ML060310572

Final - Section C Operating Exam (Folder 3)

ML060310572
Person / Time
Site:	Susquehanna
Issue date:	10/07/2005
From:	Chin R E Susquehanna
To:	Fish T H Operations Branch I
Conte R J
References
Download: ML060310572 (89)

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PPL-SUSQUEHANNA, LLC Prepared By: Richard E. Chin Instructor TRAINING CENTER 10/07/05 Date SIMULATOR SCENARIO Suoervisina ManaaerEhift SuDervisor Scenario Title: Date ILO CertificatioflRC Exam Scenario Scenario Duration:

90 Minutes 11 Scenario Number:

ILO-505 Reviewed By: I Nuclear Ooerations Trainina Suoervisor I Date ~ ~ Approved By: Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 2 Rev. 0. 10/07/05 ILO-505 I/ THIS PAGE IS INTENTIONALLY LEFT BLANK i NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 3 Rev. 0, 10/07/05 ILO-505 II SCENARIO

SUMMARY

I The Scenario begins with both Units at 100 percent power and all systems OPERABLE. When the crew has assumed shift responsibilities, they will perform SO-184-001, QUARTERLY MSlV CLOSURE RPS INSTRUMENTATION. During the performance of the surveillance, one of the relays will fail to de-energize when the C MSL Inboard Isolation Valve is tested.

The crew will halt the surveillance, and address Technical Specification 3.3.1.1 RPS Instrumentation. Because no half-scram will be generated, the US should declare that channel INOP.

I&C can boot the relay to satisfy the Required Action, or the crew may elect to manually insert a half-scram.

After the RPS Tech Spec exercise has been evaluated, the crew will experience EHC Oscillations. The Oscillations will require them to implement ON-193-001, TURBINE EHC SYSTEM MALFUNCTION. This activity will be completed when the alternate Pressure Regulator has been successfully placed in service. After the EHC Malfunction event has been evaluated, the crew will experience an inadvertent HPCl Initiation. The HPCl LO FLOW alarm, AR-114-EO2 will fail to annunciate. The crew should recognize the change in Reactor power due to the cold water injection, as well as the change in Feedwater Level Control response due to the additional inventory from the HPCl injection.

The crew may enter ON-156-001, UNEXPLAINED REACTIVITY CHANGE in order to identify the problem.

The crew should also attempt to reduce power to the level established prior to the HPCl failure. The crew will override HPCl and declare the system INOPERABLE in accordance with T.S. 3.5.1. If power was reduced to less than 95 percent, the crew should enter GO-1 00-01 2, POWER MANEUVERS.

After the HPCl System failure has been evaluated, a leak in the Drywell over a five- minute timeframe will force the crew to scram the reactor and execute EO-1 00-1 02, RPV CONTROL, as well as EO-1 00-1 03, PRIMARY CONTAINMENT CONTROL. If HPCl is returned to service for RPV level control, it will trip on overspeed and will not be restored. Condensate Pumps will trip if started, resulting in limited high pressure feed systems. The leak will worsen over time, and eventually will exceed the ability of makeup to the vessel, resulting in RPV level lowering to Top of Active Fuel.

If HPCl is to be used for injection, it will have an overspeed trip and be unavailable for injection. When TAF is reached, the crew will execute EO-100-1 12, RAPID DEPRESSURIZATION.

The combination of high Primary Containment parameters and lowering RPV pressure will result in violation of the Saturation Curve. Because RPV Level will become Indeterminate, the crew will execute EO-100-1 14, RPV FLOODING.

When all injection sources are started, the RHR Injection Valve FO15B will fail to open, and will be required to be manually opened. When RPV FLOODING has been successfully completed as indicated by the FLOODED TO STEAM LINES table, the scenario will be terminated.

Based upon EP-TP-001 , EAL Classification Levels, a LOSS and POTENTIAL LOSS has occurred because RPV level went below -1 61 inches. The EAL is a Site Area Emergency FSl . Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 4 Rev. 0, 10/07/05 ILO-505 THIS PAGE IS INTENTIONALLY LEFT BLANK Fo~ NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 5 Rev. 0, 10/07/05 ILO-505 - SCENARIO OBJECTIVES 1 The objective of this scenario is to evaluate the Licensed Operator Candidate's ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

0 Knowledge of integrated plant operations 0 Ability to diagnose abnormal plant conditions 0 Ability to work together as a team 0 Ability to mitigate plant transients that exercise their knowledge and use of ONs and EOPs 0 Ability to utilize Technical Specifications (SRO Only) To meet this objective, the licensed operator candidates must demonstrate proficiency in the following competencies: Reactor Operator Candidates:

1. Interpretldiagnose events and conditions based on alarms, signals, and readings.

2. Comply with and use procedures, references, and Technical Specifications.

3. Operate the control boards. 4. Communicate and interact with other crew members. Senior Reactor Operator Candidates:

1. Interpretldiagnose events and conditions based on alarms, signals, and readings.

2. Comply with and use procedures and references.

3. Operate the control boards (N/A to upgrade candidates).

4. Communicate and interact with the crew and other personnel.

5. Direct shift operations.

6. Comply with and use Technical Specifications.

FOtm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 6 Rev. 0, 10/07/05 ILO-505 Fo~ NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 7 Rev. 0, 10/07/05 ILO-505 II CRITICAL TASKS I

• Perform Rapid Depressurization when RPV level drops to -1 61 ". Safety Sign if icance RPV leakage or loss of injection systems impacts the ability to provide continued adequate core cooling through core submergence based on inventory loss. Consequences for Failure to Perform Task Failure to take the EOP actions will result in uncovering the core and breach of the fuel clad due to overheating. SSES EOP Basis for: The following steps provide the operating crew guidance to line up injection systems as available to maintain level >-129 inches.

If these actions are unsuccessful, the crew receives additional direction when it is determined that level can not be maintained above TAF. RC/L-4 RESTORE AND MAINTAIN LVL BETWEEN +13" AND +54" USING TABLE 3 SYSTEMS RC/L-5 IF LVL CANNOT BE RESTORED AND MAINTAINED

> +13" MAINTAIN LVL

> -129" USING TABLE 3 SYSTEMS AUGMENTING AS DESIRED WITH TABLE 5 ALTERNATE SUBSYSTEMS RC/L-10 IRRESPECTIVE OF VORTEX LIMITS WITH TABLE 3 SYSTEMS PERFORM ALL 1 LINE UP FOR INJECTION 2 STARTPUMPS 3 INCREASE INJECTION TO MAX RC/L-11 IF LESS THAN 2 TABLE 4 SUBSYSTEMS CAN BE LINED UP COMMENCE LINING UP AS MANY AS POSSIBLE TABLE 5 ALTERNATE SUBSYSTEMS RC/L-13 WITH TABLE 5 ALTERNATE SUBSYSTEMS PERFORM ALL:LINE UP FOR INJECTION 1 STARTPUMPS 2 INCREASE INJECTION TO MAX RCR-16 WHEN LVL CANNOT BE RESTORED AND MAINTAINED

> -161" GO TO RAPID DEPRESS Form NTP-QA91.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 0 Rev. 0, 10/07/05 ILO-505 CRITICAL TASKS Rapid Depressurization is not initiated until RPV water level has dropped to -1 61 inches (TAF) because:

0 Adequate core cooling exists so long as RPV water level remains above

-161 inches (TAF). 0 The time required for RPV water level to decrease to -1 61 inches (TAF) can best be used to line up and start pumps, attempting to reverse the decreasing RPV water level trend before Rapid Depressurization is required to assure continued adequate core cooling. (

Reference:

SSES-EPG C1-4 and second override before C3-1)

IndicationdCues for Event Requiring Critical Task Reactor water level trending downward, eventually indicating less than the top of active fuel height on the Fuel Zone Level Indicator. Performance Criteria Perform a Rapid Depressurization per EO-1 00-1 12 when water level reaches the TAF -1 61 inches as read on the Fuel Zone Instrument. . Initiate ADS/Manually open all six ADS Valves. Performance Feedback Initiating a rapid depressurization causes Reactor pressure to lower to the shutoff head of the low pressure injection systems allowing water level to rise on the Fuel Zone and Wide Range level instruments.

Verify ADS Valves are open using light red light indication, acoustic monitoring and lowering Reactor pressure and rising reactor water level.

• Declare RPV level indication indeterminate due to violation of the RPV Saturation Curve. Perform RPV Floodina when RPV water level becomes indeterminate bv establishing RPV Flooded to Steamlines.

Saf etv Sign if icance Adequate core cooling may be challenged if core submergence can not be verified. Consequences for Failure to Perform Task Failure to take the EOP actions may result in uncovering the core and breach of the fuel clad due to over heating. SSES EOP Basis for:

RCR-2 IF LVL CANNOT BE DETERMINED GO TO RPV FLOODING If RPV water level cannot be determined, the actions specified in the subsequent

[EO-1021 steps cannot be performed, since RPV water level and water level trend information is required for determining which actions to take. The transition to EO-000-1 14, RPV FLOODING, is necessary to assure continued adequate core cooling under conditions where RPV water level cannot be determined.

These systems consist of all motor-driven systems, which are available to flood the RPV. As many of these systems as necessary must be used to establish and maintain the conditions required to verify RPV flooding. Establishing adequate core cooling conditions dictates that adherence to Vortex limits not be required.

IndicationdCues for Event Reauirinq Critical Task Violation of the RPV Saturation Curve is indicated by PlCSY format (RPVSAT) showing purple indication on the curve, plot on the unsafe side by the Crew and/or RPV level instrumentation failing in the upscale direction.

Rev. 0 (03/04) 2005 NRC Exam, As Given FO~ NTP-QA-31.7A Page 9 Rev. 0, 10/07/05 ILO-505 11 Performance Criteria Recognize failure of RPV level indicators due to reaching saturation conditions on the instrument runs, initiate rapid depressurization by opening ADS Valves, and then increasing RPV injection until RPV flooded as indicated by a combination of conditions as shown in FLOODED TO STEAMLINES TABLE. Performance Feedback Initiating a rapid depressurization causes Reactor pressure to lower to the shutoff head of the low pressure injection systems allowing water level to rise to the point that RPV pressure will increase to a value that is 81 psid above Suppression Chamber.

At this point injection should be stabilized to maintain the DP. Verify ADS Valves are open using light red light indication, acoustic monitoring and lowering Reactor pressure.

Verify injection from available systems raises RPV pressure to a value that is 81 psid above Suppression Chamber

• Manuallv ooen RHR F015B Valve to iniect to the RPV. Safetv Significance: Ensures Adequate Core Cooling. IndicationdCues for Event Reauirina Critical Task Auto initiation signals present as evidenced by High Drywell Pressure, Low RPV Pressure, Low RPV Water Level, initiation signal indication lamp illuminated. Consequences for Failure to Perform Task Lack of Adequate Core Cooling IndicationdCues for Event Reauirina Critical Task RPV water level less than

-129 inches, High Drywell Pressure 1.72, Low RPV Pressure, 420 psig Performance Criteria Opening the injection valve manually to ensure RHR flow to the RPV top allow RPV FLOODED TO STEAMLINES.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 10 Rev. 0, 10/07/05 ILO-505 THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 11 Rev. 0, 10/07/05 ILO-505 I SCENARIO REFERENCES n 1. Quarterly Surveillance SO-184-001 QUARTERLY MSlV CLOSURE RPS INSTRUMENTATION TECH SPECS 3.3.1.1 RPS INSTRUMENTATION Ml-C72-22 SH. 11 RPS ELEMENTARY DRAWING 2. Inadvertent HPCl Initiation ON-156-001 UNEXPLAINED REACTIVITY CHANGE GO-100-012 POWER MANEUVERS OP-AD-002 STANDARDS FOR SHIFT OPERATIONS OP-AD-004 OPERATIONS STANDARDS FOR ERROR AND EVENT PREVENTION NDAP-QA-702 CONDITION REPORT TECH SPECS 3.5.1 ECCS - OPERATING

3. EHC OSCILLATIONS ON-193-001 TURBINE EHC SYSTEM MALFUNCTION

4. LOCA EO-100-102 RPV CONTROL EO-100-103 PRIMARY CONTAINMENT CONTROL OP-149-001 RHR SYSTEM 5. RAPID DEPRESSURIZATION EO-100-112 RAPID DEPRESSURIZATION

6. RPV FLOODING EO-100-114 RPV FLOODING 7. EP-TP-001 EAL CLASSIFICATION LEVELS Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 12 Rev. 0, 10/07/05 ILO-505 THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 13 Rev. 0, 10/07/05 ILO-505 U SCENARIO SPECIAL INSTRUCTIONS II 1. Initialize to IC 20. For NRC LOC 21 Exam, IC 187 is dedicated for this setup. 2. Prepare US Turnover Sheet indicating:

0 0 NoLCO. 0 All systems Operable.

0 Both Units at 100 percent power.

Perform SO-184-001 Quarterly MSlV Closure RPS Instrumentation at beginning of shift. 3. Execute Preference File restorepref YPP.IL0-505.

4. Prepare a copy of SO-184-001 with green cover sheet. FOITI NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 14 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT DESCRIPTION FORM Initial Conditions: IC-20, Both Units at 100 Percent Power; All Systems are OPERABLE.

EVENT I TIME I DESCRIPTION 1 5 I Perform so-1 84-001, QUARTERLY MSIV CLOSURE RPS INSTRUMENTATION 2 15 Relay failure C72A-K3F 3 30 EHC Oscillations 4 45 Inadvertent HPCl Initiation/Unexplained Reactivity 5 55 Recirculation Loop B Suction Line Break DBA 6 55 ADS Auto Logic Failure 7 60 RHR injection Valve 158 Auto Logic Failure 8 65 Rapid Depressurization 9 70 RPV Flooding 75 Termination Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 15 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No: 192 Brief

Description:

SO-184-001, QUARTERLY MSlV CLOSURE RPS INSTRUMENTATlONlRELAY FAILURE POSITION PCOP TIME STUDENT ACTIVITIES 5 Establish Communications with Relay Room Operators.

~ 10 Report failure of response for Relay C72A-K3F.

I 1 PCOM Monitor Main Steam Line flow indications.

Crew Recognize Steam Line Flow decreasing, relay not de-energized.

us I Discontinue Surveillance until problem resolved.

1 Determine acceptance criteria not satisfied for the surveillance.

I Reference T.S. 3.3.1.1, RPS Instrumentation, FUNCTION 5 of Table 3.3.1 .l-1 applicable. Declare Channel INOP and app1yT.S. 3.3.1.1 CONDITION A: Place channel in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. OR Place associated trip system in trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• Denotes Simulator Critical Task. NOTES: I Fo~ NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 16 Rev. 0, 10/07/05 ILO-505 -T INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 192 Brief

Description:

SO-1 84-001, QUARTERLY MSlV CLOSURE RPS INSTRUMENTATION/RELAY FAILURE INSTRUCTOR ACTIVITY: 1. Monitor RPS format RP4 during the surveillance and observe relays de-energize and re-energize.

Note that only one-half of the RPS circuitry is available to monitor, but is sufficient for this surveillance. ROLE PLAY:

1. 2. When Relays K3A and K3B de-energize, report "TRIPPED (or "relay cycled" if PCO uses this method).

When Relays K3E and K3D de-energize, report "TRIPPED" (or "relay cycled if PCO uses this method).

3. 4. When the Inboard MSlV F022C is tested, DO NOT rePort C72A-K3F cvcled as required.

As GCC acknowledge the situation.

5. As Station Management acknowledge situation.

6. Wait 5-10 minutes and report as I&C that as best as they can determine, it must be the limit switch inside the Primary containment. In order to place the channel in the tripped condition, they can boot the relay so that it is in effect tripped.

Form NTP-QABl.7A Rev. 0 (0304) 2005 NRC Exam. As Given Page 17 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No: 3 Brief

Description:

EHC OSCILLATIONS POSITION PCOM PCOM/P TIME us PCOM/M STUDENT ACTIVITIES Recognize power/pressure/electrical channes. Perform ON-1 93-003, TURBINE EHC SYSTEM MALFUNCTION 0 0 Observe BPV 1 OSCILLATING.

0 0 Report oscillations have stopped. Reduce reactor power with recirculation flow until EITHER of following reached. Reactor power reduction of five percent (55 MWe) or Note the Setpoint of the LOAD LIMIT SET potentiometer is 8.9. Using LOAD LIMIT SET potentiometer, Decrease setting until #1 BYPASS VALVE approximately 50 percent open at BPV 1 Percent position indicator. Check Control Valve oscillations STOP. Direct performance of ON-193-001, Section

3.3. Notify

WWM to direct I&C to place the "B" Pressure Regulator in service. If time permits, notrfy GCC of reason for power reduction.

Notify Plant Management of condition. Authorize the FUS to perform transfer to alternate Pressure Regulator. Direct the activities of the FUS as per the Off-Normal Procedure.

Slowly restore LOAD LIMIT SET to 8.9. DirecVdiscuss restoration of Reactor Power to 100 percent.

• Denotes Simulator Critical Task.

NOTES: I Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 18 Rev. 0, 10/07/05 ILO-505 Event No: 3 Brief

Description:

EHC OSCILLATIONS INSTRUCTOR ACTIVITY:

1. When MSlV limit switch problem and T.S. evolution has been completed, initiate an EHC oscillation by depressing:

[PB-31 IMF TC193003 6 1:OO 0 When the PCO has transferred the oscillations to the Bypass Valve, delete the malfunction by depressing: EHC OCILLATION OF 6 RAMPED OVER 1 MINUTE 2. [PB-41 DMF TC19303 DELETE EHC MALFUNCTION ROLE PLAY: 1. As Workweek Manager, acknowledge the request for assistance, and report that a team of I&C Technicians are still working on booting the MSlV limit switch contact.

2. As FUS, report to the Control Room, and tell them that you are ready to perform the swap. 3. While in the relay room, place Pressure Regulator B in control via PCOP direction as follows:

0 0 Turn Pressure Setpoint Bias Potentiometer CLOCKWISE (INCREASING numbers)

UNTIL Pressure Regulator A and B control lights ILLUMINATED on Panel 1 C663. Turn Pressure Setpoint Bias Potentiometer CLOCKWISE UNTIL Pressure Regulator Light A EXTINGUISHES AND THEN an additional 0.7 turns.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 19 Rev. 0, 10/07/05 ILO-505 I SCENARIO EVENT FORM If crew identifies HPCl failure, entry to the Off-Normal may not occur immediately.

' l. Event No: 4 Brief

Description:

INADVERTENT HPCl INITIATION (UNEXPLAINED REACTIVITY)

TIME STUDENT ACTIVITIES Respond to REACTOR WATER LEVEL HI-LO alarm AR-101 -B17. Reduce Reactor Power to stav within the License Limit and Monitor Power/Flow Map. Recognize Steam Flow/Feed Flow mismatch, reactor power increase. Direct implementation of ON-1 56-001, UNEXPLAINED REACTIVITY Recognize HPCl initiated.

Confirm Drywell Pressure and Reactor Water level initiation signals not present.

Override HPCl Place or Check Placed HPCl AUXILIARY OIL PUMP 1 P213 switch to START. Place HPCl TURBINE FLOW CONTROL FC-E41-1 R600 in MANUAL. Adjust Flow Controller to reduce HPCl discharge pressure less than RPV pressure.

0 Ensure HPCl MIN FLOW TO SUPP POOL HV-155-F012 OPENS when HPCl flow e500 gpm and discharge pressure >125 psig.

Depress HPCl INT SIG RESET HS-E41-1 S17 RESET pushbutton.

When HPCl initiation resets, Shut Down HPCl in accordance with "Shutdown" section. Direct PCOP to Override HPCl Injection after confirming no initiation signal present.

Declare HPCl INOP in accordance with T.S. 3.5.1 .D: 0 Obtain assistance from WWM. Enter GO-1 00-012, POWER MANEUVERS, if power reduced below 95 percent. Monitor Main Steam Line and Offgas Radiation levels due to power surge. Restore RPV water level and Dower to stable conditions. Verify by administrative means RClC System is OPERABLE immediately AND Restore HPCl System to OPERABLE status within 14 days. Review ON-1 56-001 to determine if anv other conditions occurred.

Ir Denotes Simulator Critical Task.

Fo~ NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 20 Rev. 0, 10/07/05 ILO-505 Event No: 4 Brief

Description:

INADVERTENT HPCl INITIATION (UNEXPLAINED REACTIVITY)

INSTRUCTOR ACTIVITY:

1. After EHC evolution has been completed, initiate an inadvertent HPCl initiation by depressing:

[PB-1] IMF HP152204 2. After HPCI has initiated, delete HPCl malfunction (to allow the crew to remove it from service) by depressing:

[PB-21 DMF HP152004 ROLE PLAY: 1. As Workweek Manager acknowledge reque: for assistance and tell them that the !?X crew has booted the MSI\

contact as requested, and that they will head right out to investigate the HPCl problem. 2. As I&C, wait about 10 minutes and report that the start relay is making a funny sound, and that they will need more time to determine the cause of the failure. Right now it appears to be a relay failure, and it could restart if left in auto-standby.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 21 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No: 5 Brief

Description:

RECIRC LOOP "B" SUCTION LINE BREAK POSITION TIME STUDENT ACTIVITIES PCOM/P Recognizes/reports increasing Drywell pressure and temperature. Perform ON-100-101, SCRAM, SCRAM IMMINENT.

Reduce Core Flow to approximately 65 Mlbmhr. Place Mode Switch to Shutdown.

Verifies all Control Rods inserted.

us Directs performance of ON-100-101, SCRAM, SCRAM IMMINENT. Perform EO-100-102, RPV CONTROL due to RPV Water level < +13 inches. Perform EO-100-103, PC CONTROL due to High Drywell Pressure 1.72 psig. Re-enters EO-1 00-1 02, RPV CONTROL due to High Drywell Pressure 1.72 psig. Direct initiation of Suppression Chamber Sprays. I 1 PCOP Verify ECCS initiations, Containment Isolations, and DG starts. Injects with available systems as directed to maintain RPV water level +13 inches to +54 inches. Sprays the Suppression Chamber in accordance with the Hard Card when directed.

I I CREW Recognizes RPV water level lowering at a faster rate.

us Directs injection with all available systems.

Transition RPV water level monitoring to Fuel Zone RPV Level Instrumentation when RPV level indication goes below -1 45 on Wide Range Level Instrumentation.

PCOP PCOP Report RPV water level approaching Top of Active Fuel, -161 inches.

• Denotes Simulator Critical Task.

NOTES: I Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 22 Rev. 0, 10/07/05 ILO-505 Event No: 5 Brief

Description:

RECIRC LOOP "B" SUCTION LINE BREAK INSTRUCTOR ACTIVITY:

1. While the crew is recovering from the HPCl evolution, and plant conditions have stabilized, initiate a small leak in the Drywell by depressing:

[PBB] IMF RR164011 B 0.5 300 RECIRC LOOP B SUCTION LINE BREAK, 0.5% 2. When the crew has begun Suppression Chamber Sprays, increase the severity of the Recirc loop rupture to 40 percent by depressing:

[PB-61 MMF RR164011B 40 300 0.5 RECIRC LOOP B SUCTION LINE BREAK, 40% 3. If the crew decides to use HPCl for injection, initiate an overspeed trip by depressing:

[PB-71 IMF HP152011 ROLE PLAY: As necessary Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 23 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event Nos: 6,7, 8,9 Brief

Description:

ADS FAILURE, RHR 158 FAILURE, RAPID DEPRESSURIZATION, RPV FLOODING POSITION I TIME STUDENT ACTIVITIES Report RPV level at TAF, -161 inches. Perform Rapid Depressurization when RPV level drops to -161 inches. 0 Verify Suppression Pool level >5 feet.

0 Direct opening of all ADS SRVs.

0 Observe RPV pressure lowering. Recognize/report failure of automatic ADS function. Rapidly depressurize reactor by manually opening all six ADS SRVs.

0 Verify valves open. 0 Observe RPV Dressure lowerina.

Monitor RPV Pressure and Primary Containment Instrument Run Temperature.

Declare RPV level indication indeterminate due to violation of the RPV Saturation Curve.

• Performs E01 00-1 14, RPV FLOODING by increasing injection flowrate to the RPV meet RPV Flooding conditions as determined by FLOODED TO STEAMLINES Table:

0 Acoustic monitors cycling. 0 SRV Tailpipe Temperature subcooled.

0 RPV pressure rising. 0 Direct termination of Suppression Chamber and Drywell Sprays. Stop Drywell and Suppression Chamber Sprays. Recognize and report failure of RHR 15B Valve to auto open. Manually open RHR FO15B Valve to inject to the RPV. Align all injection sources to flood the reactor. Record Time Flooded Conditions met. Suppression Pool Level not lowering.

After the scenario is complete, classifies the event as a SITE AREA EMERGENCY under EAL FS1 due to a Loss or Potential Loss of the Fuel Clad Barrier and a Loss of the RCS Barrier.

• Denotes Simulator Critical Task. Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 24 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, Event No: 6, 7,8, 9 Brief

Description:

RAPID DEPRESSURIZATIOWRPV FLOODING INSTRUCTOR ACTIVITY:

N/A ROLE PLAY: As necessary TERMINATION CUE: EO-100-1 14, RPV FLOODING CONDITIONS are met. EVENT CLASSIFICATION:

After the scenario is complete, have the US classify the scenario for the HIGHEST EAL. Provide the US with any requested information needed to perform the classification.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given PPL-SUSQUEHANNA, LLC TRAINING CENTER Prepared By: SIMULATOR SCENARIO Richard E. Chin 10/08/05 Instructor Date Scenario Title: Reviewed By: Nuclear Operations Training Supervisor ILO Certification/NRC Exam Scenario Date ~~ Scenario Duration:

90 Minutes Approved By: Supervising ManagedShift Supervisor

~~ Scenario Number: ILO-305 Date RevisiodDate:

Rev. 0, 10/08/05 ~ Course: PCOO7/PCOO8 PCO17/PCO18 Initial License RO/SRO Certification Examination Initial License RO/SRO NRC Examination Ope rational Activities

Raise Reactor Power Power Supply Failure Stuck Rods Inadvertent DG Start Failed Fuel Leak Into Secondary Containment PC Isolation Valves Fail to Isolate Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 2 Rev. 0, 10/08/05 I LO-305 Form NTP-QA91.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 3 Rev. 0, 10/08/05 ILO-305 (I SCENARIO

SUMMARY

- The scenario begins with a Reactor Startup in progress at five percent power in accordance with GO-100-002, PLANT STARTUP, HEATUP AND POWER OPERATIONS. All systems are OPERABLE. During Control Rod withdrawal, a blown fuse will occur on a common power supply to Intermediate Range Monitors B&F and Source Range Monitor B. This will result in a RPS Half-Scram on Division 2, and will require a Half-Scram Reset in accordance with OP-158-001, REACTOR PROTECTION SYSTEM. Additionally, the US will need to address Technical Specifications 3.3.1.1, RPS Instrumentation, and Technical Requirements Manual 3.1.3, Control Rod Block Instrumentation. Upon investigation, I&C will identify a blown fuse and replace it, thereby allowing the crew to reset the Half-Scram.

After the fuse is replaced and the half-scram is reset, the "D" Diesel Generator will start inadvertently.

It will need to be shut down locally due to indications of overheating and smoke. If the crew does not take prompt action, the DG will trip.

If this occurs, ON-024-001, DIESEL GENERATOR TRIP will be implemented.

In either case, this event will require the crew to declare the DG INOPERABLE and enter T.S.

3.8.1 A.C. Sources Operating. The crew should provide cooling to the DG within eight minutes (unloaded) as per procedure caution, and when the ESW System is placed in service, one of the pumps will trip. The crew will need to address ON-054-001, LOSS OF ESW, and address a Dual-Unit LCO 3.7.2 for the ES W System. When the Loss of ESW and the Diesel Generator evolution has been completed, a leak in the Reactor Water Cleanup System will occur. The primary containment isolation valves will fail to isolate, resulting in an unisolable Primary System discharge into Secondary Containment.

The crew will be forced to execute EO-1 00-1 04, SECONDARY CONTAINMENT CONTROL, due to high radiation and temperature conditions generated by the leak. When the crew has determined that a Primary System is discharging into the area, the US will direct the crew to place the Mode Switch to Shutdown, as directed by the EOP. When the Mode Switch is placed to Shutdown, seven (7) Control Rods will fail to insert. The crew will enter EO-100-1 13, LEVEUPOWER CONTROL for a brief time in order to get all Control Rods fully inserted.

The stuck control rods will eventually drift in shortly after the scram, but will remain not fully inserted for a couple of minutes so as to justify the failed fuel event. As a result of the Control Rods not being fully inserted, several fuel assemblies will fail.

The resultant increase in radiation coupled with the leak in the Secondary Containment will cause two areas in Secondary Containment to exceed Max Safe Radiation levels. These conditions will then require the crew to transition to EO-100-1 12, RAPID DEPRESSURIZATION. When Rapid Depressurization has been completed and actions are underway to address Primary Containment parameters, the scenario will be terminated.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 4 Rev. 0, 10/08/05 ILO-305 II SCENARIO OBJECTIVES I The objective of this scenario is to evaluate the Licensed Operator Candidate's ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

0 Knowledge of integrated plant operations 0 Ability to diagnose abnormal plant conditions 0 0 0 Ability to work together as a team Ability to mitigate plant transients that exercise their knowledge and use of ONs and EOPs Ability to utilize Technical Specifications (SRO Only) To meet this objective, the licensed operator candidates must demonstrate proficiency in the following competencies: Reactor Operator Candidates:

1 . InterpreVdiagnose events and conditions based on alarms, signals, and readings.

2. Comply with and use procedures, references, and Technical Specifications.

3. Operate the control boards. 4. Communicate and interact with other crew members. Senior Reactor Operator Candidates:

1. 2. 3. 4. 5. 6. Interpret/diagnose events and conditions based on alarms, signals, and readings. Comply with and use procedures and references. Operate the control boards (N/A to upgrade candidates).

Communicate and interact with the crew and other personnel. Direct shift operations.

Comply with and use Technical Specifications.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 5 Rev. 0, 10/08/05 ILO-305 I THIS PAGE IS INTENTIONALLY LEFT BLANK I Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 6 Rev. 0, 10/08/05 ILO-305 CRITICAL TASKS Sr Manuallv scram the Reactor before any Secondarv Containment Area Radiation Max Safe TemDerature. Safetv Sianificance:

High-energy leakage into the Secondary Containment Area impacts the integrity of Secondary Containment. Failure of the Secondary Containment directly relates to the 10CFR100 design criteria of dose to the General Public. Action is taken to isolate systems that are discharging into secondary containment to terminate possible sources of radioactivity release. If these efforts are unsuccessful, whatever reason, or conditions are approaching Max Safe thresholds, the reactor (source term) is placed in a low energy state, or shut down. Conseauences for Failure to Perform Task:

Failure to take actions to mitigate the energy released to the secondary containment directly affects the Radiation dose to the General Public. SSES EOP Basis for: SC/R-4 BEFORE ANY RB AREA RAD REACHES MAX SAFE GO TO RPV CONTROL The Max Safe operating radiation level is the most limiting area radiation level, which will ensure personnel exposure is kept below the emergency exposure limit (25 Rem) while performing EOP actions in the Secondary Containment for a period no longer than 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (i.e., 25 Rem/2.5 hr

= 10 Remhr). A Reactor Scram through entry to EO-000-1 02, RPV CONTROL, promptly reduces to decay heat levels the energy that the RPV may be discharging to the secondary containment.

The instruction to take this action at any time between the Max Normal and the Max Safe operating value may help avoid reaching the more severe action of Rapidly Depressurizing the RPV. (

Reference:

SSES-EPG SC/R-2.1)

IndicationdCues for Event Reauirina Critical Task: Simplex Fire Detection alarms indicating High temperatures in RB Areas. Increasing area radiation and alarms for RB Areas. Increasing Steam Leak Detection System temperatures and alarms. Performance Criteria:

Manually Scram the Reactor prior to Exceeding Max Safe Temperature/Radiation as indicated by associated control room alarms and PlCSY radiation indications. Performance Feedback:

Initiating a reactor scram reduces the heat load that will be absorbed and released by the Secondary Containment as well as the radioactive source term. Rods inserted. Power lowering.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 7 Rev. 0, 10/08/05 ILO-305 II CRITICAL TASKS I Ir Rapidly depressurize the reactor when two Secondary Containment Areas exceed Max Safe Rad levels. Safetv Simificance:

High-energy leak in the Secondary Containment Area impacts the integrity of Secondary Containment.

Failure of the Secondary Containment directly relates to the 1 OCFRl 00 design criteria of dose to the General Public. Action is taken to isolate systems that are discharging into Secondary Containment to terminate possible sources of radioactivity release.

Minimizing radioactive release to Secondary Containment also helps accomplish the objective of precluding a radioactive release outside Secondary Containment under conditions where Secondary Containment integrity cannot be maintained. Previous containment control actions have not, for whatever reason, mitigated the event and now potentially large areas of the Secondary Containment have been challenged.

Conseauences for Failure to Perform Task:

Failure to take actions to mitigate the energy released to the Secondary Containment directly affects the radiation dose to the General Public. SSES EOP Basis for: IN TWO OR MORE AREAS RAPID DEPRESS IS REQ'D SC/T-9 WHEN RB AREA TEMP EXCEEDS MAX SAFE SC/R6 WHEN RB AREA RAD EXCEEDS MAX SAFE ,N TWO OR MORE AREAS RAPID DEPRESS IS REQ'D SC/L-7 WHEN RB AREA WATER LEVEL EXCEEDS MAX SAFE IN TWO OR MORE AREAS RAPID DEPRESS IS REQ'D Should Secondary Containment Area Radiation levels continue to increase to their Max Safe values in more than one area with a Primary System discharging into Secondary Containment, the RPV must be rapidly depressurized. Depressurizing the RPV promptly places the Primary System in its lowest possible energy state, rejects heat to the Suppression Pool in preference to outside the containment, and reduces the driving head and flow of Primary Systems that are not isolated and discharging into the Secondary Containment. The criteria of "2 or more areas" identifies the increase in radiation trend as a widespread problem, which may pose a direct and immediate threat to Secondary Containment integrity, equipment located in the Secondary Containment, or continued safe operation of the plant.

Indicationdcues for Event Reauirincr Critical Task: 0 Increasing Steam Leak Detection System temperatures and alarms indicating levels at Max Safe values. Increasing area radiation and alarms for RB Areas indicating levels at Max Safe values. PlCSY formats indicating radiation values greater than Max Safer values. Reactor Building room levels above high level annunciation or as confirmed by local evaluation.

FOITTI NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 8 Rev. 0, 10/08/05 ILO-305 CRITICAL TASKS 11 Performance Criteria:

Perform a Rapid Depressurization per EO-1 00-1 12 when two or more RB areas exceed max safe radiation per EO- 100-1 04, Table 9 (1 0 Rhr for all areas). Initiate ADS/Manually open all six ADS Valves. Performance Feedback: Initiating a Rapid Depressurization causes Reactor pressure to lower, which lowers the driving force of any primary system breach. Verify ADS Valves are open using light red light indication, acoustic monitoring and lowering Reactor pressure and rising reactor water level. Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 9 Rev. 0, 10/08/05 ILO-305 n SCENARIO REFERENCES I 1. 2. 3. 4. 5. 6. 7. 8. RAISE REACTOR POWER GO-1 00-002 PLANT STARTUP, HEATUP AND POWER OPERATIONS SO-1 56-007 CONTROL ROD COUPLING/FULL-IN INDICATOR CHECKS POWER SUPPLY FAILURE AR-104-A06 AR-104-A04 AR-104-A01 AR-104-606 SRM UPSCALE OR INOP AR-104-HO3 ROD OUT BLOCK OP-158-001 REACTOR PROTECTION SYSTEM T. S. 3.3.1.1 TRM 3.1.3 IRM CHAN B/D/F/H UPSCALE TRIP OR INOP NEUTRON MON CHAN B SYSTEM TRIP RPS CHANNEL 81/82 AUTO SCRAM R PS I N STRUM EN TAT ION CONTROL ROD BLOCK INSTRUMENTATION INADVERTENT DG START:

TS 3.8.1 AC SOURCES OPERATING AR-016-C03 OP-024-001 DIESEL GENERATORS ON-024-001 DIESEL GENERATOR TRIP DG D PANEL OC521 D LO PRIORITY TROUBLE LOSS OF ESW: AR-016-DlO ESW PUMP A OVERCURRENT AR-Ol6-EO8 ESW PUMP C OVERCURRENT OP-054-001 ESW SYSTEM ON-054-001 LOSS OF ESW T.S. 3.7.2 EMERGENCY SERVICE WATER RWCU LEAK INTO SECONDARY CONTAINMENT:

AR-101 -DO1 RWCU PUMP TROUBLE AR-101 -A04 RWCU SYSTEM PRE-ISOLATION HI TEMP/HI DlFF TEMP AR-SP-002 T.S. 3.6.1.3 ON-1 00-1 01 EO-1 00-1 04 EO-1 00-1 02 RPV CONTROL SIMPLEX FIRE PROTECTION FIRE DETECTION ALARM PRIORITY 2 PRIMARY CONTAINMENT ISOLATION VALVES SCRAM, SCRAM IMMINENT SECONDARY CONTAINMENT CONTROL STUCK RODS: EO-1 00-1 13 LEVEUPOWER CONTROL FAILED FUEL: OP-AD-055 ON-1 79-001 OPERATIONS PROCEDURE PROGRAM IMMEDIATE OPERATOR ACTIONS INCREASING OFFGAS MSL RAD LEVELS RAPID DEPRESSURIZATION: EO-100-1 12 RAPID DEPRESSURIZATION Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 10 Rev. 0, 10/08/05 ILO-305 THIS PAGE IS INTENTIONALLY LEFT BLANK I Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 11 Rev. 0, 10/08/05 ILO-305 I SCENARIO SPECIAL INSTRUCTIONS 1. Prepare SO-1 56-007, CONTROL ROD COUPLING/FULL-IN INDICATOR CHECKS to align with rod pull sheets.

2. Prepare Unit Supervisor turnover Sheet indicating:

0 GO-100-002 completed up to Step 282. 0 Unit 2 is 100 percent power.

0 All systems are OPERABLE.

0 Continue with Plant Startup.

3. Initialize Simulator to IC 181, five percent power. 4. Rod Sequence 82 SU Step 282. 5. Initiate Preference File: RESTOREPREF YPP.IL0-305 (See attached file copy Rev.

0, 09/17/05 for details.)

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 12 Rev. 0, 10/08/05 ILO-305 THIS PAGE IS INTENTIONALLY LEFT BLANK - Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 13 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT DESCRIPTION FORM Initial Conditions: Reactor power Five Percent Startup in Progress IAW GO-100-002 Rod Sequence B2 SU step Step 282 1 EVENT I TIME I DESCRIPTION Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 14 Rev. 0, 10/08/05 ILO-305 THIS PAGE IS INTENTIONALLY LEFT BLANK - FO~TTI NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 15 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 1 Brief

Description:

RAISE REACTOR POWER WITH CONTROL ROD WITHDRAWAL TIME 5

• Denotes Critical Task STUDENT ACTIVITIES Withdraw Control Rods in accordance with Reactor Engineer/CRC Instructions. Perform Coupling Checks per SO-156-007: When the control rod is Withdrawn to Position 48, Perform the following:

0 Depress the Withdraw pushbutton, Confirm the rod does not withdraw past Position 48 and the ROD OVERTRAVEL alarm does not come in. 0 Release the Withdraw pushbutton.

0 Depress Display Rods Full-in/Full-out test button and Confirm FULL OUT indication. Confirm the control rod remains at Position 48. 0 Record date and initials in appropriate space for the control rod in COUPLING CHECK on Attachment C, Page 1. Maintain overall control of evolution.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. As Given Page 16 Rev. 0, 10/08/05 ILO-305 Event No: 1 Brief

Description:

RAISE REACTOR POWER WITH CONTROL ROD WITHDRAWAL I NSTU CTO R ACT1 VlTY: NIA ROLE PLAY:

As necessary, to support Control Rod Withdrawals to raise power until sufficiently evaluated.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 17 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 2 Brief

Description:

BLOWN FUSE ON COMMON POWER SUPPLY TO IRMS B/D, SRM 6

• Denotes Critical Task STUDENT ACTIVITIES Recognize and respond to the following alarms: AR-104-AO6 AR-104-AO4 AR-104-AOl AR-104-806 SRM UPSCALE OR INOP AR-104-HO3 ROD OUT BLOCK OP-158-001 REACTOR PROTECTION SYSTEM IRM CHAN B/D/F/H UPSCALE TRIP OR INOP NEUTRON MON CHAN B SYSTEM TRIP RPS CHANNEL 81/82 AUTO SCRAM Recognize and report RPS Half-Scram to US. Dispatch NPO to investigate status of Power Supply breaker 1 D68204 on Elevation 771' of Control Structure.

Report NPO feedback to US When directed, reset Half-Scram in accordance with OP-158-001: Position RPS SCRAM RESET Control Switch HS-C72A-1 SO5 as follows: To GRP 1/4 position.

To GRP 2/3 position.

Observe RPS CHANNEL 81/82 AUTO SCRAM alarm CLEAR: When directed.

notifv GCCTTCC. Notify Work Week Manager (WWM) to obtain assistance in problem investigation.

Refer to T. S. 3.3.1.1 CONDITION A REQUIRED ACTIONS applicable due to loss of B and D IRM Channels Refer to T.S. 3.3.1.2 SRM Instrumentation Due to Loss of B SRM RPS INSTRUMENTATION No Action Required.

TRM 3.1.3 CONTROL ROD BLOCK INSTRUMENTATION No action required. Sufficient number of instruments operable per Trip Function.

Notify Plant Management and GCCTTCC of plant status.

Authorize reDlacement of blown fuse. Direct PCOM to reset Half-Scram when conditions permit. NOTES: Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 18 Rev. 0, 10/08/05 ILO-305 II INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 2 Brief

Description:

BLOWN FUSE ON COMMON POWER SUPPLY TO IRMWSRM INSTUCTOR ACTIVITY:

When Control Rod withdrawals and Reactor power increase has been sufficiently evaluated, insert a blown fuse failure by depressing:

[P-1] IMF NM175002B BLOWN FUSE IN POWER SUPPLY BREAKER 1 D68204 When the US has authorized replacement of the blown fuse, delete the malfunction by depressing:

[P-21 DMF NM175002B REPLACE BLOWN FUSE IN POWER SUPPLY BREAKER 1D68204 ROLE PLAY: 1. As NPO dispatched to 1068204, report no obvious problem, and that the breaker is in the closed position.

2. If requested to open and re-close breaker, wait a moment; then report the request as completed.

3. As WWM report that I&C and Electrical personnel have been dispatched to assist in the investigation.

4. As I&C, after the US has completed the TS exercise, report the F1 B Fuse is blown (as suspected) and can be replaced at this time.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 19 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 3 and 4 Brief

Description:

INADVERTENT START OF 'D' DIESEL GENERATOWESW PUMP TRIP ~___ ~~ r -[Dispatch NPO to ESSW Pumphouse to investigate conditions.

NOTE 2 Inform US to comply with T.S. 3.7.2 and TRM 3.7.1 per AR directions. Perform ON-054-001, LOSS OF ESW and start at least one ESW Pump in each loop. us Suspend Reactor Startup activities. Ensure cooling provided to DG within eight minutes. Direct PCOP to have DG tripped locally upon report of smoke and noise. Declare the DG INOPERABLE, refer to and comply with TS 3.8.1, AC SOURCES OPERATING, CONDITION B, and realize this is a Dual-Unit LCO. Ensure actions of ON-024-00and ON-054-001 are performed by PCOP. Declare ESW Pump INOPERABLE, refer to and comply with T.S. 3.7.2, ESW, CONDITION C, and realize this is also a Dual-Unit LCO.

Contact WWM to get assistance with DG and ESW problems. Notify Plant Management, TCC/GCC of Plant conditions.

~~ TStabilize Dlant conditions: conduct crew brief as time permits.

• Denotes Critical Task. Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 20 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 3 and 4 Brief

Description:

INADVERTENT START OF 'D' DIESEL GENERATOWESW PUMP TRIP INSTUCTOR ACTIVITY:

1. 2. 3. After the Half-Scram is reset, and the blown fuse evolution has been sufficiently evaluated, initiate an inadvertent start of the 'D' Diesel Generator by depressing:

Note: The following overrides simulate an inadvertent start, but simulate the PCOP had pushing the DG Start Pushbutton on the vertical Panel of OC653. Therefore, no emergency start signal is present, and no auto start of ESW will occur.

[P-3]10R ZDIHSOO051 D RESET [P-4]MOR ZDIHS00051 D NORMAL After the ESW Pump is started to supply cooling to the Diesel, wait two minutes to allow the crew sufficient time to address the DG start; then initiate the ESW Pump trip by depressing: Note: Only trip one of the two pumps in the loop. The goal is to allow the crew to have one pumplloop in service following the trip of the first pump. [PdIlMF PM03: OP504A OVERCURRENT TRIP OF "A" ESW PUMP (If started first) [P-61IMF PM03:

OP504C OVERCURRENT TRIP OF "A' ESW PUMP (If started first) When directed, perform an Emergency Manual Shutdown of DG at Local Panel OC521 D by depressing:

[P-AIOR QD143CMD LOCAL [P-8]10R QDEESD STOP [P-91MOR QDl5ESD NORMAL PLACE DG CS TO "LOCAL" AT OC521D DEPRESS EMERGENCY STOP PUSHBUTTON RELEASE EMERGENCY STOP PUSHBUTTON ROLE PLAY:

1. As NPO dispatched to the DG, report abnormal sounds from DG and a haze of blue smoke beginning to be observed.

2. 3. As NPO dispatched to the ESW Pumphouse, report no smoke or fire, but a smell of burnt wiring. As NPO dispatched to ESW Pump switchgear, report overcurrent relays tripped, standing by for further directions.

4. As WWM, inform US that teams will be dispatched to those areas as requested, will report findings as they arrive. 5. As FIN Team, wait 10 minutes; then provide initial report that the cause of DG start investigation is underway.

The Emergency Start Relays apparently did not pick up, and we will continue to troubleshoot.

6. After the DG has been shut down, report that the air is clearing, and no further abnormal indications exist.

Form NTP-QA91.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 21 Rev. 0. 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 5 and 6 Brief

Description:

RWCU PUMP ROOM LEAK /RWCU ISOLATION VALVES FAIL TO CLOSE ~~ POSITION I TIME PCOM I 50 us I + I I 70 ltPCOM + I

• Denotes Critical Task STUDENT ACTIVITIES Recognize and respond to AR-101 -A04 RWCU SYSTEM PRE-ISOLATION HI-TEMP/HI DlFF TEMP. Recoanizes/reDorts AR-lOl-A02/A03, RWCU LEAK DET IS0 LOGIC A/B HIGH TEMP. Attempt to Isolate RWCU as directed by placing the control switches to CLOSE. Reports RWCU will not isolate as indicated by dual valve position indication. Recognizes/reports 1 C614 RWCU leak detection alarmships. Recognize and respond to AR-SP-002, SIMPLEX FIRE DETECTION.

Monitors RWCU Pump Room temperature and radiation levels.

Enters EO-1 00-104, SECONDARY CONTAINMENT CONTROL based upon leak in RWCU Pump Room. Direct PCOM to manually scram the Reactor before any Secondary Containment Area reaches Max Safe Temperature.

Direct PCOM/P to manually isolate RWCU isolation valves in order to terminate a Primary System Discharging.

Manually scram the Reactor before any Secondary Containment Area reaches Max Safe Temperature.

Start all Room Coolers with a coolinrr source as directed.

I NOTES: I It Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 22 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 5 and 6 Brief

Description:

RWCU PUMP ROOM LEAWRWCU ISOLATION VALVES FAIL TO CLOSE INSTUCTOR ACTIVITY: After the Diesel Generator and ESW events have been sufficiently evaluated, initiate a leak in the RWCU System by depressing:

[P-101 IMF CU161007 1.0 1O:OO INITIATE AND RAMP LEAK IN RWCU PUPM ROOM OVER 10 MINUTES ROLE PLAY:

1. As NPO dispatched to the RWCU, report no indication of a fire, but the sound of a steam leak is present.

HP does not recommend entry into the room. 2. Report as NPO dispatched to check on the ESW, DG, etc.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 23 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 7 and 8 Brief DescriDtion:

SEVEN (7) STUCK RODS.

FAILED FUEL POSITION TIME STUDENT ACTIVITIES PCOM 70 When Mode Switch is placed to SHUTDOWN, report all rods did not insert.

Arm and depress MANUAL SCRAM PUSHBUTTONS.

I Report control rods drifting in. I When all rods fully inserted, report all rods in. ---I- PCOM/P I Report Reactor Building Area High Radiation Alarms. Recognize/report that radiation levels in the CRD and RWCU areas are above 10 Whour. Manually initiate ARI. Monitor Reactor and Turbine Building area radiation levels. us Directs monitoring of ARMS for second area >Maximum Safe Radiation levels. Direct closing MSlVs and MSL drains due to increasing radiation levels.

Enters and executes EO-1 00-1 13 LevelPower Control with report of all rods not inserted.

Exits EO-1 00-1 13 and re-enters EO-1 00-102 RPV Control when all rods inserted.

Re-enter EO-1 00-1 04 SECONDARY CONTAINMENT CONTROL due to unexpected Hiah Radiation.

PCOP Closes MSlVs and MSL drains, when directed due to high radiation conditions.

• Denotes Critical Task FOm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. As Given Page 24 Rev. 0, 10/08/05 ILO-305 Event No: 798 Brief

Description:

SEVEN (7)STUCK RODS, FAILED FUEL INSTUCTOR ACTIVITY:

1. When the Mode Switch is placed in Shutdown, initiate the failed fuel and rising area radiation levels by depressing the following seven pushbuttons in sequence:

[P-111 IMF RR179003 300 7:00 [P-131 IMF TR02:RlTl3752 13 1O:OO [P-14] IMF TROP:RIT13750 14 2:OO 0.27 [P-151 IMF TR02:RlTl3751 15 2:OO 0.27 [P-161 IMF TR02:RlTl3705 1000 2:00 [P-17] IMF TR02:RlTl3706 1000 2:OO 300 FAILED FUEL PINS RAMPED OVER 7 MINUTES RWCU AREA RADS RAMPED TO 13 WHR OVER 10 MINUTES CRD AREA NORTH RAMPED TO > 14WHR OVER 3 MINUTES CRD AREA SOUTH RAMPED TO

> 15WHR OVER 3 MINUTES CRD AREA RAMPED TO 1000 MWHR OVER 2 MINUTES CRD AREA RAMPED TO lo00 MWHR OVER 2 MINUTES [P-12] IMF TR02:RlTl3708 100 3:OO 0.15 RWCU PP AREA RAMPED TO 100 MWHR OVER 3 MINUTES

2. Delete the stuck rod malfunctions one at a time until all rods are fully inserted. ROLE PLAY:

As necessary FOm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 25 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No: 9 Brief

Description:

RAPID DEPRESSURIZATION POSITION TIME STUDENT ACTIVITIES PCOM/PCOP 80 Report RWCU area radiation level is > 10R/Hr, now have two areas greater than 1 OR/Hr. us Performs EO-1 00-1 12, RAPID DEPRESSURIZATION when two Secondary Containment Areas exceed Max Safe Rad levels. Verifies Suppression Pool Level 7 5 feet. Directs PCOP to open 6 ADS Valves.

• us *PCOP I I Opens 6 ADS SRVs to Rapidly Depressurize the reactor. Reports RPV pressure lowering.

us Ensures 6 ADS valves are open. Verifies RPV pressure lowering.

Performs EO-1 00-1 03 PRIMARY CONTAINMENT CONTROL due to high Suppression.

Pool Temperature. Directs maximizing Suppression Pool Cooling. PCOP I Places both loops of RHR in Suppression Pool Cooling in accordance with OP- 149-005. us After the scenario is complete, classifies the event as a Site Area Emergency Classification declared based on FSl. Loss or Potential loss of ANY two Fission Product Barriers.

• Denotes Critical Task FOr"I NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 26 Rev. 0, 10/08/05 ILO-305 Event No: 9 Brief

Description:

RAPID DEPRESSURIZATION INSTUCTOR ACTIVITY:

NA ROLE PLAY: As necessary TERMINATION CUE: Rapid Depressurization has been completed, RPV level is directed to be restored to +13 inches to

+54 inches, and actions are in progress to initiate Suppression Pool Cooling.

EVENT CLASSIFICATION: After the Scenario is complete, have the US classify the scenario for the HIGHEST EAL. Provide the US with any requested information needed to perform the classification. Site Area Emergency Classification declared based on FS1 Loss or Potential loss of ANY two Fission Product Barriers FOm NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given PPL-SUSQUEHANNA, LLC TRAINING CENTER SIMULATOR SCENARIO Prepared By: ~ Scenario Title:

ILO CertificatiodNRC Exam Scenario Scenario Duration:

90 Minutes Richard E. Chin 10/07/05 Instructor Date Scenario Number: ILO-602 Reviewed By: Nuclear Operations Training Supervisor

~ ~~ ~ RevisiodDate:

Rev. 0, 10/07/05 Date ~~ PCOO7/PCOO8 PCO17/PCO18 Initial License RO/SRO Certification Examination Initial License RO/SRO NRC Examination Course: Approved By: Supervising ManagedShift Supervisor Operational Activities:

Date FOI~ NTP-QA91.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 2 Rev. 0, 10/07/05 ILO-602 THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 3 Rev. 0, 10/07/05 ILO-602 The scenario begins with both Units at 100 percent power, all systems OPERABLE.

When the crew has assumed shift responsibilities, they will shift CRD Pumps in accordance with OP-155-001, SHIFTING CONTROL ROD DRIVE PUMPS to allow an awaiting Maintenance crew to record vibration data.

After the CRD Pump swap, the crew will experience a Leading Edge Flow Meter (LEFM)

Computer Failure, which will require them to perform ON-100-006, LOSS OF REACTOR HEAT BALANCE CALCULATION. Because the PlCSY is still available with pre-event reactor power level >3441 Mwth, they will declare LEFM INOPERABLE per TRO 3.10.4 and suspend any activities related to reactivity increase in the core.

After the LEFM event has been evaluated, the crew will experience a Main Steam Line Flow Transmitter failure, resulting in a steamflow/feedflow mismatch.

RPV water level will automatically lower and stabilize below the Low Level alarm setpoint. The crew will be required to perform ON-1 45-001, RPV LEVEL CONTROL SYSTEM MALFUNCTION.

This will require the crew to manually place the Feedwater Level Control System into Single-Element Control and restore RPV water level to the normal operating band. After RPV water level has been restored and stabilized, the crew will experience a Reactor Recirc Flow Unit 'D' failure downscale, which results in a RPS Half Scram and Rod-Out Block. The crew will be required to perform ON-164-001, RECIRC DRIVE FLOW INSTRUMENT FAILURE.

The crew will bypass the failed flow, remove its input to the Low Flow Auctioneer circuit, and reset the Half-Scram.

After the Flow Unit event has been evaluated, the crew will experience a trip of the running CRD. The standby pump is not designed to auto start. The crew will be required to perform ON-1 55-007, LOSS OF CRD SYSTEM FLOW. A report from the field indicates several HCU pressures are < 940 psig for the withdrawn control rods. The crew should declare the accumulators inoperable per TS 3.1.5 Shortly after this, multiple accumulator alarms will be received.

The CRD System will not be recoverable, forcing the crew to manually scram the Reactor, and enter EO-1 00-1 02, RPV CONTROL. - When the Reactor Mode Switch is placed in SHUTDOWN, the control rods will fail to insert due to a pre-inserted failure of RPS. This will require the crew to exit EO-1 00-1 0, RPV CONTROL and enter EO-100-1 13, LEVEUPOWER CONTROL due to the ATWS. When Standby Liquid Control is started, the "A" SLC pump will start, but will eventually trip on overcurrent.

The "B" SLC Pump shaft will shear, resulting in SLC injection. Division I Alternate Rod Insertion (ARI) will fail, preventing Control Rod movement.

RPV water level will be lowered to control power and alternate methods of control rod insertion will be attempted.

When RPV water level is lowered and stabilized with Feedwater, the Main Turbine will trip. Main Turbine Bypass Valves will remain closed because of a pre-inserted malfunction to keep them closed. In addition, a loss of 13.8 kV Auxiliary Buses 11 A and 11 B will occur, resulting in a loss of all normal high pressure injection sources. When started, Reactor Core Isolation Cooling (RCIC) speed control circuitry will fail at 1,000 rpm, resulting in no flow to the vessel. With the unavailability of HPCl and RCIC, RPV water level will decrease to TAF requiring performance of EO-100-1 12, RAPID DEPRESSURIZATION.

After Rapid Depressurization and subsequent RPV level restoration to >TAF, level will be restored and maintained with RHR LPCl mode between

-60 inches and -161 inches. Emergency Support Procedure ES-158-001, DE-ENERGIZING SCRAM PILOT SOLENOIDS will be successfully performed by field operators.

As a result of RPS fuses being pulled, all Control Rods will be fully inserted.

The scenario will be terminated when all Control Rods are fully inserted and actions are in progress to restore RPV water level to +13 inches to +54 inches. Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 4 Rev. 0, 10/07/05 ILO-602 THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 5 Rev. 0, 10/07/05 ILO-602 II SCENARIO OBJECTIVES I The objective of this scenario is to evaluate the Licensed Operator Candidate's ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

0 Knowledge of integrated plant operations 0 Ability to diagnose abnormal plant conditions 0 Ability to work together as a team 0 0 Ability to mitigate plant transients that exercise their knowledge and use of ONs and EOPs Ability to utilize Technical Specifications (SRO Only) To meet this objective, the licensed operator candidates must demonstrate proficiency in the following competencies: Reactor Operator Candidates:

1. InterpreVdiagnose events and conditions based on alarms, signals, and readings.

2. Comply with and use procedures, references, and Technical Specifications.

3. Operate the control boards. 4. Communicate and interact with other crew members. Senior Reactor Operator Candidates:

1. InterpreVdiagnose events and conditions based on alarms, signals, and readings.

2. Comply with and use procedures and references.

3. Operate the control boards (N/A to upgrade candidates).

4. Communicate and interact with the crew and other personnel.

5. Direct shift operations.

6. Comply with and use Technical Specifications.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 6 Rev. 0, 10/07/05 ILO-602 THIS PAGE IS INTENTIONALLY LEFT BLANK 11 FOn NTP-QA91.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 7 Rev. 0, 10/07/05 ILO-602 II CRITICAL TASKS II

• Recoanize failure to scram and inhibit ADS. Safety Significance Inhibiting ADS prevents uncontrolled injection of large amounts of relatively cold, unborated low pressure ECCS water when the reactor is not shut down with control rods. Consequences for Failure to Perform Task Failure to inhibit ADS can result in large amounts of positive reactivity addition due to boron dilution and cold water injection.

SSES EOP Basis for: The following steps provide the operating crew guidance to line up injection systems as available to maintain level >-129 inches.

If these actions are unsuccessful, the crew receives additional direction when it is determined that level can not be maintained above TAF. SSES EOP Basis for:

LQ/Q-3 IF INITIAL ATWS PWR > 5% OR CANNOT BE DETERMINED INJECT SLC (1) AND INH FADS When scram and ARI have failed, reactor power must be considered to determine if immediate boron injection is required.

If initial ATWS power was greater than five percent, then a relatively large number of control rods have failed to insert. The seriousness of this condition requires immediate injection of boron to positively terminate the ATWS event. ADS initiation may result in the injection of large amounts of relatively cold, unborated water from low-pressure injection systems.

With the reactor either critical or shut down on boron, the positive reactivity addition due to boron dilution and temperature reduction through the injection of cold water may result in a reactor power excursion large enough to cause substantial core damage. Preventing ADS is therefore appropriate whenever boron injection is required.

IndicationdCues for Event Requiring Critical Task ATWS with initial reactor power level greater than five percent APRM power. Performance Criteria Inhibit ADS by placing 1C601 Keylock Switches to INHIBIT Performance Feedback Successful ADS inhibiting is indicated by Green Indicating Light at switch illuminating.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 8 Rev. 0, 10/07/05 ILO-602 U CRITICAL TASKS I

• Insert control rods IAW EO-000-1 13, Sheet 2. Control Rod Insertion Safety Significance Control rod insertion initiates immediate power reduction.

Consequences for Failure to Perform Task Failure to insert control rods allows power to remain elevated with resultant power oscillations and potential core damage. Indicationdcues for Event Requiring Critical Task Exceeding a RPS scram setting with NO reactor scram signal, or RPS/ARI fail to fully insert all control rods. Performance Criteria Insert Control Rods by one or more of the following methods: Maximize CRD to drift control rods. Drive control rods after bypassing RWM and RSCS. Reset and Scram again by performing ES-158-002, BYPASS RPS LOGIC TRIPS. De-energizing RPS solenoids by performing ES-158-001.

Local venting of scram air header.

Performance Feedback Successful insertion of control rods will be indicated by: Rod position full-in indication for manual insertion of control rods, venting scram air header or de-energizing RPS solenoids.

Rod position full-in after resetting scram, draining scram discharge volume and re-scram.

• Lower RPV level to < -60 inches but

> -161 inches Safety Significance Core damage due to unstable operation can be prevented, or at least mitigated, by promptly reducing feedwater flow so that level is lowered below the feedwater spargers.

FO~TTI NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 9 Rev. 0, 10/07/05 ILO-602 I CRITICAL TASKS Consequences for Failure to Perform Task A General Electric Company study (NEDO-32047) indicates that the major threat to fuel integrity from ATWS is caused by large-amplitude power/flow instabilities.

These density-wave instabilities will most likely develop in the non-isolation ATWS where the Feedwater System is still available for makeup to the RPV. In this event, the Feedwater System maintains normal water level, but feedwater heating is lost due to tripping of the turbine. Without preheating of the feedwater, high levels of core-inlet subcooling develop, which can drive the reactor into a highly unstable mode of operation.

General Electric calculations indicate that power oscillations become large enough to cause melting of fuel in high-power bundles.

SSES EOP Basis for: LWL- 1 3 MAINTAIN LVL BETWEEN

-60" AND -161" USING TABLE 15 SYSTEMS BYPASSING INTERLOCKS AS NECESSARY IAW ANY: The purpose of this step is to uncover the feedwater spargers sufficiently to reduce core inlet subcooling. In the non-isolation ATWS event, core damage due to unstable operation can be prevented or at least mitigated by promptly reducing feedwater flow so that level is lowered below the feedwater spargers.

Once level drops below the sparger nozzles, which are located at -24 inches, the feedwater is sprayed into a region occupied by saturated steam. Steam will then condense on the injected feedwater, and the coolant will be heated as it falls to the liquid surface within the downcomer. Heating of the feedwater by steam condensation limits the buildup of core inlet subcooling and can prevent the onset of severe power/flo w instabilities.

This step identifies the widest acceptable water level control band. Although level fluctuations within this band are safe, it is very desirable to maintain level within the more restrictive taraet area of - 1 10 inches to -60 inches. The target area and expanded band are shown in Figure 8, Water Level Operation Guidance.

The intent of this step is to remain within the target band at all times unless prohibited by system perturbations, and remain within the expanded band at all times. Operation outside the target area has the following disadvantages:

The basis for an upper level of -60 inches is given above. A lower level of -1 10 inches is specified for the following reasons: 1. Provides a margin for core coverage.

Fo~ NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 10 Rev. 0, 10/07/05 ILO-602 CRITICAL TASKS I 2. Avoids operation near TAF where core power is more responsive to RPV pressure fluctuations.

3. Makes level control easier by maintaining level above the narrow region of the downcomer.

Below -1 70 inches the downcomer free area reduces from 300 rt.' to 88 e, resulting in increased magnitude of indicated level oscillations.

4. 5. Maintains sufficient core flow to carry liquid boron from lower plenum upward into the core.

As level is decreased below -1 10 inches, boron mixing efficiency is reduced because the natural circulation flow rate through the jet pumps is reduced, and is not as efficient at carrying the injected boron from the lower plenum upward into the core. At very low downcomer water levels near or below top of active fuel, there is little water available in the region above the jet pump throat for mixing with boron injected via RCIC.

In this situation, there is concern that boron may accumulate in the stagnant region of the downcomer which is below the jet pump throat.

5. Water level can be determined from wide range level instrumentation.

6. Avoids MSlV isolation setpoint of - 129 inches. RPV level below TAF is not, by itself, a determination of whether or not level can be maintained

> -161 inches. The determination that level cannot be maintained

> -161 inches must be made based upon: availability of high pressure injection systems, and. present level trend This decision must not be made prematurely, since depressurization of a critical core results in destabilizing affects and has a potential to cause core damage. Controlling reactor pressure, power and level with condensate and SRVs at 500 psig is difficult because all three parameters affect each other.

Therefore, rapid depressurization is recommended when high pressure injection cannot be obtained. The initial influence of reactor depressurization is stabilizing, since the additional flashing of liquid phase required for depressurization introduces excess voids in the reactor core, which can essentially terminate the fission process if the rate of depressurization is high enough. Once the depressurization is complete, however, the result is the immediate initiation of power excursions. Core damage is expected to occur from high clad stresses induced by: temperature excursions above the rewet temperature, PCI, cyclic fatigue, burnout or having the fuel enthalpy exceed the cladding failure threshold.

FOn NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 11 Rev. 0, 10/07/05 ILO-602 IndicationdCues for Event Requiring Critical Task ATWS with initial reactor power level greater than five percent APRM power. Performance Criteria Lower reactor water level by manually controlling injection rate from Feedwater, HPCI, andlor RCIC. Preventing injection such as RCIC and HPCI as level drops below

-30 inches and -38 inches respectively, may be required when Feedwater is available.

The preferred systems for use in controlling RPV water level are those Table 15 Systems which inject into the feedwater sparger or outside the core shroud. These are used because cold water is preheated by steam and the flowpath outside the core shroud mixes the relative/y cold injected water with the warmer water in the lower plenum prior to reaching the core. lnjection from SLC and CRD are always permitted during ATWS events. The operator throttles existing injection except CRD and SLC, and prevents unwanted injection as necessary to decrease level. Performance Feedback Lowering water level to -60 to -1 10 inches will result in power level lowering as indicated on the Average Power Range Monitors.

• Perform Rapid Depressurization when RPV level drops to

-161 inches Safety Significance RPV leakage or loss of injection systems impacts the ability to provide continued adequate core cooling through core submergence based on inventory loss. Consequences for Failure to Perform Task Failure to take the EOP actions will result in uncovering the core and breach of the fuel clad due to over heating.

SSES EOP Basis for: The following steps provide the operating crew guidance to line up injection systems as available to maintain level

>-129 inches. If these actions are unsuccessful, the crew receives additional direction when it is determined that level can not be maintained above TAF. Fo~ NTP-QA31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 12 Rev. 0, 10/07/05 ILO-602 RC/L- 4 RESTORE AND MAINTAIN LVL BETWEEN +13" AND +54" USING TABLE 3 SYSTEMS RC/L-5 IF LVL CANNOT BE RESTORED AND MAINTAINED

> +13" MAINTAIN LVL

> -129" USING TABLE 3 SYSTEMS AUGMENTING AS DESIRED WITH TABLE 5 ALTERNATE SUBSYSTEMS RC/L-lO IRRESPECTIVE OF VORTEX LIMITS WITH TABLE 3 SYSTEMS PERFORM ALL 1 LINE UP FOR INJECTION 2 STARTPUMPS 3 INCREASE INJECTION TO MAX RC/L-11 IF LESS THAN 2 TABLE 4 SUBSYSTEMS CAN BE LINED UP COMMENCE LINING UP AS MANY AS POSSIBLE TABLE 5 ALTERNATE SUBSYSTEMS RC/L-13 WITH TABLE 5 ALTERNATE SUBSYSTEMS PERFORM ALL: LINE UP FOR INJECTION START PUMPS INCREASE INJECTION TO MAX RCR-16 WHEN LVL CANNOT BE RESTORED AND MAINTAINED

> -161" GO TO RAPID DEPRESS Rapid Depressurization is not initiated until RPV water level has dropped to -161 inches (TAF) because: Adequate core cooling exists so long as RPV water level remains above -161 inches (TAF). The time required for RPV water level to decrease to -16linches (TAF) can best be used to line up and start pumps, attempting to reverse the decreasing RPV water level trend before Rapid Depressurization is required to assure continued adequate core cooling. (

Reference:

SSES-EPG (21-4 and second override before C3- 1) Indicationdcues for Event Requiring Critical Task Reactor water level trending downward, eventually indicating less than the top of active fuel height on the Fuel Zone Level Indicator.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 13 Rev. 0, 10/07/05 ILO-602 I CRITICAL TASKS I Performance Criteria Perform a Rapid Depressurization per EO-000-1 12 when water level reaches the TAF -1 61 inches as read on the Fuel Zone Instrument:

Initiate ADS/Manually Open all six ADS Valves. Performance Feedback Initiating a rapid depressurization causes Reactor pressure to lower to the shutoff head of the low pressure injection systems, allowing water level to rise on the Fuel Zone and Wide Range level instruments.

Verify ADS Valves are open using light red light indication, acoustic monitoring and lowering Reactor pressure and rising reactor water level.

• Slowlv increase iniection to restore and maintain RPV level to < -60 inches but > -161 inches Safety Significance Re-establishing injection into the RPV is required in order to adequately cool the core and to make up the mass of steam being rejected through open SRVs. Since the reactor may become critical during this evolution, injection into the RPV is increased slowly to preclude the possibility of large power excursions caused by rapid injection of cold unborated water. Consequences for Failure to Perform Task Failure to restore RPV level will result in uncovering the core and breach of the fuel clad due to over heating.

Failure to slowly increase injection may result in large power excursions caused by rapid injection of cold unborated water. SSES EOP Basis for: WHEN RAPID DEPRESS HAS BEEN INITIATED COMMENCE AND IRRESPECTIVE OF VORTEX LIMITS SLOWLY INCREASE INJECTION TO RESTORE AND MAINTAIN LVL BETWEEN

-60" AND -161" USING TABLE 15 SYSTEMS Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 14 Rev. 0, 10/07/05 ILO-602 1 CRITICAL TASKS The intent of this step is to re-establish injection in a controlled manner after rapid depressurization has been initiated.

"Initiated" is defined as ADS Valves have been opened, either automatically or manually.

It is not intended that the depressurization process be completed; only initiated.

As long as any number of ADS Valves has been opened in response to an automatic signal or manual action, this condition is met and injection may be slowly re-established.

Steps LQR-6, LQ/L-12 and LQ/L-13 permit use of these same Table 15 systems. Refer to RCR-6 for a complete explanation of the systems and cautions applicable to their use. Here, however, an explicit direction is given to commence injection "irrespective of vortex limits," since restoration of adequate core cooling takes precedence over adherence to normal operating limits. The undesirable consequences of uncovering the reactor core outweigh the risk of equipment damage which could result if vortex limits are exceeded. Immediate and catastrophic pump failure is not expected to occur should operation beyond these limits be required. A specific order governing the priority over use of these systems cannot be predetermined, as it will depend greatly on plant conditions. Consider the following factors to determine order: 0 0 System availability 0 0 System throttling capability/control Water quality Injection through FW spargers (preferred)

Time and manpower required to operate system Re-establishing injection into the RPV is required in order to adequately cool the core and to make up the mass of steam being rejected through open SRVs. Since the reactor may become critical during this evolution, injection into the RPV is increased slowly to preclude the possibility of large power excursions caused by rapid injection of cold unborated water. The level control band of -60 inches to -161 inches is the widest, acceptable water level control band. Although level fluctuations within this band are safe, it is very desirable to maintain level within the more restrictive target area of -1 10 inches to

-60 inches. The target area and expanded band are shown in Figure 8, Water Level Operation Guidance. Operation outside of the target area has numerous disadvantages, which are described in LQR-13. The intent of this step is to restore and maintain level within the target band at all times, unless prohibited by system perturbations or inadequate vessel injection, and remain within the expanded band at all times.

Form NTP-QA91.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 15 Rev. 0, 10/07/05 ILO-602 CRITICAL TASKS - CAUTION 12 PROLONGED OPERATION IN YELLOW AREA OF FIG 8 MAY RESULT IN ADDITIONAL CONTAINMENT LOADING AND PWR INSTABILITIES. This caution is added to alert the operator of the undesirable affects of operating outside the target band of -1 10 inches to

-60 inches. Operation outside the target band can result in increased power level, increased containment loading, and the potential for power oscillations. (

Reference:

SSES-EPG C5-5.2) Indicationdcues for Event Requiring Critical Task The intent of this step is to re-establish injection in a controlled manner after rapid depressurization has been initiated.

"Initiated is defined as ADS Valves have been opened, either automatically or manually.

It is not intended that the depressurization process be completed; only initiated.

As long as any number of ADS Valves has been opened in response to an automatic signal or manual action, this condition is met and injection may be slowly re-established. Performance Criteria Since the reactor may become critical during this evolution, injection into the RPV is increased slowly to preclude the possibility of large power excursions caused by rapid injection of cold unborated water. Performance Feedback Injection flow to the RPV is determined using the available instrumentation (Le., pressure, flow) for the respective injection system(s).

Observation of Neutron Monitoring System indications (i.e., IRMs, APRMs, and reactor period) will provide indication of the presence and severity of any power excursions.

• Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 16 Rev. 0, 10/07/05 ILO-602 Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 17 Rev. 0, 10/07/05 ILO-602 w SCENARIO REFERENCES II 1. SWAP IN-SERVICE CRD PUMPS OP-155-001 CONTROL ROD DRIVE HYDRAULIC SYSTEM, REV. 37 2. LEFM FAILURE ON-1 00-006 TRO 3.1 0.4 LOSS OF REACTOR HEAT BALANCE CALCULATION MISCELLANEOUS LEADING EDGE FLOW METER (LEFM) 3. MSL FLOW TRANSMITTER FAILURE ON-1 45-001 RPV LEVEL CONTROL SYSTEM MALFUNCTION AR-1 00-B17 RX WATER HI-LO LEVEL 4. RECIRC FLOW UNIT 'D' FAILS DOWNSCALE AR-103-AOl AR-103-C05 ON-164-001 OP-158-001 RPS SYSTEM, REV.

27 TS 3.3.1.1 RPS CHANNEL A1/A2 AUTO SCRAM, REV.

25 APRM/RBM FLOW/REFERENCE OFF NORMAL, REV. 25 RECIRC DRIVE FLOW INSTRUMENT FAILURE, REV. 9 RPS INSTRUMENTATION, AMENDMENT 178 5. LOSS OF CRD

/ INOPERABLE ACCUMULATORS AR-107-DO2 ON-1 55-007 ON-1 00-1 01 TS 3.1.5 CRD PUMP '6' TRIP, REV. 26 LOSS OF CRD SYSTEM FLOW, REV. 16 SCRAM, REV. 11 CONTROL ROD ACCUMULATORS, AMENDMENT 178 6. ATWS / SLC SYSTEM FAILURE

/ MAIN TURBINE TRIP 1 LOSS OF AUX BUSES EO-000-1 02 EO-000-1 13 OP-184-001 ES-150-002 RPV CONTROL, REV. 1 LEVEL POWER CONTROUCONTROL ROD INSERTION, REV. 1 MAIN STEAM SYSTEM, REV. 19 BORON INJECTION VIA RCIC, REV. 13 ES-158-001 DE-ENERGIZING SCRAM PILOT SOLENOIDS, REV. 6 7. RAPID DEPRESSURIZATION EO-1 00-1 12 EO-1 00-1 03 OP-149-005 RAPID DEPRESSURIZATION, REV. 1 PRIMARY CONTAINMENT CONTROL, REV. 2 RHR SUPPRESSION POOL COOLING, REV. 21 Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 18 Rev. 0, 10/07/05 ILO-602 r SCENE7 1. Set up the simulator for the scenario by performing the following:

a. Initialize the simulator to IC-20, both Units at 100 percent power EOL. 2. Type restorepref YPP.IL0-602; verify the following pre-inserts and Program Button assignments.

See attached Preference file for reference.

Verify the Environment window:

MALFS I REMFS I OVRDS I TRlGS 5:4 0 2:2 1 MALFUNCTIONS AVO1 :HV155F100 BR05:l A1 01 04 BR05:l A1 0204 PM03:l P208A (El 3:00 0) PM05:l P208B TC193025 TRIGGERWACTIONS El SLC.STRT-SW PROGRAM BUTTONS

[P-1] MRF RD155014 0 [P-21 MRF RD155014 100 60 [P-31 IMF FW145012 [P-5] IMF NM178012D

[P-61 MRF NM178008 ZERO

[P-81 IMF TC193001 [P-91 IMF RC150002 1000 60 4400 [P-4] IMF TR02:FTC321N003C 1.5 0 3.6 [P-q bat YPBJLO-W2A

[P-101 bat YPB.IL0-602C

[P-111 bat YPBJLO-602D

[P-231 BAT FWB.101 ALARM [P-241 BAT FWB.102ALARM

[P-251 BAT FWB.103ALARM LOSS OF POWER TO HPCl HV-F1 00 SOLENOID AUX BUS 11A BREAKER FAILURE AUX BUS 11 B BREAKER FAILURE 'A' SLC PUMP MOTOR SHORT CIRCUIT

'B SLC PUMP SHAFT SHEAR ALL BYPASS VALVES FAIL CLOSED SLC START SWITCH IN START HV-146F014B CLOSED HV-146F014B OPEN (Ramps OPEN over 60 sec) LEFM COMPUTER FAILURE MSL FLOW TRANSMITTER FAILS TO 1.5 MLBM/HR 'D FLOW UNIT FAILS DOWNSCALE 'D' FLOW UNIT SWITCH TO ZERO ATWS-ELEC/LOSS OF CRDI4 ACCUM ALARMS MAIN TURBINE TRIP RClC SPEED CONTROL FAILURE

@ 1,000 RPM ES-158-001 DIV 1 FUSES PULLEDELEAR HCU ALARMS ES-158-001 DIV 2 FUSES PULLED FW HEATER PANEL ALARM RESET FW HEATER PANEL ALARM RESET FW HEATER PANEL ALARM RESET

3. Verify LEFM is selected as the Feedwater flow input to PlCSY IAW 01-TA-021.

4. Prepare a turnover sheet indicating:

a. Unit 1 is at 100 percent power EOL.

b. Swap CRD Pumps at beginning of shift to allow Maintenance to record vibration data on

'B' CRD Pump. The Maintenance crew is standing by at the pump. c. Unit 2 is in MODE 1 at 100 percent power EOL. 5. Make several copies of Pages 1 and 2 of Shutdown Control Rod Sequence 82 to use as page replacements following completion of the scenario. 6. Run Simulator to stabilize conditions.

Rev. 0 (03/04) 2005 NRC Exam, As Given Form NTP-OA91.7A Page 19 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT DESCRIPTION FORM Initial Conditions:

EVENT TIME DESCRIPTION

-- 1 5 SHIFT IN-SERVICE CRD PUMPS 2 15 LEFM COMPUTER FAILURE 3 30 "C" MSL FLOW TRANSMITER FAILURE 4 45 RECIRC FLOW UNIT D FAILS DOWNSCALE 5 55 LOSS OF CRD/INOPERABLE ACCUMULATORS 6 60 FAILURE TO SCRAM/ATWS 7 SLC SYSTEM FAILURE 8 MAIN TURBINE TRIIP/LOSS OF AUX BUSES I I 9 RClC SPEED CONTROL FAILURE 10 90 RAPID DEPRESSURIZATION Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 20 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No: 1 Brief

Description:

SHIFT IN-SERVICE CRD PUMPS POSITION SRO PCOP us TIME STUDENT ACTIVITIES Directs PCO to place '8' CRD Pump in service and shut down the 'A CRD Pump. Implements OP-155-001, Section 2.9. 1. 2. 3. 4. 5. Directs NPO to: 0 Perform equipment pre-start checks. 0 Start 'B' CRD Pump by placing control switch to RUN. Direct NPO to:

0 Slowly open 146F0148, CRD Pump 'B' Discharge, to FULL OPEN position.

0 On PI-146068, Check 1 P132B, CRD Pump 6, Gear Box oil pressure - 20 psig. 0 Check 1 P132B CRD Pump B, Gear Box oil temperature - 100 OF. Stop CRD Pump A by placing control switch to STOP. On PI-Cl2-1 R601, Panel 1 C601, Check CRD discharge pressure

-1,450 psig. Close CRD Pump 'B Discharge Valve. 6. Ensure PDI-(212-1 R602, Drive Water Diff Pressure, - 250 psid. Notifies Maintenance that "6 CRD Pump is in service.

• Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. As Given Page 21 Rev. 0, 10/07/05 ILO-602 Event No: 1 Brief

Description:

SHIFT IN-SERVICE CRD PUMPS INSTRUCTOR ACTIVITY:

NOTE: Monitor P&ID RD1. When directed to close 'B' CRD Pump Discharge Valve HV-146F0148, Depress P-1:

[P-1] MRF RD155014 0 146F014B CLOSED When directed to reopen HV-146F0148, Depress P-2:

[P-21 MRF RD155014 100 60 146F014B OPEN (Ramps open over 60 seconds.)

ROLE PLAY: As Plant Operator at the CRD Pumps: 1. Report oil levels in the pump, motor and speed increaser are normal. 2. When requested after pump start, report local parameter values are: Gearbox Oil Pressure - 22 psig, and Gearbox Oil Temperature - 95 OF. FOITII NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 22 Rev. 0, 10/07/05 ILO-602 I ' SCENARIO EVENT FORM I Event No: 2 Brief

Description:

LEFM COMPUTER FAILURE POSITION 1 TIME 1 STUDENT ACTIVITIES PCOM Recognize and respond to Computer Alarm and indications. Report LEFM failure to SRO. I Utilize APRMs for indication of Reactor Power. When directed, ensure RX power S3489 Mwth by reducing Core flow by 0.5 Mlb/hr. Direct performance of ON-100-006, LOSS OF REACTOR HEAT BALANCE CALCULATION. Declare LEFM INOPERABLE per TRO 3.1 0.4. A.l A.2 Acknowledge the following requirements: Within six hours: Contact STA to select Venturi flow elements for input to OD-3 IAW THERMAL POWER. Contact Workweek Manaaer/lC for assistance.

SRO Immediately suspend any and all activities related to reactivity increase in the core, including control rod withdrawal and recirculation pump speedblow increase. Within six hours, reduce the indicated THERMAL POWER to less than or equal to 3441 MWt. 01-TA-021, SELECTION OF FEEDWATER INPUTS FOR CALCULATION OF CORE PCOM/P Direct NPO to investigate locally. Direct NPO to attempt to re-close 1 Y128 Breaker 38. Report breaker failure to SRO.

• Denotes Simulator Critical Task.

Form NTP-QA91.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 23 Rev. 0, 10/07/05 I LO-602 Event No: 2 Brief

Description:

LEFM COMPUTER FAILURE INSTRUCTOR ACTIVITY:

After 'B' CRD Pump is in service, initiate a LEFM computer Failure by depressing:

P-31 IMF FW145012 LEFM COMPUTER FAILURE ROLE PLAY: As NPO: 0 0 0 When directed to investigate, report that the Display Monitor (1C1107) appears de-energized. Report 1 Y128 Breaker 38 is tripped.

If asked to re-close breaker, report breaker will not stay closed.

As I&C: After about 10 minutes report there appears to be an internal problem, and that LEFM data is not valid.

As STA: If/when requested select Venturi flow elements for input to OD-3 IAW 01-TA-021, SELECTION OF FEEDWATER INPUTS FOR CALCULATION OF CORE THERMAL POWER.

Form NTP-QABl.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 24 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

3 Brief

Description:

"C" MSL FLOW TRANSMITTER FAILURE

~~ STUDENT ACTIVITIES Respond to lowering RPV water level alarm. Perform ON-145-001, RPV LEVEL CONTROL SYSTEM MALFUNCTION, Section 3.6: 0 Place in AUTOMATIC.

Place in Single-Element Control as follows:

Verify LIC-C32-1 R600 controller responding correctly and maintaining level

<+54 inches and >+13 inches. Place LIC-C32-1 R600 controller in MANUAL. Raise and maintain RPV water level >30 inches as indicated on the operable Level Indicators LIC-C32- 1 R606A( 8) (C). Depress Green 1 ELEM PB for 1 OR 3 ELEMENT LEVEL CONTROL HS-106102.

Null FW LEVEL CTUDEMAND SIGNAL LIC-C32-1 R600 controller. Adjust LIC-C32-1 R600 to maintain RPV water level - 35 inches.

Directs implementation of ON-145-001, RPV LEVEL CONTROL SYSTEM MALFUNCTION.

• Denotes Simulator Critical Task. ~ ~~ NOTES: FOITI NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 25 Rev. 0, 10/07/05 ILO-602 Event No: 3 Brief

Description:

"C" MSL FLOW TRANSMITTER FAILURE INSTRUCTOR ACTIVITY:

When LEFM failure has been adequately evaluated, initiate a failure of the "C MSL Flow Transmitter by depressing P-4: [P4] IMF TR02:FK321 N003C 1.5 0 3.6 MSL FLOW TRANSMITTER FAILS TO 1.5 MLBMRlR ROLE PLAY: As I&C: Report that isolation logic relays are not affected by this failure. This is only feeding into the Feedwater Level Control circuit. Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 26 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No: 4 Brief

Description:

RECIRC FLOW UNIT 'D' FAILS DOWNSCALE 11 POSITION I TIME STUDENT ACTIVITIES Reports RPS half-scram condition on Division 2. Refers to AR-103-AO1, RPS CHANNEL Al/M AUTO SCRAM.

Refers to AR-103-CO5, APRWRBM FLOW/REFERENCE OFF-NORMAL.

Refers to ON-1 64-001, RECIRC DRIVE FLOW INSTRUMENT FAILURE. 1. 2. 3. Verifies control rod block. Determines

'C' and 'D Flow Units COMPARE lights are illuminated.

Determines and reports 'D' Flow Unit has failed downscale.

Dispatches a Plant Operator to LPR to investigate Flow Unit status. Directs implementation of ON-1 64-001, RECIRC DRVlE FLOW INSTRUMENT FAILURE.

1. 2. Directs bypassing

'D Flow Unit on Panel 1 C651 using the joystick. Directs resetting the Division 2 RPS half-scram signal. Contacts WWM to investigate failure of Recirc Flow Unit 'A'. BvDasses Flow Unit

'A' at 1 C651 bv Dlacina JOYSTICK in 'A' Dosition.

PCOM Dispatches Plant Operator to bypass Flow Unit 'A' at Panel 1 C608. At Panel 1 C651, verifies APRM flow-biased half-scram and rod block signals clear.

Resets half-scram IAW OP-158-001:

1. Resets RPS Trip System by Momentarily Positioning RPS SCRAM RESET Control Switch HS-C72A-1 SO5 as follows:

0 To GRP 1/4 position 0 To GRP 2/3 position 2. Observes the RPS CHANNEL Al/M AUTO SCRAM alarm CLEAR.

• Denotes Simulator Critical Task. NOTES: I Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 27 Rev. 0, 10/07/05 ILO-602 - INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No: 4 Brief

Description:

RECIRC FLOW UNIT 'D' FAILS DOWNSCALE INSTRUCTOR ACTIVITY: When the plant is stable after the MSL Flow Transmitter failure, insert the

'D Recirc Flow Unit failure; Depress P-5: [P-51 IMF NM178012D 'D' FLOW UNIT FAIL DNSC When directed to place the 'D' Flow Unit to ZERO in the LRR, Depress P-6: [P-61 MRF NM178008 ZERO 'D' FLOW UNIT SWITCH TO ZERO ROLE PLAY: As Plant Operator sent to LRR Panel 1 C608, wait two minutes and call Unit 1 on the Page; report all flow unit mode switches are in operate and Flow Unit 'D' has downscale indication.

If asked about additional indications, the following exist: Amber compare light is ON for Flow Units 'C' and ID'. Along the top of Panel 1 C608: 0 0 0 0 White FLOW UNIT COMPARATOR lights are ON for 'C' and 'D' Flow Units.

Red UPSC THERM TRIP lights are ON for all Division 2 APRMs. Red THERM FIRST lights are ON for all Division 2 APRMs. Amber UPSC lights are ON for all Division 2 APRMs. As Workweek Manager acknowledge the direction to investigate the flow unit failure.

No other feedback will be provided.

Fo~ NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, As Given Page 28 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event Nos: 5 Brief

Description:

LOSSOF CRDANOPERABLE ACCUMULATORS POSITION TIME STUDENT ACTIVITIES PCOP PCOM Reports trip of 'B CRD Pump. Refers to AR-107-DO2, CRD PUMP 'B TRIP. Implements ON-1 55-007, LOSS OF CRD SYSTEM FLOW: 1. 2. Dispatches Plant Operators to check CRD Pumps and Breakers 1 A201 07 and 1 A20407. Reports charging water header pressure.

Reports control rod accumulator trouble alarms.

Reports all trouble alarms are for withdrawn control rods. Dispatches a Plant Operator to the HCU to report status.

Closes Flow Control Valve FV-l46-F002 using FC-C12-1 R600 in MANUAL. Attempts to START 'A CRD Pump; reports 'A CRD Pump failed to start. us Directs implementation of ON-155-007, LOSS OF CRD SYSTEM FLOW. Directs placing mode switch to S/D within 20 minutes when two or more accumulators are determined inoperable for withdrawn control rods with steam dome pressure

>900 psig. Contracts WWM to investigate the accumulator troubles.

Discusses or requests Unit 2 CRD X-tie to Unit 1. Directs actions IAW ON-100-101, SCRAM/SCRAM IMMINENT. Refers to TS 3.1 5, CONTROL ROD ACCUMULATORS CONDITIONS A AND 6. Reduces RRP speeds to minimum if directed and time permits.

PCOM

• Denotes Simulator Critical Task.

!I 11 NOTES: I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 29 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, Event No: 5 Brief

Description:

INSTRUCTOR ACTIVITY:

LOSS OF CRDANOPERABLE ACCUMULATORS After RPS half-scram is RESET, insert a loss of CRD System flow with several control rod accumulator trouble alarms; Depress P-7: [P-A bat YPBJLO-602A ATWS-ELECLOSS OF CRD/4 ACCUMULATOR ALARMS NOTE: Times for accumulator alarms are: Rod 46-47 45 seconds Rod 38-39 2 minutes Rod 10-1 9 Rod 14-23 3 minutes 2 minutes 30 seconds Monitor Display RD-11 for CRD Temperatures if required. ROLE PLAY:

As Plant Operator sent to 'B' CRD Pump Breaker 1A20407, wait two minutes, and report overcurrent relay 50/51 has a target dropped. As Plant Operator sent to 'A' CRD Pump Breaker 1A20107, wait two minutes and report no abnormal conditions exist on the breaker.

As Plant Operator sent to

'B' CRD Pump, wait two minutes and report no abnormal conditions exist on the pump. As Plant Operator sent to HCUs for accumulator trouble alarms report pressures less than 900 psig for Rod 46-47, Rod 38-39, Rod 10-19, and Rod 14-23.

NOTE: In order to avoid a long time delay in candidate activities, make it clear that neither pump appears to be capable for return to service any time soon. As NPO sent to CRD X-tie, report difficulty opening valve, requesting Maintenance support to get it open.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 30 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM STUDENT ACTIVITIES When directed, places Mode Switch to SHUTDOWN.

Recognizes and reports failure to scram. Inserts manual scram via scram pushbuttons; reports continued failure to scram. Inserts SRMS and IRMs. Enters EO-000-102, RPV CONTROL, exits EO-000-1 02 and enters EO-000-1 13, LEVEL POWER CONTROL. Directs initiatina SLC. Directs ADS inhibited.

Directs tripping Recirc Pumps one at a time while monitoring RPV level for swell.

Initiates ARI; reports failure to scram via ARI. Inhibits ADS. Depress ADS Logic A and B Timer Reset Switches HS-B21-1 S13A and HS-B21-1 S13B. Places ADS A and B Logic Control Keylock Switches to INHIBIT. Initiates SLC; reports 'B'; SLC Pump tripped on start; 'A' Pump running. Dispatches Plant Operator to investigate

'B SLC Pump failure.

Recognizes and reports subsequent trip of 'A' SLC Pump (three minutes). Directs insertion of control rods IAW EO-000-1 13, Sheet 2, CONTROL ROD INSERTION.

1. Directs venting the scram air header

2. Inserts control rods IAW EO-000.1 13, Sheet 2, CONTROL ROD INSERTION. Directs Plant Operator to vent scram air header.

Directs performance of ES-158-001, DE-ENERGIZING SCRAM PILOT SOLENOIDS.

• Denotes Simulator Critical Task. Form NTP-QABl.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 31 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, Event No: 697 Brief

Description:

ATWWSLC SYSTEM FAILURE INSTRUCTOR ACTIVITY:

Ensure trigger El activates to trip 'A SLC Pump three minutes after start.

IMF PM03:1P208A (El 3:W 0) A SLC PUMP MOTOR SHORT CIRCUIT ROLE PLAY:

As Plant Operator directed to investigate SLC Pumps, wait - 2 minutes and report

'B' Pump shaft is sheared.

When directed to investigate the

'A' SLC Pump trip, wait - 2 minutes and report the motor appears to have scorching.

FOtm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 32 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

899 Brief

Description:

MAIN TURBINE TRIP/LOSS OF AUX BUSEWRCIC SPEED CONTROL FAILURE POSITION TIME STUDENT ACTIVITIES

• us Directs lowering RPV water level to 40 inches but >-161 inches. Directs a target RPV water level band of -60 inches to -1 10 inches using Feedwater.

Directs overriding RClC System injection.

Directs RPV pressure stabilized below 1,087 psig with SRVs.

Directs bypassing MSlV and CIG interlocks IAW OP-184-001, MAIN STEAM SYSTEM. Directs SLC injection with RClC IAW ES-150-002, BORON INJECTION VIA RCIC. Directs Workweek Manager to investigate Aux Bus problem. NOTE 1 I *PCOM Lowers RPV water level to <-60 inches but >-161 inches.

1. Reduces RFP speed/discharge pressure to lower RPV level. At Panel 1 C645 place HS-621 -S38A and HS-B21 -S38C to BYPASS. PCOPrn Recognize and report failure of Auxiliary Buses 1 1 A and 11 B.

• Denotes Simulator Critical Task. Form NTP-QA31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 33 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, Event No:

8Y9 Brief

Description:

MAIN TURBINE TRIPROSS OF AUX BUSEWRCIC SPEED CONTROL FAILURE INSTRUCTOR ACTIVITY:

If necessary, when RPV water level is lowered into the target band insert a trip of the Main Turbine; Depress P-8: [P-81 IMF TC193001 MAIN TURBINE TRIP NOTE: When the Main Generator lockout occurs, Aux Buses 11N11 B should fail to transfer and will de-energize, and cause a loss of Feedwater and Condensate. Following the Turbine Trip when Reactor Vessel Water Level is being maintained with RCIC, Depress P-9:

[P-91 IMF RC150002 lo00 60 4400 RCIC TURBINE SPEED CONTROL FAILURE ROLE PLAY: 1. As Workweek Manager if directed to investigate Aux Bus breaker problem, wait - 3 minutes, and inform crew that a bus fault is present; you are continuing to investigate.

2. As NPO directed to vent the Scram Air Header, wait - 3 minutes, and report that you are unable to get the cap off of the 147007 Valve's vent line and that the cap appears to be galled; you request Mechanical Maintenance assistance.

3. As FUS directed to perform ES-150-002, acknowledge the direction and perform no further actions.

Fo~ NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. As Given Page 34 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION TIME STUDENT ACTIVITIES Directs Rapid Depressurization when RPV level drops to -1 61 inches. 1. 2. 3. 4. 5. Performs Rapid Depressurization by opening all ADS SRVs. 1. 2. 3. Enters EO-100-1 12, RAPID DEPRESSURIZATION.

Directs STOPPING and PREVENTING injection, except for SLC, CRD, RClC and HPCI. Verifies Suppression Pool level >5 feet. Directs opening all ADS SRVs. Verifies all ADS SRVs are open. Arms and depresses Division 1 and/or Division 2 ADS manual pushbuttons and verifies six red lights lit for ADS solenoids, or Places individual control switch to open for each ADS SRV (G, J, K, L, M and N) and verifies red light lit and amber light not lit for each valve solenoid. Verifies six ADS SRVs are open: 0 Observes six ADS SRVs open on acoustic monitor status light indication.

0 Observes RPV pressure decrease.

0 Observes elevated tailpipe temperatures on TRS-B21-1 R614. Directs slowly increasing injection to restore and maintain RPV level to <-60 inches but >-161 inches using LPCI. Directs LPCI injection through the RHR heat exchangers as soon as possible.

Slowly increases injection to restore and maintain RPV level to e60 inches but >- 161 inches using LPCI. 1. Throttles open HV-151 -F017A(B) at a rate to preventlminimize power oscillations.

2. Injects through the RHR heat exchangers as soon as possible.

• Denotes Simulator Critical Task. Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. As Given Page 35 Rev. 0, 10/07/05 ILO-602 AND INSTRUCTOR'S PERSONAL NOTE Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION INSTRUCTOR ACTIVITY: When the Rapid Depressurization is begun, cross tie Unit 1 and Unit 2 CRD; Depress P-10:

[P-lo] MRF RD155005 OPEN 146016 CRD CROSS TIE FROM UNIT 2.

When RPV water level is restored

> -161 inches following RD proceed with pulling RPS Fuses. To pull Division 1 RPS fuses, Depress P-1 1 : [P-113 bat YPBJLO-602C ES-158-001 DIV 1 FUSES PULLED AND CLEARS HCU TROUBLE ALARMS. NOTE: The above file also deletes the Accumulator Fault malfunctions so that all rods can be fully inserted later in the scenario.

When RPV level is stabilized at e 40 inches but

> -1 61 inches following Rapid Depressurization, pull Division 2 RPS fuses to complete ES-158-001; Depress P-12: [P-121 bat YPB.IL0-6MD ES-158-001 DIV 2 FUSES PULLED ROLE PLAY: As Shift Manager report that Unit 1Nnit 2 CRD is crosstied.

As Operator dispatched to perform ES-158-001, wait - 2 minutes and report that you are ready to pull the Division 1 RPS fuses. As Operator pulling fuses, call the Control Room and report you have completed pulling Division 1 RPS fuses, and you are now going to the LRR to pull Division 2 fuses. As Operator pulling fuses, wait - 2 minutes and report that RPS Division 2 fuses have been pulled, and ES-158-001 is now completed.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam. As Given Page 36 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION POSITION TIME STUDENT ACTIVITIES

• PCOM Inserts control rods IAW EO-000-1 13, Sheet 2, CONTROL ROD INSERTION.

1. Coordinates ES-158-001 with FUSMPO. 2. 3. 4. Verifies indications as Division 1 RPS Fuses are pulled. Verifies control rod insertion as Division 2 RPS Fuses are pulled. Verifiesheports all rods fully inserted.

us Directs SLC injection terminated and restoration from ES-150-002.

Exits EO-000-1 13, Sheets 1 and 2; re-enters EO-000-102.

Directs establishing RPV water level +13 inches to +54 inches. PCOP Establishes RPV water level +13 inches to +54 inches with LPCI.

us Enters EO-000-1 03, PRIMARY CONTAINMENT CONTROL.

Directs RHR placed in Suppression Pool Cooling. I I

• Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, As Given Page 37 Rev. 0, 10/07/05 ILO-602 Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION INSTRUCTOR ACTIVITY:

As necessary ROLE PLAY: As necessary Form NTP-QA91.7A Rev. 0 (03/04) 2005 NRC Exam, As Given Page 38 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION TIME STUDENT ACTIVITIES Places both loops of Suppression Pool Cooling in service IAW OP-149-005, RHR SUPPRESSION POOL COOLING. 1. Places ESW in service. 2. 3. 4. 5. 6. 7. 8. Places RHRSW in service to RHR Heat Exchangers AB. Opens Suppression Chamber Test Shutoff Valve HV-151 -F028 AB. Starts RHR Pump 1 P202A(C)/B(D).

Throttles open Test Line Control Valve HV-F024NB to achieve 51 0,000 gpm on Observes Minimum Flow Valve HV-151 -F007 A/B closes at -3,000 gprn. Closes Heat Exchanger Bypass HV-151 -F048 NB. Checks RHR Pump Room Coolers 1V210 A(C)/B(D).

FI-Ell-1 R603 AB. After the scenario is complete, classifies the event as a SITE AREA EMERGENCY under EAL MS3 due to RPV and ARI failure, OR classifies the event as a SITE AREA EMERGENCY under EAL FS1 due to a Loss or Potential Loss of the Fuel Clad Barrier and a Loss of the RCS Barrier.

• Denotes Simulator Critical Task. Form NTP-QAB1.7A Rev. 0 (OW04) 2005 NRC Exam, As Given Page 39 Rev. 0, 10/07/05 ILO-602 Event No: 10 Brief

Description:

RAPID DEPRESSURIZATION INSTRUCTOR ACTIVITY:

As necessary ROLE PLAY:

As necessary TERMINATION CUE: All control rods are inserted and actions are in progress to restore RPV water level to +13 inches to +54 inches. EVENT CLASSIFICATION:

After the scenario is complete, have the US classify the scenario for the HIGHEST EAL. Provide the US with any requested information needed to perform the classification.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, As Given