ML053210433

Draft - Section C Operating Exam (Folder 2A)

ML053210433
Person / Time
Site:	Susquehanna
Issue date:	08/19/2005
From:	Roush K Susquehanna
To:	Todd Fish Operations Branch I
Conte R
References
Download: ML053210433 (150)

Text

Scenario

Title:

Prepared By:

Reviewed By:

PPL-SUSQUEHANNA, LLC TRAINING CENTER ILO Certification/NRC Exam Scenario Richard E. Chin 10/07/05 Instructor Date Nuclear Operations Training Supervisor Date SIMULATOR SCENARIO Approved By:

Supervising ManagedShift Supervisor Scenario Duration:

80 Minutes Scenario Number:

ILO-304 Date RevisiodDate:

Rev. 0, 10/07/05 Course:

PCOO7/PCOO8 PCO17/PCO18 Initial License RO/SRO Certification Examination Initial License RO/SRO NRC Examination Operational Activities:

Drywell Cooler Fan Trip Core Power Reduction RRP MG Set Temp Controller Failure Loss of Extraction Steam Rods Drift RPS Failure/ARI Success Loss of 13.8 kV Aux Bus 11 A and 11 B Instrument Line Break in DW HPCl Auto Start Failure RHWSumression Pool Leak

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SCENARIO

SUMMARY

n The scenario begins with both Units at 100 percent power. A Minimum Generation Emergency has been declared, and Unit 2 has previously reduced load by 200 Mwe. The next down-power will be taken on Unit 1 if needed. Fuel handling is in progress in Unit 1 Spent Fuel Pool.

Shortly after the crew has assumed shift responsibilities, Drywell Cooling Fan 1 V416B will trip. This will require the crew to perform Alarm Response Procedure AR-128-803. Because this fan is required by Technical Specifications, the US will need to comply with T.S. 3.6.3.2.

Following the Drywell Cooling Fan evolution, the crew will be notified by Generation Control Center (GCC) that a Minimum Generation Emergency has been declared and that a 100 MW reduction on Unit 1 is requested ASAP. The crew will reduce power in accordance with 01-AD-029, EMERGENCY LOAD CONTROL and GO-100-012, POWER MANEUVERS.

Following the load reduction, the RRP 'A' MG-Set Hydraulic Fluid Temperature Controller auto output fails, resulting in high oil temperature supply to the MG-Set bearings and fluid coupler. The crew will perform Alarm Response Procedure AR-102-C05 in order to take manual control of the controller to lower the oil temperature. If no action is taken, a Drive Motor breaker trip and Scoop Tube Lock will occur due to the high temperature. Sufficient time is allotted for the crew to correct the failure.

When MG-Set hydraulic Fluid Temperature Controller is restored and temperatures have stabilized, a loss of Extraction Steam to 4B Heater occurs. The crew will respond by lowering power to 75 percent in accordance with ON-147-001, LOSS OF FEEDWATER HEATER EXTRACTION STEAM. Additionally, the crew will need to address the MCPR LCO 3.2.2.

After power has been stabilized and the Feedwater Heaters are isolated, three Control Rods will drift from Position 48, requiring an IMMEDIATE OPERATOR Scram in accordance with OP-AD-055, Attachment B, Immediate Operator Action List. The Reactor Mode Switch fails to scram the reactor, but ARI will successfully insert the control rods. The crew will enter EO-1 00-1 02, RPV CONTROL, and when report that all rods did not insert, the Unit Supervisor may recognize an entry to EO-1 00-1 13, LEVEUPOWER CONTROL. Once ARI has been determined to be successful, EO-1 00-1 13 will no longer be applicable, and re-entry to EO-1 00-1 02 will occur. If the US decides to monitor for the success of ARI and observes all rods inserted, EO-1 00-1 13 entry would not be required.

Immediately following the scram, the Aux Buses 11A and 11 B will fail to transfer, resulting in a loss of Condensate and Feedwater Systems and a low RPV water level (-38 inches) isolation initiation and an auto initiation of HPCI. HPCl will fail to auto start, but should be started using the component by component start up method. RPV water level will be maintained with injection from HPCI, RCIC, CRD and SLC. RPV pressure will be controlled by SRV actuation.

After HPCl is manually started and RPV water level restoration is in progress, an Instrument Line Rupture occurs inside the Drywell. This rupture will cause a loss of Division 1 RPV Level and Pressure indications, as well as a rise in Primary Containment temperature and pressure. The crew will enter EO-100-103, PRIMARY CONTAINMENT CONTROL to address those issues.

When RHR is initiated to perform Suppression Chamber/Drywell Spray, a pipe break between the suction valve and the RHR Pump will occur, requiring the crew to stop the RHR Pumps and isolate the suction valves. The crew will enter EO-100-104, SECONDARY CONTAINMENT CONTROL, due to the RHR ROOM FLOODED alarm. The crew may also re-enter EO-1 00-103, PRIMARY CONTAINMENT CONTROL, if Suppression Pool level lowers to below 22 feet.

The crew should re-commence Suppression Chamber Sprays, and when Suppression Chamber Pressure exceeds 13 psig, initiate Drywell Sprays, limiting flow to between 1,000 gpm and 2,800 gpm for the first 30 seconds.

The scenario will terminate when RPV water level is being maintained >TAF, Suppression Chamber and Drywell sprays have been initiated to control Primary Containment parameters.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

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Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 5 Rev. 0, 10/07/05 ILO-304 1

SCENARIO OBJECTIVES 1

The objective of this scenario is to evaluate the Licensed Operator Candidates ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

0 Knowledge of integrated plant operations 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 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 (03/04) 2005 NRC Exam, Rev. 1

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 7 Rev. 0, 10/07/05 ILO-304 I

CRITICAL TASKS L

Manually Initiate HPCl Component bv component to iniect to the RPV.

Safety Significance:

Ensures Adequate Core Cooling.

IndicationdCues for Event Reauirinq Critical Task Auto initiation signals present as evidenced by High Drywell Pressure, Low RPV Pressure, Low RPV Water Level, initiation signal indication lamp illuminated.

Conseauences for Failure to Perform Task Lack of Adequate Core Cooling IndicationdCues for Event Requiring Critical Task RPV water level less than -38 inches, High Drywell Pressure 1.72.

Performance Criteria Align HPCl to ensure flow to the RPV.

• Manuallv Initiate ARI 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 initiating ARI.

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.

Manuallv isolate Suppression Pool pipinq leak dischamina into Reactor Buildinq Area.

Safety Significance:

High-energy leakage into the Secondary Containment Area impacts the integrity of Secondary Containment.

Failure of the Secondary Containment directly relates to the 1 OCFR100 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.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 8 Rev. 0, 10/07/05 ILO-304 1

CRITICAL TASKS 1

1 Consequences 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:

SCIT-4 WHEN ANY RB AREA TEMP EXCEEDS MAX NORMAL ISOLATE ALL SYSTEMS DISCHARGING INTO AREA EXCEPT SYSTEMS REQ'D TO:

SUPPORT EOP/DSP ACTIONS OR SUPPRESS A FIRE SC/R-1 WHEN ANY RB AREA RAD EXCEEDS HI ALARM ISOLATE ALL SYSTEMS DISCHARGING INTO AREA EXCEPT SYSTEMS REQ'D TO:

SUPPORT EOP/DSP ACTIONS OR SUPPRESS A FIRE Purpose of the Secondary Containment Control procedure is to:

0 Protect equipment in secondary containment 0

Limit radioactivity release to secondary containment, and either:

Maintain secondary containment integrity, or 0

Limit radioactivity release from secondary containment Secondary Containment Control establishes and maintains control over three key secondary containment parameters: area temperatures, area radiation levels and area water levels. Operator actions are performed concurrently to stabilize and control these parameters.

Normal systems and methods are used to maintain secondary containment parameters at or below maximum normal operating values. If a parameter exceeds its Max Normal operating value, action is taken to isolate primary systems discharging into secondary containment, except those systems required to support EOP/DSP actions or suppress a fire. Actions taken above the Max Normal operating value are dependent on determining if the parameter is elevated as a result of a primary system discharging into Secondary Containment "areas" as defined in this procedure.

An area temperature above its maximum normal operating level is an indication that steam or water from a primary system may be discharging into the secondary containment. As temperatures continue to increase, the continued operability of equipment needed to carry out EOP actions may be compromised. High area temperatures also present a danger to personnel, a consideration of significance, since access to the secondary containment may be required by actions specified in the EOPs.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 9 Rev. 0, 10/07/05 ILO-304 A radiation level above Max Normal may be indicative that water or steam from a primary system (or from a primary to secondary system leak) may be discharging into the secondary containment. Max Normal operating radiation levels are equal to ARM high alarm setpoints.

Indicationdcues for Event Requiring Critical Task:

HPCl Room flooded alarm Suppression Pool level decreasing Performance Criteria:

Close F004A( B) and F004C( D).

Performance Feedback:

Suppression Pool level STOPS decreasing Sprav the D w e l l when Suppression Chamber pressure exceeds 13 psi&

Safety Significance:

Maintenance of primary containment integrity Actions are taken to spray the Drywell during a LOCA when the Suppression Chamber pressure exceeds 13 psig. From the Susquehanna Emergency Operating Procedures basis document, EO-000-1 03, The value of 13 psig is the lowest suppression chamber pressure which can occur when 95 percent of the non-condensables (Nitrogen) in the Drywell have been transferred to the suppression chamber. At 13 psig suppression chamber pressure, five percent of the non-condensables remain in the Drywell. This five percent value is the limit established to preclude chugging - the cyclic condensation of steam at the downcomer openings of the Drywell vents. Values in excess of 13 psig are indicative of more non-condensables in the Drywell, meaning chugging is more probable.

Chugging (Steam bubble collapse at the downcomer exit, resulting in a water in-rush to fill the voided areas) induces stresses at the junction of the downcomers and the Drywell floor. Repeated such stresses may result in failure of these joints, creating a direct bypass from Drywell to Suppression Chamber. Bypassing the suppression pool will directly pressurize the primary containment during a LOCA may result in failure.

By requiring Drywell sprays at 13 psig in the suppression chamber (five percent non-condensables in the Drywell), a Drywell non-condensable value of 21 percent will be maintained and chugging should not occur.

From Appendix D of NUREG-1 021, Draft Revision 9, the critical task listed above has essential safety action that correctly completed, will prevent degradation of any barrier to fission product release and the crew will take action to effectively direct or manipulate engineered safety feature (ESF) controls that would prevent any condition describe in the previous paragraph.

Consequences of Failure to Perform the Task Potential failure of primary containment SSES EOP Basis for:

PC/P-5 WHEN SUPP CHMBR PRESS > 13 PSlG CONTINUE

[Directions to initiate Drywell sprays]

Drywell spray operation may affect the availability of electrical equipment located in the Drywell. Therefore, suppression chamber sprays are given the maximum time allowable to reduce primary containment pressure before operation of Drywell sprays is required. The allowable time is determined by the suppression chamber pressure, which is equated to the amount of non-condensables remaining in the Drywell.

Form NTP-CIA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 10 Rev. 0, 10/07/05 ILO-304 The value of 13 psig is the lowest suppression chamber pressure which can occur when 95 percent of the non-condensables (N2) in the Drywell have been transferred to the suppression chamber. That is, at least five percent non-condensables remain in the Drywell when suppression chamber pressure reaches 13 psig. This non-condensable concentration limit is established to preclude chugging - the cyclic condensation of steam at the downcomer openings of the Drywell vents. A suppression chamber pressure greater than 13 psig could be indicative of a lower concentration of non-condensables in the Drywell, thereby meaning that chugging is more probable.

Chugging occurs when a steam bubble collapses at the exit of the downcomers; the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the Drywell floor.

Repeated occurrence of such stresses could cause fatigue failure of these joints, thereby creating a direct path between the Drywell and suppression chamber. Steam discharged through the downcomers could then bypass the suppression pool and directly pressurize the primary containment. Scale model tests have demonstrated that chugging will not occur so long as the Drywell contains at least one percent non-condensables. To preclude conditions under which chugging may occur, Drywell sprays are conservatively required when at least five percent non-condensables remain in the Drywell, Le., suppression chamber pressure reaches 13 psig.

Both wide range and narrow range suppression chamber pressure indication is available in the Control Room.

Wide range suppression chamber pressure indication is available locally on Containment H2/02 Analyzer Panel if analyzer is selected to suppression chamber.

IndicationdCues for the Event Requiring Critical Task:

Multiple control board and Control Room indications of suppression chamber and Drywell pressures.

Performance Criteria:

Start an operable RHR loop.

Perform a valve alignment to provide a flowpath for spray.

Performance Feedback:

RHR Pump, valve and system flow indications are available.

Multiple indications of Drywell pressure dropping.

Limits D w e l l Sprav flow to between 1,000 and 2.800 apm for the first 30 seconds.

Safety Significance:

Maintenance of primary containment integrity Actions are taken to limit the system flowrates when first initiating Drywell sprays (1,000 to 2,800 gpm for the first 30 seconds). The reason for this restriction is to limit the magnitude of the Drywell pressure reduction such that it will not go less than atmospheric (prevents air from being drawn in to containment), and ensures a margin to the negative design pressure of the containment.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 11 Rev. 0, 10/07/05 ILO-304 The BWR Owners Group Emergency Operating Procedures Basis document discusses Drywell spray limitations, utilizing a Drywell Spray Initiation Limit Curve to protect against containment damage from exceeding the design Drywell to suppression chamber differential pressure. From the Susquehanna Emergency Operating Procedures basis document, EO-000-1 03, A Drywell spray initiation limit has been developed by PPL, which provides the same protection guarantees without necessitating the use of an additional curve on the EOP flowcharts. By limiting Drywell spray flow to between 1,000 and 2,800 gpm for the first 30 seconds of Drywell spray operation, Drywell sprays can be initiated without concern in all regions of the BWR Owners Group curve. After 30 seconds of operation, the Drywell atmosphere contains sufficient vapor to allow full Drywell sprays flow. In other words, spraying the Drywell within these limits will not result in a Drywell pressure rapid reduction such that the differential pressure limit would be challenged.

From Appendix D of NUREG-1021, Draft Revision 9, the critical task listed above has essential safety action that correctly completed, will prevent degradation of any barrier to fission product release and the crew will take action to effectively direct or manipulate engineered safety feature (ESF) controls that would prevent any condition describe in the previous paragraph.

Consequences of Failure to Perform the Task Potential failure of primary containment SSES EOP Basis for:

PUP-7 SHUT DOWN DW COOLERS SHUT DOWN RECIRC PUMPS INITIATE DW SPRAYS UNLESS PUMPS CONTINUOUSLY NEEDED FOR ADEQUATE CORE COOLING LIMITING FLOW TO BETWEEN 1,000 AND 2,800 GPM FOR FIRST 30 SEC A DWSIL (Drywell Spray Initiation Limit) has been developed by PPL, which provides protection against containment damage from exceeding the design differential pressure, yet does not restrict operation of the Drywell sprays. By limiting Drywell spray flow to between 1,000 and 2,800 gpm for the first 30 seconds of Drywell spray operation, Drywell sprays can be initiated without concern in all regions of this curve. After 30 seconds, the Drywell atmosphere contains sufficient vapor to allow full Drywell sprays flow. For this reason, the curve is not included.

Indicationdcues for the Event Requiring Critical Task:

The Unit Supervisor will direct Drywell sprays be initiated, limiting flow to between 1,000 and 2,800 gpm for the first 30 seconds. The PCO will initiate Drywell sprays monitoring the flowrate on available digital and analog indications on 1 C601, limiting flow to between 1,000 and 2,800 gpm for at least the first 30 seconds of operation before increasing flow.

Performance Criteria:

Manually throttle HV151 -F016A and B and monitor Drywell spray.

Use clock to determine 30 seconds has elapsed.

Performance Feedback:

Monitor Drywell spray flow indications during first 30 seconds of Drywell spray operation.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 12 Rev. 0, 10/07/05 ILO-304

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SCENARIO REFERENCES

1. DRYWELL COOLING FAN 1 V416B TRIP

a. AR-128-803 DRWL FAN 1V416B FAILED

b. T.S. 3.6.3.2 DRYWELL AIR FLOW SYSTEM

2. REDUCE REACTOR POWER PER MINIMUM GENERATION EMERGENCY

a. 01-AD-029 EMERGENCY LOAD CONTROL

b. GO-1 00-01 2 POWER MANEUVERS

3. RRP 'A' MG-SET HYD FLUID TEMP CTLR FAILURE

b. AR-102-CO3 RECIRC MG A FLUID DRIVE OIL Hl/LO TEMP

a. AR-102-CO5 RECIRC MG SET A/B BRG OR FLUID DRIVE HI TEMP

4. LOSS OF EXTRACTION STEAM TO 48 FEEDWATER HEATER

a. ON-147-001 LOSS OF FEEDWATER EXTRACTION STEAM

c. T.S.

MCPR LCO 3.2.2

b. GO-100-012 POWER MANEUVERS

5. 3 CONTROL RODS DRIFT FROM THEIR TARGET POSITION

a. ON-1 55-001 CONTROL ROD PROBLEMS

b. OP-AD-055 IMMEDIATE OPERATOR ACTIONS LIST

6.

FAILURE OF MODE SWITCH

a. ON-100-101 SCRAM, SCRAM IMMINENT

b. EO-1 00-1 02 RPV CONTROL

7.

LOSS OF AUX BUSES 11 A AND 11 B

a. ON-103-003 13.8 KV BUSES 11A AND 11B LOSS OF BUS LOAD SHEDDING ON UNDERVOLTAGE

8. INSTRUMENT LINE BREAK INSIDE THE DRYWELL

a. EO-100-1 03 PRIMARY CONTAINMENT CONTROL

b. ON-145-004 RPV WATER LEVEL ANOMALY

9. HPCl AUTO START FAILURE

a. OP-152-001 HIGH PRESSURE COOLANT INJECTION SYSTEM

10. RHR/SUPPRESSION POOL LEAK

a. EO-1 00-1 04 SECONDARY CONTAINMENT CONTROL

b. EO-100-1 03 PRIMARY CONTAINMENT CONTROL C. AR-109-H8 RHR LOOP 'A' ROOM FLOODED Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 13 Rev. 0, 10/07/05 ILO-304 SCENARIO SPECIAL INSTRUCTIONS

1. Initialize simulator to IC-20, both Units at 100 percent power.

2. Preference File: restorepref Y PP.ILO304 (See attached Preference File for details.)

a. Verify environment window

3. MALFS REMFS OVRDS TRG 7:7 0

2:2 0

b. Ensure 12 function buttons lit.

4. Prepare a turnover sheet indicating:

a. Unit 1 is in MODE 1 at 100 percent power.

b. A Minimum Generation Emergency is in effect. Unit 2 has already reduced output by 200 Mwe. The next downpower will be on Unit 1 if needed.

c.

Fuel handling is in progress in Unit 1 Spent Fuel Pool.

d. All Systems are OPERABLE.

e.

Unit 2 is in MODE 1.

5.

Place simulator in RUN.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

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Page 15 Rev. 0, 10/07/05 ILO-304 9

SCENARIO EVENT DESCRIPTION FORM Initial Conditions: Initialize the Simulator to IC-20. Place the Simulator to RUN. Ensure the Proaram Buttons are assianed as indicated on the Swcial Instructions Sheet via the appropriate Preference File. Assian Shift positions. Direct the start of the five-minute panel walkdown.

65 HPCl Auto Start Failure.

EVENT TIME DESCRIPTION 1

5 Drywell Cooling Fan 1 V4166 Trip.

10 15 Reduce Reactor Power due to Minimum Generation Emergency.

70 RHR/Suppression Pool Leak.

I1 I

I 80 3

25 RRP 'A MG-Set Hyd Fluid Temp Controller Failure.

4 40 Loss of Extraction Steam to 4B Feedwater Heater.

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Termination 50 3 Control Rods Drift From Their Target Position.

6 55 Failure of RPS Mode Switch.

7 55 Loss of Auxiliary Buses 1 1 A and 1 1 B.

8 60 Instrument Line Break inside the Drywell.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 16 Rev. 0, 10/07/05 ILO-304 NOTES: I Drywell temperature will rise slightly and will stabilize at approximately 122 OF.

SCENARIO EVENT FORM L

Event No:

1 Brief

Description:

DRYWELL COOLING FAN 1V416B TRIP TIME 5

STU DENT ACT1 VlTl ES Recognize and respond to back panel alarm.

Perform AR-128-BO3 DRWL FAN 1V416B FAILED.

0 0

0 CHECK DRWL CLR 1 V416B tripped.

CHECK DRW L CLR 1 V416A starts.

CHECK DRWL CLRr 1V416B Supply Breaker 18246081, thermal overloads and control power fuse.

Inform US that procedure says to comply with T.S. 3.6.3.2.

Dispatch a NPO to investiQate Breaker 18246081.

When reDort from NPO is received that the breaker is tripped. reDort findinas to US.

If ordered bv US. direct NPO to attemDt one re-closure.

If needed, refer to T.S. Bases to determine which Drywell Cooling Fan Pairs must be OPERABLE.

Recognize entry to T.S. 3.6.3.2 DRYWELL AIR FLOW SYSTEM.

CONDITION A:

Restore required Drywell cooling fan to OPERABLE status within 30 Days.

Direct PCOM and PCOP to monitor Drywell Temperature and Pressure for any major changes due to the loss of the cooling fan.

If average Drywell temperature reaches 135 OF, COMPLY with TS 3.6.1.5.

Notifv Work Week Manaaer to obtain assistance with the troubleshooting of the breaker.

Notifv Shift Manaaer/ODerations SuDervision of LCO entry.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 17 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

1 Brief

Description:

DRYWELL COOLING FAN 1V416B TRIP INSTRUCTOR ACTIVITY:

When the crew has assumed shift responsibilities, trip Drywell Cooling Fan 1 V416B by depressing:

[PB-11 MRF DE106394 OPEN DRYWELL COOLING FAN 1V416B TRIP NOTE: (No further action will be taken on this event.)

ROLE PLAY:

1. As NPO, report breaker is tripped and a burnt odor is coming from it. If asked to try one re-closure, report that the breaker will not reset.

2. As Work Week Manager, report that you will get Electrical Maintenance to investigate the problem.

3. As Electrical Maintenance, wait about 10 minutes; then report that the control power fuse is burnt, and before they replace it with a new one, they need to see why it failed. This fuse was replaced only two months ago, and it could be a breaker wiring problem.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 18 Rev. 0, 10/07/05 ILO-304 SCENARIO EVENT FORM Event No:

2 Brief

Description:

REDUCE REACTOR POWER DUE TO MINIMUM GENERATION EMERGENCY

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POSITION PCOP PCOM

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TIME us I

_i_

PCOM/P STUDENT ACTIVITIES Respond to GCC notification and acknowledge request for 100 Mwe power reduction due to Minimum Generation Emergency.

Initiate actions to reduce Reactor Power to lower Mwe in accordance with the CRC Book.

Lower Reactor Recirculation Pumps speeds, maintaining pump speeMlow mismatch within the T.S. limit of 5 Mlbmhr.

Maintain Load Set 50 to 100 Mwe above actual Load.

Maintain Voltage Regulator Automanual Transfer nulled.

Enter GO-100-012, POWER MANEUVERS; ifhvhen power is reduced below 95 percent.

Provide short Reactivity Brief per OP-AD-338:

If three or more control rods have drifted and/or scrammed from their target positions -

SCRAM the reactor.

Only one method of changing Reactivity shall be performed at a time.

Monitor all applicable Nuclear Instrumentation during reactivity manipulations.

Use Reactivity Control System Bypass Form(s) when appropriate.

If a controlled shutdown is required, perform actions as delineated per the Reactor Engineering Instructions in the CRC Book or per Reactor Engineering recommendation if present or needed.

If an error occurs in a reactivity manipulation, the evolution shall STOP immediately and Shift Supervision shall be notified to address/direct corrective actions.

Review any recentkelated SSES andor Industry Events, as applicable.

Ensure GO-1 00-01 2 is implemented:

Notify Operations Management, Reactor Engineer, Chemistry, etc.

Direct and monitor overall activity.

Respond to TRA Sentinel Trip and recognize cause is Control Valve Position Changes.

Denotes Simulator Critical Task.

NOTES:

Form NTP-CIA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 19 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

2 Brief

Description:

REDUCE REACTOR POWER DUE TO MINIMUM GENERATION EMERGENCY INSTRUCTOR ACTIVITY:

N/A ROLE PLAY:

Using the GCC connection, request a 100 MWe reduction due to a Minimum Generation Emergency. If asked, reply that Unit 2 has already reduced power by 200 Mwe while Unit 1 was involved with Drywell Cooling Fan issues.

Form NTP-CIA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 20 Rev. 0, 10/07/05 ILO-304 SCENARIO EVENT FORM Event No:

3 Brief

Description:

RRP A MG-SET HYD FLUID TEMP CONTROLLER FAILURE POSITION PCOM PCOP us TIME STUDENT ACTIVITIES Recognizeheport AR-102-C03 RECIRC MG A FLUID DRIVE OIL HI-LO TEMP.

Recognize/report AR-102-CO5 RECIRC MG SET A/B BRG OR FLUID DRIVE OIL HI TEMP.

Check MG SET B HYD FLUID TEMP TI-l4020A to determine if alarm due to high OR low temperature.

Check TlCl 101 6A for proper operation.

Dispatch NPO to investigate locally.

Determines controller TIC-1 101 6A is at minimum output.

Takes manual control of TIC-1 1016A and lowers oil temps before RRP A Drive Motor Breaker trips.

Ensure MG SET B HYD FLUID CLR TEMP TIC-21 01 6A set at 125 OF AND functioning to control temperature.

Check Recirc MG Set lube oil system operation in accordance with OP-164-001 Reactor Recirculation System.

IF high temperature occurs AND unable to control increasing oil temperature with TIC1 101 66, THEN Throttle/Control oil temperature using TlCl 101 66 BPV 11 0038 as necessary.

Directs restoration of RRP A MG Set Hyd Fluid Temperature.

Notifies WWM to obtain assistance from I&C.

Denotes Simulator Critical Task.

NOTES:

I II Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 21 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTOR'S PERSONAL NOTES Event No:

3 Brief

Description:

RRP 'A' MG-SET HYD FLUID TEMP CONTROLLER FAILURE INSTRUCTOR ACTIVITY:

When crew has completed power reduction per GCC request, initiate RRP ' A MG-Set Hyd Fluid Temp Controller Auto output failure:

[PB-2] IMF CNO2:TIC11016A 0 ROLE PLAY:

1. As NPO dispatched to TV-11016A: Wait approximately - 2 minutes, and report the valve appears closed (unless the valve was opened by manual operation of the TIC on 1 C668).

2. As I&C investigating TV-11016A problem: Wait - 5 minutes, and report the controller auto output circuitry has failed to minimum. The controller must be replaced.

NOTE:

Do pJ allow temperature to reach 210 OF, as this will result in a RRP MG Set trip. If necessary, modify the valve position to allow cooling sufficient to remain below trip setpoint.

Form NTP-QA91.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 22 Rev. 0, 10/07/05 ILO-304 SCENARIO EVENT FORM Event No:

4 Brief

Description:

LOSS OF EXTRACTION STEAM TO 48 HEATER POSITION PCOP us PCOM PCOP us PCOP PCOM TIME STUDENT ACTIVITIES Recognize/report Extraction Steam to 48 Heater Isolation Valve HV-10241 B going closed.

Dispatch NPO to Local Alarm/Control Panel 1 C102 to investigate.

Direct implementation of ON-1 47-001, LOSS OF FEEDWATER HEATING EXTRACTION STEAM and GO-100-012, POWER MANEUVERS.

Immediately Reduce Reactor Power IAW RE Instructions in CRC Book to 5 75 percent RTP.

0 0

On 1C600, Monitor Monitor position and Comply with Stability Region Requirements on Power/Flow Map.

Adjust Load Set as necessary.

Main Steam Line Radiation Monitor RR-D12-1 R603.

Offgas Pretreatment Log Radiation Monitor RR-D12-1 R601.

Notify Work Week Manager, Plant Management, Chemistry, Reactor Engineering.

Refer to and address MCPR LCO 3.2.2 Action Statement, as applicable.

Any MCPR not within limits, restore MCPR(s) to within limits within two hours.

0 IF operating with only two feedwater heater strings in service at greater than 25 percent power, all extraction steam to those heaters must be in service or MCPR LCO 3.2.2 Action Statement is applicable and power must be reduced to < 25 percent.

Direct implementation of CORE FLUX OSCILLATIONS upon receipt of OPRM TRIP ENABLE alarm caused by power reduction.

Ensure HTR 58 HP EXTR IS0 HV-102428 CLOSED.

Ensure HTR 4B LP EXTR IS0 HV-10241 BC CLOSED.

Ensure MSEP A and B DRN TO HTR 48 HV-102168, HV-102138 CLOSED.

Respond to AR-103-DO5 OPRM TRIP ENABLE by referring to ON-1 78-002.

Select a non-peripheral rod to monitor LPRMs for oscillations.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 23 Rev. 0, 10/07/05 ILO-304 AND INSTRUCTORS PERSONAL NOTES Event No:

4 Brief

Description:

LOSS OF EXTRACTION STEAM TO 46 HEATER INSTRUCTOR ACTIVITY:

When crew has completed actions for RRP temperature Controller failure, initiate isolation of Extraction Steam to 48 Heater:

[PB-3] IMF MV05:HV10241 B ROLE PLAY:

1. As NPO dispatched to 1C102: Wait - 2 minutes, and report no apparent reason for 41 B Valve closure; feedwater heating system is responding as expected.

2. As I&C investigating extraction steam isolation: Wait - 5 minutes, and report no obvious reason for isolation has been found, continuing to investigatehroubleshoot.

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 24 Rev. 0. 10/07/05 ILO-304 SCENARIO EVENT FORM Event No:

5,637 Brief

Description:

THREE (3) RODS DRIFT/FAILED MODE SWITCH /LOSS OF AUX BUSES 11 A AND B STUDENT ACTIVITIES Recognize/report Control Rod Drift alarm AR-104-HO5, ROD DRIFT.

Perform Alarm Response Operator Actions:

0 IF three (3) Control rods have drifted OR scrammed from their target position, Manually Scram Reactor IAW ON-1 00-1 01, SCRAM.

Perform ON-1 55-001, CONTROL ROD PROBLEMS.

0 Direct implementation of ON-1 55-001, CONTROL ROD PROBLEMS.

Ensure PCOMIPCOP aware of requirement to Scram Reactor if three or more Control Rods have drift OR scram from their target position.

0 Recognize/report three Control Rods have drifted from their target position.

0 Place Mode Switch to SHUTDOWN in accordance with OP-AD-055, IMMEDIATE OPERATOR ACTIONS LIST, and ON-1 55-001.

Direct PCOM to SCRAM reactor.

Perform ON-100-101 SCRAM, SCRAM IMMINENT and EO-100-102, RPV CONTROL.

0 0

0 Recognize/report ATWS.

0 0

0 Insert SRMs and IRMs.

0 Initiate Manual Scram using Manual Scram Push Buttons.

Recognize/report failure of Manual Scram Pushbuttons.

Report rods driftinghully inserted when ARI successfully bleeds off Scram Air Header.

Initiate ARI to insert Control Rods by bleeding off Scram Air Header.

Ensure Isolations, ECCS Initiations and DG Starts Recognize/report loss of Aux Buses 1 1 A and 1 1 B.

Restore and maintain RPV water level between +13 inches and +54 inches using available systems.

Maintain RPV Pressure between 800 and 1,087 psig.

Exit EO-100-102, RPV CONTROL and Enter EO-100-1 13, LEVEUPOWER when ATWS occurs.

Exit EO-100-1 13 and re-enter EO-100-102 when all rods have been inserted via ARI.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 25 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 4

Event No:

5, 697 Brief

Description:

THREE (3) RODS DRIFT/FAILED MODE SWITCWLOSS OF AUX BUSES 11 A AND B INSTRUCTOR ACTIVITY:

1. After power reduction has been performed, and actions to isolate extraction steam to 5B Heater are complete per ON-147-001, initiate three (3) Control Rods Drifting by depressing:

[PB-41 IMF RD1550040639 100 0 100 Wait two minutes

[PB-51 IMF RD1550043835 100 0 100 Wait one minute

[PB-61 IMF RD1550042619 100 0 100

2. When the ARI Push Buttons are depressed, insert instrument line break malfunction:

[PB-A IMF RR180001 100 15:OO 0 Instrument line break (reference leg to 1CO04 Div 2)

ROLE PLAY:

As Electrical Maintenance/FIN sent to investigate the Aux Buses: Wait - 5 minutes, and report that it appears there is a failure in the breaker logic for 1Al0104 and 1 A10204, preventing breaker closure. More time is needed for investigation/troubleshooting.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 26 Rev. 0. 10/07/05 ILO-304 I

SCENARIO EVENT FORM I

Event No:

8,9 Brief

Description:

INSTRUMENT LINE BREAK SIDE THE DRYWELUHPCI AUTO START FAILURE POSITION PCOM/P us

• PCOM TIME STUDENT ACTIVITIES Recognize/report loss of Division 2 level and pressure indications.

Verify RPV level using redundant level indicators.

Recognize/report Drywell pressure and temperature rising slowly.

implement ON-145-004, RPV WATER LEVEL ANOMALY as time permits.

Implement EO-1 00-1 03, PRIMARY CONTAINMENT CONTROL.

Re-enter EO-1 00-1 02 due to High Drywell pressure.

Recoanize/reDort failure of HPCl to auto start.

Start HPCl component by component per OP-152-001 Hard Card.

Ensure HPCl TEST LINE TO CST IS0 HV-155-F008 CLOSED.

Ensure HPCl TEST LINE TO CST IS0 HV-155-F011 CLOSED.

Ensure HPCl PUMP DSCH HV-155-F007 OPEN.

Place HPCl TURBINE FLOW CONTROL FC-E41-1 R600 in MANUAL set at minimum.

Open HPCl L-0 CLG WTR HV-156-F059.

Start HPCl BARO CDSR VACUUM PP 1 P216.

SIMULTANEOUSLY Start HPCl AUXILIARY OIL PUMP 1 P213, AND open HPCl Accelerate HPCl turbine with HPCl TURBINE FLOW CONTROL FC-E41-1 R600 until HPCl Pump discharge pressure within 50 psig of reactor pressure.

Ensure HPCl MIN FLOW TO SUPP POOL HV-155-F012 OPENS when HPCl Pump discharge pressure > 125 psig with flow < 500 gpm, then CLOSES when flow > 600 gpm.

Open HPCl INJECTION HV-155-FOO6.

Establish flow using HPCl TURBINE FLOW CONTROL FC-E41-1 R600.

Null HPCl TURBINE FLOW CONTROL FC-E41-1 R600 and place in AUTO.

OBSERVE HPCl BARO CDSR COND PP 1 P215 Automatically Starts as needed.

Ensure HV-155-FO28, F029 HV-156-FO25, F026 CLOSE.

Ensure HPCl PUMP ROOM UNIT COOLER 1V209A(B) STARTS.

TURBINE STEAM SUPPLY HV-155-FO01.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 27 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

899 Brief

Description:

INSTRUMENT LINE BREAK SIDE THE DRYWELUHPCI AUTO START FAILURE INSTRUCTOR ACTIVITY:

After HPCl is started and RPV level is recovering, increase severity of Drywell leak:

[PB-81 IMF RR164010 15 8:OO Bottom Head Drain leak ROLE PLAY:

As necessary FOm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 28 Rev. 0, 10/07/05 ILO-304 SCENARIO EVENT FORM Event No:

10 Brief

Description:

RHWSUPPRESSION POOL LEAK STUDENT ACTIVITIES Direct initiation of Suppression Chamber Sprays.

Align RHR for Suppression Chamber Sprays; starts an RHR Pump.

Recognize/report RHR LOOP A (6) ROOM FLOODED alarm.

Verify Suppression Pool level decreasing.

Dispatch NPO to investigate room flood.

Stop RHR Pump.

Isolate F004A (6) and F004CIDLwhen directed.

Perform EO-1 00-104, SECONDARY CONTAINMENT CONTROL based on ROOM FLOODED alarm.

Direct isolation of A (B) Loop RHR to terminate Suppression Pool leak.

Direct start of ESW and Reactor Building room coolers with a cooling source.

Request assistance from Work Week Manager.

Direct B Loor, RHR placed in Containment Spravs.

Start ESW and Reactor Buildina room coolers with a coolino source.

Directs PCOP/M to re-initiate Suppression Chamber Sprays Direct PCOP/M to initiate Drywell Sprays when Suppression Chamber pressure exceeds 13 psig.

Initiate Drywell Sprays when Suppression Chamber pressure exceeds 13 psig.

Limit Drywell Spray Flow to 1000 to 2,800 gpm for the first 30 Seconds.

After the scenario is complete, classify the event as:

FA 1 Any loss or potential loss of either the Fuel clad or RCS OR MA3 based on RPS setpoint exceeded and RPS automatic scram did not reduce reactor power to < 5 percent AND ARI initiated to reduce power below five percent.

Denotes Simulator Critical Task.

NOTES:

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 29 Rev. 0, 10/07/05 ILO-304 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

10 Brief

Description:

RHWSUPPRESSION POOL LEAK INSTRUCTOR ACTIVITY:

1. If A Loop of RHR is placed in service, insert a break on the RHR suction line:

[PB-91 IMF RH149004A 20 4:OO

2.

When RHR Suppression Pool Suction Valves F004A and B are closed, delete the RHR leak by depressing:

[PB-lo] DMF RH149004A

3. If B Loop of RHR is placed in service, insert a break on the RHR suction line:

[PB-111 IMF RH1490046 20 4:00

4. When RHR Suppression Pool Suction Valves F004C and D are closed, delete the RHR leak by depressing:

[P6-12] DMF RH1490046 ROLE PLAY:

1. As NPO dispatched to verify the RHR Room flood alarm: Wait - 3 minutes, and report that there is at least four inches of water on the floor; you have exited the area and closed the water tight door.

TERMINATION CUE:

When RPV level is being maintained above TAF with available sources and Suppression Chamber/Drywell Sprays have been utilized for containment control, the scenario may be terminated.

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:

FA1 Any loss or potential loss of either the Fuel clad or RCS OR MA3 based on RPS setpoint exceeded and RPS automatic scram did not reduce reactor power to < 5 percent AND ARI initiated to reduce power below five percent.

FO~TTI NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

PPL-SUSQUEHANNA, LLC TRAINING CENTER Prepared By:

SIMULATOR SCENARIO Richard E. Chin 10/08/05 Instructor Date Scenario

Title:

Nuclear Operations Training Supervisor ILO CertificatiodNRC Exam Scenario Date Scenario Duration:

90 Minutes 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 Operational Activities:

Raise Reactor Power Power Supply Failure Stuck Rods Inadvertent DG Start Failed Fuel Loss of ESW Rapid Depressurization Leak Into Secondary Containment PC Isolation Valves Fail to Isolate Reviewed By:

Approved By:

I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 2 Rev. 0, 10/08/05 ILO-305 u

THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 3 Rev. 0, 10/08/05 ILO-305 SCENARIO

SUMMARY

I 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 ESW 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.

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 4 Rev. 0, 10/08/05 ILO-305 L

e The objective of this scenario is to evaluate the Licensed Operator Candidates ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

Knowledge of integrated plant operations Ability to diagnose abnormal plant conditions 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. Interpret/diagnose 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 Ouerator Candidates:

1.

2.

3.

4.

5.

6.

InterpreVdiagnose 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 (03/04) 2005 NRC Exam. Rev. 1

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1 Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 6 Rev. 0, 10/08/05 ILO-305

-11 CRITICAL TASKS II Ir Manually scram the Reactor before any Secondaw Containment Area Radiation Max Safe Temperature.

Safety Sisnificance:

High-energy leakage into 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. 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.

Consequences 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-000-102, 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)

Indications/Cues for Event Requirinq 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.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 7 Rev. 0, 10/08/05 ILO-305 Sr Rapidly depressurize the reactor when two Secondary Containment Areas exceed Max Safe Rad levels.

Safetv Sianificance:

High-energy leak in the Secondary Containment Area impacts the integrity of Secondary Containment. Failure of the Secondary Containment directly relates to the 10CFRlOO 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 SCTT-9 WHEN RB AREA TEMP EXCEEDS MAX SAFE SC/R-6 WHEN RB AREA RAD EXCEEDS MAX SAFE IN 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 Reauirina Critical Task:

0 0

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.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 8 Rev. 0, 10/08/05 ILO-305 I 1 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 Whr 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 (OW04) 2005 NRC Exam, Rev. 1

Page 9 Rev. 0, 10/08/05 ILO-305 SCENARIO REFERENCES 11

1. RAISE REACTOR POWER GO-1 00-002 PLANT STARTUP, HEATUP AND POWER OPERATIONS SO-1 56-007 CONTROL ROD COUPLING/FULL-IN INDICATOR CHECKS

2. POWER SUPPLY FAILURE AR-104-A06 AR-104-A04 AR-104-A01 AR-104-BO6 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 B1/B2 AUTO SCRAM R PS I NSTRUM ENTATION CONTROL ROD BLOCK INSTRUMENTATION

3. 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

4. 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

5. RWCU LEAK INTO SECONDARY CONTAINMENT:

AR-101 -DO1 RWCU PUMP TROUBLE 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 AR-101 -A04 RWCU SYSTEM PRE-ISOLATION HI TEMP/HI DlFF TEMP SIMPLEX FIRE PROTECTION FIRE DETECTION ALARM PRIORITY 2 PRIMARY CONTAINMENT ISOLATION VALVES SCRAM, SCRAM IMMINENT SECONDARY CONTAINMENT CONTROL

6. STUCK RODS:

EO-1 00-1 13 LEVEUPOWER CONTROL

7. FAILED FUEL:

OP-AD-055 ON-179-001 OPERATIONS PROCEDURE PROGRAM IMMEDIATE OPERATOR ACTIONS INCREASING OFFGAS MSL RAD LEVELS

8. RAPID DEPRESSURIZATION:

EO-1 00-1 12 RAPID DEPRESSURIZATION F

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 10 Rev. 0. 10/08/05 ILO-305 II II THIS PAGE IS INTENTIONALLY LEFT BLANK 1

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 11 Rev. 0, 10/08/05 ILO-305 SCENARIO SPECIAL INSTRUCTIONS 1

1. Prepare SO-156-007, CONTROL ROD COUPLING/FULL-IN INDICATOR CHECKS to align with rod pull sheets.

2. Prepare Unit Supervisor turnover Sheet indicating:

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, Rev. 1

Page 12 Rev. 0, 10/08/05 ILO-305 A

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Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 13 Rev. 0. 10/08/05 ILO-305 SCENARIO EVENT DESCRIPTION FORM Initial Conditions:

Reactor power Five Percent Startup in Progress IAW g0-1 00-002 Rod Sequence 82 SU step Step 282 11 EVENT I

TIME I DESCRIPTION F

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 14 Rev. 0, 10/08/05 LO-305 THIS PAGE IS INTENTIONALLY LEFT BLANK II Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 15 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No:

1 Brief

Description:

RAISE REACTOR POWER WITH CONTROL ROD WITHDRAWAL POSITION PCOM/P us TIME 5

• Denotes Critical Task STUDENT ACTIVITIES Withdraw Control Rods in accordance with Reactor Engineer/CRC Instructions.

Perform Coupling Checks per SO-1 56-007:

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.

0 Confirm the control rod remains at Position 48.

Record date and initials in appropriate space for the control rod in COUPLING CHECK on Attachment C, Page 1.

Maintain overall control of evolution.

NOTES:

I n

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 16 Rev. 0, 10/08/05 ILO-305 I

INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

1 Brief

Description:

RAISE REACTOR POWER WITH CONTROL ROD WITHDRAWAL INSTUCTOR ACTIVITY:

N/A ROLE PLAY:

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

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

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 B STUDENT ACTIVITIES Recognize and respond to the following alarms:

AR-104-AO6 AR-104-A04 AR-104-A01 AR-104-BO6 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 Bl/B2 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, notify GCC/TCC.

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 GCCnCC of plant status.

Authorize replacement of blown fuse.

Direct PCOM to reset Half-Scram when conditions permit.

_ _ _ ~

Sr Denotes Critical Task NOTES:

F o

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 18 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES INSTUCTOR ACTIVITY:

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

[P-1]

IMF NMl75002B BLOWN FUSE IN POWER SUPPLY BREAKER 1D68204 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 1 D68204 ROLE PLAY:

1. As NPO dispatched to 1 D68204, 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.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

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 POSITION TIME STUDENT ACTIVITIES PCOP 30 Recognize and respond to AR-Ol6-CO3 DG D PANEL OC521 D LO PRIORITY TROUBLE.

DisDatch NPO to investiaate locallv for cause of start.

Report DG status to US.

Ensure D Diesel Generator did not start from a valid signal.

Check ESS Bus 1 D energized, Drywell Pressure normal, etc.

Direct NPO to perform a Manual Emergencv Shutdown of the DG from OC527 D due to U

I abnormal indications. (OP-024-001, DIESEL GENERATORS, Section 2.7)

I I

Refer to ON-024-001, DIESEL GENERATOR TRIP.

Place ESW Pumps in service to provide Cooling.

Recognize and respond to AR-Ol6-DlO (E-08) ESW PUMP A(C) Overcurrent Alarm Dispatch NPO to ESSW Pumphouse to investigate conditions.

NOTE 1 35

~

~~

11

~ N O

T F

r

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1 lnform u s t o 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.

~~

~

• Denotes Critical Task.

n II therefore. NA here.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 20 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

3 and 4 Brief

Description:

INSTUCTOR ACTIVITY:

INADVERTENT START OF D DIESEL GENERATOWESW PUMP TRIP

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 06653. Therefore, no emergency start signal is present, and no auto start of ESW will occur.

[P-3]10R ZDIHS00051 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 pump/loop in service following the trip of the first pump.

[P-5]IMF 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-qIOR QD143CMD LOCAL

[P-gIMOR QD15ESD NORMAL PLACE DG CS TO LOCAL AT OC521 D RELEASE EMERGENCY STOP PUSHBUTTON

[P-8]10R QD15ESD STOP DEPRESS 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-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

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 Sr Denotes Critical Task STUDENT ACTIVITIES Recognize and respond to AR-101 -A04 RWCU SYSTEM PRE-ISOLATION HI-TEMP/HI DlFF TEMP.

Recognizesheports AR-lOl-A02/A03, RWCU LEAK DET IS0 LOGIC NB 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.

Recognizedreports 1 C614 RWCU leak detection alarms/trips.

Recognize and respond to AR-SP-002, SIMPLEX FIRE DETECTION.

Monitors RWCU Pump Room temperature and radiation levels.

Enters EO-1 00-1 04, 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 cooling source as directed.

Form NTP-QA91.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 22 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 1

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-QA91.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 23 Rev. 0, 10/08/05 ILO-305

~~

Exits EO-1 00-1 13 and re-enters EO-1 00-1 02 RPV Control when all rods inserted.

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

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

Event No:

7 and 8 Brief

Description:

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.

Report control rods drifting in.

I When all rods fully inserted, report all rods in.

PCOM/P I PCOP I

Report Reactor Building Area High Radiation Alarms.

Recognize/report that radiation levels in the CRD and RWCU areas are above 10 FVhour.

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 LeveVPower Control with report of all rods not inserted.

• Denotes Critical Task NOTES:

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 24 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

7, 8 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-113 IMF RR179003 300 7:OO

[P-131 IMF TR02:RlTl3752 13 1O:OO

[P-141 IMF TR02:RlTl3750 14 2:OO 0.27

[P-151 IMF TR02:RlTl3751 15 2:W 0.27

[P-161 IMF TR02:RIT13705 1000 2:OO

[P-17] IMF TR02:RIT13706 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 1000 MWHR OVER 2 MINUTES

[P-121 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 Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 25 Rev. 0, 10/08/05 ILO-305 SCENARIO EVENT FORM Event No:

9 Brief

Description:

RAPID DEPRESSURIZATION

• Denotes Critical Task STUDENT ACTIVITIES Report RWCU area radiation level is > lOR/Hr, now have two areas greater than 1 OR/Hr.

Performs EO-1 00-1 12, RAPID DEPRESSURIZATION when two Secondary Containment Areas exceed Max Safe Rad levels.

Verifies Sumression Pool Level > 5 feet.

Directs PCOP to open 6 ADS Valves.

Opens 6 ADS SRVs to Rapidly Depressurize the reactor.

Reports RPV pressure lowering.

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.

Places both loops of RHR in Suppression Pool Cooling in accordance with OP-149-005.

After the scenario is complete, classifies the event as a Site Area Emergency Classification declared based on FS 1.

Loss or Potential loss of ANY two Fission Product Barriers.

11 NOTES:

I I

NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 26 Rev. 0, 10/08/05 ILO-305 INSTRUCTOR ACTIVITIES, ROLE PLAY, 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 Form NTP-QA31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

UNIT SUPERVISOR TURNOVER SHEET UNIT 1

Today Date S H I R 1900 to 0700 start End MODE 2

POWER LEVEL 0

GENERATOROUTPUT MWe CASK STORAGE GATE INSTALLED: YESNO SHIFT 0700 to 1900 Start End MODE 2

POWER LEVEL 5

GENERATOROUTPUT MWe CASK STORAGE GATE INSTALLED: YES/NO POTENTIAL LCOTTRO's:

REMARKS:

1.

2.

All svstems are OPERABLE Plant Startup in progress GO-1 00-002 completed up to step 282.

3.

Continue with Plant Startup

4.

5.

Unit 2 is

6.

11.

12.

COMMON

1.

All Systems OPERABLE

2.

3.

5.

6.

7.

8.

9.

10.

Form NTP-QAB1.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

1900-0 0700-

-I#

POST RELIEF L OFFGOING UNIT SUPERVISOR CHECKLIST:

1.

2.

Evolutions in progress and items to be completed during next shift, as noted in remarks, have been discussed with oncoming Unit Supervisor.

Problems encountered during past shift and abnormal plant conditions, as noted in remarks, have been discussed with oncoming Unit Supervisor.

1900 - 0700 0700 - 1900 Offgoing Unit Supervisor ONCOMING UNIT SUPERVISOR CHECKLIST:

I.

LCOlTRO Log reviewed.

0700 - 1900 1900 - 0700 Oncoming Unit Supervisor

1.

Walk down Control Room panels with Unit Responsible K O.

2.

3.

4.

Unit Log reviewed for entries made in past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Reactor Engineering Instructions (CRC Book) reviewed.

Completed System Status Operable audit for open PMT this shift.

0700 - 1900 1900 - 0700 Oncoming Unit Supervisor Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

PPL-SUSQUEHANNA, LLC TRAINING CENTER SIMULATOR SCENARIO Supervising ManagedShift Supervisor I

Scenario

Title:

ILO Certification/NRC Exam Scenario Scenario Duration:

75 Minutes Scenario Number:

ILO-504 I

Revision/Date:

Rev. 0, 10/06/05 Course:

PCOO7/PCOO8, Initial License RO/SRO Certification Examination PCO17/PCO18, Initial License RO/SRO NRC Examination Operational Activities:

RRP Speed Controller Failure HPCl ROOM Steam Leak FWLC Failure Inadvertent SRV Opening RClC Failure Primary System Break in the Drywell ADS Auto Logic Failure Rapid Depressurization Prepared By:

Reviewed By:

Approved By:

Richard E. Chin Instructor 10/06/05 Date Nuclear Operations Training Supervisor Date Date Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam. Rev. 1

Page 2 Rev. 0, 10/06/05 ILO-504 THIS PAGE IS INTENTIONALLY LEFT BLANK

-I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 4 Rev. 0, 10/06/05 ILO-504 After the RFPT Controller failure has been evaluated, an inadvertent opening of the D Safety Relief Valve will occur.

This will require the crew to implement Off-Normal Procedure ON-1 83-001, STUCK OPEN SAFETY RELIEF VALVE.

The failure will be accompanied by Control Room alarm AR-1 1 0-EO, MAIN STEAM SRV LEAKING. Because the SRV has failed, the indicating lamps for the SRV will NOT change state. The crew must evaluate plant response and indications on various Control Room panels to correctly diagnose the failure. Confirmation of the open SRV will be through Acoustic Monitoring System, loss of electric load, Steam flow/Feedflow mismatch, and SRV/ADS Temp Recorder TRS-621-1 R614. When Operator Actions have failed to close the SRV, and it has become evident that the valve will not close, the crew will place the Mode Switch to Shutdown. Following the Scram, the Stuck Open SRV malfunction will be deleted and the valve will close.

1 Reactor water level will lower to less than +13 inches when the scram occurs. This will require the performance of EO-100-102, RPV CONTROL. During the RPV transient, a break on the Reactor Water Cleanup line in the Drywell will occur. Drywell Pressure will exceed 1.72 psig, requiring the crew to perform EO-100-103, PRIMARY CONTAINMENT CONTROL. The size of the leak will gradually increase as the scenario continues. The combination of High Drywell Pressure and a Main Generator Lockout will result in a Plant Auxiliary Loadshed. This Loadshed logic will result in a loss of all Condensate, Feedwater, Service Water and Circulating Water Pumps. This loss of equipment will cause RPV level to continue to lower.

Drywell and Suppression Chamber Sprays will be required due to the rise in Primary Containment temperature and pressure.

RClC will fail to automatically start at -30 inches RPV water level, and Manual Initiation will NOT be successful. RClC may be started using a component by component start, but will trip on overspeed. With no high pressure injection available, RPV water level will lower to -1 61 inches (Top of Active Fuel) requiring the crew to perform EO-1 00-1 12, RAPID DEPRESSURIZATION. Because the ADS auto logic is failed, the crew must manually open the ADS Valves, and follow-up with actions to maintain the valves open via keylock switches in the Upper/Lower Relay Rooms.

If Primary Containment parameters and RPV pressure meet RPV Flooding Requirements as indicated by crossing the Saturation Curve, then the crew will be required to perform EO-100-1 14, RPV FLOODING.

The scenario termination criteria will be met when the RPV Flooding is in progress if required. Otherwise, it will be when the RPV has been depressurized, RPV level has been restored above TAF, and actions have been taken to spray the Drywell and Suppression Chamber.

FOm NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 5 Rev. 0, 10/06/05 ILO-504 SCENARIO OBJECTIVES 1

The objective of this scenario is to evaluate the Licensed Operator Candidates ability to respond to the scenario events. These events will require each candidate to demonstrate the following:

Knowledge of integrated plant operations 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. 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 ODerator 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 (0304) 2005 NRC Exam, Rev. 1

Page 6 Rev. 0, 10/06/05 ILO-504 I THIS PAGE IS INTENTIONALLY LEFT BLANK 11 Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 7 Rev. 0, 10/06/05 ILO-504 I1 A

CRITICAL TASKS Perform Rapid Depressurization when RPV level drops to -161 inches.

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:

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 RCR-10 IRRESPECTIVE OF VORTEX LIMITS WITH TABLE 3 SYSTEMS PERFORM ALL 1

LINE UP FOR INJECTION 2

START PUMPS 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 1

LINE UP FOR INJECTION 2

STARTPUMPS 3

INCREASE INJECTION TO MAX RCR-16 WHEN LVL CANNOT BE RESTORED AND MAINTAINED > -161 GO TO RAPID DEPRESS Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 8 Rev. 0, 10/06/05 ILO-504 CRITICAL TASKS (Continued) 1 Rapid Depressurization is not initiated until RPV water level has dropped to -1 61 " (TAF) because:

1 Adequate core cooling exists so long as RPV water level remains above -1 61 " (TAF).

2 The time required for RPV water level to decrease to -161" (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", 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.

Manually isolate HPCl System discharqinq into Reactor Buildina Area.

Safety Significance:

High-energy leakage into the Secondary Containment Area impacts the integrity of Secondary Containment.

Failure of the Secondary Containment directly relates to the 1 OCFRlOO 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.

Consequences 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.

F o

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NTP-OA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 9 Rev. 0, 10/06/05 ILO-504 SSES EOP Basis:

scrr-4 WHEN ANY RB AREA TEMP EXCEEDS MAX NORMAL ISOLATE ALL SYSTEMS DISCHARGING INTO AREA EXCEPT SYSTEMS REQ'D TO:

SUPPORT EOP/DSP ACTIONS OR SUPPRESS A FIRE SC/R-l WHEN ANY RB AREA RAD EXCEEDS HI ALARM ISOLATE ALL SYSTEMS DISCHARGING INTO AREA EXCEPT SYSTEMS REQ'D TO:

SUPPORT EOP/DSP ACTIONS OR SUPPRESS A FIRE Purpose of the Secondary Containment Control procedure is to:

0 Protect equipment in secondary containment 0

Limit radioactivity release to secondary containment, and either:

0 Maintain secondary containment integrity, or 0

Limit radioactivity release from secondary containment Secondary Containment Control establishes and maintains control over three key secondary containment parameters: area temperatures, area radiation levels and area water levels. Operator actions are performed concurrently to stabilize and control these parameters.

Normal systems and methods are used to maintain secondary containment parameters at or below maximum normal operating values. If a parameter exceeds its Max Normal operating value, action is taken to isolate primary systems discharging into secondary containment, except those systems required to support EOP/DSP actions or suppress a fire. Actions taken above the Max Normal operating value are dependent on determining if the parameter is elevated as a result of a primary system discharging into Secondary Containment "areas" as defined in this procedure.

An area temperature above its maximum normal operating level is an indication that steam or water from a primary system may be discharging into the secondary containment. As temperatures continue to increase, the continued operability of equipment needed to carry out EOP actions may be compromised. High area temperatures also present a danger to personnel, a consideration of significance, since access to the secondary containment may be required by actions specified in the EOPs.

A radiation level above Max Normal may be indicative that water or steam from a primary system (or from a primary to secondary system leak) may be discharging into the secondary containment. Max Normal operating radiation levels are equal to ARM high alarm setpoints.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 10 Rev. 0, 10/06/05 ILO-504 Indications/Cues for Event Requiring Critical Task:

Simplex Fire Detection alarms indicating High temperatures in RB Areas Room high temperature annunciation Feedback from plant personnel High radiation annunciation for affected areas/rooms Other indication of steam line break: lowering steam supply pressure, depressurization of the RPV; level transient on RPV Performance Criteria:

Initiate an isolation of the affected system before parameter of concern (temperature or radiation) reaches max safe value as noted in EOP-104 Tables.

Initiate system isolation using isolation pushbuttons, as appropriate.

Manually close steam supply/steam isolation valve(s), as appropriate.

Note: Successful completion of this task can occur either before or after temperatures/radiation levels reach max safe dependent on plant conditions and pace of the transient. This philosophy is justified by the EPG definition of the term Before or Prior to:

Indicates that the step should be performed, if possible, in advance of the specified condition, but that the timing of the action is event-dependent. No particular margin to the identified action level is intended. If the condition has already occurred when the instruction is reached, the action should still be performed unless expressly prohibited.

(BWROG EPGs/SAGs, Appendix 8-3)

Performance Feedback:

Initiating an isolation of the affected system results in Control Room/PICSY indications of lowering room temperatures and radiation levels. Successful isolation is indicated on Control Room panel by full closed light indication for operated valves Sprav the D w e l l when Suppression Chamber pressure exceeds 13 psiq.

Safety Significance Maintenance of primary containment integrity Actions are taken to spray the Drywell during a LOCA when the Suppression Chamber pressure exceeds 13 psig. From the Susquehanna Emergency Operating Procedures basis document, EO-000-1 03, The value of 13 psig is the lowest suppression chamber pressure which can occur when 95 percent of the non-condensables (Nitrogen) in the Drywell have been transferred to the suppression chamber. At 13 psig suppression chamber pressure, five percent of the non-condensables remain in the Drywell. This five percent value is the limit established to preclude chugging - the cyclic condensation of steam at the downcomer openings of the Drywell vents. Values in excess of 13 psig are indicative of more non-condensables in the Drywell, meaning chugging is more probable.

Chugging (Steam bubble collapse at the downcomer exit, resulting in a water in-rush to fill the voided areas) induces stresses at the junction of the downcomers and the Drywell floor. Repeated such stresses may result in failure of these joints, creating a direct bypass from Drywell to Suppression Chamber. Bypassing the suppression pool will directly pressurize the primary containment during a LOCA may result in failure.

Form NTP-QAB1.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 11 Rev. 0, 10/06/05 ILO-504 By requiring Drywell sprays at 13 psig in the suppression chamber (five percent non-condensables in the Drywell), a Drywell non-condensible value of >1 percent will be maintained and chugging should not occur.

From Appendix D of NUREG-1021, Draft Revision 9, the critical task listed above has essential safety action that correctly completed, will prevent degradation of any barrier to fission product release and the crew will take action to effectively direct or manipulate engineered safety feature (ESF) controls that would prevent any condition describe in the previous paragraph.

Consequences of Failure to Perform the Task Potential failure of primary containment SSES EOP Basis for:

PC/P-5 WHEN SUPP CHMBR PRESS > 13 PSlG CONTINUE

[Directions to initiate Drywell sprays]

Drywell spray operation may affect the availability of electrical equipment located in the Drywell. Therefore, suppression chamber sprays are given the maximum time allowable to reduce primary containment pressure before operation of Drywell sprays is required. The allowable time is determined by the suppression chamber pressure, which is equated to the amount of non-condensables remaining in the Drywell.

The value of 13 psig is the lowest suppression chamber pressure which can occur when 95 percent of the non-condensables (N2) in the Drywell have been transferred to the suppression chamber. That is, at least five percent non-condensables remain in the Drywell when suppression chamber pressure reaches 13 psig. This non-condensible concentration limit is established to preclude chugging - the cyclic condensation of steam at the downcomer openings of the Drywell vents. A suppression chamber pressure greater than 13 psig could be indicative of a lower concentration of non-condensables in the Drywell, thereby meaning that chugging is more probable.

Chugging occurs when a steam bubble collapses at the exit of the downcomers; the rush of water drawn into the downcomers to fill the void induces stresses at the junction of the downcomers and the Drywell floor.

Repeated occurrence of such stresses could cause fatigue failure of these joints, thereby creating a direct path between the Drywell and suppression chamber. Steam discharged through the downcomers could then bypass the suppression pool and directly pressurize the primary containment. Scale model tests have demonstrated that chugging will not occur so long as the Drywell contains at least one percent non-condensables. To preclude conditions under which chugging may occur, Drywell sprays are conservatively required when at least five percent non-condensables remain in the Drywell, Le., suppression chamber pressure reaches 13 psig.

Both wide range and narrow range suppression chamber pressure indication is available in the Control Room.

Wide range suppression chamber pressure indication is available locally on Containment H2/02 Analyzer Panel if analyzer is selected to suppression chamber.

IndicationdCues for the Event Requiring Critical Task Multiple control board and Control Room indications of suppression chamber and Drywell pressures.

Performance Criteria Start an operable RHR loop.

Perform a valve alignment to provide a flowpath for spray.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam. Rev. 1

Page 12 Rev. 0, 10/06/05 ILO-504 Performance Feedback RHR pump, valve and system flow indications are available.

Multiple indications of Drywell pressure dropping.

Limits Drywell Sprav flow to between 1,000 and 2,80, capm for the first 30 seconds.

Safety Sign if icance Maintenance of primary containment integrity Actions are taken to limit the system flowrates when first initiating Drywell sprays (1,000 to 2,800 gpm for the first 30 seconds). The reason for this restriction is to limit the magnitude of the Drywell pressure reduction such that it will not go less than atmospheric (prevents air from being drawn in to containment), and ensures a margin to the negative design pressure of the containment.

The BWR Owners Group Emergency Operating Procedures Basis document discusses Drywell spray limitations, utilizing a Drywell Spray Initiation Limit Curve to protect against containment damage from exceeding the design Drywell to suppression chamber differential pressure. From the Susquehanna Emergency Operating Procedures basis document, EO-000-1 03, A Drywell spray initiation limit has been developed by PPL, which provides the same protection guarantees without necessitating the use of an additional curve on the EOP flowcharts. By limiting Drywell spray flow to between 1,000 and 2,800 gpm for the first 30 seconds of Drywell spray operation, Drywell sprays can be initiated without concern in all regions of the BW R Owners Group curve. After 30 seconds of operation, the Drywell atmosphere contains sufficient vapor to allow full Drywell sprays flow. In other words, spraying the Drywell within these limits will not result in a Drywell pressure rapid reduction such that the differential pressure limit would be challenged.

From Appendix D of NUREG-1 021, Draft Revision 9, the critical task listed above has essential safety action that correctly completed, will prevent degradation of any barrier to fission product release and the crew will take action to effectively direct or manipulate engineered safety feature (ESF) controls that would prevent any condition describe in the previous paragraph.

Consequences of Failure to Perform the Task Potential failure of primary containment SSES EOP Basis for:

PC/P-7 SHUT DOWN DW COOLERS SHUT DOWN RECIRC PUMPS INITIATE DW SPRAYS UNLESS PUMPS CONTINUOUSLY NEEDED FOR ADEQUATE CORE COOLING LIMITING FLOW TO BETWEEN 1,000 AND 2,800 GPM FOR FIRST 30 SEC A DWSIL (Drywell Spray Initiation Limit) has been developed by PPL, which provides protection against containment damage from exceeding the design differential pressure, yet does not restrict operation of the Drywell sprays. By limiting Drywell spray flow to between 1,000 and 2,800 gpm for the first 30 seconds of Drywell spray operation, Drywell sprays can be initiated without concern in all regions of this curve. After 30 seconds, the Drywell atmosphere contains sufficient vapor to allow full Drywell sprays flow. For this reason, the curve is not included.

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NTP-QA91.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 13 Rev. 0, 10/06/05 ILO-504 Indicationdcues for the Event Requiring Critical Task The Unit Supervisor will direct Drywell sprays be initiated, limiting flow to between 1,000 and 2,800 gpm for the first 30 seconds. The PCO will initiate Drywell sprays monitoring the flowrate on available digital and analog indications on 1 C601, limiting flow to between 1,000 and 2,800 gpm for at least the first 30 seconds of operation before increasing flow.

Performance Criteria Manually throttle HVl51 -FO16A and B and monitor Drywell spray.

Use clock to determine 30 seconds has elapsed.

Performance Feedback Monitor Drywell spray flow indications during first 30 seconds of Drywell spray operation.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 14 Rev. 0, 10/06/05 ILO-504

- I SCENARIO REFERENCES

1. A REACTOR RECIRCULATION PUMP SPEED INCREASE:

ON-100-004 REACTOR POWER GREATER THAN LICENSE LIMIT ON-156-001 U N EXPLAl N ED REACTIVITY CHANGE NDAP-QA-0720 STATION REPORT MATRIX AND REPORTABILITY EVALUATION AR-105-106 MAIN TURBINE BYPASS VALVES OPEN OP-164-001 REACTOR RECl RCULATION SYSTEM 0

T.S. 3.4.1.1 REACTOR COOLANT SYSTEM - RECIRCULATION LOOPS OPERATING

2. HPCl SYSTEM STEAM LEAK EO-100-104 SECONDARY CONTAINMENT CONTROL ON-169-002 FLOODING IN REACTOR BUILDING AR-114-E05 HPCl LEAK DETECTION HI TEMP/HI DlFF TEMP AR-Ol6-Gl5 FIRE PROTECTION PANEL OC650 SYSTEM TROUBLE AR-SP-001 FIRE SUPP X228-27 ALM OP-155-001 HPCl SYSTEM 0

T. S. 3.5.1 ECCS - OPERATING

3. C RFPT SPEED CONTROL FAILURE AR-101-B16 RFPT A(B)(C) CONTROL SIGNAL FAILURE ON-145-001 RPV LEVEL CONTROL SYSTEM MALFUNCTION

4. D SRV INADVERTENT OPENING ON-183-001 STUCK OPEN SAFETY RELIEF VALVE AR-110-E01 MAIN STEAM SRV LEAKING

5. RClC FAILURE TO AUTO START OP-150-001 RClC SYSTEM

6. PRIMARY SYSTEM BREAK INSIDE DRYWELUADS AUTO LOGIC FAILURE EO-100-102 RPV CONTROL EO-100-103 PRIMARY CONTAINMENT CONTROL EO-100-112 RAPID DEPRESSURIZATION FOm NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 15 Rev. 0, 10/06/05 ILO-504

- I SCENARIO SPECIAL INSTRUCTIONS

1.

Initialize the Simulator to IC-20. Both Units at 100 percent power.

2.

Load the following malfunctions and overrides USING restorepref YPP.IL0-504:

(See attached Preference File.)

RESTOREPREF YPP.PRGBUTCLR IMF RLO1 :E51 1 K3 IMF RC150002 IMF AD183001 IOR ZDIHSB211S3OAA Ask IOR ZDIHSB211 S31 AA Ask IOR ZDIHSB211 S30BA Ask IOR ZDIHSB211 S31 BA ASIS IMF AV03:HVl55F100 IMF AV06:HV155F003 IMF AV06:HV155F002 PFS 1 IMF CN03:SYB311R621A 100 30 90.36 PFS 2 MRF RR164017 OFF PFS 3 IMF HP152009 0.25 0 0 PFS 4 IMF FW145004C PFS 5 DMF FW145004C PFS 6 IMF RVO1 :PSV141 F13D PFS 7 IMF MS18301 OD 100 0 0 PFS 8 DMF MS18301 OD PFS 9 IMF RR164010 10 0 0 PFS lOMMF RR164010 10 20 2:00 PFS 11 MMF RR164010 30 3:OO PFS 12 IMF PM03:1P102A PFS 13 IMF PM03:1P102B PFS 14 IMF PM03:1P102C PFS 15 IMF PM03:1P102D RClC Failure to Auto Start RClC Overspeed ADS Auto Logic Failure ADS Pushbuttons Fail As-is ADS Pushbuttons Fail As-is ADS Pushbuttons Fail As-is ADS Pushbuttons Fail As-is HPCl FlOO Valve Auto Logic Failure HPCl F003 Valve Auto Logic Failure HPCl F002 Valve Auto Logic Failure A

POWER TO SCOOP TUBE POSlTlONER OFF HPCl ROOM STEAM LEAK RFPT CONTROL SIGNAL FAILURE DELETE RFPT CONTROL SIGNAL FAILURE INADVERTENT OPENING OF D SRV STUCK OPEN D SRV DELETE STUCK OPEN D SRV RX VESSEL BOTTOM HEAD DRAIN LEAK INCREASE BOTTOM HEAD DRAIN LEAK INCREASE BOTTOM HEAD DRAIN LEAK RRP SPEED CONTROLLER FAIL TO 100%

Condensate Pump 1A Overcurrent trip Condensate Pump 1 B Overcurrent trip Condensate Pump 1 C Overcurrent trip Condensate Pump 1 D Overcurrent trip

3. Place the Simulator in RUN.

4. Ensure environment window displays 5 Malfunctions and 4 Overrides

5. Prepare a turnover package with turnover sheets.

0 0

Both Units at 100 percent power.

Maintain reactor power at 100 percent.

0 All systems OPERABLE.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 16 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT DESCRIPTION FORM Initial Conditions:

Both Units are at 100 Percent.

FOMI NTP-QA91.7A Rev. 0 (03104) 2005 NRC Exam. Rev. 1

Page 17 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM Event No:

1 Brief

Description:

A REACTOR RECIRCULATION PUMP SPEED INCREASE POSITION PCOM CREW SRO See Note TIME 5

STUDENT ACTIVITIES Respond to AR-105-106, MAIN TURBINE BYPASS VALVES OPEN.

Perform ON-1 00-00, REACTOR POWER GREATER THAN LICENSE LIMIT 0

0 0

Plot Power/Flow Position Recognize reactor power is greater than 100 percent.

Initiate Prompt action to Reduce Core Thermal Power to I Licensed Limit (3489 MWth -

100 percent APRM with LEFM in service)

Reduce B Reactor Recirc Pump Speed to get below 100 percent power Perform ON-1 56-001, UNEXPLAINED REACTIVITY CHANGE 0

IF > 3475 MWth without LEFM in service OR > 3502 MWth with LEFM in service as indicated on computer point NBA 100 (one-minute average), Perform ON-1 00-004, REACTOR POWER GREATER THAN LICENSE LIMIT Plot position on Power/Flow Map, Form NDAP-QA-0338-10.

AS REQUIRED, Take Action to correct any apparent change in any following variable which could affect reactivity: Recirculation flow or scram.

Check offgas monitors at discharge of SJAE for change in activity.

Check that applicable nuclear safety limits were not exceeded.

Check that applicable thermal-hydraulic limits were not exceeded.

Comply with Technical Specification 3.1.2.

0 0

0 Notify Reactor Engineering.

0 0

0 Direct PCOM to reduce power to 11 00 percent by performing ON-1 00-004, REACTOR POWER GREATER THAN LICENSE LIMIT.

Instruct the STA to Determine the maximum thermal power excursion by using PlCSY to obtain a plot of NBAO1, One Second Core Thermal Power, during the transient period.

Direct STA to review NDAP-QA-0720, STATION REPORT MATRIX AND REPORTABILITY EVALUATION, Attachments H and L (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Reportability)

Recognize T.S. 3.4.1 Condition B.l applicability.

Recirculation loop flow mismatch not within limits, declare the Recirculation loop with lower flow to be not in operation. within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Discuss local manual control of the Scoop Tube Positioner to reduce the speed of the pump.

Request Work Week Manager (WWM) to have I&C investigate.

Notify Rx Engineer, Station Management, Generation Control Center.

• Denotes Critical Task Form NTP-QA61.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 18 Rev. 0, 10/06/05 ILO-504 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES U

Event No:

1 Brief

Description:

A REACTOR RECIRCULATION PUMP SPEED INCREASE a

INSTRUCTOR ACTIVITY:

Once the crew has assumed shift responsibilities, initiate an increase of the A RRP speed by inserting:

PB 1 IMF CN03:SYB311R621A 100 30 90.36 A RRP SPEED CONTROLLER FAIL TO 100%

If requested, remove power to the Scoop Tube Positioner by depressing:

PB 2 MRF RR164017 OFF POWER TO SCOOP TUBE POSlTlONER OFF ROLE PLAY:

1. If dispatched as NPO to investigate RRP MG Set, report no abnormal activity or indications.

2. As WWM, tell SRO that I&C will be sent right away to investigate and troubleshoot.

3. As I&C, wait approximately 10 minutes, (or upon evaluators cue) and report that a the speed feedback circuit appears to be the problem, and if that is the case, then it will be at least a couple of hours to fix the problem.

4. As GCC, acknowledge power reduction, but request return to power as conditions permit due to heavy winter load.

5.

As STA. determine the maximum thermal power excursion by using PlCSY to obtain a plot of NBAO1, One Second Core Thermal Power, during the transient period, 3571 maximum instantaneous power.

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NTP-QA91.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 19 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM Event No:

293 Brief

Description:

HPCl SYSTEM STEAM LEAK, HPCl ISOLATION LOGIC FAILURE POSIT I OT(TI M ET STUDENT ACTIVITIES PCOP 20 Perform SIMPLEX Fire Alarm AR-SP-001, and report HPCl Room area as cause.

Fire Supp x 228_27ALM, 25/28 - 645 DS115 HPCl Dispatch Fire Brigade Leader/FUS to investigate locally and to proceed with caution.

PCOM/P I

I Determine Fire Pre-Plan 1 13-1 03 not applicable when report of NO FIRE from FBUFUS.

PCOP I

I Monitor and report HPCl Room Temperature on CR Panel 1 C614 instrumentation.

See Note # 1 Place ESW in service and start all Room Coolers when directed by SRO.

Perform AR-114-EO5, HPCl LEAK DETECTIONHI TEMP/HI DlFF TEMP:

DETERMINE cause of alarm by observing URS-G33-1N604 on Panel 1C614.

OPERATE HPCl Riley Module Temperature Read/Set switch.

When complete RETURN HS-B21 B-1 S4A(B) to NORMAL position.

EVALUATE conditions for EO-1 00-1 04 entry.

CHECK for HPCl System leaks.

CHECK Rx Building Ventilation System and HPCl Room Unit Coolers operation.

Perform AR-114-HO3 HPCl ROOM FLOODED as necessary:

ENTER ON-169-002, FLOODING IN REACTOR BUILDING ENTER EO-1 00-1 04, SECONDARY CONTAINMENT CONTROL PCOM/P SRO

• SRO See Note #2 Direct NPO to restore Fire Suppression System when no longer required, and notify SRO that Fire Suppression to HPCl Room will be momentarily INOP to allow restoration.

Perform EO-100-104 due to Reactor Building Area Temp (HPCI) above Max Normal and re-enter upon ROOM FLOODED.

Direct PCOP to manually isolate HPCl Steam Supply Isolation Valves F002 and F003.

Direct PCOP to start ESW and all room coolers with a cooling source.

• PCOP I

I Manually isolate HPCl Steam Supply Isolation Valves when directed by SRO.

SRO Comply with T.S. 3.5.1 CONDITION D due to HPCl INOPERABLE:

Verify by administrative means RClC System is OPERABLE immediately AND Restore HPCl System to OPERABLE status within 14 days.

Note 1 : ON-1 69-002 is a lower tiered document and will not supersede actions directed by the EOP. Therefore, the performance of this ON and segments of it may be delayed, or be performed when the emergency actions of a higher priority have been addressed.

Note 2: HPCl Isolation valves auto logic is failed, requiring manual isolation via key switches.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 20 Rev. 0, 10/06/05 ILO-504 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES

+

Event No:

2, 3 Brief

Description:

HPCl SYSTEM STEAM LEAK, HPCI ISOLATION LOGIC FAILURE INSTRUCTOR ACTIVITY:

When Recirc Speed Controller failure has been completed, insert the HPCl System Steam Leak by depressing:

PB 3 IMF HP152009 0.25 0 0 HPCI ROOM STEAM LEAK ROLE PLAY:

1.

2.

3.

4.

As the Fire Brigade Leader/FUS, report NO FIRE in HPCl Room, but a loud hissing noise as if a steam leak is in progress.

When HPCl STEAM Isolation Valves are closedklosing, report that the loud noise has stopped, and that the only thing you hear is water coming in from the Fire Protection Deluge System.

As necessary, when contacted as WWM.

Ifwhen directed, reset and restore the Fire Protection System IAW OP-013-001. See Instructor station format FD4.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 21 Rev. 0, 10/06/05 ILO-504 NOTES:

SCENARIO EVENT FORM Event No:

4 Brief

Description:

C REACTOR FEEDWATER PUMP CONTROLLER FAILURE POSITION PCOM PCOP SRO TIME 40 Sr Denotes Critical Task STUDENT ACTIVITIES Perform AR-101 -B16 RFPT CONTROL SIGNAL FAILURE.

Determine C RFPT Controller has failed because the green indicating light above CTL SIG FAIL RESET HS-C32-1 S05C switch is illuminated.

Implement ON-145-001, Section 3.4.

0 Place SIC-C32-1 R601C controller in MANUAL.

0 Adjust SICC32-1 R601 C to control RPV Water Level = 35 AND Equalize discharge flows on operating pumps.

0 When SIC-C32-1R601C does not take control, lower RFPT C MTR SPD CHANGER using HS-l2730C1 SLOW pushbutton until RFPT Speed LOWERS.

0 Depress HYD JACK RFPT A(B)(C) HS-12772C ON pushbutton.

0 Adjust RFPT C MTR SPD CHANGER HS-12730 C1 and C2 using SLOW pushbuttons to control RPV Water Level = 35 inches AND Equalize discharge flows on operating pumps.

Provide Peer Check if requested.

Continue with follow-up actions from HPCl Steam Leak event.

Directs implementation of ON-1 45-001, Section 3.4.

Request assistance from WWM.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 22 Rev. 0, 10/06/05 ILO-504 Event No:

4 Brief

Description:

C REACTOR FEEDWATER PUMP CONTROLLER FAILURE INSTRUCTOR ACTIVITY:

When the HPCl evolution has been adequately addressed, insert C Reactor Feedwater Pump Controller Failure by depressing:

PB 4 IMF FW145004C RFPT CONTROL SIGNAL FAILURE If desired, following fuse replacement by I&C, remove RFPT Control Signal failure by depressing:

PB 5 DMF FW145004C Delete malfunction RFPT Control signal Failure ROLE PLAY:

As I&C:

When notified of problem, take five minutes; then report that a fuse blew in the local controller. Wait until the operator actions have been addressed; then report that it can be replaced, but will take about half an hour to get the paperwork in place and the correct fuse reinstalled.

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 23 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM Event No:

5 Brief

Description:

D SRV INADVERTENT OPENING, REACTOR SCRAM POSITION TIME PCOP 50 PCOM SRO I

PCOP/M +

I

• Denotes Critical Task STUDENT ACTIVITIES Perform AR-1 1 0-EO1 MAIN STEAM SRV LEAKING.

0 OBSERVE following on Panel 1 C601:

SRV Open PSV-F013 VI-1 41 81 A.

SRV Open PSV-FOl3 VI-1 41 81 B.

OBSERVE SRV/ADS Temperature TR-B21-1 R614 on Panel 1C614 to determine relief valve indicating temperature increase in discharge piping.

OBSERVE relief valve solenoid energized/de-energized status lights at Panel 1 C601 0

0 Determine the D SRV is the failed component, and should NOT be open at this time.

Perform ON-1 83-001 STUCK OPEN SRV.

0 PLACE D SRV control switch to OFF.

CHECK SRV closed by observing: Acoustic Monitor red light status.

0 CHECK RPV pressure and pressure trend.

0 When determined SRV did not close, PLACE SRV control switch to OPEN.

0 RETURN SRV control switch to OFF.

0 CHECK for SRV closure.

Monitor Suppression Pool Temperature to SCRAM prior to 1 10 O F.

0 On Panel 1 C614, CHECK SRV/ADS Temp Recorder TRS-B21-1 R614, if directed.

0 On Panel 1 C690A(B), confirm SRV is open/still open by observing MSRVs Acoustic Monitor green light ILLUMINATED.

0 CHECK generator MWe and reactor MWTH.

When directed, PLACE REACTOR MODE SWITCH TO SHUTDOWN.

Verify All Control Rods Inserted, insert SRM and IRM detectors Direct performance of ON-1 83-001 Recognize TS 3.6.2.1 to test Vacuum Breakers applicability as time permits.

Direct MODE SWITCH TO SHUTDOWN when determined that efforts to close the SRV are unsuccessful or prior to reaching 1 10 O F Suppression Pool Temperature.

Enter and perform EO-1 00-1 02, RPV CONTROL when RPV water level goes less than

+13 inches.

Perform EO-1 00-1 02, RPV Control actions as directed:

0 0

Verify all Rods inserted.

0 0

0 Ensure all Isolations, ECCS Initiations, DG Starts.

Place Feedwater in Startup Level Control Lineup.

Maintain awareness of stuck open SRV and the possibility of exceeding the RPV Cooldown Rate of less than 100 Fhr if it remains open.

Report SRV closed when identified.

11 NOTES: I I

U I

FO~TTI NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 24 Rev. 0, 10/06/05 ILO-504 Event No:

5 Brief

Description:

ID SRV INADVERTENT OPENING INSTRUCTOR ACTIVITY:

When RFPT C Controller failure event has been completed, initiate an inadvertent opening and sticking of D SRV by depressing sequentially.

PB 6 IMF RVOl:PSV141F13D INADVERTENT OPENING OF D SRV PB 7 IMF MS1830101 OD 100 0 0 D SRV STUCK OPEN NOTE: RPV pressure must remain above shutoff head of Condensate/RHR/Core Spray.

When RPV Pressure is reduced to 800 PSIG, delete Stuck Open SRV by depressing.

PB 8 DMF MS18301 OD ROLE PLAY:

Role play as necessary.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 25 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM Event No:

7 Brief

Description:

FAILURE OF RClC TO INITIATE/TRIP ON OVERSPEED TIME 55 STUDENT ACTIVITIES Recognize/report failure of RClC to automatically initiate.

Attempt Manual Pushbutton Initiation - Determine COMPONENT BY COMPONENT startup required.

Perform RClC Startup, COMPONENT BY COMPONENT PER OP-150-001, Section 2.4, or using the HARD CARD.

0 0

0 0

0 Observe RClC speed rises.

0 ReDort Failure of RClC to SRO.

Place RClC Flow Controller in MANUAL; set for minimum speed.

Start Barometric Condenser Vacuum Pump.

Open Cooling Water Valve, HV-150-FO46.

Open Steam to RCIC, HV-150-FO45.

Diagnose RClC trips on overspeed.

Dispatch a NPO to investigate locally to determine the cause of the trip.

Monitor RPV Pressure, RPV Water Level lowering.

Continue to perform and direct EO-1 00-1 02 actions as necessary to stabilize the plant:

0 0

Restore and maintain RPV level between +13 inches and +54 inches using CRD Maximized and SBLC.

Maintain RPV pressure between 800 and 1,087 psig.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 26 Rev. 0. 10/06/05 ILO-504 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES A

Event No:

7 Brief

Description:

FAILURE OF RClC TO INITIATE/TRIP ON OVERSPEED INSTRUCTOR ACTIVITY:

Ensure the following pre-inserted malfunctions are active:

IMF RLO1 :E51 1 K3 IMF RC150007 RClC Overspeed RClC Failure to Auto Start ROLE PLAY:

As NPO, report overspeed condition is not mechanical. It must be the electrical overspeed signal causing the trip.

Perform no other actions with RCIC.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 27 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM L.

Event No: \\

Brief

Description:

PRIMARY SYSTEM BREAK INSIDE DRYWELUADS FAILURURAPID DEPRESSURIZATION STUDENT ACTIVITIES Recognize and report change in Drywell Temperature and Pressure due to leak.

Ensure all DG start when Drywell Pressure reaches 1.72 psig.

Transition to Fuel Zone RPV Level instrumentation per EOP CAUTION 1.

Report RPV Level decrease to TAF using Fuel Zone Level Instrumentation.

Report loss of all Condensate, Feedwater, Service Water and Circ Water Pumps due to Plant Auxiliary Load Shed Logic actuation (Main Gen Lockout + 1.72 DW Pressure).

Assist PCOP as necessary.

Ensure all available injection sources are started.

Enter EO-1 00-1 03, PRIMARY CONTAINMENT CONTROL due to Drywell Temperature and Pressure conditions.

Re-enter EO-1 00-1 02. RPV CONTROL due to D w e l l Pressure greater than 1.72 psiw.

Recognize RPV Water Level cannot be maintained above -161 inches.

Perform and direct EO-100-1 12, RAPID DEPRESSURIZATION actions when RPV water level reaches -1 61 inches (TAF).

Direct PCOP to manually open six ADS Valves when Auto/Manual PB fail to actuate ADS Logic.

Recognize failure of ADS to Auto Initiate.

Manually open six ADS Valves.

Control injection to restore RPV water level above TAF.

Ensure all Low Pressure ECCS Systems (RHR and CORE SPRAY) are available and in service to restore RPV water level above TAF.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 28 Rev. 0, 10/06/05 ILO-504 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES I

Event No:

6, 8, 10 Brief

Description:

PRIMARY SYSTEM BREAK INSIDE DRYWELUADS FAILURE/RAPID DEPRESSURIZATION INSTRUCTOR ACTIVITY:

When Mode Switch is placed to Shutdown, initiate the Drywell leak by depressing:

PB 9 IMF RR164010 10 0 0 RPV Bottom Head Drain Line LeaklRupture (1 0 percent) Inside Containment Increase RWCU leak as necessary to ensure RPV level reaches TAF and Drywell Sprays will be required by depressing:

PB 10 MMF RR164010 20 2:OO RPV Bottom Head Drain Line LeaklRupture (20 percent) Inside Containment ramped over two minutes When Suppression Chamber Sprays have been initiated, increase the Drywell leak to 20 percent by depressing:

PB 11 MMF RR164010 30 3:OO RPV Bottom Head Drain Line LeaklRupture (30 percent) Inside Containment ramped over three minutes To Ensure Operators are unable to use Condensate Pumps following Plant Auxiliary Loadshed, trip them on overcurrent if started by depressing:

PB 12 IMF PM03:1P102A 1A Condensate Pump overcurrent trip PB 13 IMF PM03:1P102B 1 B Condensate Pump overcurrent trip PB 14 IMF PM03:1P102C 1 C Condensate Pump overcurrent trip PB 15 IMF PM03:1P102D 1 D Condensate Pump overcurrent trip ROLE PLAY:

As necessary Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 29 Rev. 0, 10/06/05 ILO-504 SCENARIO EVENT FORM Event No:

N/A Brief

Description:

PRIMARY CONTAINMENT CONTROL/TERMINATION

)I POSITION 1 TIME I

~

_ _ _ _ _ _ _ _ _ _ _ ~

STUDENT ACTIVITIES

• PCOP

• PCOP When Suppression Chamber pressure exceeds 13 psig, sprays the Drywell.

Limits Drywell spray flow to between 1,000 to 2,800 gpm for first 30 seconds.

SRO Continues EO-1 00-1 03, PRIMARY CONTAINMENT CONTROL.

• SRO When Suppression Chamber pressure exceeds 13 psig, directs initiation of Drywell Sprays.

I CREW I

I Verifies conditions are met for initiation of Suppression Chamber and DW Spray.

Monitors Containment pressure trends after Suppression Chamber and DW Spray have I

been initiated.

u pcop I Sprays the Suppression Chamber IAW OP-149-004, RHR OPERATION IN CONTAINMENT SPRAY MODE.

us Directs terminating Drywell Spray before Drywell pressure drops to 0 psig.

is complete, classifies the event as a SlTE AREA EMERGENCY e to a Loss or Potential Loss of the Fuel Clad Barrier and a Loss

• Denotes Critical Task NOTES: I Scenario duration may not be long enough to allow crew to terminate Drywell Spray.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 30 Rev. 0, 10/06/05 ILO-504 AND INSTRUCTORS PERSONAL NOTES Event No:

6, 8, 10 Brief

Description:

PRIMARY CONTAINMENT CONTROL/TERMINATION INSTRUCTOR ACTIVITY:

NIA ROLE PLAY:

As necessary TERMINATION CRITERIA:

Reactor depressurized, RPV level has been restored above TAF, and Primary Containment parameters have been addressed.

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, Rev. 1

PPL-SUSQUEHANNA, LLC TRAINING CENTER Prepared By:

Reviewed By:

Approved By:

SIMULATOR SCENARIO Richard E. Chin 10/07/05 Instructor Date Nuclear Operations Training Supervisor Date Supervising ManagerEhift Supervisor Date I

Scenario

Title:

ILO CertificationhIRC Exam Scenario Scenario Duration:

90 Minutes Scenario Number:

ILO-505 RevisiodDate:

Rev. 0, 10/07/05

~

Course:

PCOO7/PCOO8 Initial License RO/SRO Certification Examination PCO17/PCO18 Initial License RO/SRO NRC Examination Operational Activities:

Quarterly Surveillance LOCA Relay Failure ADS Logic Failure EHC Oscillations Inadvertent HPCl Initiation Rapid Depressurization RHR Injection Valve Failure RPV Flooding Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 2 Rev. 0, 10/07/05 ILO-505 THIS PAGE IS INTENTIONALLY LEFT BLANK 11 -

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 3 Rev. 0, 10/07/05 ILO-505 SCENARIO

SUMMARY

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-1 84-001, 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-1 93-001, 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 HPCI 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 HPCI 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-100-012, 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 FS1.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 4 Rev. 0, 10/07/05 ILO-505 I

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Page 5 Rev. 0, 10/07/05 ILO-505 H

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:

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 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. Interpret/diagnose 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. Interpret/diagnose 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 (03/04) 2005 NRC Exam, Rev. 1

Page 6 Rev. 0, 10/07/05 ILO-505 Form NTP-QA-3 1.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 7 Rev. 0, 10/07/05 ILO-505 CRITICAL TASKS Perform Rapid Depressurization when RPV level drops to -1 61.

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 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 RC/L-16 WHEN LVL CANNOT BE RESTORED AND MAINTAINED > -161 GO TO RAPID DEPRESS Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 8 Rev. 0. 10/07/05 ILO-505 CRITICAL TASKS I]

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).

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 Reauirina 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 0-1 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 Floodinq when RPV water level becomes indeterminate bv establishina RPV Flooded to Steamlines.

Safety Siq n if icance Adequate core cooling may be challenged if core submergence can not be verified.

Conseauences 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:

RC/L-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.

Form NTP-QA-31.7A 2005 NRC Exam, Rev. 1 Rev. 0 (OW04)

Page 9 Rev. 0, 10/07/05 ILO-505 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 open RHR F015B Valve to iniect to the RPV.

Safetv Siqnif icance:

Ensures Adequate Core Cooling.

IndicationdCues for Event Reauirinp Critical Task Auto initiation signals present as evidenced by High Drywell Pressure, Low RPV Pressure, Low RPV Water Level, initiation signal indication lamp illuminated.

Conseauences for Failure to Perform Task Lack of Adequate Core Cooling Indicationdcues for Event Reauirina Critical Task RPV water level less than -1 29 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.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 10 Rev. 0, 10/07/05 ILO-505 I/

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 11 Rev. 0, 10/07/05 ILO-505

- I SCENARIO REFERENCES 1

1. Quarterly Surveillance SO-184-001 QUARTERLY MSlV CLOSURE RPS INSTRUMENTATION 0

TECH SPECS 3.3.1.1 RPS INSTRUMENTATION Ml-C72-22 SH. 1 1 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 (0304) 2005 NRC Exam, Rev. 1

Page 12 Rev. 0, 10/07/05 ILO-505 THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 13 Rev. 0, 10/07/05 ILO-505 I

SCENARIO SPECIAL INSTRUCTIONS I

1. Initialize to IC 20. For NRC LOC 21 Exam, IC 187 is dedicated for this setup.

2. Prepare US Turnover Sheet indicating:

NoLCO.

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-1 84-001 with green cover sheet.

Form NTP-CIA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 14 Rev. 0, 10/07/05 ILO-505 1

SCENARIO EVENT DESCRIPTION FORM Initial Conditions: IC-20, Both Units at 100 Percent Power; All Systems are OPERABLE.

5 Perform SO-1 84-001, QUARTERLY MSlV CLOSURE RPS INSTRUMENTATION 11 EVENT I

TIME I 4

DESCRIPTION 45 Inadvertent HPCl Initiation/Unexplained Reactivity 5

II 55 Recirculation Loop B Suction Line Break DBA 2

6 I

15 I Relav failure C72A-K3F 55 ADS Auto Logic Failure II 7

3 60 RHR Injection Valve 158 Auto Logic Failure I

30 I EHC Oscillations a

65 Rapid Depressurization 9

I 70 I RPV Floodina 75 Termination Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 15 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No:

112 Brief

Description:

SO-184-001 QUARTERLY MSlV CLOSURE RPS INSTRUMENTATIONFIELAY FAILURE Y

STUDENT ACTIVITIES Establish Communications with Relay Room Operators.

Report failure of response for Relav C72A-K3F.

I Monitor Main Steam Line flow indications.

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

Discontinue Surveillance until problem resolved.

Y Determine acceDtance criteria not satisfied for the surveillance.

I Reference T.S. 3.3.1.1, RPS Instrumentation, FUNCTION 5 of Table 3.3.1.1 -1 applicable. 11

~

Declare Channel INOP and apply T.S. 3.3.1.1 CONDITION A:

ll I 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 />.

Notify Workweek Manager of problem and request assistance.

I Notify Plant Management of LCO.

I Notify GCC of LCO.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 16 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES I

Event No:

192 Brief

Description:

SO-1 84-001, QUARTERLY MSlV CLOSURE RPS INSTRUMENTATIONlRELAY 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.

3.

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).

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

4.

5.

6.

As GCC acknowledge the situation.

As Station Management acknowledge situation.

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.

F o

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NTP-OA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 17 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No:

3 Brief

Description:

EHC OSCILLATIONS POSITION

~~

PCOM PCOMP us

~

PCOMN PCOP us

~ ~ _ _ _ _ _ _ _ _ _

STUDENT ACTIVITIES Recognize power/pressure/electrical changes.

Perform ON-1 93-003, TURBINE EHC SYSTEM MALFUNCTION 0

0 0

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-1 93-001, Section 3.3.

Notify WWM to direct I&C to place the B Pressure Regulator in service.

If time permits, notify 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.

Slowlv restore LOAD LIMIT SET to 8.9.

Direct/discuss restoration of Reactor Power to 100 percent.

Denotes Simulator Critical Task.

NOTES:

I I

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 18 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES

1.

1 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 EHC OCILLATION OF 6 RAMPED OVER 1 MINUTE

2.

When the PCO has transferred the oscillations to the Bypass Valve, delete the malfunction by depressing:

[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.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 19 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM NOTES: I 1.

If crew identifies HPCl failure, entry to the Off-Normal may not occur immediately.

li Event No:

4 Brief

Description:

INADVERTENT HPCl INITIATION (UNEXPLAINED REACTIVITY)

STUDENT ACTIVITIES Respond to REACTOR WATER LEVEL HI-LO alarm AR-101 -B17 Reduce Reactor Power to stay 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 0

Place or Check Placed HPCl AUXILIARY OIL PUMP 1 P213 switch to START.

0 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.

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

0 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 0

Verify by administrative means RClC System is OPERABLE immediately AND Restore HPCl System to OPERABLE status within 14 days.

Obtain assistance from WWM.

Enter GO-100-012, 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 power to stable conditions.

Review ON-1 56-001 to determine if any other conditions occurred.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam. Rev. 1

Page 20 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 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 HPCl 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 request for assistance and tell them that the I&C crew has booted the MSlV 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 (03/04) 2005 NRC Exam, Rev. 1

Page 21 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event No:

5 Brief

Description:

RECIRC LOOP B SUCTION LINE BREAK POSITION I TIME

_s us I

STUDENT ACTIVITIES Recognizedreports increasing Drywell pressure and temperature.

Perform ON-1 00-1 01, SCRAM, SCRAM IMMINENT.

Reduce Core Flow to approximately 65 Mlbmhr.

Place Mode Switch to Shutdown.

Verifies all Control Rods inserted.

Directs performance of ON-1 00-1 01, SCRAM, SCRAM IMMINENT.

Perform EO-1 00-1 02, RPV CONTROL due to RPV Water level c +13 inches.

Perform EO-1 00-1 03, PC CONTROL due to High Drywell Pressure 1.72 psig.

Re-enters EO-100-102, RPV CONTROL due to High Drywell Pressure 1.72 psig.

Direct initiation of Suppression Chamber Sprays.

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.

Recognizes RPV water level lowering at a faster rate.

Directs injection with all available systems.

Transition RPV water level monitoring to Fuel Zone RPV Level Instrumentation when RPV level indication goes below -145 on Wide Range Level Instrumentation.

Report RPV water level approaching Top of Active Fuel, -1 61 inches.

Denotes Simulator Critical Task.

1 Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 22 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 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:

[PB-5] 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-6] 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-A IMF HPl52011 ROLE PLAY:

As necessary Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 23 Rev. 0, 10/07/05 ILO-505 SCENARIO EVENT FORM Event Nos:

6979 899 Brief

Description:

ADS FAILURE, RHR 158 FAILURE, RAPID DEPRESSURIZATION, RPV FLOODING STUDENT ACTIVITIES 0

Verify Suppression Pool level >5 feet.

0 Verify valves open.

0 Acoustic monitors cycling.

0 SRV Tailpipe Temperature subcooled.

0 RPV pressure rising.

After the scenario is complete, classifies the event as a to a Loss or Potential Loss of the Fuel Clad Barrier and a Loss Denotes Simulator Critical Task.

L.

U Y

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 24 Rev. 0, 10/07/05 ILO-505 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

697,899 Brief

Description:

RAPID DEPRESSURIZATIOWRPV FLOODING INSTRUCTOR ACTIVITY:

N/A ROLE PLAY:

As necessary TERMINATION CUE:

EO-1 00-1 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.

F o

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

PPL-SUSQUEHANNA, LLC TRAINING CENTER SIMULATOR SCENARIO Prepared By:

Richard E. Chin Instructor

~~~~~~

~

~

~

~

Scenario

Title:

ILO CertificatiodNRC Exam Scenario 10/07/05 Date Scenario Duration:

90 Minutes Reviewed By:

Nuclear Operations Training Supervisor Approved By:

Supervising ManagedShift Supervisor Scenario Number:

ILO-602 Date Date RevisiodDate:

Rev. 0, 10/07/05 PCOO7/PCOO8 PCO17/PCO18 Initial License RO/SRO Certification Examination Initial License RO/SRO NRC Examination Course:

Operational Activities:

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 2 Rev. 0, 10/07/05 ILO-602 Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 3 Rev. 0, 10/07/05 ILO-602 SCENARIO

SUMMARY

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 3.1 0.4 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-145-001, 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-1 64-001, 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 e 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 0-1 00-1 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 11A and 1 1 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.

NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 4 Rev. 0, 10/07/05 ILO-602 F

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 5 Rev. 0, 10/07/05 ILO-602 SCENARIO OBJECTIVES d

The objective of this scenario is to evaluate the Licensed Operator Candidates 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. 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.

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NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 6 Rev. 0, 10/07/05 ILO-602 I

1)

THIS PAGE IS INTENTIONALLY LEFT BLANK Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 7 Rev. 0, 10/07/05 ILO-602 Recoqnize failure to scram and inhibit ADS.

Safety Sign if icance 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 INHFADS When scram and ARl have failed, reactor power must be considered to determine if immediate boron injection is required. If initial A W S 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 A W S 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 (03/04) 2005 NRC Exam. Rev. 1

Page 8 Rev. 0. 10/07/05 ILO-602 CRITICAL TASKS

• 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 dam age.

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~TI NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 9 Rev. 0, 10/07/05 ILO-602 CRITICAL TASKS 1

I 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:

LQ/L-1 3 MAINTAIN LVL BETWEEN -60" AND -161 I' 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/flow 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 taruet 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.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam. Rev. 1

Page 10 Rev. 0, 10/07/05 ILO-602

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 10 inches the downcomer free area reduces from 300 rt' to 88 rt', resulting in increased magnitude of indica fed 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 lowerplenum 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 > - 761 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.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

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, andor 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 relatively 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 Dewessurization when RPV level drops to -1 61 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.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

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 AUGMENTING AS DESIRED WITH TABLE 5 ALTERNATE SUBSYSTEMS MAINTAIN LVL > -129 USING TABLE 3 SYSTEMS RC/L-1 0 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 RC/L-l6 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 starf pumps, attempting to reverse the decreasing RPV water level trend before Rapid Depressurization is required to assure continued adequate core cooling.

(

Reference:

SSES-EPG Cl-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 (03/04) 2005 NRC Exam, Rev. 1

Page 13 Rev. 0, 10/07/05 ILO-602 I

CRITICAL TASKS 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, Rev. 1

Page 14 Rev. 0, 10/07/05 ILO-602 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 LQ/L-6, LQ/L-12 and LQ/L-13 permit use of these same Table 15 systems. Refer to RC/L-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 System availability 0

0 System throttling capabilitykontrol 0

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 -1 61 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 LQ/L-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-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 15 Rev. 0. 10/07/05 ILO-602 1

CRITICAL TASKS

-1 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.

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NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

THIS PAGE IS INTENTIONALLY LEFT BLANK Page 16 Rev. 0, 10/07/05 ILO-602 Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 17 Rev. 0, 10/07/05 ILO-602 SCENARIO REFERENCES 1

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-A01 AR-103-C05 ON-1 64-001 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 B 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 / 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-100-1 12 EO-1 00-1 03 OP-149-005 RAPID DEPRESSURIZATION, REV. 1 PRIMARY CONTAINMENT CONTROL, REV. 2 RHR SUPPRESSION POOL COOLING, REV. 21 Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 18 Rev. 0, 10/07/05 ILO-602 MALFS II SCENARIO SPECIAL INSTRUCTIONS U

REMFS I

OVRDS 1

TRlGS

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.

5:4 I

0

2. Type restorepref YPP.IL0-602; verify the following pre-inserts and Program Button assignments.

See attached Preference file for reference.

2:2 1

Verify the Environment window:

MALFUNCTIONS AVO1 :HV155FlOO BR05:1A10104 BR05:1Al0204 PM03:l P208A (El 3:OO 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-51 IMF NM178012D

[P-61 MRF NM178008 ZERO

[P-81 IMF TC193001

[P-91 IMF RC150002 1000 60 4400

[P-4] IMF TRM:FTC321 N003C 1.5 0 3.6

[P-A bat YPB.IL0-602A

[P-1 01 bat YPB.IL0-602C

[P-111 bat YPB.IL0-602D

[P-231 BAT FWB.101ALARM

[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 TRANSMllTER FAILS TO 1.5 MLBMMR D FLOW UNIT FAILS DOWNSCALE D FLOW UNIT SWITCH TO ZERO MAIN TURBINE TRIP RClC SPEED CONTROL FAILURE Q 1,000 RPM ATWS-ELECROSS OF CRD/4 ACCUM ALARMS ES-158-001 DIV 1 FUSES PULLEDKLEAR 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 B2 to use as page replacements following completion of the scenario.

6. Run Simulator to stabilize conditions.

Rev. 0 (03104) 2005 NRC Exam, Rev. 1 Form NTP-QA91.7A

Page 19 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT DESCRIPTION FORM Initial Conditions:

EVENT I TIME I DESCRIPTION 1

5 SHIFT IN-SERVICE CRD PUMPS 2

15 LEFM COMPUTER FAILURE 3

30 C MSL FLOW TRANSMITTER 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 a

MAIN TURBINE TRIIPROSS OF AUX BUSES 9

RClC SPEED CONTROL FAILURE 10 90 RAPID DEPRESSURIZATION Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

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 TIME STUDENT ACTIVITIES Directs PCO to place 6 CRD Pump in service and shut down the A CRD Pump.

Implements OP-155-001, Section 2.9.

1.

2.

3.

4.

5.

6.

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-l4606B, Check 1 P132B, CRD Pump B, 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.

Ensure PDI-C12-1 R602, Drive Water Diff Pressure, - 250 psid.

Notifies Maintenance that B CRD Pump is in service.

Denotes Simulator Critical Task.

NOTES:

- I I

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NTP-QA91.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 21 Rev. 0, 10/07/05 ILO-602

- II INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

1 Brief

Description:

SHIFT IN-SERVICE CRD PUMPS INSTRUCTOR ACTIVITY:

NOTE: Monitor P&ID RD1.

When directed to close 3 CRD Pump Discharge Valve HV-146F0148, Depress P-1:

[P-1] MRF RD155014 0 146F014B CLOSED When directed to reopen HV-l46FO14B, 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 O F.

Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 22 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

2 Brief

Description:

LEFM COMPUTER FAILURE 11 POSITION I TIME I STUDENT ACTIVITIES I

PCOM I

I Recognize and respond to Computer Alarm and indications.

I I

I ReDOrt LEFM failure to SRO.

Utilize APRMs for indication of Reactor Power.

When directed, ensure RX power 53489 Mwth by reducing Core flow by 0.5 Mlbhr.

Direct performance of ON-1 00-006, LOSS OF REACTOR HEAT BALANCE CALCULATION.

SRO I Declare LEFM INOPERABLE per TRO 3.10.4.

I PCOM/P T 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.

Immediately suspend any and all activities related to reactivity increase in the core, including control rod withdrawal and recirculation pump speedhlow 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 Contact Workweek Manager/lC for assistance.

Direct NPO to investigate locally.

Direct NPO to attempt to re-close 1Y128 Breaker 38.

Report breaker failure to SRO.

Y I

I I

Denotes Simulator Critical Task.

F O n NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 23 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES

/

1 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

When directed to investigate, report that the Display Monitor (1 C1107) appears de-energized.

Report 1 Y 128 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-QA-31,7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 24 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

3 Brief

Description:

C MSL FLOW TRANSMITTER FAILURE POSITION I TIME STUDENT ACTIVITIES PCOM/P I

Respond to lowering RPV water level alarm.

us Denotes Simulator Critical Task.

Perform ON-1 45-001, RPV LEVEL CONTROL SYSTEM MALFUNCTION, Section 3.6:

0 0

0 0

0 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(B)(C).

Depress Green 1 ELEM PB for 1 OR 3 ELEMENT LEVEL CONTROL HS-106102.

Null FW LEVEL CTUDEMAND SIGNAL LICC32-1 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.

I 11 NOTES:

I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam. Rev. 1

Page 25 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 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:

[P-4] IMF TR02:FTC321 N003C 1.5 0 3.6 MSL FLOW TRANSMITER FAILS TO 1.5 MLBM/HR 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, Rev. 1

Page 26 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

4 Brief

Description:

RECIRC FLOW UNIT D FAILS DOWNSCALE POSITION PCOM

~

us PCOM TIME STUDENT ACTIVITIES Reports RPS half-scram condition on Division 2.

Refers to AR-103-AO1, RPS CHANNEL Al/A2 AUTO SCRAM.

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

Refers to ON-164-001, RECIRC DRIVE FLOW INSTRUMENT FAILURE.

1.

2.

3.

Verifies control rod block.

Dispatches a Plant Operator to LPR to investigate Flow Unit status.

Determines C and D Flow Units COMPARE lights are illuminated.

Determines and reports D Flow Unit has failed downscale.

Directs implementation of ON-164-001, 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.

Bypasses Flow Unit A at 1 C651 by placing JOYSTICK in A position.

Dispatches Plant Operator to bypass Flow Unit A at Panel 1 C608.

At Panel 1C651. verifies APRM flow-biased half-scram and rod block sianals 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 A1/A2 AUTO SCRAM alarm CLEAR.

Denotes Simulator Critical Task.

NOTES: I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 27 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS 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:

[P6] 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 1C608, 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.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

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 us Refers to AR-107-DO2, CRD PUMP B TRIP.

Implements ON-155-007, 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.

ReDorts all trouble alarms are for withdrawn control rods.

Closes Flow Control Valve FV-146-FO02 using FC-C12-1 R600 in MANUAL.

Attempts to START A CRD Pump; reports A CRD Pump failed to start.

Dispatches a Plant Operator to the HCU to report status.

Directs implementation of ON-1 55-007, 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-1 00-1 01. 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.

NOTES: I I

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 29 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

5 Brief

Description:

LOSS OF CRDANOPERABLE ACCUMULATORS INSTRUCTOR ACTIVITY:

After RPS half-scram is RESET, insert a loss of CRD System flow with several control rod accumulator trouble alarms; Depress P-7:

[P-q bat YPBJLO-602A ATWS-ELECROSS 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 6 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 10-1 9, 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, Rev. 1

Page 30 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

6, 75 Brief

Description:

ATWS/SLC SYSTEM FAILURE PO3SI:N I TIME

• PCOM +

• us PCOP

• PCOP PCOP 4-

• PCOM STUDENT ACTIVITIES When directed. Dlaces 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-000-1 02, RPV CONTROL, exits EO-000-1 02 and enters EO-000-1 13, LEVEL POWER CONTROL.

Directs initiating 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.

2.

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

Directs Plant Operator to vent scram air header.

Directs venting the scram air header.

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

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 31 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 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:OO 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.

F o

~

NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam. Rev. 1

Page 32 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

8,9 Brief

Description:

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

• us NOTE 1 SrPCOM TIME STUDENT ACTIVITIES Directs lowering RPV water level to <-60 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.

Lowers RPV water level to <-60 inches but >-161 inches.

1.

2.

Reduces RFP speed/discharge pressure to lower RPV level.

Maintains RPV level c-60 inches but >-161 inches (>-llO inches) using Feedwater.

At Panel 1 C645 place HS-B21 -S38A and HS-B21 -S38C to BYPASS.

Denotes Simulator Critical Task.

Form NTP-QA-31.7A Rev. 0 (0304) 2005 NRC Exam, Rev. 1

Page 33 Rev. 0, 10/07/05 ILO-602 AND INSTRUCTOR'S PERSONAL NOTES Event No:

899 Brief

Description:

MAIN TURBINE TRIP/LOSS OF AUX BUSES/RCIC 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 1 1 A/11 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 1000 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.

Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

SCENARIO EVENT FORM Page 34 Rev. 0, 10/07/05 ILO-602 Event No:

10 Brief

Description:

RAPID DEPRESSURIZATION POSITION

• us

• PCOP

• us

• PCOP TIME P

STUDENT ACTIVITIES Directs Rapid Depressurization when RPV level drops to -161 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 <-60 inches but >-

161 inches using LPCI.

1.

Throttles open HV-151 -FO17A(B) at a rate to prevent/minimize power oscillations.

2.

Injects through the RHR heat exchangers as soon as possible.

Denotes Simulator Critical Task.

NOTES: I Form NTP-QA-31.7A Rev. 0 (03104) 2005 NRC Exam, Rev. 1

Page 35 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES 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 14601 6 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-111 bat YPB.IL0-602C 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 < -60 inches but > -161 inches following Rapid Depressurization, pull Division 2 RPS fuses to complete ES-158-001; Depress P-12:

[ P-1 21 bat Y PB. I LO-602 D ES-158-001 DIV 2 FUSES PULLED ROLE PLAY:

As Shift Manager report that Unit 1/Unit 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 (OW04) 2005 NRC Exam, Rev. 1

Page 36 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM Event No:

10 Brief

Description:

RAPID DEPRESSURIZATION

• PCOM 7

+-

STUDENT ACTIVITIES Inserts control rods IAW EO-000.1 13, Sheet 2, CONTROL ROD INSERTION.

1.

Coordinates ES-158-001 with FUSNPO.

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.

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

Exits EO-000-1 13, Sheets 1 and 2; re-enters EO-000-1 02.

Directs establishing RPV water level +13 inches to +54 inches.

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

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

Directs RHR placed in Suppression Pool Cooling.

Denotes Simulator Critical Task.

F o

~

NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam, Rev. 1

Page 37 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES Event No:

10 Brief

Description:

RAPID DEPRESSURIZATION A

INSTRUCTOR ACTIVITY:

As necessary ROLE PLAY:

As necessary Form NTP-QA-31.7A Rev. 0 (03/04) 2005 NRC Exam. Rev. 1

Page 38 Rev. 0, 10/07/05 ILO-602 SCENARIO EVENT FORM POSITION PCOP us 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 A B.

Opens Suppression Chamber Test Shutoff Valve HV-151 -F028 AA.

Starts RHR Pump 1 P202A(C)/B(D).

Throttles open Test Line Control Valve HV-FO24AA to achieve 51 0,000 gpm on Observes Minimum Flow Valve HV-151 -F007 AB closes at -3,000 gpm.

Closes Heat Exchanger Bypass HV-151 -F048 AB.

Checks RHR Pump Room Coolers 1V210 A(C)/B(D).

FI-Ell-1 R603 AA.

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.

NOTES: I Form NTP-QA-31.7A Rev. 0 (OW04) 2005 NRC Exam, Rev. 1

Page 39 Rev. 0, 10/07/05 ILO-602 INSTRUCTOR ACTIVITIES, ROLE PLAY, AND INSTRUCTORS PERSONAL NOTES c

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, Rev. 1