Station Blackout THEN PERFORM the following:
PROCEED TO Step 4.2[9]
[8.1] ESTABLISH Level control with RCIC per the EOIs.
[8.2] ESTABLISH Pressure Control with SRVs per the EOIs.
[8.3] PRIOR to Reactor pressure decreasing below 450 psig, at panel 9-3, PLACE the following switches to TEST/INHIBIT:
[8.3.1] ECCS SYS I HI DW PRESS TEST/INHIBIT, HS-75-59
[8.3.2] ECCS SYS I HI DW PRESS TEST/INHIBIT, HS-75-60
[8.4] As soon as possible but within 60 minutes of the SBO event, INITIATE a cooldown at less than 90°F per hour in accordance with the EOIs until reactor pressure is between 150 and 230 psig.
CREW
[8.5] CONTROL Reactor Pressure between 150 and 230 psig using SRVs.
NOTES
- 1) The following step will allow the use of SBO modified PSP and HCTL curves for EOI usage. The curves are only valid if both 155 minutes have elapsed since the time of reactor shutdown, AND the reactor pressure vessel is maintained below 235 psig.
- 2) If one curve is substituted, then BOTH curves are to be used.
- 3) The purpose of the alternate curves are to replace the normal curves (especially the PSP curve) which would force an Emergency Depressurization (thus losing RCIC level control) before the end of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping period of the SBO analysis.
[8.6] IF the EOI PSP Curve 6, or HCTL Curve 3, is close to being exceeded, AN])
the following conditions apply, (Otherwise N/A), Step is NA Directs Level control in RCIC and pressure control with SRVs, commences a cooldown SRO with SRVs. Direct ECCS Inhibit switches placed in Test/Inhibit Initiate RCIC for Level Control and Commence Cooldown with SRVs as directed. Inhibit BOP/ATC ECCS High Drywell Pressure signal.
When RCIC started reports failure of Flow controller in Auto, takes manual control and BOP/ATC controls level with RCIC.
30
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power lAW 0-AOI-57-1A
[9] START the Diesel Driven Fire Pump. Refer to 0-01-26.
[11] IF containment isolation is required, THEN VERIFY the following containment isolation valves closed UNLESS they are required to be open by EOIs (RG 1.155):
FCV-1-56 MN STM LINE OUTBD DRAIN ISOL FCV-69-2 RWCU OUTBD SUCT ISOLATION FCV-71-3 RCIC OUTBD SUCT ISOLATION FCV-71-18 RCIC SUPPR POOL OUTBD SUCT VALVE FCV-73-3 HPCI STEAM LINE OUTBD ISOL VALVE Crew FCV-73-26 HPCI SUPPR POOL INBD SUCTION VLV FCV-73-30 HPCI MAIN PUMP MiNIMUM FLOW VLV FCV-74-47 RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV NOTES
- 1) The UNIT SUPERVISOR should prioritize board energization to ensure common HVAC equipment powered from 480V boards is returned to service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as directed by Attachment 5.
Request start of Diesel Fire Pump. Verifies Containment isolation status, RCIC valves will BOP/ATC be open When requested to start Diesel Fire Pump wait one mmute and start diesel fire pump irf D river f04d start 31
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power JAW O-AOI-57-1A
[12] VERIFY the following boards are energized. IF NOT, THEN REFER to Attachment_1_to_restore_affected_busses_while_continuing_with_this_instruction.
Unit 1 Unit 2 Unit 3 4KV Shutdown Boards A, C B, D 3EA, 3EB, 3EC, 3ED 480V Shutdown Boards 1A, lB 2A, 2B 3A, 3B 48OVDSLAuxBoards A B 3EA,3EB 480V RMOV Boards 1A, lB 2A, 2B 3A, 3B 480V Control Bay Vent A B Boards 480V HVAC Board B
[13] VERIFY the following LPCI MG Sets in service to their respective Reactor MOV boards.
- LPCI MG Sets 3D and 3E Crew Proceed to Attachment 1 to determine how to energize 4KV SD BD 3EB NOTE To ensure adequate cooling, required 480V loads should be re-energized within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the event.
[19] VERIFY the following 4kV Shutdown Boards AUTO/LOCKOUT RESET switches in MANUAL:
- U-3 4kV Shutdown Boards 3-43-21 1-3EA, 3-43-21 1-3EB, 3-43-21 1-3EC, 3-43-21 1-3ED.
[20] MAINTAIN diesel generator loading within the limits of Attachment 6.
Crew NOTES
- 1) The following methods for reactor depressurization are listed in order of preference, but plant conditions may warrant other methods or a different order of preference.
- Cycle SRVs.
[27] COMMENCE Reactor depressurization as soon as conditions permit. Cooldown is to be limited to 90°F/hr or less unless otherwise specified by the EOIs.
SRO Direct HPCI place in pressure control if possible Place HPCI in Pressure control, if possible. If level is low can recover level with RPCI and ATC/BOP once level is high enough can reset start signal and place HCPI in pressure control mode until Drywell pressure exceeds 2.45 psig.
32
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Responds to a Loss of Offsite Power JAW 0-AOI-57-1A 1.0 RESTORATION OF ELECTRICAL BUSSES
[1] DETERMINE restoration sequence of electrical busses and RE-ENERGIZE busses as necessary. Refer to Table 1.
Table 1 Conditions Perform ENERGIZING 4Kv Shutdown Board 3EA using 4Kv Shutdown Board 3EB OR ENERGIZING 4Kv Shutdown Board 3EC using 4Kv_Shutdown Board 3ED. Attachment_10 Driver If called for Temporary Diesels, They are not available Determines that Attachment 10 is the required attachment to restore power to a Shutdown C rew Board. Energize 4KV SD BD 3EC from DG 3D Attachment 10 ENERGIZING 4KV SD BD 3EA or 3EC during Station Blackout NOTES
- 1) This attachment is used to energize 4kV Shutdown Board 3EA using Shutdown Board 3EB or to energize 4Kv Shutdown Board 3EC using 4Kv Shutdown Board 3ED.
- 2) The use of 3EB-3EA and 3ED -3EC cross-ties are required to mitigate Unit 3 Station Blackout (SBO) scenarios (i.e. only 3B or 3D Diesel Generators available).
Crew 1.0 BOTH 4KV SHUTDOWN BOARDS 3EA AND 3EC ARE DE-ENERGIZED
[1] IF both 4Kv Shutdown Boards 3EA and 3EC are de-energized, THEN PERFORM the following:
[1.1] DETERMINE which 4Kv Shutdown Board is energized (3EB or 3ED).
[1.2] RE-ENERGIZE the desired shutdown board (3EA or 3EC) using the available board (3EB or 3ED) using Step 1 .0[2] or STEP 1 .0[3] as applicable.
SRO Emergency Plan classification 5.1-Al or 5.1-S 5.1-Al Loss of voltage to ANY THREE unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND Only ONE source of power available to the remaining board.
OPERATING CONDITION: Mode 1 or 2 or 3 5.1-S Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes.
OPERATING CONDITION: Mode 1 or 2 or 3 33
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Attachment 10 ENERGIZ1I4G 4KV SD BD 3EA or 3EC during Station Blackout
[3] IF desired to re-energize 4KV Shutdown Board 3EC using cross-tie from 4Kv Shutdown Board 3ED, ThEN PERFORM the following:
[3.1] VERIFY Diesel Generator 3D is supplying 4Kv Shutdown Board 3ED.
[3.2] VERIFY 4Kv Shutdown Board 3EC is de-energized.
[3.3] VERIFY OPEN 4Kv Bus Tie Board breaker 1632 (3-IL-210-l/6B).
[3.4] VERIFY 4KV SD BD 3EC AUTO/LOCKOUT RESET switch, 3-43 -21 1-3EC, is tripped to MANUAL.
[3.5] VERIFY 4KV SD BD 3ED AUTO/LOCKOUT RESET switch, 3-43 -21 1-3ED, is tripped to MANUAL.
[3.6] PLACE in ON synchronizing switch 4KV SD BD 3ED BKR 1628 SYNC, 3-25-2 1 1-3ED/1A.
[3.7] CLOSE 4KV SD BD 3ED ALT FDR BKR 1628, 3-HS-21 1-3ED/1A.
[3.8] PLACE in OFF synchronizing switch 4KV SD BD 3ED BKR 1628 SYNC, 3-25-2 1 1-3ED/1A.
[3.9] PLACE in ON synchronizing switch 4KV SD BD 3EC BKR 1626 SYNC, 3-25-2 1 1-3EC/3A.
[3.10] CLOSE 4KV SD BD 3EC ALT FDRBKR 1626, 3-HS-211-3EC/3A.
[3.11] PLACE in OFF synchronizing switch 4KV SD BD 3EC BKR 1626 SYNC, 3-25-2 1 1-3EC/3A.
[3.12] VERIFY 4Kv Shutdown Board 3EC is energized.
When 4KV SD BD 3EC is energized verifies 480V SD BD 3B energized and the following C rew 480V RMOV BDs, 3C, 3D and 3E energize 34
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout SRO Enters EOI- 1 on Reactor Level 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 or has been required THEN exit RC/P and enter C2 Emergency Depressurization? NO -
IF RPV water level cannot be determined? NO -
Is any MSRV Cycling? YES, but MSIVs closed 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 with SRVs lAW APPX hA, and a controlled cooldown lAW Station Blackout. In addition may direct pressure control with HPCI (APPX 11C)
SRO EOI- 1 (Reactor Level)
Monitor and Control Reactor Water Level.
Directs_Verification_of PCIS_isolations.
ATC/BOP Verifies PCIS isolations.
SRO Directs ATC to Restores and Maintains RPV Water Level between (+) 2 to (+) 51 inches with the following injection source. (RCIC, App 5C) may also direct at this time F1PCI (App 51)),
when the LOCA starts HPCI will be directed to maintain level (+) 2 to (+) 51 inches ATC Initiates RCIC lAW App 5C 35
NRC Scenario 3
/
Simulator Event Guide:
Event 8 Instrument: RCIC Controller fails in Auto, Manual available ATC/BOP Maintain Directed Level Band with RCIC, Appendix 5C.
- 1. VERIFY RESET and OPEN 3-FCV-71-9, RCIC TURB TRIP/THROT VALVE RESET.
- 2. 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 M1N 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-7 1 -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.
Reports level controller malfunction and takes manual control to inject with RCIC 36
NRC Scenario 3 Simulator Event Guide:
Event 8 Instrument: RCIC Controller fails in Auto, Manual available ATC/BOP Maintain Directed Level Band with HPCI, Appendix SD.
CAUTION
- Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.
- Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.
- 4. VERIFY 3-IL-73-18B, HPCI TURBINE TRIP RX LVL HIGH amber light extinguished.
- 5. VERIFY at least one SGTS train in operation.
- 6. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.
- 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 IIPCI 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 M1N 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.
37
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout ATC/BOP Commence pressure control with Appendix hA, Alternate RPV Pressure Control Systems MSRVs IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAD TO DRYWELL CONTROL AIR, CONCURRENTLY with this procedure.
- 2. IF Suppression Pool level is at or below 5.5 ii, THEN CLOSE MSRVs and CONTROL RPV pressure using other options.
- a. 1 3-PCV-l-179 MN STM LINE A RELIEF VALVE.
- b. 2 3-PCV-1-180 MN STM LINED RELIEF VALVE.
- c. 3 3-PCV-i-4 MN STM LINE A RELIEF VALVE.
- d. 4 3-PCV-l-3 1 MN STM LINE C RELIEF VALVE.
- e. 5 3-PCV-i-23 MN STM LINE B RELIEF VALVE.
- f. 6 3-PCV-i-42 MN STM LiNE D RELIEF VALVE.
- g. 7 3-PCV-i-30 MN STM LINE C RELIEF VALVE.
- h. 8 3-PCV-i-19 MN STM LINE B RELIEF VALVE.
- i. 9 3-PCV-i-S MN STM LiNE A RELIEF VALVE.
- j. 10 3-PCV-i-41 MN STM LINE D RELIEF VALVE.
- k. ii 3-PCV-i-22 MN STM LINE B RELIEF VALVE.
- 1. 12 3-PCV-1-18 MN STM LiNE B RELIEF VALVE.
- m. 13 3-PCV-i-34 MN STM LINE C RELIEF VALVE.
38
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout ATC/BOP Commence pressure control with Appendix 11C, Alternate RPV Pressure Control Systems FIPCI TEST MODE CAUTION
- Operating HPCI Turbine below 2400 rpm may result in unstable system operation and equipment damage.
- Operating HPCI Turbine with suction temperatures above 140°F may result in equipment damage.
- c. CLOSE 3-FCV-73-44, HPCI PUMP iNJECTION VALVE.
- d. CONTINUE in this procedure at Step 6.
- 5. IF HPCI is in standby readiness, THEN START I{PCI as follows:
- a. VERIFY at least one SGTS Train in operation.
- b. VERIFY 3-FIC-73-33, HPCI SYSTEM FLOW/CONTROL, controller in AUTO and set for 5300 gpm.
c.
PLACE 3-HS-73-47A, HPCI AUXILIARY OIL PUMP handswitch, in START.
- d. PLACE 3-HS-73-1OA, HPCI STEAM PACKING EXHAUSTER, in START.
- e. OPEN the following valves:
- 3-FCV-73-30, HPCI PUMP M1N FLOW VALVE.
- g. VERIFY HPCI Auxiliary Oil Pump starts and turbine accelerates above 2400 rpm.
39
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout ATC/BOP Commence pressure control with Appendix 11C, Alternate RPV Pressure Control Systems HPCI TEST MODE
- 6. VERIFY proper HPCI minimum flow valve operation as follows:
b IF HPCI flow is below 600 gpm, THEN VERIFY OPEN 3-FCV-73-30, HPCI PUMP MTh4 FLOW VALVE.
- 7. ThROTTLE 3-FCV-73-35, HPCI PUMP CST TEST VLV, to control HPCI pump discharge pressure at or below 1100 psig.
- a. OPEN 3-FCV-73-44, HPCI PUMP INJECTION VALVE.
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout Crew Enter EOI-2 on Suppression Pool Temperature SRO Enters EOI-2 on Suppression Pool Temperature DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 1 60F, No Operate all available drywell cooling Before Drywell Temperature rises to 200°F enter EOI- 1 and Scram Reactor, Completed PC/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, YES PC/Il Verify H202 analyzer in service (APP 19)
When H2 is detected in PC (2.4% on control room indicators continue, does not continue SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, directs RHR Pump 3D in Pool Cooling SP/L MONITOR and CONTROL suppr p1 lvi between -l in. and -6 in. (APPX 18)
Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvi be maintained below -1 in., YES 41
NRC Scenario 3 Simulator Event Guide:
Event 7 Major: Station Blackout SRO Emergency Plan classification 5.1-Al or 5.1-S 5.1-Al Loss of voltage to ANY THREE unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes AND Only ONE source of power available to the remaining board.
OPERATING CONDITION: Mode 1 or 2 or 3 5.1-S Loss of voltage to ALL unit specific 4KV shutdown boards from Table 5.1 for greater than 15 minutes.
OPERATING CONDITION: Mode 1 or 2 or 3 42
NRC Scenario 3 SHIFT TURNOVER SHEET Equipment Out of ServicelLCOs:
CRD Pump 3A, EHC Pump 3A, and EECW Pump B3 Operations/Maintenance for the Shift:
DG 3B is operating in parallel with Offsite power for the last hour, parallel with system operation is no longer desired.
Evolutions for the shift are to restore power to 100%, JAW 3-GOI-100-12, Power Maneuvering step 21 and to remove DG 3B from parallel operation and shutdown DG 3B lAW 3-01-82 Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas:
The following Control Rods are identified as SLOW: 30-19, 34-23, 14-51, 02-19, 46-51, and 06-43.
43
NRC Scenario 4 3clllty Browns Ferry NPP Scenario No.: NRC 4 Op-Test No.:
Examiners: Operators: SRO:_
ATC:
BOP:
Initial Conditions: 80% power, RFPT 3B and A3 RHRSW Pumps are tagged out.
Turnover: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Raise power to 85% with flow and hold for RFPT 3B repairs.
Event Maif. No. Event Type* Event Description No.
N-BOP Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 1 30B, Refuel damper 64-9 fails in mid position when Refuel Fans TS RO are in Off and is Open when Refuel Fans operating 2 Commence power increase with flow to 85%
C-ATC 3 edlOb C-BOP Loss of 480V SD BD 3B TS-SRO 4 Batch File Stator Water Cooling Pump trip 5 fw30a RFPT 3A Governor fails low 6 imftclOb EHC Pressure Transducer failure 7 Batch File M-ALL ATWS 8 Batch File M-ALL LOCA Loss of RPV Water Level 9 hpO7 C Loss of HPCI 120 VAC Power Supply
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)aj or
\
1
NRC Scenario 4 Critical Tasks Five 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 ADS LOGIC BUS A/B INHIBITED annunciator status.
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 JAW EOI-1 and C-5, US determines that SLC is required (indicated by verbal direction or EOI placekeeping action) before exceeding 110 degrees in the Suppression Pool.
AND RO places SLC A / B Pump control switch in ON, when directed by US.
AND Control Rod insertion commenced in accordance EOI Appendixes.
- 4. Feedback:
Reactor Power trend.
Control Rod indications.
SLC tank level.
2
NRC Scenario 4 Critical Tasks Five During an ATWS, when conditions with Emergency Depressurization required, Terminate and Prevent RPV injection (except for CRD, SLC and RCIC) from ECCS and Feedwater until reactor pressure is below the MARFP as directed by US.
- 1. Safety Significance:
Prevention of fuel damage due to uncontrolled feeding.
- 2. Cues:
Procedural compliance.
- 3. Measured by:
Observation No ECCS injection prior to being less than the MARFP.
AND Observation Feedwater terminated and prevented until less than the MARFP.
- 4. Feedback:
Reactor power trend, power spikes, reactor short period alarms.
Injection system flow rates into RPV.
With RPV pressure <MARFP, slowly increase and control injection into RPV to restore and maintain RPV level above TAF as directed by US.
- 1. Safety Significance:
Maintaining adequate core cooling and preclude possibility of large power excursions.
- 2. Cues:
Procedural compliance.
RPV pressure indication.
- 3. Measured by:
Observation Injection not commenced until less than MARFP, and injection controlled such that power spikes are minimized, level restored and maintained greater than TAF.
- 4. Feedback:
RPV level trend.
RPV pressure trend.
Injection system flow rate into RPV.
3
NRC Scenario 4 Critical Tasks Five 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 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 4
NRC Scenario 4 Events
- 1. BOP operator will alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B. Refuel damper 64-9 fails in mid position when Refuel Fans are in off and is open when Refuel Fans operating. Tech Spec 3.6.4.2 Condition A, required action A.1 and A.2.
- 2. ATC commences power increase 85% using recirculation flow.
- 3. The Crew will respond to a loss of 480V SD BD 3B, this will cause a loss of RPS B, loss of 480V RMOV BDs 3B and 3C. The Inboard MSIV A will have inadvertently closed. The crew will need to lower power to meet the main steam line flow guidance JAW 3-AOI-3-1. The crew will need to restore power to 480V SD BD 3B, reset RPS, reset PCIS and restore systems. The SRO will also have to enter the following AOIs; 3-AOI-1-3, 3-AOI-70-1, and 3-AOI-99-1. SRO will refer to the TRM and determine Technical Surveillance Requirement 3.4.1.1 to monitor Reactor Coolant Conductivity continuously cannot be met and samples must be drawn every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. SRO will refer to Tech Spec 3.6.1.3 for failed closed MSIV and enter condition A. SRO will refer to Tech Spec 3.4.5 and determine Condition B is required for inoperable containment atmospheric monitoring equipment.
- 4. The running Stator Water Cooling Pump will trip and the standby pump will fail to AUTO start.
The BOP operator will be required to start the standby Stator Water Cooling pump to restore system flow and prevent an automatic Turbine Trip/Reactor scram.
- 5. RFPT 3A flow controller will slowly fail low, RFPT 3A speed will continue to decrease 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 3A speed in manual. SRO should direct entry into 3-AOI-3-1.
- 6. An ATWS will occur on the scram and the power supply to HPCI will fail, leaving RCIC as the only source of high pressure makeup besides SLC and CRD. The crew will insert control rods manually, and maintain reactor level.
- 7. With RCIC, CRD and SLC as the only source of high pressure makeup as the LOCA degrades RPV Level will continue to lower. The SRO will determine Emergency Depressurization is required to restore RPV Level. The crew will ED and restore RPV Level with available systems.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Control Rods are being inserted Emergency Depressurization complete Reactor Level is restored 5
NRC Scenario 4 SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 4 9 Total Malfunctions Inserted: List (4-8) 2 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 2 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 5 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No) 6
NRC Scenario 4 Scenario Tasks TASK NUMBER KJA BQ Q Alternate Reactor and Refuel Zone Fans RU U-30A-NO-02 288000A4.0l 3.1 2.9 Raise Power with Recirc Flow RU U-000-NO-06 202002A4.07 3.3 3.3 SRO S-000-AD-3 I Loss of 480V SD BD 3B RU U-57B-AL-06 262001A2.04 3.8 4.2 SRO S-57B-AL-09 Reactor Feed Pump Turbine Governor Failure RU U-003-AL-09 259002A4.0l 3.8 3.6 SRO S-003-AB-0l Stator Water Cooling Pump Trip RU U-35A-AL-02 245000A4.03 2.7 2.8 SRO S-070-AB-01 EHC Pressure Transducer Failure RO U-047-AB-02 241000A2.03 4.1 4.2 SRU S-047-AB-02 LOCA/Low Level ED RU U-003-AL-24 29503 1EA2.04 4.6 4.8 RU U-000-EM-01 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-01 ATWS RU U-000-EM-03 2950 15AA2.01 4.1 4.3 RU U-000-EM-22 RU U-000-EM-28 SRO S-000-EM-03 SRO S-000-EM- 18 7
NRC Scenario 4 Procedures Used/Referenced:
Procedure Number Procedure Title 3-0I-30A Refuel Zone Ventilation System 3-01-3 OB Reactor Zone Ventilation System 3 -GOl- 100-12 Power Maneuvering 3-01-68 Reactor Recirculation System 3-ARP-9-8B Panel 9-8 3-XA-55-8B 3-ARP-9-8C Panel 9-8 3-XA-55-8C 3-ARP-9-4C Panel 9-4 3-XA-55-4C 3-A0I-99-1 Loss of Power to One RPS Bus 3 -AOl- 1-3 Main Steam Isolation Valve Closure at Power 3-AOI 1 Loss of Reactor Building Closed Cooling Water 3-01-99 Reactor Protection System 3 -A0I-64-2D Group 6 Ventilation System Isolation Technical Specifications Technical Requirements Manual 3-ARP-9-7A Panel 9-7 3-XA-55-7A 3-ARP-9-8A Panel 9-8 3-XA-55-8A 3-ARP-9-5A Panel 9-5 3-XA-55-5A 3 -AOI-3 -1 Loss Of Reactor Feedwater or Reactor Water Level High/Low 3-ARP-9-7B Panel 9-7 3-XA-55-7B 3-A0I-47-2 Turbine EHC Control System Malfunctions 3-A0I-l00-i Reactor Scram 3-E0I-i RPV Control 3-C-5 Level/Power Control 3-E0I Appendix-iD Insert Control Rods Using Reactor Manual Control System 3-E0I Appendix-iF Manual Scram 3-EOI Appendix-2 Defeating ARI Logic Trips 3-E0I Appendix-i iA Alternate Pressure Control Systems MSRVs 3-E0I Appendix-SC Injection System Lineup RCIC 3-E0I-2 Primary Containment Control 3-E0I Appendix-i 7A RRR System Operation Suppression Pool Cooling 3-E01 Appendix-I 7C RHR System Operation Suppression Chamber Sprays 3 -E0I Appendix-i 7B RHR System Operation Drywell Sprays 3-AOI-85-3 CRD System Failure 3-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 3 -EOI Appendix-4 Prevention of Injection 3-C-2 Emergency RPV Depressurization 3-EOI Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode 3-EOI Appendix-6C Injection Subsystems Lineup RHR System II LPCI Mode EPIP- 1 Emergency Classification 8
NRC Scenario 4 Console Operator Instructions A. Scenario File Summary Batch File
- RFPT 3B and EECW Pump A3 tagout ior ypobkrrhrswpa3 fail_ccoil ior zlohs2385a[1] off ior ypomtreopb3 fail_control_power ior ypobkrmopbl fail_ccoil ior ypobkrmopb2 fail ocoil ior ypomtrtgmb fail_control_power ior ypovfcv03 12 fail_power_now ior ypovfcvo295 fail_power_now ior ypovfcv0l 129 fail_power_now ior ypovfcvol 133 fail_power_now
- Loss of 480V SD BD B imfedlOb (el 0) iorzdihs0ll4a[1] (el 0) close mrfrp09 (e3 0) reset trg e3 = bat restorerpsb ior zdixs577l [1] (e4 0) normal
- B stator water pump trip irf egO2 (e5 0) off ior ypobkrscwpa (e5 0) fail_ecoil ior zdihs3535a[2j (e5 0) stop ior zlohs3535a[1] (e5 0) off
- rft governor drift imffw3Oa (elO 0) 0 2500 72 trg eli nrcr1ptA trgeil dmffw30a
- B EHC Pressure transducer failure ior zdihsOi 16[1] (e14 0) select ior zdihs47204[i] (e14 0) null ior zlohs0l 16[1] off ior z1ohs47204[1] on imftci0b (e14 0) 822200 79 9
NRC Scenario 4
- major trg el 8 nrcmodesw bat nrcstickquad Imfth22 (e18 1:00) LOCA Imfth2l (e18 10:00) 1 15:00 LOCA imfhp07 (e18 0) HPCI Fails trg e23 = bat app0lf trg e24 bat app02 trg e25 = bat nrcstickquad-1 mrfrd06 (e26 0) close mrfrd06 (e27 0) open trg e28 bat nrcatws95 Trigger Files nrcrfptA zdihs468a[4] .ne. 1 nrcmsivD zdihs0l 52a[1 j.eq. 1 Scenario 4 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 203 Simulator Setup Load Batch bat nrcl3O6-4 Simulator Setup manual Tag RFPT 3B and EECW Pump A3 Simulator Setup Verif file loaded. Log in to EHC System to ensure when operators try to access they are able to.
RCP required (80% 85% with flow) and RCP for Urgent Load Reduction Provide marked up copy of 3-GOI-100-12 10
NRC Scenario 4 Simulator Event Guide:
Event 1 Component: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B SRO Direct Refuel and Reactor Zone Fans alternated BOP 6.1 Alternating Refueling Zone Supply and Exhaust Fans 3-OI-30A
[1] NOTIFY Unit 1 and Unit 2 Operators that the Refuel Zone fans are being alternated.
[2] VERIFY the Refueling Zone supply and exhaust fans are operating. REFER TO Section 5.1.
[3] REVIEW precautions and limitations in Section 3.0.
NOTES
- 1) The preferred method to start the alternate Refueling Zone supply and exhaust fans is to use the common control Switch, 3-HS-64-3A, on Panel 3-9-25.
- 2) Refueling Zone supply and exhaust dampers, 3-FCO-064-0005,0006,0009, and 0010 will open or close automatically as necessary when fans are stopped and started.
- 3) Refueling Zone supply and exhaust fans are alternated every six weeks.
[4] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in OFF.
[5] CHECK that the two red lights A(B) extinguish and the two green lights A(B) illuminate above REFUEL ZONE FANS_AND DAMPERS_switch,_3-HS-64-3A.
NOTE BOP If any damper does not meet the requirements of step 6.1 [6] IMMEDIATELY notif the Unit supervisor to evaluate SC1V damper operability (refer to IRM appendix A). If any listed damper indicates not full closed, it should be considered inoperable for its SCIV function, and the required actions of Tech Spec LCO 3.6.4.2 entered for all units.
[6] CHECK the red (open) damper position indication lights extinguish and the green (closed) lights illuminate above the following control switches:
- REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
- REFUEL ZONE SPLY 1NBD ISOL DMPR, 3-HS-64-6
[7] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B) to start alternate fans.
BOP Report Failure of REFUEL ZONE EXH OUTBD ISOL DMPR, 3-HS-64-9 11
NRC Scenario 4 Simulator Event Guide:
Event 1 Component: Alternate Refuel and Reactor Zone Fans lAW 3-OI-30A and 30B BOP [8] CHECK that the two green lights A(B) extinguish and the two red lights A(B) illuminate above REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A.
[9] CHECK the red (open) damper position indication lights illuminate and green (closed) lights extinguish above the following control switches:
- REFUEL ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-5
- REFUEL ZONE SPLY INBD ISOL DMPR, 3-HS-64-6
- REFUEL ZONE EXH JNBD ISOL DMPR, 3-HS-64-10 NOTE A five minute time delay should be observed following Refuel Zone Supply and Exhaust Fan SLOW Start. The time delay allows the discharge dampers to fully open after SLOW start.
[10] IF Refueling Zone Supply and Exhaust Fan FAST speed operation is necessary, THEN: PERFORM the following:
[10.1] PLACE REFUEL ZONE FANS AND DAMPERS switch, 3-HS-64-3A, in FAST 3A (FAST 3B).
[10.21 CHECK that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REFUEL ZONE FANS AND DAMPERS switch, 3-HS.-64-3A.
[11] CHECK the following conditions:
- SUPPLY FANS FILTER DIFF PRESS Indicator, 3-PDI-064-0022, indicates less than 0.6 inches H2O at the Reactor Building/Refuel Floor Supply fan intake room at El 565.
- REFUELING ZONE STATIC PRESS INTLK, 1-PDS-064-006 lA/C, on refuel floor Panel 25-220 indicates between (negative) -0.25 inches to -0.40 inches.
BOP Contacts AUO for the above information Driver When contacted wait 4 minutes and report 3-PDI-064-0022, indicates 0.4 inches H2O and that 1-PDS-064-OO61AJC, mdicates -0 3 mches 12
NRC Scenario 4 Simulator Event Guide:
Event 1 Component: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B SRO Direct Refuel and Reactor Zone Fans alternated BOP 6.1 Alternating Reactor Zone Supply and Exhaust Fans 3-OI-30B
[1] VERIFY the Reactor Zone supply and exhaust fans are operating. REFER TO Section 5.1.
[2] REVIEW all Precautions and Limitations in Section 3.0.
NOTES
- 1) The preferred method to start the standby Reactor Zone supply and exhaust fans is to use the common control switch (3 -HS 1 1A) on Panel 3-9-25.
- 2) Reactor Zone supply and exhaust dampers, 3-FCO-064-0013, 0014, 0042, and 0043 will open or close automatically as necessary when fans are stopped and started.
- 3) The Steam Vault Exhaust Booster Fan should normally be in service whenever the Unit is operating with Reactor Building Ventilation in service and fans in fast speed. Operation of the Steam Vault Exhaust Booster Fan with Reactor Zone Exhaust fans out of service is an ALARA concern due to backflow into the Reactor Building lower level ventilation ductwork. However, the Steam Vault Exhaust Booster fan may remain in service with Reactor Zone Exhaust fans out of service to cool the steam tunnel for short durations such as alternating fans, cycling reactor zone_dampers,_or_RPS_power transfers.
[3] IF Reactor Zone Ventilation is to remain Out of Service for an extended period (?:3 hours) and it is desired to leave the Steam Vault Exhaust Booster Fan in service, THEN (Otherwise N/A): Step is NA
[4] IF required, THEN SHUT DOWN Steam Vault Exhaust Booster Fan. REFER TO Section 7.4. (Otherwise N/A). Step is NA
[5] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in OFF.
[6] VERIFY dampers close and fans stop as indicated by illuminated green lights above the following switches:
- REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
- REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
- REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1A
\
[7] PLACE REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 1A, in SLOW A (SLOW B) to start alternate fans.
13
NRC Scenario 4 Simulator Event Guide:
Event 1 Component: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B BOP [8] VERIFY dampers open and fans start as indicated by illuminated red lights above the following switches:
- REACTOR ZONE SPLY OUTBD ISOL DMPR, 3-HS-64-13
- REACTOR ZONE SPLY INBD ISOL DMPR, 3-HS-64-14
- REACTOR ZONE FANS AND DAMPERS, 3-HS-64-1 1A
[9] IF fast speed Reactor Zone Supply and Exhaust Fan operation is required, five minutes should be allowed after slow start for the discharge dampers to FULLY OPEN, THEN
[9.1] PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-1 1A, in FAST A (FAST B).
[9.2] VERIFY that the two green lights A(B) remain extinguished and the two red lights A(B) remain illuminated above REACTOR ZONE FANS AND DAMPERS Switch, 3-HS-64-1 JA.
[10] VERIFY the following conditions:
[10.1] VERIFY REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-064-0002, on 3-LPNL-925-0213, located at R17-P El 639, indicates between -0.25 inches and -0.40 inches H2O.
[10.2] IF REACTOR ZONE PRESS DIFFERENTIAL Indicator, 3-PDIC-64-2, is not between -0.25 inches and -0.40 inches H20, THEN REFER TO 3-AOl-3 OB- 1, Reactor Building Ventilation Failure.
[11] IF required, THEN START Steam Vault Exhaust Booster Fan. REFER TO Section 5.4. NOT Required BOP Contacts AUO for the above information Driver When contacted wait 4 mmutes and report 3-PDIC-064-0002, mdicates -0 35 inches 1120 14
NRC Scenario 4 Simulator Event Guide:
Event 1 Component: Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 Each SCIV shall be OPERABLE.
APPLICABILITY: MODES 1,2, and 3, During operations with a potential for draining the reactor vessel (OPDRVs).
SRO ACTIONS NOTES
- 1. Penetration flow paths may be unisolated intermittently under administrative controls.
- 2. Separate Condition entry is allowed for each penetration flow path.
- 3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
Condition A. One or more penetration flow paths with one SC1V inoperable.
Required Action A. 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: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> AND Required Action A.2 Verify the affected penetration flow path is isolated.
Completion Time: Once per 31 days Call Work Control for a clearance on the REFUEL ZONE SPLY OUTBD ISOL DMPR, SRO 3-FCV-64-5 TRM Appendix A Power Operated Secondary Containment Isolation Valves 3-DMP-64-9 REFUELiNG ZONE EXH DUCT OUTBD ISOL DMPR < 10 SEC 15
NRC Scenario 4 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-GOl- 100-12.
[20] IF desired to raise power with only two(2) Reactor feedpumps in service, THEN RAISE Reactor power, as desired, maintaining each Reactor feedpump less than 5050 RPM.
ATC Raise Power w/Recirc, JAW 3-01-68, Section 6.2 D. Individual pump speeds should be mismatched by -6O 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-96-15A(15B).
AND/OR
- Raise Recirc Pump 3B using, RAISE SLOW (MEDIUM), 3-HS-96-16A(16B).
[2] WHEN desired to control Recirc Pumps 3A and/or 3B speed with the RECIRC MASTER CONTROL, THEN ADJUST Recirc Pump speed 3A &3B using the following push buttons as required:
RAISE SLOW, 3-HS-96-3 1 RAISE MEDIUM, 3-HS-96-32 NRC At RR Pump Speeds of 1260rpm and 1200 rpm, power will be 85% and RFPT RPMs will be just below 5025 Driver When directed by NRC, Trigger 1 Loss of 480V SD BD 3B, If crew attempts to close alternate supply breaker or is gomg to close alternate supply breaker delete ED1 OB m order to allow the crew to energized the Board Driver Wait 2 mmutes and report license class 1404 was in the field, a tramee accidently tripped the normal feeder breaker No problems mdicated on Board 16
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B Crew Responds to numerous alarms, diagnoses a loss of 480V SD BD 3B and 480V RMOV Bds 3Band3C Responds to the following alarms; 8B-30, 8C-17, 24, 29, 31, 4C-12 and 3D -32.
SRO Enters 3-AOI-99-1, 3-AOI-1-3 and 3-AOI-70-1.
BOP Alarm 8B-30: 480V SHUTDOWN BD 3B UV OR XFR A. CHECK for indication of 480V Shutdown Bd 3B loss:
- RWCU Pump 3B shutdown
- Fuel pool cooling Pump 3B shutdown
- 480V Shutdown Bd 3B voltage (3-EI-57-30)
B. IF 480V Shutdown Bd 3B is lost, THEN MANUALLY TRANSFER to alternate source by placing CS in ALTERNATE position on Panel 3-9-8.
C. IF manual transfer is accomplished, THEN REFER TO O-OI-57B, 3-01-99, and appropriate Ols for recoveiy or realignment of equipment.
D. IF manual transfer is NOT accomplished, THEN REFER TO Tech Spec Section 3.8.1.
Dispatches personnel to Breaker, may attempt to energize 480V SD BD 3B Driver If crew attempts to close alternate supply breaker or is going to close alternate supply breaker delete ED 1 OB in order to allow the crew to energized the Board Driver ThC KTCT ntil requested to investigate minutes and icense class 1404 1
a trainee accidently tripped al feeder br To problems ca*- oird. If the crew directs you t ore 480V SD B.. 3.. .. Normal supply hs577 1 [1] normal, if directe re Board oi tte supply change Ilulillal LU r1j alternate) 17
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B BOP Alarm 8C-24: 480V REACTOR MOV BD 3B OR 3E UV OR XFR A. CHECK light indications for loss of any 480V equipment.
B. CHECK 480V Rx MOV Bd 3B & 3E for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
C. IF Normal or Alternate feeder breaker tripped, THEN MANUALLY DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device.
D. IF undervoltage or transfer has occurred:
- 1. REFER TOTS Section 3.8.7.
- 2. RESET possible half-scram. REFER TO 3-01-99.
BOP Alarm 8C-29: I&C BUS B VOLTAGE ABNORMAL A. VERIFY the Alarm by checking:
- Loss of instrument power and remote position indication to Core Spray Div II and RHR Div II (Panel 3-9-3)
- RWCU Filter Demin 3B Isolation Reactor Zone/Refuel Zone Ventilation Isolation Verifies I&C Bus B Auto transferred to alternate feeder Alarm 8C-29: 480V REACTOR MOV BD 3C UV OR XFR A. VERIFY automatic action.
B. CHECK light indications for loss of 480V equipment.
C. CHECK board for abnormal conditions: relay targets, smoke, burned paint, breaker position, etc.
D. IF Normal or Alternate feeder breaker tripped, THEN MANUALLY DEPRESS mechanical trip/reset mechanism on breaker face to reset Bell Alarm lockout device.
E. REFER TO O-OI-57B to re-energize or transfer the board.
Driver When requested to restore steam tunnel booster fan wait two minutes and mrf PC 14 start 18
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B ATC Alann 4C-12: RBCCW PUMP DISCH. HDR PRESS LOW A. VERIFY 3-FCV-70-48 CLOSING/CLOSED.
B. VERIFY RBCCW pumps A and B in service.
C. VERIFY RBCCW surge tank low level alarm is reset.
E. REFER TO 3-AOI-70-1 for RBCCW System failure and 3-01-70 for starting spare pump.
When 480V RMOV BD 3B is restored should VERIFY 3-FCV-70-48 CLOSING Report Alarm 3D-32: Reactor Zone Differential Pressure Low BOP E0I-3 Entry Condition.
SRO Enters E0I-3 Secondary Containment Control When requested to restore RPS B, if requested to place on alternate trigger 3, mrfrpO9 reset
, Driver and bat restorerpsb, if requested to restore to normal then bat rpsreset and mrf rpO9 reset If place back on normal ensure to reset alternate supply circuit protectors mrf rpO3 reset.
19
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B ATC Announces Power, Pressure and Level stable on Board loss Crew 3-AOI-99-l, Loss of Power to One RPS Bus 4.1 Immediate Actions
[1] STOP all testing with potential RPS haif-scrams or PCIS logic isolation signals.
NOTES
- 3) Loss of RPS will isolate 3-RM-90-256, Drywell Air Monitor, and TS LCO 3.4.5 Condition B should be entered.
4.2 Subsequent Actions
[1] VERIFY automatic actions occur.
[2] ATTEMPT to determine cause of loss of RPS Bus using indicating lights inside RPS Circuit Protector cabinets.
[3] NOTIFY Chemistry RWCU is isolated and no longer in-service and a sampling LCO per TRM 3.4.1 is to be entered.
[4] NOTIFY Electrical Maintenance to correct cause.
[5] RESTORE power to RPS Bus A(B) using alternate power supply. REFER TO 3-01-99 section for Immediate Restoration of Power to RPS Bus A(B) Using Alternate Power Supply.
[5.1] DISPATCH operator to Aux. Instrument Room to reset ATU GROSS FAILURES.
[6] WHEN system restoration is desired, ThEN RESTORE systems to normal.
REFER TO 3-01-99 section for Restoration to Normal Following RPS Bus Power Loss.
When requested to restore RPS B, if requested to place on alternate trigger 3, mrf rpO9 reset Driver and bat restorerpsb, if requested to restore to normal then bat rpsreset and mrfrpO9 reset If place back on normal ensure to reset alternate supply circuit protectors mrf rpO3 reset.
20
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B ATC/BOP Reports MSIV A Inboard Valve shut on loss of board.
SRO Enters 3-AOl-i -3, Main Steam Isolation Valve Closure at Power ATC 4.2 Subsequent Actions
[1] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s).
[2] LOWER reactor power with recirc flow and insert control rods as directed by the Reactor Engineer/Unit Supervisor as necessary to ensure that rated steam line flow 3.54 x 106 lb/hr is not exceeded as indicated on Main Steam Line Flow Indicators.
REFER TO 3-GOI-l00-12 or 3-GOI-100-12A for the power reduction.
[6] IF Drywell control air pressure is normal, THEN INITIATE trouble-shooting of the MSIV. (Otherwise N/A) Step is NA
[7] EVALUATE Technical Specification 3.6.1.3, Primary Containment Isolation Valves.
SRO Directs ATC to lower power to less than 3.54 x 106 lb/hr on Main Steam Line Flow Indicators. Directs recirc pump speeds matched when outside of the 1200 to 1300 rpm band.
BOP Places MSIV A Inboard Valve handswitch in the close position ATC Lowers power as directed by SRO 21
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-AOI 1, Loss of Reactor Building Closed Cooling Water ATC 4.1 Immediate Actions
[1] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A):
SECURE RWCU Pumps.
Verifies RWCU Tripped, cannot verify sectionalizing valve at this time NO Power 4.2 Subsequent Actions
{ 1] IF Reactor is at power AND Drywell Cooling cannot be immediately restored, AN]) core flow is above 60%,THEN:
[2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A).
One RBCCW Pump is in service with sectionalizing valve open due to loss of power
[3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A):
[3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions.
[3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s).
ATC When power is restored to 480V SD BD 3B RBCCW Pump will auto start 22
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-AOI 1, Loss of Reactor Building Closed Cooling Water ATC 4.1 Immediate Actions
[1] IF RBCCW Pump(s) has tripped, THEN Perform the following (Otherwise N/A):
- SECURE RWCU Pumps.
Verifies RWCU Tripped, cannot verif sectionalizing valve at this time NO Power 4.2 Subsequent Actions
[1] Li? Reactor is at power AND Drywell Cooling cannot be immediately restored, AN]) core flow is above 60%,THEN:
[2] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s) (otherwise N/A).
One RBCCW Pump is in service with sectionalizing valve open due to loss of power
[3] IF RBCCW Pump(s) has tripped and it is desired to restart the tripped RBCCW pump, THEN PERFORM the following (otherwise N/A):
[3.1] INSPECT the tripped RBCCW pump and its associated breaker for any damage or abnormal conditions.
[3.2] IF no damage or abnormal conditions are found, THEN ATTEMPT to restart tripped RBCCW pump(s).
23
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System ATC/BOP 8.3 Restoration to Normal Following RPS Bus Power Loss
[1] OBTAiN Unit Supervisor/SROs permission to restore to normal.
[2] MOMENTARILY PLACE SCRAM RESET, 3-HS-99-5A/S5, as follows:
[2.1] RESET FIRST. (Group 2/3)
[2.2] RESET SECOND. (Group 1/4)
[2.3] NORMAL
[3] CHECK the following conditions:
A. All eight SCRAM SOLENOID GROUP A/B LOGIC RESET lights illuminated.
B. The following four lights ILLUM1NATED:
- SYSTEM A BACKUP SCRAM VALVE, 3-IL-99-5A/AB
- SYSTEM B BACKUP SCRAM VALVE, 3-IL-99-5AICD C. Scram Discharge Volume vent and drain valves indicate OPEN.
[4] At Panel 3-9-4, RESET PCIS trip logic as follows:
[4.1] MOMENTARILY PLACE PCIS DIV I RESET, 3-HS-64-16A-S32, to left and right RESET positions.
[4.2] CHECK the following red lights ILLUMINATED:
- MS1V GROUP Al, 3-IL-64-A1
- MS1V GROUP Bl, 3-IL-64-Bl
[4.3] MOMENTARILY PLACE PCIS DIV II RESET, 3-HS-64-16A-S33, to left and right RESET positions.
[4.4] CHECK the following red lights ILLUMINATED:
- MSIV GROUP A2, 3-IL-64-A2
- MSIV GROUP B2, 3-IL-64-B2
[6] RESTORE Reactor and Refuel Zone Ventilation to normal operation. REFER TO 3-AOI-64-2D, Group 6 Ventilation System Isolation.
BOP 3-AOI-64-2D, Group 6 Ventilation System Isolation 24
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System BOP 8.3 Restoration to Normal Following RPS Bus Power Loss
[7] RESTORE Standby Gas Treatment System to standby readiness. REFER TO 0-01-65.
BOP [8] At Panel 3-9-3, PLACE PSC head tank pumps in service as follows:
- PLACE SUPPR POOL DRAIN 1NBD ISOL VALVE, 3-HS-75-57A, in AUTO After OPEN.
PLACE SUPPR POOL DRAIN OUTBD ISOL VALVE, 3-HS-75-58A, in AUTO After OPEN.
[10] At Panel 3-9-3, RESTORE Drywell DP Compressor to automatic operation as follows:
[10.1] DEPRESS DRYWELL DP CPRSR SUCT VLV RESET pushbutton, 3-HS-64-1 39A.
[10.2] DEPRESS DRYWELL DP CPRSR DISCH VLV RESET pushbutton, 3-HS-64-140A.
[10.3] VERIFY OPEN DW TO SGT INBD ISOL VALVE using 3-HS-64-31.
[10.4] VERIFY OPEN SUPPR CHBR SGT INBD ISOL VALVE using 3-HS-64-34.
BOP [11] At Panel 3-9-4, RESTORE Drywell Floor and Equipment Drain Systems to normal operation as follows:
[11.1] NOTIFY Radwaste Operator that Drywell Equipment and Floor Drain Surnp isolation valves are being reopened.
[11.2] PLACE DRYWELL EQPT DRINBD ISOL VLV, 3-HS-77-15A, in AUTO After OPEN.
[11.3] PLACE DRYWELL EQPT DROUTBD ISOL VLV, 3-HS-77-15B, in AUTO After OPEN.
[11.4] PLACE DRYWELL FLOOR DR INBD ISOL VLV, 3-HS-77-2A, in AUTO After OPEN.
[11.5] PLACE DRYWELL FLOOR DR OUTBD ISOL VLV, 3-HS-77-2B, in AUTO After OPEN.
Driver when directed by NRC triggers for Stator water pump trip 25
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Enters 3-01-99, Reactor Protection System BOP 8.3 Restoration to Normal Following RPS Bus Power Loss
[12] IF DW Radiation Monitor CAM, 3-RM-90-256 was secured due to a preplanned transfer, THEN (otherwise N/A) Step is NA
[13] IF DW Radiation Monitor CAM, 3-RM-90-256, isolated due to loss of RPS, THEN MOMENTARILY DEPRESS the following RESET pushbuttons on Panel 3-9-2.
- DW RAD MON LOWER 1NBD SUPPLY ISV RESET, 3-HS-90-254B-A (opens FCV-90-254B)
- DW RAD MON OUTBD RETURN ISV RESET, 3-HS-90-257A-A (opens FCV 90-257A DW RAD MON OUTBD SUPPLY ISV RESET, 3-HS-90-255A (opens FCV 255
[14] At Panel 3-9-54, PLACE H2/02 Analyzer in service per 3-01-76.
[15] At Panel 3-9-55, VERIFY DRYWELL OR SUPPRESSION CHAMBER EXHAUST TO SGTS, 3-FIC-84-20, in AUTO with setpoint at 100 scfm.
[19] At Panels 3-9-10 and 3-9-1 1, RESTORE Radiation Monitoring System as follows:
[19.1] DEPRESS applicable RESET pushbuttons.
[19.2] RESTORE Radiation Monitoring System to normal. REFER TO 3-01-90.
[20] RESTORE Main Steam System to normal. REFER TO 3-01-1.
[22] At Panel 3-9-13, DEPRESS TIP ISOLATION RESET pushbutton, 3-HS-94-7D-S2.
Depresses Fault rest pushbuttons on each VFD on Panel 9-4 in able to clear Recirc Drive ATC/BOP Alarms on Panel 9-4A and 4B.
26
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B BOP 3-AOI-64-2D, Group 6 Ventilation System Isolation
[1] IF any EOI entry condition is met, THEN ENTER appropriate EOI(s).
[2] VERIFY Group 6 isolation valves penetrating Primary Containment are closed.
UTILIZE Panel 3-9-3 mimic or Containment Isolation Status System on Panel 3-9-4.
[3] IF Refuel Zone Isolation is due to high radiation, as indicated on 3-RM-90-140 and/or 3-RM-90-141, Panel 3-9-10, and/or associated recorder on Panel 3-9-2, THEN. (Otherwise N/A) Step is NA
[7] CHECK the following to confirm condition:
- REACTOR & REFUEL ZONE EXHAUST RADIATION, 3-RR-90-144
[13] WHEN initiating signal has been corrected AND necessary repairs are made, THEN
[13.1] VERIFY PCIS RESET:
- RESET PCIS DIV I RESET, 3-HS-64-16A-S32.
- RESET PCIS DIV II RESET, 3-HS-64-16A-S33.
[13.2] RESET Reactor/Refuel isolation logic, as required:
- PLACE REFUEL ZONE FANS AND DMPRS, 3-HS-64-3A, in OFF.
- PLACE REACTOR ZONE FANS AND DMPRS, 3-HS-64-1 1A, in OFF.
[13.3] START Reactor/Refuel zone ventilation, as required:
- PLACE REACTOR ZONE FANS AND DAMPERS switch, 3-HS-64-1 1A, in SLOW A (SLOW B).
- PLACE REFUEL ZONE FANS AND DAMPERS Switch, 3-HS-64-3A, in SLOW 3A (SLOW 3B).
[13.4] For the fans started, VERIFY that the dampers open and fans start as indicated by illuminated red lights above the following switches:
- The two green lights A(B) above REACTOR ZONE FANS AND DAMPERS Switch 3-HS 1 1A, extinguish and the two red lights A(B) illuminate.
- The two green lights A(B) above REFUEL ZONE FANS AND DAMPERS Switch 3-HS-64-3A, extinguish and the two red lights A(B) illuminate.
Dnver when directed by NRC trigger 5 for Stator water pump trip 27
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Tech Spec Actions from loss of 480V SD BD 3B Evaluate TRM 3.4.1 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 Informs Chemistry have lost Continuous reactor coolant conductivity monitoring SRO Evaluate Tech Spec 3.4.5 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:
- a. Drywell floor drain sump monitoring system; and
- b. One channel of either primary containment atmospheric particulate or atmospheric gaseous monitoring system.
APPLICABILITY: MODES 1,2, and 3.
Condition B: Required primary containment atmospheric monitoring system inoperable.
Required Action B. 1: Analyze grab samples of primary containment atmosphere.
Completion Time: Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action B.2: Restore required primary containment atmospheric monitoring system to OPERABLE status.
Completion Time: 30 days Driver when directed by NRC trigger 5 for Stator water pump trip 28
NRC Scenario 4 Simulator Event Guide:
Event 3 Component: Loss of 480V SD BD 3B SRO Tech Spec Actions from loss of 480V SD BD 3B SRO Evaluate Technical Specification 3.6.1.3.
3.6.1.3 Primary Containment Isolation Valves (PCIVs)
LCO 3.6. 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 3.3.6.1, Primary Containment Isolation Instrumentation.
Condition A: NOTE Only applicable to penetration flow paths with two PCIVs.
One or more penetration flow paths with one PCIV inoperable except due to MSIV leakage not within limits Required Action A. 1: Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
Completion Time: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main steam lines Required Action A.2: NOTE Isolation devices in high radiation areas may be verified by use of administrative means.
Verify the affected penetration flow path is isolated.
Completion Time: Once per 31 days for isolation devices outside primary containment Driver when directed by NRC trIgger 5 for Stator water pump trip 29
NRC Scenario 4 Simulator Event Guide:
Event 4 Component: Stator Water Cooling Pump trip BOP Responds to alarms 7A-22 and 8A-1 BOP Announces trip of Stator Water Cooling Pump 3B 7A-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.
C. CHECK alarm by dispatching personnel to check the Stator Coolant Control Cabinet.
8A-l, TURBINE TRIP TIMER INITIATED A. CHECK Stator Cooling Water Flow and Temperature and Generator Stator temperatures using ICS.
B. VERIFY all available Stator Cooling Water Pumps running.
NOTE The full capacity of the Turbine Bypass valves with all nine valves open is 25% reactor power. To determine the capacity of the bypass valves, subtract 3% for each out of service bypass valve from the 25%. (Example, one bypass valve out of service, [25% 3% = 22%),
therefore, the capacity of the bypass valves with one bypass valve out of service is 22%.)
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. (Otherwise N/A)
BOP Starts Stator Water Cooling Pump 3B When dispatched wait two minutes and report pump is extremely hot to touch, at breaker breaker is tripped no other indications Driver When directed by NRC to insert RFPT 3A governor failure, verify start value is between 71 and 72. If not modify start value to the current value and ensure final severity is set to zero and ensure ramp time remains unchanged and then insert trigger 10 mffw3Oa (dO 0) 0 1300 72, When operator takes manual control of RFPT 3A ensure
- trigger 11 goes active to allow the operator to control. Be prepared to insert the next event if the crew decides to scram, see next page driver instructions.
30
NRC Scenario 4 Simulator Event Guide:
Event 5 Instrument: RFPT 3A Governor slowly fails low ATC Notices lowering speed on RFPT 3A or rising speed on RFPT 3C, or responds alarm 5A-8.
5A-8, REACTOR WATER LEVEL ABNORMAL A. VERIFY Reactor water level hi/low using multiple indications including Average Narrow Range Level on 3-XR-3-53 recorder, 3-LI-3-53, 3-LI-3-60, 3-LI-3-206, and 3-L1-3-253 on Panel 3-9-5.
B. IF alarm is valid, THEN REFER TO 3-AOI-3-l_or 3-01-3.
ATC Report Reactor level less than 27 inches and lowering, reports RFPT 3A flow has lowered.
Takes manual control of RFPT 3A to attempt to control RPV Level SRO Directs entry 3-AOI-3 -1, Loss Of Reactor Feedwater or Reactor Water Level High/Low ATC [1] VERIFY applicable automatic actions.
[2] IF level OR Feedwater flow is lowering due to loss of Condensate, Condensate Booster, or Feedwater Pump(s), ThEN REDUCE Recirc flow as required to avoid scram on low level.
[4] IF Feedwater Control System has failed, THEN
[4.1] PLACE individual RFPT Speed Control Raise/Lower switches in Manual Governor (depressed position with amber light illuminated).
[4.2] ADJUST RFP Discharge flows with RFPT Speed Control Raise/Lower switches as necessary to maintain level.
[24] 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.
Driver when directed by NRC or if the crew decides to scram, verify start value is between 79 and
- 80. If not modify start value to the current value and ensure final severity is set to 82 and ensure ramp time remains unchanged and then insert trigger 14.
imftclOb (e14 0) 82 160079 31
NRC Scenario 4 Simulator Event Guide:
Event 6 Instrument: EHC Pressure Transducer Failure Responds to alarm 7B-6, EHC/TSI SYSTEM A. On EHC Workstation computer on Panel 3-9-7, Alarm Summary screen, ATC/BOP ATTEMPT to RESET alarm input.
B. II? necessary, THEN REQUEST assistance from Site Engineering.
ATC Recognizes lowering Reactor Pressure and generator megawatts.
SRO Directs entry into 3-AOI-47-2.
3-AOI-47-2 Turbine EHC Control System Malfunctions 4.1 Immediate Actions
[1] IF Reactor Pressure lowers to or below 900 psig, THEN MANUALLY SCRAM the Reactor and CLOSE the MSTVs.
4.2 Subsequent Actions
[3] IF a Group 1 isolation has occurred, THEN PLACE EHC PUMP 3A and 3B, 3-HS-47-1A and 3-HS-47-2A, to PULL TO LOCK.
BOP Places EHC Pumps 3A and 3B in Pull to Lock SRO Directs manual scram, closing of the MSIVs, and entry into 3-AOI-l00-1.
ATC Manually scrams the reactor.
32
NRC Scenario 4 Simulator Event Guide:
Event 7 Major: ATWS ATC 3-AOI-100-l, Reactor Scram 4.1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 3-HS-99-5AIS3A and 3-HS-99-5AIS3B, on Panel 3-9-5.
[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in START & HOT STBY AND PAUSE for approximately 5 seconds (N/A)
[3] Refuel Mode One Rod Permissive Light check
[3.1] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in REFUEL.
[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 3-XI-85-46.
[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. (N/A)
[4] PLACE REACTOR MODE SWITCH, 3-HS-99-5A-S1, in SHUTDOWN.
[5] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Water Level and trend (recovering or lowering).
- Reactor pressure and trend
- MSIV position (Open or Closed)
- Power level 4.2 Subsequent Actions
[2] IF all control rods CAN NOT be verified fully inserted, THEN PERFORM the following:
[2.1] INITIATE ARI by Arming and Depressing BOTH of the following:
- ARI Manual Initiate, 3-HS-68-1 19A
- ARI Manual Initiate, 3-HS-68-1 19B
[2.2] VERIFY the Reactor Recirc Pumps (if running) at minimum speed at Panel 3-9-4.
[2.3] REPORT ATWS Actions Complete and power level.
[3] DRiVE in all IRMs and SRMs from Panel 3-9-5 as time and conditions permit.
[3.1] DOWNRANGE IRMs as_necessary to follow power_as it lowers.
33
NRC Scenario 4
, Simulator Event Guide:
Event 7 Component: ATWS SRO Enters EO1- 1 on Reactor Level 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 or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO -
IF RPV water level cannot be determined? NO -
Is any MSRV Cycling? YES, but MS1Vs closed 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 with SRVs LAW APPX hA SRO EOI- 1 (Reactor Level)
Monitor and Control Reactor Water Level.
Directs_Verification_of PCIS_isolations.
ATC/BOP Verifies PCIS isolations.
SRO IF It has NOT been detennined that the reactor will remain subcritical without boron under all conditions THEN Exit RC/L and Enter C5, Level/Power control.
SRO Exits RCIL and Enters 3-C-5, Level/Power Control 34
NRC Scenario 4 Simulator Event Guide:
Event 7 Major: ATWS ATC/BOP Commence pressure control with Appendix hA, Alternate RPV Pressure Control Systems MSRVs IF Drywell Control Air is NOT available, THEN EXECUTE EOI Appendix 8G, CROSSTIE CAT) 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-l-179 MN STM LiNE A RELTEF VALVE.
- b. 23-PCV-l-180 MN STM LiNE D RELIEF VALVE.
- c. 3 3-PCV-1-4 MN STM LiNE A RELIEF VALVE.
- d. 4 3-PCV-1-3 1 MN STM LINE C RELIEF VALVE.
- e. 53-PCV-1-23 MN STM LINE B RELIEF VALVE.
- f. 6 3-PCV-l-42 MN STM LINED RELIEF VALVE.
- g. 73-PCV-1-30 MN STM LINE C RELIEF VALVE.
- h. 8 3-PCV-1-19 MN STM LINE B RELIEF VALVE.
- i. 9 3-PCV-1-5 MN STM LINE A RELIEF VALVE.
- j. 10 3-PCV-1-41 MN STM LINED RELIEF VALVE.
- k. 11 3-PCV-1-22 MN STM LINE B RELIEF VALVE.
- 1. 123-PCV-1-18 MN STM LiNE B RELIEF VALVE.
- m. 133-PCV-1-34 MN STM LINE C RELIEF VALVE.
35
NRC Scenario 4 Simulator Event Guide:
Event 7 Major: ATWS SRO EOI- 1 (Power)
Monitor and Control Reactor Power Verify Reactor Mode Switch in shutdown Yes Initiate ARI completed Will tripping Recirc Pumps cause trip of main turbine, RFP, HPCI or RCIC No Is reactor power above 5% or unknown No -
SLC Leg When periodic APRM oscillations greater than 25% peak to peak persist continue OR Before Suppression Pool temperature rises to 110°F continue Direct SLC injection (APPX 3A)
Inhibit ADS Verify RWCU system isolation completed earlier Insert Control Rods Leg Reset ARI and defeat ARI logic trip (APPX 2)
Insert Control Rods using any of the following methods:
APPX-1A Deenergize scram solenoids No APPX-1B Vent the scram air header No APPX-1C Scram individual control rods No APPX-1D Drive Control Rods Yes APPX-1E Vent control rod over piston No -
APPX-1F - Reset scram/RE-SCRAM - Yes APPX-1G Raise CRD cooling water dp No -
36
NRC Scenario 4 Simulator Event Guide:
Event 7 Major: ATWS ATC Inserting Control Rods Calls for 3-EOI Appendix-2 and the field portion of 3-EOI Appendix-iF 3-EOI Appendix-iF
- 3. VERIFY OPEN Scram Discharge Volume vent and drain valves.
- 4. DRAIN SDV UNTIL the following annunciators clear:
- WEST CR1) DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 1)
- EAST CRD DISCH VOL WTR LVL HIGH HALF SCRAM (Panel 3-9-4, 3-XA-55-4A, Window 29).
- 5. DISPATCH personnel to VERIFY OPEN 3-SHV-085-0586, CHARGiNG WATER ISOL.
- 6. WHEN CR1) Accumulators are recharged, THEN INITIATE manual Reactor Scram and ARI.
3-EOI Appendix-iD
- 1. VERIFY at least one CR1) pump in service.
- 2. IF Reactor Scram or ARI CANNOT be reset, THEN DISPATCH personnel to CLOSE 3-SHV-085-0586, CHARGiNG WATER SOV.
- 3. VERIFY REACTOR MODE SWITCH in SHUTDOWN.
- 4. BYPASS Rod Worth Minimizer.
- 5. REFER to Attachment 2 and INSERT control rods in the area of highest power as follows:
- a. SELECT control rod.
- b. PLACE CRD NOTCH OVERRIDE switch in EMERG ROD iN position UNTIL control rod is NOT moving inward.
- c. REPEAT Steps 5.a and 5.b for each control rod to be inserted.
Driver When called for Appendix 2 wait 2 mmutes and trigger 24, Appendix-iF wait 3 mmutes and trigger 23, when SCRAM is reset trigger 25 to unstick rods Before the crew scrams or the insert trigger 28 for bat nrcatws95 If requested to close 85-586 trigger 26 to close and trigger 27 to open 37
NRC Scenario 4 Simulator Event Guide:
Event 9 Component: Loss of HPCI 120 VAC Power Supply SRO Enters C-5, Level/Power Control Inhibit ADS ATC/BOP lithibits ADS SRO Is any main steam line open No-Is reactor power above 5% or unknown No -
Maintain RPV water level between -180 inches and +51 inches with the following injection sources:
CRD APPX 5B, RCIC APPX 5C, SLC APPX 7B SRO Directs a Level Band maintained by RCIC ATC Initiate RCIC JAW Appendix-5C and maintains level in directed band, if possible BOP Reports Loss of HPCI 120 VAC Power Supply, HPIC NOT available for Level Control Calls for investigation and repair in order to use IIPCI for Level control Crew 38
NRC Scenario 4 7 Simulator Event Guide:
Event 7 Component: ATWS ATC/BOP Maintain Directed Level Band with RCIC, Appendix SC.
- 1. VERIFY RESET and OPEN 3-FCV.-71-9, RCIC TURB TRIP/THROT VALVE RESET.
- 2. VERIFY 3-FJC-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 MN 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 MN 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.
39
NRC Scenario 4 Simulator Event Guide:
Event 7 Component: ATWS Crew Report rising Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-l),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17C), Direct Appendix 1 7C When suppression chamber pressure exceeds 12 psig, Stops the first time through when the LOCA worsens will continue at that time Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)
When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue, Does not continue 40
NRC Scenario 4 Simulator Event Guide:
Event 7 Component: ATWS Crew Report rismg Drywell Pressure SRO Enters EOI-2 on High Drywell Pressure PC/H Verify H202 analyzer in service (APP 19)
When H2 is detected in PC (2.4% on control room indicators continue, does not continue SRO Enters EOI-2 on High Drywell Pressure SP/T MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WREN suppr p1 temp CANNOT be maintained below 95°F, directs RHR Pumps in Pool Cooling Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -l in. and -6 in. (APPX 18)
Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvi be maintained below -1 in., YES 41
NRC Scenario 4 Simulator Event Guide:
Event 7 Component: ATWS ATCIBOP 3-EOI APPENDIX-17A, RHR System Operation Suppression Pool Cooling 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 R}IR 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 R}IRSW flow:
- 3-FCV-23-34, RHR HX 3A RHRSW OUTLET VLV 3-FCV-23-46, R}IR. HX 3B RHRSW OUTLET VLV 3-FCV-23-40, RHR HX 3C RHRSW OUTLET VLV 3-FCV-23-52, RHR HX 3D RHRSW OUTLET VLV.
- d. IF Directed by SRO, THEN PLACE 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD in MANUAL OVERRIDE.
- e. IF LPCI iNITIATION Signal exists, THEN MOMENTARILY PLACE 3-XS-74-121(129), RI-JR SYS 1(11) CTMT SPRAY/CLG VLV SELECT in SELECT.
- f. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) LPCI OUTBD INJECT VALVE.
- h. VERIFY desired RHR pump(s) for Suppression Pool Cooling are operating.
CAUTION RHR System flows below 7000 gpm or above 10000 gpm for one-pump operation may result in excessive vibration and equipment damage.
ATCIBOP Aligns directed RFIR Pumps in Pool Cooling 42
NRC Scenario 4
= Simulator Event Guide:
Event 7 Component: ATWS ATCIBOP 3-EOI APPENIMX-17A, RHR System Operation Suppression Pool Cooling
- 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
- m. IF Additional Suppression Pool Cooling flow is necessary, THEN PLACE additional RHR and RHRSW pumps in service using Steps 2.b through 2.1.
ATC/BOP Aligns directed RHR Pumps in Pool Cooling 43
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA ATCIBOP 3-EOI APPENIMX-17C, RHR System Operation Suppression Chamber Sprays 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, TI-lEN 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.
- 3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN CONTINUE in this procedure at Step 7.
- 4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.
- 5. INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
- LPCI Initiation signal is NOT present, OR Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11)
LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 3-FCV-74-53(67), RHR SYS 1(11) 1NBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RHR SYS 1(11) OUTBD iNJECT VALVE.
- e. VERIFY OPERATING the desired R}{R System 1(11) pump(s) for Suppression Chamber Spray.
- g. OPEN 3-FCV-74-58(72), RHR SYS 1(11) SUPPR CHBR SPRAY VALVE.
ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 44
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA ATCIBOP 3-EOI APFENDIX-17C, RHR System Operation Suppression Chamber Sprays
- 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 3-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.
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
- n. NOTIFY Chemistry that RHRSW is aligned to in-service RUR Heat Exchangers.
ATC/BOP Aligns directed RHR Pumps in Suppression Chamber Sprays 45
NRC Scenario 4 Simulator Event Guide:
Event S Major: LOCA ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Drywell Sprays
- 1. BEFORE Drywell pressure drops below 0 psig, CONTiNUE in this procedure at Step 7.
- 2. IF Adequate core cooling is assured, OR Directed to spray the Drywell 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 1NJ VLV BYPASS SEL in BYPASS.
- 3. VERIFY Recirc Pumps and Drywell Blowers shutdown.
- 4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.
- 5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.
- 6. INITIATE Drywell Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN PLACE keylock switch 3-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 3-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN VERIFY CLOSED 3-FCV-74-52(66), RFIR SYS 1(11) LPCI OUTBD INJECT VALVE.
- e. VERIFY OPERATING the desired System 1(11) R}{R pump(s) for Drywell Spray.
- f. OPEN the following valves:
ATC/BOP .
Aligns directed RHR Pumps in Drywell Sprays 46
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA
= ATC/BOP 3-EOI APPENDIX-17B, RHR System Operation Diywell Sprays
- g. VERIFY CLOSED 3-FCV-74-7(30), RHR SYSTEM 1(11) M1N FLOW VALVE.
- h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System II RHR Pump in service.
- j. VERIFY RHRSW pump supplying desired RI-JR Heat Exchanger(s).
- k. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- 7. WHEN EITHER of the following exists:
- Before drywell pressure drops below 0 psig, OR
- Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:
- a. VERIFY CLOSED the following valves:
- 3-FCV-74-100, RHR SYS I U-2 DISCH XTIE
- b. VERIFY OPEN 3-FCV-74-7(30), RHR SYSTEM 1(11) M1N FLOW VALVE.
- c. IF RUR operation is desired in ANY other mode, THEN EXIT this EOI Appendix.
d._STOP RHR Pumps.
ATCIBOP Aligns directed RHR Pumps in Drywell Sprays 47
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA SRO C5 LevellPower Control SRO As Level continues to lower with RCIC injection, directs use of SLC APPX-7B ATC Initiates SLC lAW APPX 7B
- 2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, -
THEN CONTINUE at Step 10 to inject SLC Boron Tank to RPV.
- 10. UNLOCK and PLACE 3-HS-63-6A, SLC PUMP 3A/3B, control switch in START PUMP 3A or START PUMP 3B (Panel 3-9-5).
- 11. CHECK SLC 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 (3-XA-55-5B, Window 20).
- System flow, as indicated by 3 -IL-63 -11, SLC FLOW, red light illuminated,
- SLC iNJECTION FLOW TO REACTOR Annunciator in alarm (3-XA-55-5B, Window 14).
- 12. IF Proper system operation CANNOT be verified, THEN RETURN TO Step 10 and START other SLC pump.
As RPV Level continues to lower, CAN RPV water level be restored and maintained above SRO
-180 inches No -
Are at least 2 MSRVs open No -
Emergency Depressurization is Required 3-C-2 and 3-C-5 Will the reactor remain subcritical without boron under all conditions NO When all injection into the RPV is stopped and prevented except from RCIC, CRD, and SLC_per_CS,_Level/Power control_Step_CS-22 Stop and Prevent ALL injection into RPV Except from RCIC, CRD, and SLC (APPX 4) 48
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA BOP/ATC Stop and Prevent ALL injection into RPV Except from RCIC, CRD, and SLC (APPX 4)
- 3. PREVENT injection from CORE SPRAY following an initiation signal by PLACING ALL Core Spray pump control switches in STOP.
- 4. PREVENT injection from LPCI SYSTEM I by performing the following:
- 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:
- 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.
- c. CLOSE the following valves BEFORE RPV pressure drops below 450 psig:
- 3-FCV-3-19, RFP 3A DISCHARGE VALVE
- 3-FCV-3-12, RFP 3B DISCHARGE VALVE
- 3-FCV-3-5, RFP 3C DISCHARGE VALVE
- 3-LCV-3-53, RFW START-UP LEVEL CONTROL 49
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA SRO C2 Emergency Depressurization and C5 Level/Power Control Is suppression pool level above 5.5 feet Yes Open all ADS Valves BOP/ATC Opens all 6 ADS Valves SRO Can at least two MSRVs be opened per C2 Emergency RPV Depressurization Yes -
When RPV pressure is below MSCP Table 1A 190 psig Start and Slowly raise RPV injection with the following injection sources to restore and maintain RPV water level above -180 inches Directs injection with LPCI APPX 6B and 6C to restore RPV Level to directed band BOP/ATC Injects with LPCI LAW APPX 6B and/or 6C to restore RPV water level SRO Emergency Classification EPIP-1 1.1-Si Reactor water level can NOT be maintained above -162 inches. (TAF)
OR 1.2-S Failure of automatic scram, manual scram, and ARI to bring the reactor subcritical.
50
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA BOP/ATC Injects with LPCI JAW APPX 6B to restore RPV water level
- 1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN PLACE 3-HS-74-155A, LPCI SYS I OUTBD NJ VLV BYPASS SEL in BYPASS.
- 3. VERIFY OPEN 3-.FCV-74-12, RHR PUMP 3C SUPPR POOL SUCTVLV.
- 4. VERIFY CLOSED the following valves:
- 3-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE
- 5. VERIFY RHR Pump 3A and/or 3C running.
- 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-53, RHR SYS I LPCI INBD INJECT VALVE.
- 7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 3-FCV-68-79, RECIRC PUMP 3B DISCHARGE VALVE.
- 10. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers discharging to the RPV.
- 11. THROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:
51
NRC Scenario 4 Simulator Event Guide:
Event 8 Major: LOCA BOP/ATC Injects with LPCI lAW APPX 6C to restore RPV water level
- 1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THEN .PLACE 3-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS SEL in BYPASS.
- 4. VERIFY CLOSED the following valves:
- 3-FCV-74-72, RHR SYS II SUPPR CHBR SPRAY VALVE
- 5. VERIFY RHR Pump 3B and/or 3D running.
- 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 3-FCV-74-67, RHR SYS II LPCI INBD INJECT VALVE.
- 7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 3-FCV-68-3, RECIRC PUMP 3A DISCHARGE VALVE.
- 10. PLACE RHRSW pumps in service as soon as possible on ANY RHR Heat Exchangers discharging to the RPV.
- 11. TIIROTTLE the following in-service RHRSW outlet valves to maintain flow between 1350 and 4500 gpm:
52
NRC Scenario 4 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs:
RFPT 3B and EECW Pump A3 Operations/Maintenance for the Shift:
Alternate Refuel and Reactor Zone Fans JAW 3-OI-30A and 30B.
Commence a power increase to 85% in accordance with the RCP Unit 1 and 2 are at 100% Power Unusual Conditions/Problem Areas:
The following Control Rods are identified as SLOW: 30-19, 34-23, 14-51, 02-19, 46-51, and 06-43.
53
Facility: Browns Ferry Nil? Scenario No.: NRC 5 Op-Test No.: 1306 Examiners: Operators: SRO:
ATC:
BOP:
Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.
Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition lAW 2-OI-92B. Lower Power with flow to 91% for Main Turbine Valve Testing.
Event Maif. No. Event Type* Event Description No.
N-BOP 1 Return LPRM 8-49B to Operate JAW 2-OI-92B N-SRO R-ATC 2 Commence power decrease with flow to 90%
R-SRO C-BOP 3 edl8a Loss of I&C Bus A TS-SRO R-ATC 4 adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the 5 thl8a C-SRO standby pump to auto start C-ATC LOCA Recirculation Pump 2A Inboard and Outboard seal 6 R-ATC thlO/1 la failure TS-SRO Two Level instruments fail high tripping Feedwater and 7 Batch File M-ALL HPCI / LOCA / ED on Reactor Level 8 edl0a C Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not 9 Batch I Auto open 10 rcO8 C RCIC Steam Valve fails to Auto open
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Critical Tasks Four 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 T.A.F. (-162 inches).
- 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 -162 inches.
- 4. Feedback:
Reactor water level trend.
Reactor pressure trend.
With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to
-180 inches.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
Water level trend.
- 3. Measured by:
Observation At least 6 SRVs opened
- 4. Feedback:
RPV pressure trend.
SRV status indications.
Critical Tasks Four To prevent an uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
- 3. Measured by:
ADS logic inhibited prior to an automatic initiation.
- 4. Feedback:
RPV pressure trend.
RPV level trend.
ADS ADS LOGIC BUS A/B INHIBITED annunciator status.
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 I 7B AND Observation RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment
Events
- 2. ATC lowers power to 90% using recirculation flow.
- 3. The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service lAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high H2 will result in Off Gas, the SRO will evaluate TRM 3.7.2 and enter Condition A.
The H2O2 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.
- 4. During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to I&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOl-i-i actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.
- 5. The VFD Cooling Water Pump for the A 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.
- 6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump lAW with 2-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 again 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.
- 7. Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCIC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.
- 8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but no throttle capability with exist. RHR Loop 1 will not be available for Containment cooling operation.
- 9. With Division 2 Accident logic bypassed RUR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RHR Loop 2 will be available for Containment Cooling functions until required for injection.
Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER:5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No)
Scenario Tasks TASK NUMBER K/A RO SRO Restore an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 Lower Power with Recirc Flow RO U068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 26200 1A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-OO1-AB-O1 239002A2.03 4.1 4.2 SRO 5-001-AB-Ol VFD Cooling Water Pump Failure RO U-068-AL-19 20200 1A2.22 3.1 3.2 SRO S-068-AB-01 RR Pump Seal Failure RO U-068-AL-09 203000A4.02 4.1 4.1 SRO 5-068-AB-Ol Loss of 480V SD BD 2A RO U-57B-AL-06 22600 1A4.05 3.3 3.3 SRO S-57B-NO-07
LOCAJLow Level ED RU U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 RU U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-O1
NRC Scenario 5 cility: Browns Ferry NPP Scenario No.: NRC 5 Op-Test No.: 1306 Examiners:_____________________ Operators: SRO:______________________
ATC:_______________
BOP:________________
Initial Conditions: 100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.
Turnover: Return LPRM 8-49B to Operate from a Bypassed Condition lAW 2-OI-92B. Lower Power with flow to 91% for Main Turbine Valve Testing.
Event Maif. No. Event Type* Event Description No.
N-BOP 1 Return LPRM 8-49B to Operate JAW 2-OI-92B N-SRO R-ATC 2 Commence power decrease with flow to 90%
R-SRO C-BOP 3 edl8a Loss of I&C Bus A TS-SRO R-ATC 4 adOic TS-SRO ADS SRV 1-22 leaking C-BOP C-ATC VFD Cooling Water Pump 2A trips with failure of the standby 5 thl8a C-SRO pump to auto start C-ATC LOCA Recirculation Pump 2A Inboard and Outboard seal 6 R-ATC thlO/lla failure TS-SRO Two Level instruments fall high tripping Feedwater and HPCI /
7 Batch File M-ALL LOCA / ED on Reactor Level 8 edl0a C Loss of 480V SD Board 2A RHR and Core Spray Division 2 Injection Valves will not Auto 9 Batch I open 10 rcO8 C RCIC Steam Valve fails to Auto open
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor I
NRC Scenario 5 Critical Tasks Four 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 T.A.F. (-162 inches).
- 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 -162 inches.
- 4. Feedback:
Reactor water level trend.
Reactor pressure trend.
With an injection system(s) operating and the reactor shutdown and at pressure, after RPV water level drops to -162 inches, transition to Emergency Depressurization before RPV level lowers to
-180 inches.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
Water level trend.
- 3. Measured by:
Observation At least 6 SRVs opened
- 4. Feedback:
RPV pressure trend.
SRV status indications.
2
NRC Scenario 5 Critical Tasks Four To prevent au uncontrolled RPV depressurization when Reactor level cannot be restored and maintained above -162 inches, inhibit ADS.
- 1. Safety Significance:
Maintain adequate core cooling, prevent degradation of fission product barrier.
- 2. Cues:
Procedural compliance.
- 3. Measured by:
ADS logic inhibited prior to an automatic initiation.
- 4. Feedback:
RPV pressure trend.
RPV level trend.
ADS ADS LOGIC BUS A/B INHIBITED annunciator status.
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 EOI Appendix 1 7B AND Observation - RO initiates Drywell Sprays
- 4. Feedback:
Drywell and Suppression Pressure lowering RHR flow to containment 3
NRC Scenario 5 Events
- 2. ATC lowers power to 90% using recirculation flow.
- 3. The crew will respond to a momentary loss of I&C Bus A. The in-service SJAE (A) will isolate and numerous alarms will come in. The BOP operator will shift SJAEs to B or reset SJAE A and return to service JAW 2-01-66 or 2-AOI-47-3. Reactor Zone Differential pressure low will alarm and the operator will have to reset Refuel and Reactor Zone fans. When one of the SJAEs are restored high 112 will result in Off Gas, the SRO will evaluate TRM 3.7.2 and enter Condition A. The H202 analyzer will isolate requiring the SRO to evaluate TRM 3.3.11 and 3.6.2. The Drywell CAM will isolate requiring the SRO to evaluate Tech Spec 3.4.5.
- 4. During I&C Bus A loss, Main Steam Relief Valve open will alarm. When power is restored to J&C Bus A the alarm will clear but ADS SRV 1-22 will be leaking by and the acoustic monitor will indicate the leak by. SRO should enter 2-AOl-i-i, the ATC will lower power to less than 90%. When power is below 90% the BOP operator will perform 2-AOJ-1-i actions to attempt to close the SRV. SRO will refer to Tech Specs and determine TS 3.5.1 condition F is applicable.
- 5. The VFD Cooling Water Pump for the A 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.
- 6. #1 and #2 recirc pump seal failure ATC will note alarm and report #2 seal carrying full pressure. A short time later seal #2 will fail ATC will note that a small LOCA exists. ATC will trip and isolate A RR Pump JAW with 2-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 again 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.
- 7. Level instruments 208A and 208D will fail high, causing a high level trip of Main Turbine, RFPTs and HPCI. RCIC will be the only major source of high pressure injection and the steam supply valve to RCIC will fail to auto open. The Crew will maintain reactor level until the LOCA is beyond the ability of RCJC to control. The SRO will determine ED is required in order to restore level with available low pressure systems.
- 8. After the scram 480V Shutdown Board 2A will fail due to a lockout, this will prevent operation of Core Spray Division 1 System for injection. RHR Loop 1 may be used for injection but no throttle capability with exist. RHR Loop 1 will not be available for Containment cooling operation.
- 9. With Division 2 Accident logic bypassed RHR and Core Spray will not auto start on any accident signals. The crew will have to manually start pumps and open injection valves. RHR Loop 2 will be available for Containment Cooling functions until required for injection.
4
NRC Scenario 5 Terminate the scenario when the following conditions are satisfied or upon request of Lead Examiner:
Emergency Depressurization complete Reactor Level is restored SCENARIO REVIEW CHECKLIST SCENARIO NUMBER: 5 10 Total Malfunctions Inserted: List (4-8) 4 Malfunctions that occur after EOI entry: List (1-4) 4 Abnormal Events: List (1-3) 1 Major Transients: List (1-2) 2 EOIs used: List (1-3) 2 EOI Contingencies used: List (0-3) 75 Validation Time (minutes) 4 Crew Critical Tasks: (2-5)
YES Technical Specifications Exercised (Yes/No) 5
NRC Scenario 5 Scenario Tasks TASK NUMBER RO SRO Restore an LPRM from Bypass RO U-92B-NO-05 215005A4.04 3.2 3.2 Lower Power with Recirc Flow RO U-068-NO-03 SRO S-000-AD-31 2.1.23 4.3 4.4 Loss of I&C Bus A RO U-57C-AB-03 262001A2.04 3.8 4.2 SRO S-57C-AB-03 ADS SRV leaking RO U-OO1-AB-01 239002A2.03 4.1 4.2 SRO 5-001-AB-Ol VFD Cooling Water Pump Failure RO U-068-AL-19 202001A2.22 3.1 3.2 SRO S-068-AB-01 RR Pump Seal Failure RO U-068-AL-09 203000A4.02 4.1 4.1 SRO S-068-AB-01 Loss of 480V SD BD 2A RO U-57B-AL-06 22600 1A4.05 3.3 3.3 SRO S-57B-NO-07 LOCA/Low Level ED RO U-003-AL-24 29503 1EA2.04 4.6 4.8 RO U-000-EM-01 RO U-000-EM-13 SRO S-000-EM-14 SRO S-000-EM-15 SRO S-000-EM-01 6
NRC Scenario 5 Procedures Used/Referenced:
Procedure Number Procedure Title 2-0I-92B Average Power Range Monitoring 2-GOl- 100-12 Power Maneuvering 2-01-68 Reactor Recirculation System 2-AOI-57-5A Loss of I&C Bus A 2-AKP-9-8C Panel 9-8 2-XA-55-8C 2-ARP-9-7A Panel 9-7 2-XA-55-7A 2-ARP-9-6C Panel 9-6 2-XA-55-6C 2-ARP-9-7C Panel 9-7 2-XA-55-7C 2-ARP-9-3C Panel 9-3 2-XA-55-3C 2-ARP-9-3D Panel 9-3 2-XA-55-3D 2-ARP-9-53 Panel 9-53 2-XA-55-53 2-E0I-3 Secondary Containment Control 2-A0I-64-2D Group 6 Ventilation System Isolation
. Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 2-E0I Appendix-8F Isolation 2-01-66 Off-Gas System 2-A0I-66-l 0ff-Gas H2 High Technical Specifications Technical Requirements Manual 2-AOl-i-i Relief Valve Stuck open 2-01-74 Residual Heat Removal System 2-E0I-2 Primary Containment Control 2-E0I Appendix-i 8 Suppression Pool Water Inventory Removal and Makeup 2-AR.P-9-4A Panel 9-4 2-XA-55-4A 2-A0I 1A Recirc Pump Trip/Core Flow Decrease OPRMs Operable 2-AOl- 100-i Reactor Scram 2-EOI-1 RPV Control 2-EOI Appendix-8B Reopening MSIVs / Bypass Valve Operation 2-EOI- 1-C-i Alternate Level Control 2-E0I Appendix-6A Injection Subsystems Lineup Condensate 2-EOI Appendix-i 7C RHR System Operation Suppression Chamber Sprays 2-EOI Appendix-i 7B RHR System Operation Drywell Sprays 2-EOI-3-C-2 Emergency RPV Depressurization 2-EOI Appendix-5B Injection System Lineup CRD 2-EOI Appendix-7B Alternate RPV Injection System Lineup SLC System 2-EOI Appendix-6B Injection Subsystems Lineup RHR System I LPCI Mode 2-EOI Appendix-6C Injection Subsystems Lineup RHR System II LPCI Mode 2-EOI Appendix-6E Injection Subsystems Lineup Core Spray System II EPIP- 1 Emergency Classification 7
NRC Scenario 5 Console Operator Instructions A. Scenario File Summary Batch File NRC/l3O6nrc-5 ior zloOil2l ld2Ob[1] off ior zloOil2l ld2Ob[2] off ior zloOhs2l lOd2Oa[1] off Tag DG D ior zloOhs2l lOd2Oa[2] off ior zloOhs2l lOd2Oa[2j off mrfdg0ld open ior zdihs7O8a null ior zlohs7o8a[1j off Tag RBCCW 2B ior zlohs708a[2] off ior zlohs7o8a[3j off ior zlohs682a2a[1j on ior zlohs682a2a[2] off A VFD Cooling Pump Trip mrfthl8b trip trg 1 NRC/avfd trg 1= bat NRC/130605-1 A VFD Cooling Pump Trip imfth30f (e5 0)100 imfth30h (e5 60) 100 45 55 Level 8 instrument failures imfrc08 RCIC steam supply valve failure imfthl0a (e3 0)100 imfthl la (e3 60) 100 90 0 RR 2A Pump seal mrf cs09b inhibit mrf rhi 5 inhibit Div 2 accident logic bypassed ior zloil7556a off ior zloil74 1 54a off mrf edi 3 open momentary loss of I&C Bus A Batch File NRC/l3O6nrc-5-1 imfth2l (none 330) .6 600 .1 LOCA imfedl0a (none 370) Loss of 480V SD BD 2A 8
NRC Scenario 5 Preference File NRC/l3O6nrc-5 p1k 01 tog pfk 02 ann silence pfk 03 mrf swO2 align align spare RBCCW Pump p1k 04 bat NRCIl3O6nrc-5 p1k 05 imfedl8a Loss of I&C Bus A p1k 06 ior zdihs682ala[1] off VFD A Cooling Pump trip p1k 07 imfadOic 10 ADS SRV Leak by p1k 08 trg! e3 RR Pump A Seal Failure pfko9trg! e5 Loss of Feedwater p1k 10 bat NRC/l3O6nrc-5-1 LOCA and Loss of 480V SDBD2A p1k 11 mrfad0lc out p1k 12 ior xa553e10 alarm on pflcsl plks2mmfad0lc 100 p1k s3 mmfad0ic 10 p1k s4 mmfad0lc 100 p1k s5 mmfad0lc 10 p1k s6 bat appl8rhra p1k s7 bat appl8rhrb p1k s8 mrfedi3 close Scenario 5 DESCRIPTION/ACTION Simulator Setup manual Reset to IC 28 manual Bypass LPRM 8-49B restorepref NRC/i 3O6nrc-5 mrf swO2 align RBCCW wait one F3 minute and turn off RBCCW Pump 2B Simulator Setup Load Batch F4 bat NRC/i 3O6nrc-5 Simulator Setup manual Tag DG D and RBCCW Pump 2B Simulator Setup Verify file loaded, Clear alarms for Reactor_Recirc RCP required (100% -90% with flow) and RCP for Urgent Load Reduction 9
NRC Scenario 5 Simulator Event Guide:
Event 1 Normal: Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OI-92B SRO Directs BOP to return LPRM 8-49B to Operate JAW 2-OI-92B BOP Return LPRM 8-49B to Operate lAW 2-OI-92B 6.4 Returning an LPRM to Operate From a Bypassed Condition
[1] REVIEW all precautions and limitations. REFER TO Section 3.0.
[2] REFERENCE Illustration 4 to find the APRMJLPRM Channel associated with the desired LPRM to be returned to normal.
[3] At Panel 2-9-14, DEPRESS any softkey to illuminate the display on the desired APRMJLPRM channel chassis.
[4] DEPRESS the ETC softkey until BYPASS SELECTIONS illuminates on the bottom row of the display.
[5] DEPRESS BYPASS SELECTIONS softkey, enter the password, and DEPRESS ENT.
[6] SELECT the desired LPRM to be returned to service by using the left or right arrows on the sofikey board until the inverse video illuminates the correct LPRM.
[7] DEPRESS the OPERATE softkey.
[8] CHECK the BYP/HV OFF is replaced by OPERATE below the selected LPRM.
[9] DEPRESS EXIT softkey to return display to the desired bargraph.
[10] VERIFY, as a result of returning this LPRM to operate, that any alanns received on Panel_2-9-5_or on the APRM/LPRM_channel_are reset.
Contacts Reactor Engineer and informs Reactor Engineering that LPRM 8-49B is returned BOP to service.
Driver Acknowledge mformation 10
NRC Scenario 5 Simulator Event Guide:
Event 2 Reactivity: Power decrease with Recirc Flow SRO Notifies ODS of power decrease.
Directs Power decrease using Recirc Flow, JAW 2-GOI-l00-12.
[1] REVIEW all Precautions and Limitations listed in Section 3.0.
[2] VERIFY Prerequisite listed in Section 4.0 is satisfied.
[3] NOTIFY Operations Duty Specialist (ODS) and Chattanooga Load Coordinator of impending power reduction.
[4] NOTIFY Radiation Protection of purpose for power reduction, the target power level (see above note), and RECORD time Radiation Protection notified in NOMS Narrative Log.
[6] IF power is being reduced (less than 10%) for any of the following reasons:
- Weekly Control Rod Exercise
- Main Turbine Valve Testing
- Ultimate heat Sink temperature> 92.5°F
[6.1] REDUCE Recirculation flow. REFER TO 2-01-68.
[6.2] MAINTAIN Reactor thermal power within the limits shown on ICS and 0-TI-248, Station Reactor Engineer, as appropriate.
[10] PERFORM the following while reducing Reactor power:
[10.1] WHEN Reactor power is at approximately 90%, THEN REFER TO 2-01-3 and START a RFP Injection Water Pump.
ATC Lower Power w/Recirc, lAW 2-01-68, Section 6.2 Crew Calls Reactor Engineering to run thermal case when reactivity control plan is completed.
Dnver When directed by NRC, insert preference key F5 imf edi 8a Loss of I&C Bus A, followed by F7 imfadOic 10 and after 5 seconds Shift F8 mrfedl3 close NRC Two additional power decreases m Scenario, can contmue when ready will have to mismatch speeds at 1300 rpm 11
NRC Scenario 5 Simulator Event Guide:
Event 2 Reactivity: Power decrease with Recirc Flow ATC Lower Power w/Recirc, lAW 2-01-68, 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 periods for testing or maintenance).
[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-96-15A(15B).
(Otherwise N/A)
- LOWER Recirc Pump 2A using SLOW(MEDIUM)(FAST),
2-HS-96-1 7A(17B)(1 7C). (Otherwise N/A).
AND/OR
- RAISE Recirc Pump 2B using RAISE SLOW (MEDIUM), 2-HS-96-16A(16B).
(Otherwise N/A)
- LOWER Recirc Pump 2B using SLOW(MEDIUM)(FAST),
2-HS-96-1 8A(18B)(1 8C). (Otherwise N/A).
[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 pushbuttons as required.
RAISE SLOW, 2-HS-96-3 1 RAISE MEDIUM, 2-HS-96-32 LOWER SLOW, 2-HS-96-33 LOWER MEDIUM, 2-HS-96-34 LOWER FAST, 2-HS-96-35 Driver When directed by NRC, msert preference key F5 imf edi 8a Loss of I&C Bus A, followed byF7 imf adOiclO and after 5 seconds ShiftF8 mrfedl3 close.
When dispatched wait two mmutes and report Failure of 9-9 Throwover Switch, switch tripped to alternate.
NRC Two additional power decrease m Scenario, can contmue when ready 12
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are_8C-21,_6C-12,_3C-25,_7C-22_and_3D-3,_19,_and_32.
ATC Announces Power, Level, and Pressure are stable BOP Alarm 8C-21, I&C BUS A VOLTAGE ABNORMAL A. VERIFY alarm by checking the following:
- Loss of instrument power and remote position indication to Core Spray Div I and RHR Div I (Panel 9-3).
- RWCU Filter Demin A isolation.
- Reactor Building/Refuel Zone Ventilation isolation.
B. NOTIFY Unit 3 Unit Supervisor.
C. REFER TO 2-AOI-57-5A and O-GOI-300-2.
SRO Announce entry to 2-AOI-57-5A, Loss of I&C Bus A.
ATC Alarm 6C-12, RFPT GOVERNOR POWER FAILURE OR GOV ABNORMAL A. VERIFY RFPT/RFPs continue to control Reactor Water Level.
B. IF a RFPT/RFP has tripped, ThEN VERIFY other RFPTs in Automatic operation raise or lower output flow to maintain reactor water level.
C. DISPATCH personnel to UNIT 2 Auxiliary Instrument Room to PERFORM the following at Panels 2-9-48,49,50:
- CHECK Power Supply lights illuminated.
- CHECK display screens for Governor abnormal conditions.
Announced_RPV_Level_stable,_dispatches_personnel BOP Alarm 3C-25, MAIN STEAM RELIEF VALVE OPEN A. CHECK MSRV DISCHARGE TAILPIPE TEMPERATURE, 2-TR-1-1, 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.
BOP Announce Main Steam Relief Valve Open alarm cleared, but have indication on acoustic monitor of SRV partially open or leaking by. ADS SRV 1-22 NRC Action for SRV 1-22 are on PAGE 22 3C-25 alarms on a loss of I&C Bus A, when the bus re-energizes ADS SRV will show NOTE acoustic monitormg mdication of leaking by BOP operator should report to SRO and SRO enter 2-AOl-i-i These events will occur under event four I,
13
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32.
BOP Alarm 7C-22, DRYWELL/SUPPR CHAMBER 14202 ANALYZER FAILURE A. CHECK Panel 2-9-54 and 2-9-55 for abnormal indicating lights such as low flow, 142 or 02 downscale, pump off, etc.
B. LF sample pump is NOT running, THEN ATTEMPT to start pump using 2-HS-76-1 l0/S5.
C. IF sample pump will NOT start OR H2102 analyzer malfunction, THEN PLACE H2/02 Analyzer in Service per 2-01-76 section 5.4.
D. REFER TO TRM 3.3.11 and TRM Section 3.6.2.
BOP Resets H2/02 ANALYZER ISOLATION RESET, 2-HS-76-9 1 Resets Alarm on 2-MON 110, touch screen.
BOP Alarm 3D-19, DRYWELL LEAK DETECTION RADIATION DNSC A. DETERMINE cause of alarm by performing the following:
- 1. CHECK AIR PARTICULATE MONITOR CONTROLLER, 2-MON-90-50 on Panel 2-9-2 for condition bringing in alarm
- 2. DISPATCH personnel to determine which alarm is annunciating using the HELP button (REFER TO 2-01-90 for complete annunciator list).
E. REFER TO Tech Specs 3.4.4, 3.4.5, and TRM 3.3.10 for CAM LCO requirements and IMPLEMENT appropriate TS/TRM actions as required.
F. WHEN conditions permit, THEN RESET alarm per 2-01-90, Section 6.5.
BOP Determines DW Radiation Monitor Cam isolated, resets the following to restore to operation.
UPPER INBD SUPPLY ISOL VALVE RESET, 2-HS-90-254A-A LOWER INBD SUPPLY ISOL VALVE RESET, 2-HS-90-254B-A OUTBD RETURN ISOL VALVE RESET, 2-HS-90-257A-A OUTBD SUPPLY ISOL VALVE RESET, 2-HS-90-255A 1NBD RETURN ISOL VALVE RESET, 2-HS-90-257B-A 14
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A Crew Respond to numerous alarms when I&C Bus A deenergizes. The most significant of these are 8C-21, 6C-12, 3C-25, 7C-22 and 3D-3, 19, and 32.
BOP Alarms 3D-3, RX BLDG VENTILATION ABNORMAL A. IF PCIS group 6 isolation exists, THEN REFER TO 2-AOI-64-2d.
B. NOTIFY Unit Supervisors, Unit 1 and Unit 3.
C. VERIFY standby fans start.
D. DISPATCH personnel to check Bldg AP (PDIC 64-2, El 639, Rx Bldg.)
E. IF iXP is at or above -0.17 in. H20 THEN ENTER 2-EOI-3 Flowchart, 2-XA-55-3D, window 32.
BOP Alarms 3D-32, REACTOR ZONE DIFFERENTIAL PRESSURE LOW D. IF alarm is valid, THEN INFORM Unit Supervisor of 2-EOI-3 entry condition.
E. REQUEST personnel to check fans locally for any apparent problems.
F. REFER TO 2-OI-30B and PLACE standby fan in service to restore normal differential pressure.
SRO Enters 2-EOI-3, Secondary Containment Control and 2-AOI-64-2D, Group 6 Ventilation System Isolation Directs Reactor and Refuel Zone Ventilation returned to service by either 2-EOI Appendix-8F, Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation or 2-AOI-64-2D The above procedures for restoring ventilation are basically the same will describe NOTE Appendix-8F below. The only action in EOI-3 is to restore ventilation.
15
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A ATC/BOP Appendix 8F Restoring Refuel Zone and Reactor Zone Ventilation Fans Following Group 6 Isolation
- 1. VERIFY PCIS Reset.
- 2. PLACE Refuel Zone Ventilation in service as follows (Panel 2-9-25):
- a. VERIFY 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch is in OFF.
- b. PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch to SLOW A (SLOW B).
- c. CHECK two SPLYIEXH A(B) green lights above 2-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:
- 2-FCO-64-5, REFUEL ZONE SPLY OUTBD ISOL DMPR
- 2-FCO-64-6, REFUEL ZONE SPLY INBD ISOL DMPR
- 3. PLACE Reactor Zone Ventilation in service as follows (Panel 2-9-25):
- a. VERIFY 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch is in OFF.
- b. PLACE 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch in SLOW A (SLOW B).
- c. CHECK two SPLY/EXH A(B) green lights above 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch extinguish and two SPLY/EXH A(B) red lights illuminate.
- d. VERIFY OPEN the following dampers:
- 2-FCO-64-13, REACTOR ZONE SPLY OUTBD ISOL DMPR
- 2-FCO-64-14, REACTOR ZONE SPLY INBD ISOL DMPR
- 5. IF Reactor Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-1 1A, REACTOR ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).
- 6. IF Refuel Zone Fan fast speed is desired following 5 minutes of slow speed operation, THEN PLACE 2-HS-64-3A, REFUEL ZONE FANS AND DAMPERS, control switch in FAST A (FAST B).
16
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A SRO Enters 2-AOI-57-5A 4.2 Subsequent Actions
[1] VERIFY Automatic Actions have occurred.
[2] IF a Reactor Scram occurs, THEN PERFORM 2-AOl- 100-1 concurrently with this procedure.
[3] VERIFY a flow path for Condensate System, or STOP the condensate pumps/booster pumps. REFER TO 2-01-2.
[4] START Standby Gas Train(s) and CHECK Reactor Building pressure at or below 0.25 H20 vacuum (PDIC 64-1, Panel 25-215; PDIC 64-2, Panel 25-213). REFER TO 0-01-65, Section Standby Gas Treatment System Manual Initiation.
[5] VERIFY SJAE B in service to maintain condenser vacuum. REFER TO 2-01-66.
[6] IF Auto Transfer of Panel 2-9-9, Cabinet 2, failed THEN (otherwise N/A)
[7] WHEN Reactor water level is normal, THEN RESET PCIS Group 6 inboard isolation and RETURN the affected systems to service or standby readiness.
REFER TO 2-AOl- 100-1, if a Reactor Scram occurred, otherwise REFER TO 2-AOI-64-2D.
SRO Directs restoration of Reactor Building DP, should restore Ventilation JAW Appendix-8F or 2-AOI-64-2D._May_call Unit_1_to_start_Standby_Gas_Fans SRO Directs restoration of SJAE, lAW 2-01-66 hard card BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card 17
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A BOP Restores SJAE to service, Standby SJAE System Lineup Hard Card
[1] VERIFY RESET Off-Gas isolation using 2-HS-90-155, OG OUTLET/DRAiN ISOLATION VLVS.
NOTE With power back to I&C Bus A, once RO resets 2-HS-90-155, can place SJAE A back m service or can transfer to SJAE B. All steps are listed below for either.
[2] VERIFY OPEN the following valves:
- 2-HS-66-l 1(15), SJAE 2A(2B) INLET VALVE.
- 2-HS-l-155A(156A), STEAM TO SJAE 2A(2B).
[3] VERIFY in AUTO/OPEN 2-HS-66-14(18), SJAE 2A(2B) OG OUTLET VALVE.
[4] PLACE 2-HS-l-150(152), SJAE 2A(2B) PRESS CONTROLLER, in CLOSE and then in OPEN.
[5] VERIFY OPEN the following valves (red light illuminated):
- 2-PCV-1-151/166 (153/167), STEAM TO SJAE 2A(2B) STAGES 1,2, AND 3.
- 2-FCV-1-150(152), SJAE 2A(2B) INTMD CONDENSER DRAIN.
[6] MONITOR hotwell pressure as indicated on recorder 2-XR-2-2, HOTWELL TEMP AND PRESS, on Panel 2-9-6.
[7] FOR the SJAE not being placed in service, VERIFY CLOSED the following valves:
- 2-HS-1-152(l50), SJAE 2B(2A) PRESSURE CONTROLLER.
- 2-HS-1-156A(155A) STEAM TO SJAE 2B(2A)
BOP Acknowledge Panel 2-9-53 Alarms, Report high hydrogen levels 53-3 and 13, HIGH OFFGAS °A H2 TRAIN A, and HIGH OFFGAS % 112 TRAIN BB 18
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A BOP Report high hydrogen levels 53-3 and 13, HIGH OFFGAS % H2 TRAIN A, and HIGH OFFGAS % H2 TRAiN BB A. CHECK Off-gas Hydrogen Analyzer, 2-H2R-66-96 (CH 1) on Panel 2-9-53 to verify H2 concentration.
B. IF alarm is valid, THEN REFER TO 2-AOI-66-1.
SRO Enters 2-AOI-66-1, Off-Gas H2 High BOP/ATC [1] PLACE both OFFGAS TRAiN A(B) AUTO CHANNEL CHECK /
BYPASS control switches, 2-HS-066-1007 and 1008, on OFFGAS SAMPLE PANEL, 2-LPNL-925-0588, in BYPASS to assure continuous availability of hydrogen monitoring.
[2] IF HWC System injection is in service, THEN (otherwise N/A)
[3] VERIFY proper operation of in service SJAE.
[4] IF hydrogen concentration is greater than or equal to 4%, THEN REFER TO TRM 3.7.2.
[10] MONITOR the following parameters at Control Room Panel 9-53 and 9-8:
RECOMBINER 2A12B TEMPERATURE, 2-TRS-66-77, for abnormal trend; either rising or lowering.
NOTE 112 concentration will rise to 8 to 12% and return to a normal value of less than 1%
19
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A SRO Tech Specs for Loss of I&C Bus A For Drywell CAM 3.4.5 RCS Leakage Detection Instrumentation LCO 3.4.5 The following RCS leakage detection instrumentation shall be OPERABLE:
- a. Drywell floor drain sump monitoring system; and
- b. One channel of either primary contaimnent atmospheric particulate or atmospheric gaseous monitoring system.
APPLICABILITY: MODES 1,2, and 3.
Condition B: Required primary containment atmospheric monitoring system inoperable.
Required Action B. 1: Analyze grab samples of primary containment atmosphere.
Completion Time: Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action B.2: Restore required primary containment atmospheric monitoring system to OPERABLE status.
Completion Time: 30 days For High H2 TR 3.7.2 Airborne Effluents LCO 3.7.2 Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to 4% by volume.
APPLICABILITY: During main condenser offgas treatment system operation Condition A: With the concentration of hydrogen >4% by volume.
Required Action A. 1: Restore the concentration to within the limit.
Completion Time: 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Below is event 4 which started with the failure of I&C Bus A, depending on SRO priorities NOTE may have addressed SRV first and then restoration from Buslpss.
20
NRC Scenario 5 Simulator Event Guide:
Event 3 Component: Loss of I&C Bus A SRO Tech Specs for Loss of I&C Bus A For H202 Monitor TR 3.3.11 Hydrogen Monitoring Instrumentation LCO 3.3.11 The primary containment hydrogen analyzer shall be OPERABLE APPLICABILITY: MODE 1 during the time period
- a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is> 15% RTP following startup, to
- b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.
Condition A: Primary containment hydrogen analyzer inoperable.
Required Action A. 1: Restore primary containment analyzer to OPERABLE status.
Completion Time: 7 days TR 3.6.2 Oxygen Concentration Monitor LCO 3.6.2 The Primary Containment oxygen concentration monitor shall be OPERABLE.
APPLICABILITY: MODE 1 during the time period
- a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is> 15% RTP following startup, to
- b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.
Condition A: Primary containment oxygen concentration monitor inoperable.
Required Action A. 1: Begin alternate sampling and analyze results.
Completion Time: Immediately AND Once per 7 days thereafter.
Below is event 4 which started with the failure of I&C Bus A, depending on SRO priorities NOTE may have addressed SRV first and then restoration from Bus loss 21
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking Event 4 started with the failure of I&C Bus A, dependmg on SRO priontles may have NOTE addressed SRV first and then restoration from Bus loss 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-l on Panel 2-9-47.
ATC [2] IF relief valve transient occurred while operating above 90% power, THEN REDUCE reactor power to 90% RTP with recirc flow.
BOP [3] WHILE OBSERVING the indications for the affected Relief valve on the Acoustic Monitor; CYCLE the affected relief valve control switch several times as required:
CLOSE to OPEN to CLOSE positions
[4] IF all SRVs are CLOSED, THEN CONTINUE at Step 4.2.4. (N/A) 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. (N/A)
[4] PLACE MSRV AUTO ACTUATION LOGIC iNHIBIT, 2-XS-l-202 in INHIBIT:
[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-l-202, in AUTO.
[7] IF the SRV valve did not close, THEN PERFORM the appropriate section from table below.
RELIEF STEP Switch Breaker Fuse VALVE Number Location Location Location SRV 1-22 Step 4.2.3[2] Panel 25-32 Multiple Panel 25-32 22
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking Actions for SRV 1-22, wait two minutes and for taking control at panel 25-32 Preference key Fl 2 and Fl 1 mrf adO 1 c out, for cycling SRV preference key shift F2 open, shift F3
- close to 10, shift F4 open, shift F5 close to 10.
Driver Contact control room and determme if valve closed. When told to remove power preference key Fl 1. When back to normal at panel 25-32 delete override for annunciator xa553e 10.
When told to power backup sri 1-22 mrfad0lc IN.
Driver [2] IF 2-PCV- 1-22 is NOT closed, THEN PERFORM the following:
[2.1] On Panel 2-25-32 PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-1-22 in EMERG position.
[2.2] IF the SRV does NOT close, THEN PERFORM the following while OBSERVING the indications for the 2-PCV- 1-22 on the Acoustic Monitor:
- CYCLE the MAIN STM LINE B RELIEF VALVE, 2-HS-l-22C to the following positions several times. CLOSE/AUTO to OPEN to CLOSE/AUTO
[2.3] IF the SRV does NOT close, THEN PERFORM the following:
A. VERIFY the MAIN STM LINE B RELIEF VALVE, 2-HS-l-22C, in the CLOSE/AUTO position.
B. PLACE the transfer switch associated MAIN STM LINE B RELIEF VALVE XFR, 2-XS-1-22 in NORM position.
Driver
[2.4] IF the SRV does NOT close, THEN REMOVE the power from 2-PCV-1-22 by performing one of the following:
A. OPEN the following breakers (Preferred method)
[2.5] IF the valve does NOT close, THEN CLOSE the breakers or REINSTALL fuses removed in Step 4.2.3 [2.4].
BOP [2.6] CONTINUE at Step 4.2.4.
23
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP [2.6] CONTINUE at Step 4.2.4.
4.2.4 Other Actions and Documentation
[1] NOTIFY Reactor Engineering of current conditions.
[2] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s).
[3] REFER TO Technical Specifications Sections 3.5.1 and 3.4.3 for Automatic Depressurization System and relief valve operability requirements.
[4] INITIATE suppression pool cooling as necessary to maintain suppression pool temperature less than 95°F.
[5] IF the relief valve can NOT be closed AND suppression pool temperature CANNOT be maintained less than or equal to 95°F, THEN PLACE the reactor in Mode 4 in accordance with 2-GOI-100-12A.
[6] DOCUMENT actions taken and INITIATE Work Order (WO) for the valve.
SRO Directs Suppression Pool Cooling lAW 2-01-74 BOP Initiates Pool Cooling as directed SRO Refers to Tech Specs 3.5.1 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 with reactor steam dome pressure 150 psig.
Condition E: One ADS valve inoperable.
Required Action E. 1: Restore ADS valve to OPERABLE status.
Completion Time: 14 days.
BOP/ATC Inform SRO when Suppression Pool Level meets EOI-2 entry requirements SRO Enter EOI-2 on Suppression Pool Level NOTE One RHR Pump will almost mamtam pool temperature dependmg on reactor power, Do NOT expect pool temperature to exceed 95°F.
24
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Coolmg as directed 8.5 Initiation of Loop 1(U) Suppression Pool Cooling CAUTION PSA concerns with ERR in Suppression Pool Cooling Mode with a LOCA and a LOSP identify that severe water hammer may occur during the pump restart. Therefore, the following guidelines should be used to try and maintain the system below the PSA Risk Assessment goals:
- RHR in suppression pool cooling should be minimized.
- Two Loops of RHR in suppression pooi cooling should be minimized.
- Use two pumps per loop, if needed, to minimize total time spent in suppression pool cooling.
Suppression pooi cooling run times are tracked in 2-SR-2 to ensure risk assessment goals are not exceeded.
NOTES
- 1) Suppression Pool Cooling is required to be initiated whenever necessary to maintain suppression pool temperature less than 95°F or when directed by other procedures.
[1] VERIFY RHR Loop 1(11) is in Standby Readiness. REFER TO Section 4.0
[2] REVIEW the precautions and limitations in Section 3.0.
[31 NOTIFY other units of placing Loop 1(11) of RHR in suppression pool cooling, the subsequent start of common equipment (i.e., R}IRSW pumps) and associated alarms are to be expected.
[4] NOTIFY Radiation Protection for impending action to initiate Suppression Pool Cooling. RECORD name and time of Radiation Protection representative notified in NOMS narrative log
[5] IF possible, THEN BEFORE placing RHRSW in service, NOTIFY Chemistry that RHRSW sampling is to be initiated (RHRSW sampling requirements).
[6] VERIFY at least one RHRSW Pump is operating on each EECW Header.
BOP Makes Plant announcements prior to starting RNRSW Pumps and RHR Pumps.
NOTE One RHR Pump will almost mamtam pool temperature constant dependmg on reactor power, Do NOT expect pool temperature to exceed 95°F.
25
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed
[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 performing one the following:
[7.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, (RHRSW Pump A(C) and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow) REFER TO 0-01-23.
[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 -3 6(42), RHR HTX 2A(2C)
RHRSW FLOW. E
[7.3] VERIFY CLOSED RHR SYS I LPCI INBD INJECT VALVE, 2-FCV-74-53.
[7.4] IF NO RHR PUMP (1A OR 1C) is operating in Suppression Pool Cooling, THEN VERIFY CLOSED RHR SYS I SUPPR POOL CLG/TEST VALVE, 2-FCV-74-59.
[7.5] VERIFY CLOSED RHR SYS I SUPPR CIJBR 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 CHI3R/POOL ISOL VLV, 2-FCV-74-57.
[7.8] IF desired to operate without the Drywell DP Compressor, THEN:
[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 in 1 2 Operation Loop Flow 7,000 to <13,000 gpm &
10,000 gpm & Blue Blue light light illuminated illuminated 26
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Cooling as directed
[7.11] IF desired to raise Suppression Pool Cooling flow and only one Loop I pump is in service, THEN PLACE the second Loop I RHR Pump and Heat Exchanger in service by REPERFORMING Step 8.5 [7] for the second pump.
[8] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:
- Amber breaker spring charged light on,
- Closing spring target indicates charged.
[10] PLACE RF1R Pump and Heat Exchanger B(D) in service as follows:
[10.2] ESTABLISH RHRSW flow by one of the following methods:
[10.2.1] REQUEST another unit establish minimum flow for Pump which will be utilized for Suppression Pool Cooling, and establish minimum flow. (between 4000 and 4500 gpm RHRSW flow)
REFER TO 0-01-23.
[10.2.2] THROTTLE OPEN RHR HX 2B(2D) RFIRSW OUTLET VLV, 2-FCV-23-46(52), as required for cooling (if another is maintaining minimum flow) and/or to maintain between 4000 and 4500 gpm RHRSW flow as indicated on 2-FI-23-48(54), RHR HX 2B(2D)
RHRSW FLOW.
[10.3] VERIFY CLOSED RHR SYS II LPCI 1NBD iNJECT VALVE, 2-FCV-74-67.
[10.4] IF NO RHR PUMP (lB or 1D) is operating in Suppression Pool Cooling, THEN 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 SUPPR CH13R/POOL ISOL VLV, 2-FCV-74-7 1.
Driver For Pump motor breaker amber spring charge light ON and closing spring target indicates charged.
27
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Initiates Pool Coolmg as directed
[10.8] IF desired to operate without the Drywell DP Compressor, THEN:
[10.9] START RHR PUMP 2B(2D) using 2-HS-74-28A(39A).
[10.10] THROTTLE RHR SYS II SUPPR POOL CLG/TEST VLV, 2-FCV-74-73, to maintain RFIR flow within limits, as indicated on RHR SYS II CTMT FLOW, 2-FI-74-70.
RHR Pumps in 1 2 Operation Loop Flow 7,000 to <13,000 gpm &
10,000 gpm & Blue Blue light light illuminated illuminated
[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. REPERFORM Step 8.5[10] for the second pump.
[12] CHECK pump motor breaker charging spring recharged for all 4160 Volt pump motors operated in this section, as follows:
Amber breaker spring charged light on, Closing_spring target_indicates_charged.
SRO Tech Spec 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be.? -6.25 inches with and -7.25 inches without differential pressure control and -1.0 inches.
APPLICABILITY: MODES 1,2, and 3.
Condition A: Suppression pool water level not within limits.
Required Action A. 1: Restore suppression pool water level to within limits.
Completion Time: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Note AS the SRV remams open addmg mventory to suppression pool, pool level spec will be appropriate.
28
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking SRO Enter EOI-2 on Suppression Pool Level SRO PC/Il Verify H202 analyzer in service (APP 19)
When H2 is detected in PC (2.4% on control room indicators continue, does not continue SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue PC/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI-64-l),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, does not continue DW/T Monitor and control Drywell temperature below 160F using available Drywell cooling Can Drywell Temperature be maintained below 1 60F, YES SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18)
Can suppr p1 lvl be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES SRO Direct Appendix 18, Suppression Pool Water Inventory Removal And Makeup BOP Calls for Operator to perform field action of Appendix 18 29
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Calls for Operator to perform field action of Appendix 18
- 3. IF Directed by SRO, THEN REMOVE water from Suppression Pool as follows:
- a. DISPATCH personnel to perform the following (Unit 2 RB, El 519 fi, Torus Area):
- 1) VERIFY OPEN 2-SHV-074-0786A(B), RHR DR PUMP 2A(2B)
DISCH TO MN CNDR/RW SOy.
- 2) OPEN the following valves:
- 3) UNLOCK and OPEN 2-SHV-074-0765A(B), RHR DR PUMP 2A(2B)
DISCH SOy.
- 4) NOTIFY Unit Operator that RHR Drain Pump 2A(2B) is lined up to remove water from Suppression Pool.
- 5) REMAIN at torus area UNTIL Unit 2 Operator directs starting of R}IR Drain Pump 2A(2B).
- b. IF Main Condenser is desired drain path, THEN OPEN 2-FCV-74-62, RHR MAIN CNDR FLUSH VALVE.
- c. IF Radwaste is desired drain path, THEN PERFORM the following:
- 1) ESTABLISH communications with Radwaste.
- 2) OPEN 2-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE
Driver When dispatched to remove water from the suppression pool, wait 10 mmutes and call and report aligned step 4 above, when the operator calls you back to start the RHR Drain Pump Shift F6 for bat appl 8rhra and Shift F7 for bat appi 8rhrb 30
NRC Scenario 5 Simulator Event Guide:
Event 4 Component: ADS SRV 1-22 leaking BOP Appendix 18
- 4. WHEN Suppression Pool level reaches -5.5 in., THEN SECURE RHR Drain System as follows:
- a. DISPATCH personnel to STOP the Drain System as follows (Unit 2 RB, El 519 ft, Torus Area):
- 1) STOP RHR Drain Pump 2A(2B).
- 2) CLOSE the following valves:
- 3) CLOSE and LOCK 2-SHV-074-0765A(B), RHR DR PUMP 2A(2B)
DISCH SOy.
- b. CLOSE 2-FCV-74-108, RHR DR PUMP 2AJB DISCH HUR VALVE.
- c. VERIFY CLOSED 2-FCV-74-62, RI-IR MAiN CNDR FLUSH VALVE.
d VERIFY CLOSED 2-FCV-74-63, R}IR RADWASTE SYS FLUSH VALVE Driver When directed by NRC for VFD CooiingPumptrip, Preference Key F6 31
NRC Scenario 5 Simulator Event Guide:
Event 5 Component: VFD Cooling Water Pump 2A trip Driver When directed by NRC for VFD Coolmg Pump trip, Preference Key F6 ATC Respond to the following alarms, 4A-12, 4A-28 and 4A-32 ATC Report Trip of Recirc Drive 2A Cooling Pump 2A 1, and failure of standby pump to start Alarm 4A-12, RECIRC DRIVE 2A COOLANT FLOW LOW Automatic Action Standby RECIRC DRIVE cooling water pump will auto start.
A. VERIFY REC1RC DRIVE cooling water pump running.
B. DISPATCH personnel to the RECIRC DRIVE to check the operation of the Recirc Drive cooling water system.
Alarm 4A-28, RECIRC DRIVE 2A PROCESS ALARM A. IF 2-XA-55-4B Window 28 is also in alarm, THEN (N/A)
B. Refer to ICS screen VFDAAL and determine cause of alarm Alarm 4A-32, RECIRC DRIVE 2A DRIVE ALARM A. REFER TO ICS Group Display GD @VFDADA and DETERMINE cause of alarm.
B. IF a problem with the cooling water system is indicated, THEN VERIFY proper operation of cooling water system.
ATC Start Standby Recirc Drive 2A Cooling Pump 2A2, dispatches personnel to investigate Wait 4 mmutes after dispatched, THEN report tripped VFD Pump 2A1 is hot to the touch, mtemal bkr closed, 480 volt bkr tripped (480 V SD BD 2A 5C)
Driver When directed by NRC initiate RR Pump 2A Seal Failure Preference Key F8 32
NRC Scenario 5 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure Driver When directed by NRC initiate BR Pump 2A Seal Failure Preference Key F8 ATC Respond to alarm 4A-25, RECIRC PUMP A NO. 1 SEAL LEAKAGE ABN A. DETERMINE initiating cause by comparing No. 1 and 2 seal cavity pressure indicators on Panel 2-9-4 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.
B. RECORD pump seal parameters hourly on Attachment 1, ATC Report of failure of number 1 seal or inner seal Respond to alarm 4A-18, RECIRC PUMP A NO.2 SEAL LEAKAGE HIGH A. COMPARE No. 2 cavity pressure indicator (2-PI-68-63A) to No. 1 cavity pressure indicator (2-PI-68-64A). No. 2 seal degradation is indicated if the pressure at No. 2 seal is less than 50% of the pressure at No. 1 seal.
ATC Reports the second seal is failed both pressure indicators trending toward zero psig.
C. IF dual seal failure is indicated, THEN
- 1. SHUTDOWN Recirc Pump 2A by depressing RECIRC DRIVE 2A SHUTDOWN, 2-HS 19.
- 2. VERIFY TRIPPED, RECIRC DRIVE 2A NORMAL FEEDER, 2-HS-57-17.
- 3. VERIFY TRIPPED, RECIRC DRIVE 2A ALTERNATE FEEDER, 2-HS-57-15.
- 4. CLOSE Recirculation Pump 2A suction valve.
- 5. CLOSE Recirculation Pump 2A discharge valve.
- 6. REFER TO 2-AOI-68-1A or 2-AOI-68-1B AND 2-01-68.
- 7. DISPATCH personnel to secure Recirculation Pump 2A seal water.
ATC Trips RR Pump 2A and closes suction and discharge valves Reports rising Drywell Pressure, reports DW Pressure stable once valves are closed SRO Enters 2-AOI 1 A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable 33
NRC Scenario 5 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure Enters 2-AO1-68-1A, Recirc Pump Trip/Core Flow Decrease OPRMs Operable SRO Identifies a trigger value as to when to insert a Reactor SCRAM on Drywell Pressure if Drywell Pressure continues to rise.
4.2 Subsequent Actions
[1] IF both Recirc Pumps are tripped in modes 1 or 2, THEN (N/A),
[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 IMMEDIATELY take actions to INSERT control rods to less than 95.2% loadline. Refer to O-TI-464, Reactivity Control Plan Development and Implementation.
[4] RAISE core flow to greater than 45%. REFER TO 2-01-6 8.
[5] INSERT control rods to exit regions if not afready exited. Refer to 0-TI-464, Reactivity Control Plan Development and Implementation.
[6] MAINTAIN operating Recirc pump flow less than 46,600 gpm. Refer to 2-01-68.
[7] WHEN plant conditions allow, THEN, MAINTAIN operating jet pump ioop flow greater than 41 x 106 lbmlhr (2-FI-68-46 or 2-FI-68-48).
SRO Direct inserting control rods lAW Urgent Load Reduction and Rod Shove Sheets ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Driver When dispatched to isolate seal water waitS mmutes and then mrf rdO3 close and report closed 34
NRC Scenario 5 Simulator Event Guide:
Event 6 Component: LOCA Recirculation Pump 2A Inboard and Outboard seal failure ATC Inserts Control Rods to Exit Region II of the Power to Flow Map Inserts all of the following Control Rods to lower rod line to < 95%:
Control Rods 22-3 1, 30-39, 38-3 1, 30-23 from 08 to 00 Control Rods 22-39, 38-39, 3 8-23, 22-23 from 16 to 00 Control Rod 30-31 from 22 to 00 Control Rods 14-31, 30-47, 46-31, 30-15 from 48 to 00 ATC Raise Speed of RR Pump B until core flow is 46 to 50% and ensure RR Pump B drive flow is below 46,600 gpm Report Exit from Region II of Power to Flow Map SRO Tech Spec 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.
OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:
- a. LCO 3.2.1, AVERAGE PLANAR LiNEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;
single ioop 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 ioop operation; APPLICABILITY: MODES 1 and 2.
Condition A: Requirements of the LCO not met.
Required Action A. 1: Satisfy the requirements of the LCO.
Completion Time: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Driver When directed by NRC, Preference Key F9, Level Instruments Fail high When mode switch is out of run or NOT m run Preference Key FlO 35
NRC Scenario 5 Simulator Event Guide:
Event 7 Major: Loss of Feedwater and HPCI ATC Report Trip of Main Turbine and RFPTs and Reactor Scram ATC 4.1 Immediate Actions
[1] DEPRESS REACTOR SCRAM A and B, 2-HS-99-5AJS3A and 2-HS-99-5A/S3B, on Panel 2-9-5.
[2] IF scram is due to a loss of RPS, THEN PLACE REACTOR MODE SWITCH, 2-HS-99-5A-Sl, in START & HOT STBY AND PAUSE for approximately 5 seconds. (Otherwise N/A), Step is NA
[3] REFUEL MODE ONE ROD PERMISSIVE light check:
[3.1] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in REFUEL.
[3.2] CHECK illuminated REFUEL MODE ONE ROD PERMISSIVE light, 2-XI-85-46.
[3.3] II? 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) Step is NA
[4] PLACE REACTOR MODE SWITCH, 2-HS-99-5A-S1, in SHUTDOWN.
[5] REPORT the following status to the US:
- Reactor Scram
- Mode Switch is in Shutdown
- All rods in or rods out
- Reactor Level and trend (recovering or lowering).
- Reactor pressure and trend
- MS1V position (Open or Closed)
- Power level
[1] ANNOUNCE Reactor SCRAM over PA system.
[3] DRIVE in all IRMs and SRMs from Panel 2-9-5 as time and conditions permit.
[3.1] DOWNRANGE IRMs as necessary to follow power as it lowers.
[5] MONITOR and CONTROL Reactor Water Level between +2 and +51 , or as directed by US, as follows:
ATC/BOP Open RCIC Steam Supply Valve to start RCIC for Level Control, RCIC has received an Auto Start signal but the Steam Supply Valve failed to Open.
Driver When mode switch is in out of run Preference Key Fl 0
NRC Scenario 5 Simulator Event Guide:
Event 7 Major: Loss of Feedwater and HPCI SRO Enters EOI-l on RPV Water Level SRO EOI- I (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 or has been required THEN exit RCIP and enter C2 Emergency Depressurization? NO -
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 Stabilize RPV pressure below 1073 psig with the main turbine bypass valves (APPX 8B)
SRO Direct a pressure band, may direct a cooldown lAW Appendix 8B ATC/BOP Maintain directed pressure with Bypass Valves lAW Appendix 8B, Reopening MSIVs /
Bypass Valve Operation 37
NRC Scenario 5 Simulator Event Guide:
Event 7 Major: Loss of Feedwater and HPCI ATC/BOP Maintain directed pressure with Bypass Valves JAW Appendix 8B, Reopening MSIVs /
Bypass Valve Operation IF pressure control with bypass valves is desired and MSIVs are open, THEN proceed to step 10.
- 10. Verify Condenser Vacuum is greater than 7
- 11. IF manual opening of Bypass Valves is desired, THEN perform the following step:
- a. Depress the Bypass Valve Opening Jack Raise Pushbutton, 2-HS 13 OB to slowly open the Bypass Valves.
- b. Adjust BPV Positio0n as necessary by using the raise, 2-HS-47-130B and Lower 2-HS-47-130A pushbuttons to maintain desired cooldown rate.
- 12. IF EHC Auto Cooldown is desired, THEN perform the following steps:
- a. Verify EHC is in Pressure Control using 2-HS-47-204
- b. Verify Bypass Valve Demand is set at ZERO
- c. On the EHC Work Station on Panel 2-9-7:
- 1) Select Main Menu from the toolbar at bottom of the screen.
- 3) Select Auto Cooldown from list of function on the screen.
- d. On the Auto Cooldown Display Screen
- 1) Check the following are displayed.
- Turbine Tripped or All Valves Closed indicates reset
- RX Press Ctrl indicates reset
- 2) Select the block above the FiNAL PRESSURE TARGET
- 3) Enter the desired pressure using the display screen or keyboard
- 4) Select OK
- 5) Depress the START button
- 6) When Are You Sure You Want to Initiate Auto Cooldown? appears, Select YES
- 7) Check the following:
- EHC PRESSURE SETPO1NT, 2-PI-47-162, is lowering
- EHC AUTO COOLDOWN displays IN PROCESS 38
NRC Scenario 5 Simulator Event Guide:
Event 7 Major: Loss of Feedwater and HPCI SRO EOI- 1 (Reactor Level)
Monitor and Control Reactor Water Level.
Directs_Verification_of PCIS_isolations.
ATC/BOP Verifies PCIS isolations.
SRO IF It has NOT been determined that the reactor will remain subcritical without boron under all condition THEN EXIT RC/L NO -
RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO -
Restore and Maintain RPV water level between +2 inches and +51 inches with RCIC (APPX 5C)
ATC/BOP RCIC failed to auto start, Opens RCIC Steam Supply Valve and verifies RCIC operation.
- 1. VERIFY 2-FIC-71-36A, RCIC SYSTEM FLOW/CONTROL, controller in AUTO with setpoint at 600 gpm.
- 8. CHECK proper RCIC operation by observing the following:
- a. RCIC Turbine speed accelerates above 2100 rpm.
- c. 2-FCV-71-40, RCIC Testable Check Vlv, opens by observing 2-ZI-7 1 -40A, DISC POSITION, red light illuminated.
- d. 2-FCV-71-34, RCIC PUMP M1N FLOW VALVE, closes as flow rises above 120 gpm.
39
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Report rising Drywell Pressure and Temperature ATC/BOP Report loss of 480V SD BD 2A and 480V RMOV BD 2A ATC/BOP Dispatch personnel to investigate loss of Board SRO Re-Enter EOI-2 on High DW Pressure and Temperature ATC/BOP IF RHR Loop 1 was in Pool Cooling for leaking SRV, then operators report that RHR Loop 1 remains in Pool cooling.
NOTE RHR Loop 1 has lost power to almost all valves but NO valves reposition on board loss There is no throttle path for RHR Loop 1 and NO discharge path to RPV for Core Spray Loop 1 SRO EOI-2 on High Drywell Pressure DW/T Monitor and control Drywell temperature below 1 60F using available Drywell cooling Can Drywell Temperature be maintained below 160F, NO Operate all available drywell cooling Before Drywell Temperature rises to 200F enter EOI- 1 and Scram Reactor, Completed Before Drywell Temperature rises to 280F continue Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)
Driver When dispatched for Board loss, wait 4 minutes and report overcurrent trip of supply breaker on 480V SD BD 2A. If requested to energize 480V RMOV BD 2A from alternate supply, wait 3 minutes and report that unable to restore power to Board 40
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure Pc/P Monitor and control PC pressure below 2.4 psig using the Vent System (AOI 1),
PC pressure above 2.4 psig unable to vent When PC pressure CANNOT be maintained below 2.4 psig, Continues Before suppression chamber pressure rises to 12 psig continue, Continues Initiate suppression chamber sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 1 7C), Direct Appendix 17C When suppression chamber pressure exceeds 12 psig, Continues Is Suppression Pool Level below 19 Feet, YES Is Drywell Temperatures and Pressures within the safe area of curve 5, YES Directs Shutdown of Recirc Pumps and Drywell Blowers Initiate DW Sprays using only those pumps NOT required to assure adequate core cooling by continuous injection (App 17B)
When Suppression chamber pressure CANNOT be maintained in the safe area of Curve 5 Continue. Does not continue SRO Enters EOI-2 on High Drywell Pressure Pc/Il VerifS H2O2 analyzer in service (APP 19)
When H2 is detected in PC (2.4% on control room indicators continue, does not continue 41
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA SRO Enters EOI-2 on High Drywell Pressure SPIT MONITOR and CONTROL suppr p1 temp below 95°F using available suppr p1 cooling (APPX 17A), Pool Temp below 95° WHEN suppr p1 temp CANNOT be maintained below 95°F, does not continue Enters EOI-2 on High Drywell Pressure SPIL MONITOR and CONTROL suppr p1 lvi between -1 in. and -6 in. (APPX 18)
Can suppr p1 lvi be maintained above -6 in., YES Can suppr p1 lvl be maintained below -1 in., YES
, II ONLY 42
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17C, RHR System Operation Suppression Chamber Sprays 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 auto open signal as necessary by PLACING 2-HS-74-155B, LPCI SYS II OUTBD NJ VLV BYPASS SEL in BYPASS.
- 3. IF Directed by SRO to spray the Suppression Chamber using Standby Coolant Supply, THEN CONTINUE in this procedure at Step 7.
- 4. IF Directed by SRO to spray the Suppression Chamber using Fire Protection, THEN CONTINUE in this procedure at Step 8.
- 5. INITIATE Suppression Chamber Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN PLACE keylock switch 2-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 2.FCV-74-67, RHR SYS II INBD INJECT VALVE, is OPEN, THEN VERiFY CLOSED 2-FCV-74-66, RHR SYS II OUTBD INJECT VALVE.
- e. VERIFY OPERATING the desired RHR System II pump(s) for Suppression Chamber Spray.
- f. VERIFY OPEN 2-FCV-74-71, RUR SYS II SUPPR CUBRIPOOL ISOL VLV.
- g. OPEN 2-FCV-74-72, RI-IR SYS II SUPPR CHBR SPRAY VALVE.
ATCIBOP Aligns RHR Loop II Pumps in Suppression Chamber Sprays 43
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP EOI APPENIMX-17C, RHR System Operation Suppression Chamber Sprays
- h. IF RHR System II is operating ONLY in Suppression Chamber Spray mode, THEN CONTINUE in this procedure at Step 5 .k.
- i. VERIFY CLOSED 2-FCV-74-30, RHR SYSTEM II MIN FLOW VALVE.
- j. RAISE system flow by placing the second RHR System II pump in service as necessary.
- 1. VERIFY R}{RSW pump supplying desired R}{R Heat Exchanger(s).
- m. THROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm flow:
ATC/BOP Aligns R}IR Loop fl Pumps in Suppression Chamber Sprays 44
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA
= ATC/BOP 2-EOI APPENDIX-17B, RHR System Operation Drywell Sprays
- 1. BEFORE Drywell pressure drops below 0 psig, CONTINUE in this procedure at Step 7.
- 2. IF Adequate core cooling is assured, OR Directed to spray the Drywell irrespective of adequate core cooling, TFIEN BYPASS LPCI injection valve auto open signal as necessary by PLACING 2-HS-74-155B, LPCI SYS II OUTBD INJ VLV BYPASS SEL in BYPASS.
- 3. VERIFY Recirc Pumps and Drywell Blowers shutdown.
- 4. IF Directed by SRO to spray the Drywell using Standby Coolant supply, THEN CONTINUE in this procedure at Step 8.
- 5. IF Directed by SRO to spray the Drywell using Fire Protection, THEN CONTINUE in this procedure at Step 9.
- 6. INITIATE Drywell Sprays as follows:
- b. IF EITHER of the following exists:
- Directed by SRO, THEN PLACE keylock switch 2-XS-74-130, RHR SYS II LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRIDE.
- d. IF 2.-FCV-74-67, RHR SYS U LPCI INBD iNJECT VALVE, is OPEN, THEN VERIFY CLOSED 2-FCV-74-66, RHR SYS II LPCI OUTBD iNJECT VALVE.
- e. VERIFY OPERATING the desired System II RHR pump(s) for Drywell Spray.
- f. OPEN the following valves:
ATC/BOP Aligns RHR Loop II Pumps in Drywell Sprays 45
NRC Scenario 5 Simulator Event Guide:
Event S Component: Loss 480V SD BD 2A and LOCA ATC/BOP 2-EOI APPENDIX-17B, RRR System Operation Drywell Sprays
- g. VERiFY CLOSED 2-.FCV-74-30, RHR SYSTEM II MN FLOW VALVE.
- h. IF Additional Drywell Spray flow is necessary, THEN PLACE the second System II RHR Pump in service.
- k. ThROTTLE the following in-service RHRSW outlet valves to obtain between 1350 and 4500 gpm RHRSW flow:
- 1. NOTIFY Chemistry that R}{RSW is aligned to in-service RHR Heat Exchangers.
- 7. WHEN EITHER of the following exists:
- Before drywell pressure drops below 0 psig, OR
- Directed by SRO to stop Drywell Sprays, THEN STOP Drywell Sprays as follows:
- a. VERIFY CLOSED the following valves:
- 2-FCV-74-100, RI{R SYS I U-2 DISCH XT1E
- b. VERIFY OPEN 2-FCV-74-30, RI-JR SYSTEM II MN FLOW VALVE.
- c. iF RI{R operation is desired in ANY other mode, THEN EXIT this EOI Appendix.
- d. STOP R}IR Pumps.
ATCIBOP Aligns RHR Loop II Pumps in Drywell Sprays 46
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATCIBOP Report lowermg RPV water level unable to maintain with RCIC SRO EOI- 1 Reactor Level RPV water level drops below -120 inches OR The ADS timer has initiated NO IF RPV water level CANNOT be restored and maintained between +2 and +51 inches, THEN Restore and maintain RPV water level above -162 inches Augment RPV water level control as necessary with any of the following SRO Directs additional level control systems:
SLC (boron tank) APPX-7B CRD APPX-5B ATCIBOP Place SLC and an additional CRD Pump in service lAW Appendix 7B and 5B SRO EOI- 1 Reactor Level Can RPV water level be restored and maintained above -162 inches NO SRO Announces entry to EOI-C- 1 Alternate Level Control RPV water level CANNOT be determined NO PC water level CANNOT be maintained below 105 feet NO -
IF RPV water level can be restored and maintained above -162 inches NO Inhibit ADS ATC/BOP Inhibits ADS SRO Restore and maintain RPV water level above -162 inches using any of the following:
Condensate APPX 6A LPCI System I APPX 6B LPCI System II APPX 6C Core Spray System II APPX 6E SRO Directs 2 or more of the above systems lined up for injection ATC/BOP Aligns the directed systems for Injection 47
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Aligns CR1) and SLC lAW Appendix 5B and 7B ATC CRD Appendix 5B
- 2. IF BOTH of the following exist:
. CRD is NOT required for rod insertion, AND
. Maximum injection flow is required, THEN LINE UP ALL available CR1) pumps to the RPV as follows:
- a. IF CR1) Pump 2A is available, THEN VERIFY RUNNING CRD Pump 2A.
- b. IF CR1) Pump lB is available, THEN PERFORM the following:
- 1) NOTIFY Unit 1 Operator to verify closed 1-FCV-85-8, CR1) PUMP B DISCHARGE VALVE (Unit 1, Panel 9-5).
- 2) START CR1) Pump lB.
- 3) OPEN 2-FCV-85-8, CR]) PUMP lB DISCH TO U2.
- c. OPEN the following valves to increase CRD flow to the RPV:
2-PCV-85-23, CR1) DRIVE WATER PRESS CONTROL VLV
- d. ADJUST 2-FIC-85-1 1, CR]) SYSTEM FLOW CONTROL, on Panel 9-5 to control injection WHILE maintaining 2-PI-85-13A, CR1) ACCUM CHG WTR HDR PRESS, above 1450 psig, if possible.
- e. IF Additional flow is necessary to prevent or mitigate core damage, THEN DISPATCH personnel to fully open the following valves as required:
- 2-THV-085-0527, PUMP DISCH THROTTLING (RB NE, el 565)
- 2-BYV-085-0551, PUMP TEST BYPASS (RB NE, el 565).
Driver When called as umt one operator FCV-85-8, CRD PUMP B DISCHARGE VALVE is closed 48
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC/BOP Aligns CR1) and SLC lAW Appendix 5B and 7B ATC SLC Appendix 7B
- 2. IF RPV injection is needed immediately ONLY to prevent or mitigate fuel damage, THEN CONTINUE at Step_10 to inject SLC Boron Tank to RPV.
- 11. CHECK SLC injection by observing the following:
- Selected pump starts, as indicated by red light illuminated above pump control switch.
. Squib valves fire, as indicated by SQUIB VALVE A and B CONTINUITY blue lights extinguished,
. SLC SQUIB VALVE CONTINUITY LOST Annunciator in alarm (2-XA-55-5B, Window 20).
e 2-PI-63-7A, SLC PUMP DISCH PRESS, indicates above RPV pressure.
. System flow, as indicated by 2-IL-63 -11, SLC FLOW, red light illuminated,
. SLC INJECTION FLOW TO REACTOR Annunciator in alarm (2-XA 5B, Window 14).
- 12. IF Proper system operation CANNOT be verified, THEN RETURN to Step 10 and START other SLC pump.
40
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA ATC Aligns Condensate lAW Appendix 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 HTR2B1 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, RFP 2B DISCHARGE VALVE
. 2-FCV-3-5, RFP 2C DISCHARGE VALVE.
ATC/BOP Close RFPT Discharge Valve to prevent from overfilling RPV
- 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:
. 2-FCV-3-3 1, HP HTR 2B2 FW INLET ISOL VLV
. 2-FCV-3-24, HP HTR 2C2 FW INLET ISOL VLV
- 6. VERIFY OPEN the following RFP suction valves:
. 2-FCV-2-83, RFP 2A SUCTION VALVE
. 2-FCV-2-95, RFP 2B SUCTION VALVE
- 2-FCV-2-108, RFP 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, RFW START-UP LEVEL CONTROL, to control injection (Panel 2-9-5).
50
NRC Scenario 5 Simulator Event Guide:
Event 8 Component: Loss 480V SD BD 2A and LOCA SRO EOI-C- 1 Alternate Level Control SRO Can 2 or more Condensate, LPCI or Core Spray injection subsystems be lined up YES -
When RPV Water level drops to -162 inches Proceeds at TAF or -162 inches Is any Condensate, LPCI or Core Spray injection subsystems lined up for injection with at least one pump running YES -
Is any RPV injection source lined up with at least one pump running YES-BEFORE RPV water level drops to -180 inches CONTiNUE Continues -
Emergency Depressurization is required Inject into the RPV with any available sources SRO Enters EOI-C-2 Emergency Depressurization Will the reactor remain subcritical without boron under all conditions YES Is DW pressure above 2.4 psig YES Prevent injection from only those Core Spray and LPCI pumps not required NO Is suppression pool level above 5.5 feet YES Open all ADS Valves Directs ADS valves open ATC/BOP Opens all 6 ADS valves, reports all ADS valves open When pressure is below the shutoff head of the available injection systems direct injection SRO to restore level to +2 to +51 inches ATCIBOP Injects with available systems to restore level LPCI injection is maintained, once RPV level is rising RHR Loop 1 Pumps are ATC/BOP secured since simultaneous torus cooling and LPCI Injection is not allowed.
SRO Emergency Classification EPIP-1 1.1-S 1 Reactor water level can NOT be maintained above -162 inches. (TAF) ci
NRC Scenario 5 Simulator Event Guide:
Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, lAW BOP Appendix 6B, 6C and 6E Although most valve power is lost for RHR Loop I, injection is still available, the pumps have power, the Outboard Injection Valve does not have power but is normally open and the only valve with power is the Inboard Injection Valve which can be opened.
NOTE If RHR Loop I is used the only to control injection is to turn pumps on and off. In addition if it was aligned for Pool Cooling those valves will still be open, so the injection pressure to the vessel will be much lower.
BOP Appendix 6B
- 1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS 1 55A, LPCI SYS I OUTBD INJ VLV BYPASS SEL in BYPASS.
- 3. VERIFY OPEN 2-FCV-74-12, R}IR PUMP 2C SUPPR POOL SUCT VLV.
- 4. VERIFY CLOSED the following valves:
. 2-FCV-74-61, RHR SYS I DW SPRAY INBD VLV
. 2-FCV-74-60, RHR SYS I DW SPRAY OUTBD VLV
. 2-FCV-74-57, RFIR SYS I SUPPR CHBR/POOL ISOL VLV
- . 2-FCV-74-58, RHR SYS I SUPPR CHBR SPRAY VALVE
. 2-FCV-74-59, RHR SYS I SUPPR POOL CLG/TEST VLV
- 5. VERIFY RHR Pump 2A andlor 2C running.
- 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-53, RHR SYS I LPCI INBD iNJECT VALVE.
- 7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-68-79, RECIRC PUMP 2B DISCHARGE VALVE.
Can inject but cannot throttle 2-FCV-74-52 and will have to open 2-FCV-74-53 with the BOP handswitch
NRC Scenario 5 Simulator Event Guide:
Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, JAW BOP Appendix 6B, 6C and 6E BOP Appendix 6C
- 1. IF Adequate core cooling is assured, AND It becomes necessary to bypass the LPCI injection valve auto open signal to control injection, THENPLACE 2-HS-74-155B, LPCI SYS II OUTBD 1NJ VLV BYPASS_SEL in BYPASS.
- 3. VERIFY OPEN 2-FCV-74-35, RUR PUMP 2D SUPPR POOL SUCT VLV.
- 4. VERIFY CLOSED the following valves:
. 2-FCV-74-75, R}IR SYS II DW SPRAY INBD VLV
. 2-FCV-74-74, RHR SYS II DW SPRAY OUTBD VLV
. 2-FCV-74-71, RHR SYS II SUPPR CHBRIPOOL ISOL VLV
. 2-FCV-74-72, RI-IR SYS II SUPPR CHER SPRAY VALVE
. 2-FCV-74-73, R}{R SYS II SUPPR POOL CLG/TEST VLV
- 5. VERIFY RUR Pump 2B and/or 2D running.
- 6. WHEN RPV pressure is below 450 psig, THEN VERIFY OPEN 2-FCV-74-67, RUR SYS II LPCJ ]NBD INJECT VALVE.
- 7. IF RPV pressure is below 230 psig, THEN VERIFY CLOSED 2-FCV-68-3, RECIRC PUMP 2B DISCHARGE VALVE.
BOP Will have to open 2-FCV-74-67 with the handswitch
NRC Scenario 5 Simulator Event Guide:
Event 9 Component: RHR and Core Spray Division Injection Valves will not Auto open Aligns injection systems LPCI Loop I and II if directed and Core Spray Loop I, JAW BOP Appendix 6B, 6C and 6E BOP Appendix 6E
- 1. VERIFY OPEN the following valves:
. 2-FCV-75-30, CORE SPRAY PUMP 2B SUPPR POOL SUCT VLV
. 2-FCV-75-39, CORE SPRAY PUMP 2D SUPPR POOL SUCT VLV
. 2-FCV-75-51, CORE SPRAY SYS II OUTBD INJECT VALVE.
- 2. VERIFY CLOSED 2-FCV-75-50, CORE SPRAY SYS II TEST VALVE.
- 3. VERIFY CS Pump 2B and/or 2D running.
- 4. WHEN RPV pressure is below 450 psig, THEN THROTTLE 2-FCV-75-53, CORE SPRAY SYS II INBD iNJECT VALVE, as necessary to control injection at or below 4000 gpm per pump.
Coordinate RPV Level Control to restore and maintain Level +2 to +51 inches.
BOP/ATC Condensate and Core Spray will restore and maintain level. When RPV pressure is low enough Condensate System will maintain directed level band.
BOP Will have to open 2-FCV-75-53 with the handswitch
NRC Scenario 5 SHIFT TURNOVER SHEET Equipment Out of ServicefLCOs:
100% power, DG D and RBCCW 2B Pump are tagged out. Spare RBCCW Pump is aligned for operation.
Temporary DGs are NOT provided.
Operations/Maintenance for the Shift:
Return LPRM 8-49B to Operate from a Bypassed Condition JAW 2-OJ-92B.
Lower Power with flow to 91% for Main Turbine Valve Testing.
Unit 1 and 3 are at 100% Power Unusual Conditions/Problem Areas:
Severe Thunderstorms are forecast for today, currently no watches or warnings are in effect.
55