Text

Final As-Administered SRO Written Examination for the Duane Arnold Initial Examination - Jan/Feb 2005
	ML051250432
Person / Time
Site:	Duane Arnold
Issue date:	01/31/2005
From:	Hironori Peterson; NRC/RGN-III/DRS/OLB
To:	; Nuclear Management Co
References
	50-331/05-301 50-331/05-301
	Download: ML051250432 (161)
	v • d • e

QF-1030-03 Rev. 3 (FP-T-SAT-30)

WRITTEN EXAMINATION COVERSHEET Trainee Name:

Employee Number:

Site:

DAEC Examination Number/Title: 57_05-ILC-SRO NRC-Written Training Program: Operations Course/Lesson Plan Number(s): Senior Reactor Operator, 60006 GRADE:

Total Points Possible: 25 PASS CRITERIA: 70%

Grade: /25= %

Graded by:

Date:

Co-graded by (not required if Scantron graded):

Date:

EXAMINATION RULES

1. References may not be used during this examination, unless otherwise stated.

2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.

3. Conversation with other trainees during the examination is prohibited.

4. Partial credit will not be considered, unless otherwise stated. Show all work and state all assumptions when partial credit may be given.

5. Rest room trips are limited and only one examinee at a time may leave.

6. For exams with time limits, you have minutes to complete the examination.

7. Feedback on this exam may be documented on QF-1040-13, Exam Feedback Form. Contact Instructor to obtain a copy of the form.

EXAMINATION INTEGRITY STATEMENT Cheating or compromising the exam will result in disciplinary actions up to and including termination.

I acknowledge that I am aware of the Examination Rules stated above, Further, I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination.

Examinees Signature:

Date:

REVIEW ACKNOWLEDGEMENT I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.

Examinees Signature:

Date:

Retention: Life of plant insurance policy + 10 yr.

Retain in: Training Records 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Rev. 1

QF-1030-02 Rev. 3 (FP-T-SAT-30)

WRITTEN EXAMINATION COVERSHEET Training program: Operations Examination Number/Title: 57_05-ILC-SRO-NRC-Written Course/lesson plan Number(s):

Reactor Operator, 60006 Retention: 6 years Retain in: Training Records 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Rev. 1

1 Point

1.

Given the following:

The plant has been shutdown.

The A loop of RHR has been placed in the Shutdown Cooling mode.

B Recirc Pump remained in service in parallel with Shutdown Cooling.

RPV temperature was 160°F and lowering slowly.

RPV level was being maintained 190 to 195 with CRD and RWCU.

The B Core Spray Pump was tagged out for maintenance.

The B Loop of RHR was tagged out for maintenance.

At this point, outage workers inadvertently struck instrument rack 1C56 with their equipment causing a LO-LO Level and LO-LO-LO Level initiation signals. Among other alarms and indications:

Annunciator 1C03A (A-8), A CORE SPRAY SYSTEM AUTO INITIATED, activated.

Annunciator 1C03B (A-6), LPCI LOOP SELECT RX LO-LO-LEVEL, activated.

Annunciator 1C03B (A-4), LPCI RX LO-LO-LO-LEVEL INITIATION, activated.

The amber light is ON above the B LOOP SELECTED RHR mimic label.

A Core Spray pump is injecting with RPV level 200 and rising.

In this condition, what is the status of forced circulation in the RPV and what actions are necessary regarding the A Core Spray pump?

a.

There would be no forced circulation in the RPV.

Allow the Core Spray Pump to inject until RPV level is >214 per AOP 149 Loss of Decay heat Removal, before securing it because RPV level must be higher than normal to establish natural circulation.

b.

Forced circulation would continue to be maintained by B Recirc Pump.

Allow the Core Spray Pump to inject until RPV level is >214 before securing it because RPV level should be higher than normal per OI-149, RHR, for realigning RHR Shutdown Cooling to LPCI mode.

c.

There would be no forced circulation in the RPV.

Secure the Core Spray Pump at this time to maintain RPV level 170-211 per EOP-1, RPV Control.

d.

Forced circulation would continue to be maintained by A Loop of RHR in the LPCI mode.

Secure the Core Spray Pump at this time to maintain RPV level 170-211 per EOP-1, RPV Control.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 1 Written Exam

1 Point

2.

(PCIS Mimic indications for this question are provided on the next page.)

The plant was operating at 40% thermal power with no LCOs when 1C08B (D-4), 250V DC SYSTEM TROUBLE, alarm activated and cleared. An operator was immediately dispatched to investigate.

While waiting for this report, the CRS observes the PCIS Mimic indications provided on the next page.

What do these indications tell him about the extent of the problem with 250VDC?

Also, of all the Primary Containment Isolation Valves pictured, how many must the operators close, deenergize, and tag deenergized in order to support the requirements of Technical Specifications and ACP 1410.7, Guidelines for Primary Containment Valves and Penetrations?

a.

Only MCC 1D42 has been lost.

Operators must close, deenergize, and tag deenergized one (1) Motor Operated Valve.

b.

Only MCC 1D42 has been lost.

Operators must close, deenergize, and tag deenergized two (2) Motor Operated Valves.

c.

The Main Distribution Panel 1D40 has been lost.

Operators must close, deenergize, and tag deenergized three (3) Motor Operated Valves.

d.

The Main Distribution Panel 1D40 has been lost.

Operators must close, deenergize, and tag deenergized four (4) Motor Operated Valves.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 2 Written Exam

1 Point

3.

The plant was operating at 93% thermal power. The only equipment problem was a failed FULL-IN reed switch on one control rod.

From the above condition, there was a COMPLETE LOSS OF UNINTERRUPTIBLE AC POWER that resulted in a reactor scram.

EOP-1, RPV Control, has been entered due to RPV low level during the initial transient:

All 8 RPS Scram Group A and B white lights are OFF, The 1C05 operator has reported that he cannot confirm that all rods are fully inserted.

On the 1C05 Full Core Display, all LPRM downscale lights are ON.

All IRMs are fully inserted, on range 3 or 4, reading midscale, and lowering on all available indications.

RPV pressure is 900 psig and lowering slowly with all available Main Steam Lines Drains open.

Standby Liquid Control (SBLC) was NOT injected.

There are no challenges to Containment.

Which of the following correctly describes how the CRS shall utilize the IPOI-5, Reactor Scram, EOP-1, and the ATWS EOP procedures when directing further operator actions in this situation?

a.

All operator actions will be directed from EOP-1 and IPOI-5.

NO operator actions will be directed from the ATWS EOP.

b.

Operator actions for reactivity control will be directed from the ATWS EOP.

Operator actions for RPV level and pressure will be directed from EOP-1.

c.

Operator actions for reactivity control will be directed from IPOI-5.

Operator actions for RPV Pressure and Level will be directed from the ATWS EOP.

d.

NO operator actions will be directed from either EOP-1 or IPOI-5.

All operator actions will be directed from the ATWS EOP.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 3 Written Exam

1 Point

4.

A transient has occurred that has resulted in the following plant conditions:

Torus Water Level is 10.0 ft. and is starting to lower.

Average Torus Water Temperature is 160°F and stable.

Torus pressure is 19.0 psig and stable.

RPV pressure is 500 psig and stable.

1) How much can RPV pressure INCREASE before Emergency Depressurization is required?

2) How much can Torus level DECREASE before Emergency Depressurization is required?

a.

1) 400 psig

2) 1.0 ft.

b.

1) 400 psig

2) 2.0 ft.

c.

1) 500 psig

2) 1.0 ft.

d.

1) 500 psig

2) 2.0 ft.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 4 Written Exam

1 Point

5.

In accordance with ACP 1402.3, Regulatory Reporting Activities, which of the following is considered a VALID ESF actuation?

a.

An Instrument Technician pulls the wrong relay block out of the HPCI initiation relay, during an STP, causing HPCI to initiate.

b.

A Security Guard keys his radio in the Reactor Building, near 1C58, in the NECR, causing a reactor scram on APRM flow biased signals.

c.

Operators isolate one Main Steam Line for an inoperable MSIV and then attempt to perform the Main Turbine Stop Valve test resulting in a pressure surge that scrams the reactor.

d.

A generic problem with the wide range Yarway level indicating switches, causes one to drift below the Low-Low-Low RPV setpoint, causing the initiation of RHR, CS, SBDG, portions of PCIS Groups 1 & 7.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 5 Written Exam

1 Point

6.

(SPDS indications for this question are provided on the following page.)

The plant was being shut down from 93% thermal power due to increased Drywell leakage from an unknown source. However, during this evolution, the plant experienced a complete loss of Well Water that resulted in a reactor scram and entry into EOP-1 and EOP-2.

So far, the operator on 1C03 has initiated Torus Sprays and reported the following Containment parameters:

Average Drywell Air Temperature 270°F and rising slowly Average Torus Water Temperature 80°F and stable Drywell Pressure 7.0 psig and rising slowly Torus Pressure 6.0 psig and rising slowly Torus Water Level 10.3 ft. and stable The STA has called up the applicable screen on the SPDS monitor. See the SPDS indications provided.

What is the appropriate action for the CRS to direct and which parameters are the bases for his direction?

The CRS shall direct that Drywell Sprays

a.

NOT be initiated, based on the SPDS indication.

b.

be initiated, based on the SPDS indication.

c.

NOT be initiated, based on the 1C03 parameters.

d.

be initiated, based on the 1C03 parameters.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 6 Written Exam

1 Point

7.

Initial Plant conditions:

The plant was operating at 93% thermal power.

Well Water was operating normally.

A Control Building Chiller was in service.

At the Fire Alarm Panel, 1C40, the alarm was received for Detector Zone 16 1A3 SWGR RM.

(EAST) followed shortly thereafter by a loss of Bus 1A3.

The Fire Brigade was activated and offsite assistance was called.

Light smoke was detected in the Control Room.

Unnecessary personnel were evacuated.

The operating crew donned SCBAs.

When the reactor was manually scrammed, all control rods fully inserted.

Drywell Temperature and Pressure began increasing to, and eventually above, the EOP entry conditions due to the loss of power.

The plant was stabilized using the appropriate EOPs.

The fire was extinguished shortly after the Fire Brigade arrived on the scene.

The Control Room remained habitable, however Control Room temperature is at 92°F and increasing very slowly.

The smoke has cleared and air quality has been verified in the necessary areas.

Operators have begun restoring Control Building ventilation. The B Chiller and appropriate fans and have been restarted however, the desired airflow can NOT be established and Control Room temperature is NOT lowering.

In accordance with AOP 913, Fire, which of the following is correct concerning damper operation at this point in the event?

The applicable Safe Shutdown Path subsection provides

a.

NO instructions concerning CB HVAC dampers.

b.

instructions concerning CB HVAC dampers however, the given plant conditions make the CB HVAC damper steps NOT applicable.

c.

instructions concerning CB HVAC dampers.

Direct operators to disconnect and secure damper D61-0011 in the CLOSED position.

d.

instructions concerning CB HVAC dampers.

Direct operators to disconnect and secure damper D61-0017 in the CLOSED position.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 7 Written Exam

1 Point

8.

Operators are performing the EOP Contingency RPV Flooding in response to an accident in which all RPV water level indications were lost. The following is a timeline of the accident:

0800 The reactor was successfully scrammed.

0815 All RPV level indication was lost.

0830 Operators are controlling injection to maintain 4 SRVs open, RPV pressure stable at 55 psig, Torus pressure stable at 2 psig.

0910 During a flow adjustment, RPV pressure momentarily dropped to 45 psig and was then restored to 55 psig.

0950 Operators notice one of the four SRVs is no longer open. They raise injection flow to adjust RPV pressure to 60 psig and the SRV reopens.

Determine the MINIMUM CORE FLOODING INTERVAL Finish Time.

Table 4 MINIMUM CORE FLOODING INTERVAL (MCFI)

Number of Open SRVs Time Started

=

+

Finish Time MCFI (min)

a.

0900

b.

0915

c.

0955

d.

1035 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 8 Written Exam

1 Point

9.

An unisolable coolant system leak has occurred in the Reactor Building that has resulted in RPV level lowering to the point that fuel became uncovered and fuel damage occurred.

Operators recovered RPV level and were attempting to stabilize the plant when they noticed a RED annunciator on panel 1C35 for REACTOR BLDG KAMAN 3, 4, 5,6, 7,& 8 HI RAD OR MONITOR TROUBLE.

The Reactor Building Exhaust Fans (1V-EF-11A & B) and the Main Plant Exhaust Fans (1V-EF-1, 2, & 3) responded as designed.

What could be the cause of this alarm and what actions must be directed regarding these fans to mitigate this condition?

a.

The Main Plant Exhaust Fans must still be drawing on the Reactor Building Vent Shaft.

Direct operators to TRIP the Main Plant Exhaust Fans.

b.

The Main Plant Exhaust Fans will have tripped causing a high concentration of activity at the monitors. Direct operators to RESTART the Main Plant Exhaust Fans.

c.

The Reactor Building Exhaust Fans must still be drawing on the Reactor Building Vent Shaft.

Direct operators to TRIP the Reactor Building Exhaust Fans.

d.

The Reactor Building Exhaust Fans will have tripped causing a high concentration of activity at the monitors. Direct operators to RESTART the Reactor Building Exhaust Fans.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 9 Written Exam

1 Point

10. Given the following scenario:

A HPCI Walkdown Surveillance test was in progress at 93% thermal power.

Both Loops of Torus Cooling were in service.

A leak developed in the South East Corner Room (SECR).

The Radwaste Operator has been directed to pump Reactor Building sumps.

The Radwaste Operator reported that two floor drain valves were stuck open.

The HPCI test was suspended and HPCI has been secured.

Torus water Temperature is 93°F.

SECR Room water level is 12 inches and rising.

HPCI Room water level is 4 inches and rising.

A RHRSW flow is abnormally high with MO-2046, RHR Heat Exchanger Service Water Outlet Isolation Valve barely open.

What additional actions, if any, are necessary per the EOPs?

a.

Enter EOP-1 and manually scram the reactor.

b.

Begin a reactor shutdown per IPOIs -3, 4, & 5.

c.

Secure A Loop RHR and RHRSW Pumps.

d.

No further EOP-3 actions are necessary because the leak is not from a primary system.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 10 Written Exam

1 Point

11. (Indications of PCIS Div 1 Panel 1C41 and the Miscellaneous System Status Board are provided on the following pages.)

The plant was at 93% thermal power with no LCOs. Both RPS MG Sets are powering their respective RPS Busses. All LPRMs are operable.

At the Division 1 PCIS Panel 1C41, the indications provided were present but had not yet been identified by the operating crew. RPS was as indicated on the Miscellaneous System Status Board in the Control Room (also provided).

If the A RPS Motor Generator, 1G-51, tripped the OVER-ALL PLANT RESPONSE would be a

_____(1)_____ and the CRS will direct __________________(2)__________________ to re-energize the A RPS Bus.

a.

(1) 1/2 scram (2) aligning the Alternate Power Source to Reg Transformer 1Y2A through 1Y36 and 1Y26 and then transfer the A RPS Bus to the Alternate Power Source by taking C71B-S1A on 1C15 to the ALT position

b.

(1) 1/2 scram (2) transferring the A RPS Bus to the Alternate Power Source by taking C71B-S1A on 1C15 to the ALT position

c.

(1) full scram (2) aligning the Alternate Power Source to Reg Transformer 1Y2A through 1Y36 and 1Y26 and then transfer the A RPS Bus to the Alternate Power Source by taking C71B-S1A on 1C15 to the ALT position

d.

(1) full scram (2) transferring the A RPS Bus to the Alternate Power Source by taking C71B-S1A on 1C15 to the ALT position Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 11 Written Exam

1 Point

12. The plant is at 93% Thermal power.

A problem with plant valving allows Condensate Storage Tank (CST) level to drop to 7.4 feet in both CSTs.

Level is being restored at 0.1 ft every 15 minutes and is currently at 7.5 ft.

What is the impact of this CST level on Emergency Core Cooling System (ECCS) Operating Systems?

a.

There is less than 75,000 gallons of water available for RPV makeup.

Therefore TS 3.5.1 Condition F is entered.

b.

There is less than 75,000 gallons of water available for RPV makeup.

Therefore TS 3.5.1 Condition H is entered.

c.

There is a possibility that, during initiation, water hammer will damage some discharge piping.

Therefore TS 3.5.1 Conditions F and J are entered.

d.

There is a possibility that, during initiation, water hammer will damage some discharge piping.

Therefore TS 3.5.1 Conditions H and N are entered.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 12 Written Exam

1 Point

13. The plant was operating at 93% thermal power when operators noticed that the tailpipe temperature on ADS Safety Relief Valve PSV-4400 was rising at a rate of 5°F every day.

After several days, the tailpipe temperature reached 170°F. The operating crew entered AOP 683 Abnormal Safety Relief Valve Operation for Tailpipe Temperature Baseline + 30°F. They began recording tailpipe temperatures every hour and wrote a CAP to notify plant management.

Assume that PSV-4400 tailpipe continues to rise at the same rate to 205°F. Per AOP 683, to whom, by title, does the next required notification go?

At 205°F notification must be made to the

a.

Event Response Team (ERT) Leader.

b.

Plant Manager.

c.

Site Director.

d.

Site Vice President.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 13 Written Exam

1 Point

14. (The applicable portion of P&ID M-115 is provided on the following page.)

The pressure transmitter that feeds RPV Pressure PI-4590B on 1C03 is reading erratically.

Instrument Technicians suspect air intrusion and want to vent the associated pressure transmitter, PT-4590B.

1) If the venting process is used on PT-4590B, what will be the potential impact on the other instruments sharing that sensing line?

2) Where are the requirements found for using the venting process on this pressure transmitter?

a.

1) Opening a vent path valve will have NO effect other than a change in indicated pressure on 1C03.

2) ACP 1410.1, Operations Working Standards

b.

1) Opening a vent path too much could cause the other pressure instruments that cause ESF actuations to sense a pressure that is lower than actual.

2) ACP 1408.7, Control of Permanent Plant Instrumentation

c.

1) Opening a vent path too much could cause the level instruments that cause ESF actuations to sense a level that is higher than actual.

2) ACP 1410.1, Operations Working Standards

d.

1) Closing a vent path too fast could cause a pressure surge that makes the level instruments that cause ESF actuations to sense a level that is lower than actual.

2) ACP 1408.7, Control of Permanent Plant Instrumentation Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 14 Written Exam

1 Point

15. The plant was operating at 93% thermal power when a grid instability caused a degraded voltage trip of the essential busses. The Diesel Generators successfully reenergized 1A3 & 4 but the temporary loss of bus power caused a complete loss of RPS that resulted in a reactor scram and a PCIS Group 1 Isolation.

After 5 minutes of methodically working through the EOPs, the following conditions exist:

All 8 Scram Group white lights are OFF.

Approximately 1/2 of the GREEN Rod Full In lights are OFF.

The CRS has declared that a Hydraulic ATWS exists and has initiated EOP actions.

ADS is locked out.

The Mode Switch is in Shutdown.

ARI has been initiated and the Recirc Pumps have tripped.

The operator at 1C05 has NOT started inserting IRMs yet.

All APRM Downscale Lights are OFF.

All 1C05 APRM recorders are selected to APRM and are reading downscale.

SRVs are cycling at the Low-Low Set pressures.

If the OSM were to declare an EAL AT THIS TIME, which one must be declared/reported to outside agencies?

AND What is the reason for this classification?

a.

SA2; Reactor power can be verified to be <5% by the way the SRVs are cycling.

b.

SA2; Reactor power can be verified to be <5% by the APRM recorders that are downscale.

c.

SS2; Reactor power can be verified to be >5% by the APRM Downscale lights that are OFF.

d.

SS2; Reactor power cannot be verified to be < 5% at this time.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 15 Written Exam

1 Point

16. All control rod scram times were within the limits of Table 3.1.4-1 during the last scram time surveillance.

While operating at 93% thermal power, operators received 1C05A (F-7), CRD ACCUMULATOR LO PRESSURE OR HI LEVEL.

The alarm was for HCU 02-19. The NSPEO who investigated reported that HCU 02-19 accumulator alarm was due to low nitrogen pressure at 970 psig and steady.

The CRS then directs the NSPEO to perform Precharging CRD (**-**) HCU Accumulator with Nitrogen per OI-255 CRD Hydraulic System Section 8.4.

Which other actions, if any, must the CRS perform when the NSPEO begins precharging the accumulator?

When precharging begins

a.

HCU ACCUMULATOR 02-19 will remain OPERABLE. No action is necessary.

b.

HCU ACCUMULATOR 02-19 must be declared INOPERABLE.

c.

ROD 02-19 must be declared SLOW.

d.

ROD 02-19 must be declared INOPERABLE.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 16 Written Exam

1 Point

17. A plant startup is in progress. The 1C05 operators were withdrawing a group of rods from position 12 to position 24 by group notch withdrawal. All the rods in that group were at position 14, and the first rod in the group was being withdrawn again.

That is when the solid-state timer malfunctioned, applying a withdraw signal longer than the automatic protective circuitry would allow. After the Reactor Manual Control System responded as designed, the rod was identified at position 20. No alarms were received from nuclear instrumentation.

1) Until the timer malfunction is reset, how will the ability of operators to move control rods be impacted?

2) Must AOP 255.1, Control Rod Movement/Indication Abnormal be entered because the control rod qualifies as a Mispositioned Control Rod?

a.

1) Operators will NOT be able to select control rods.

2) AOP 255.1 must be entered because the rod has withdrawn far enough to qualify as a Mispositioned Rod.

b.

1) Operators will NOT be able to select control rods.

2) AOP 255.1 need NOT be entered because the rod has NOT withdrawn far enough to qualify as a Mispositioned Rod.

c.

1) Operators will be able to select control rods but a ROD OUT BLOCK (1C05B A-6) will prevent further withdrawals.

2) AOP 255.1 must be entered because the rod has withdrawn far enough to qualify as a Mispositioned Rod.

d.

1) Operators will be able to select control rods but a ROD OUT BLOCK (1C05B A-6) will prevent further withdrawals.

2) AOP 255.1 need NOT be entered because the rod has NOT withdrawn far enough to qualify as a Mispositioned Rod.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 17 Written Exam

1 Point

18. A plant start up is in progress following an outage during which the containment was de-inerted.

The following is a time line of the startup:

0600 The reactor is made critical.

1200 The MODE Switch was taken to the RUN position.

1400 Inerting of the containment was begun IAW OI 573, Containment Atmosphere Control System.

1800 Reactor power exceeded 15% Rated Thermal Power.

0200 The Aux Boiler broke down. Estimated repair time for the Aux Boiler is 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

0200 (Present Time) Primary Containment Oxygen is currently reading 7.7% and stable.

In accordance with TECHNICAL SPECIFICATIONS, which of the following statements is correct for the oxygen concentration in the primary containment?

a.

There are 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months remaining to reduce the oxygen concentration to < 4%.

If this cannot be accomplished, enter T.S. 3.6.3.2 Condition A to restore Containment Oxygen concentration to <4% in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

There are 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months remaining to reduce the oxygen concentration to < 4%.

If this cannot be accomplished, enter T.S. 3.6.3.2 Condition B to reduce Thermal Power to 15% RTP in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c.

There are 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months remaining to reduce the oxygen concentration to < 4%.

If this cannot be accomplished, enter T.S. 3.6.3.2 Condition A to reduce Containment Oxygen concentration to <4% in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

There are 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months remaining to reduce the oxygen concentration to < 4%.

If this cannot be accomplished, enter T.S. 3.6.3.2 Condition B to reduce Thermal Power to 15% RTP in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 18 Written Exam

1 Point

19. The plant is operating at 93% thermal power when a trouble alarm, for the A SBDG, is received.

The NSPEO is sent to investigate.

The NSPEO calls the control room and informs the RO that the alarm is 1C93 (B-2), LUBE OIL MAKE UP TANK LEVEL LOW, and that it comes in at 206 gallons of lube oil remaining in the Makeup Tank, 1T-114A.

1C93 (B-1), LUBE OIL MAKE UP TANK LEVEL LOW-LOW is NOT in alarm.

The SBDG crankcase oil level is normal, 2 inches above the FULL mark on the dipstick.

The cause of the lowering oil level has already been identified and corrected. The Mechanics have been directed to fill the Make Up Tank with oil.

Which of the following statement is CORRECT concerning the A SBDG Limiting Conditions for Operation (LCO)?

a.

The total SBDG lube oil inventory of 206 gallons satisfies the Tech Spec requirements.

Therefore no LCO is applicable.

b.

The SBDG Make-Up Tank level of 206 gallons satisfies Tech Spec requirements.

Therefore no LCO is applicable.

c.

A 48 Hour LCO must be entered. If lube oil inventory cannot be restored in that time, the SBDG must be declared inoperable.

d.

The A SBDG must be declared inoperable immediately.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 19 Written Exam

1 Point

20. The plant has been shutdown for a forced outage to repair a piping rupture in the Reactor Water Cleanup System.

You are to plan a Work Order that has the potential to cause the loss of all decay heat removal, so the planning should include a calculation of the Time to Boil. The following information is necessary for the calculation:

Operators scrammed the reactor for the shutdown 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ago.

The Work Order is scheduled for this same time tomorrow.

The Control Room is expected to keep RPV water level stable at 200.

The Control Room is expected to keep RPV water temperature stable at 152°F.

Using the attachments provided from AOP 149, Loss of Decay Heat Removal, and IPOI-8, Outage and Refueling Operations, calculate the Time to Boil for this work order.

If all decay heat removal is lost during this work, the Time to Boil will be between

a.

39 and 43 minutes.

b.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 25 minutes and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 35 minutes.

c.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 30 minutes.

d.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 40 minutes and 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months , 30 minutes.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 20 Written Exam

1 Point

21. (RFP 403, Performance of Fuel Handling Activities Appendix 1 & 2 are provided on the following page.)

You are serving as the Fuel Handling Supervisor during refueling.

The Fuel Handler on the Refueling Bridge is directly in front of the mast controls.

When centered, he is facing directly South.

This is mast position 0° for the purpose of this question.

The next fuel move is a fuel assembly in the Spent Fuel Pool with a SOUTH-EAST spring clip orientation.

The Fuel Handler rotates the mast 45° counter-clockwise (Looking down). He grapples the fuel assembly, raises it to Normal Up, centers the mast controls, and starts toward the core.

The fuel assembly is going to a core location of 23-04, in the cell with control rod 22-03, which has been fully inserted.

As Fuel Handling Supervisor, you must provide concurrence that fuel assembly orientation is correct before it is lowered into the core.

Predict the mast position for the correct seating orientation in the core at location 23-04. Use the Spent Fuel Pool and Core Coordinate Maps provided.

For proper fuel assembly orientation, the mast must have a

a.

45° rotation clockwise from 0°.

b.

135° rotation clockwise from 0°.

c.

45° rotation counter-clockwise from 0°.

d.

135° rotation counter-clockwise from 0°.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 21 Written Exam

1 Point

22. The plant is operating at 93% power.

You are the CRS and planning a Work Order to repair the HPCI Auxiliary Oil Pump, 1P-218.

Part of the planning includes the final checks on HPCI performance that will be needed prior to declaring the HPCI System OPERABLE.

In which procedure do you find general guidance on the checks to be performed?

a.

ACP 1408.1, Work Orders

b.

MD-5, Conduct of Maintenance

c.

ACP 110.3, Operability Determination

d.

MD-24, Post Maintenance Testing Program Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 22 Written Exam

1 Point

23. The plant is operating at 93% thermal power.

There is NO testing or maintenance of any kind in progress.

In the Control Room, 1C08B (C-2), B DIESEL 1C94 TROUBLE has activated. The Auxiliary Operator who is responding reports that the alarm is 1C94 (D-1), CV-2081 OPEN. He believes that CV-2081, the Engine Coolant Valve, has failed open because the associated solenoid valve, SV-2081 feels colder to the touch than normal and may be deenergized.

Does the Control Room Supervisor (CRS) have any other responsibility in response to this alarm in addition to initiating a Work Request Card to have the defective component tested/repaired as necessary? If so, identify that responsibility.

a.

The CRS has no additional responsibility beyond initiating a Work Request Card.

b.

The CRS must arrange for the Chemistry Technician to take a local grab sample to prevent an unmonitored release via this pathway.

c.

The CRS must direct the Auxiliary Operator to throttle closed and Warning Tag the associated manual valve, V13-034, to prevent engine cooldown.

d.

The CRS must immediately enter the LCO 3.8.1 for one SBDG inoperable.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 23 Written Exam

1 Point

24. The plant is at 30% thermal power and is being shutdown for a Drywell entry to investigate increased Floor Drain leakage. Annunciator 1C03A (C-4), OFFGAS VENT PIPE RM4116A/B RAD MONITOR DNSCL/INOP, is active because RM4116A has been inoperable for several days.

There are no other LCOs.

Operators were about to begin venting the containment in preparation for performing an air purge (de-inerting) of the containment per OI-573, Containment Atmosphere Control System. Section 6.1 Step (2) of Normal Containment Venting has the operators verify that the above annunciator is reset or that the ARP actions have been completed. When they checked the second installed radiation monitor, RM4116B, they found that it was also INOPERABLE.

May the CRS authorize containment venting in this situation?

If NO, identify the requirement that prevents venting.

If YES, identify the additional requirements to allow venting.

a.

NO; At least one of the installed radiation monitors must be operable per Technical specifications to allow venting.

b.

NO; At least one of the installed radiation monitors must be operable per the Offsite Dose Assessment Manual (ODAM) to allow venting.

c.

YES; However, operators must continuously monitor alternate instrumentation and they must close the primary vent and purge valves within four hours.

d.

YES; However, operators must continuously monitor alternate instrumentation and they must have administrative control of the primary vent and purge valves.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 24 Written Exam

1 Point

25. (Annunciator Panel 1C07A indications for this question are provided on the following page.)

During power operation a leak developed on the bearing supply header of the Main Turbine Lube Oil System. The following is the sequence of events:

The Turning Gear Oil Pump (TGOP), 1P38, and the Emergency Bearing Oil Pump (EBOP),

1P-40, started automatically but Bearing Header pressure CONTINUED TO LOWER.

Level in the Turbine Lube Oil Tank, 1T-1, was LOWERING.

Operators entered AOP 693, Turbine Lube Oil Trouble.

In-plant operators reported that the LEAK COULD NOT BE ISOLATED.

Operators completed the AOP 693 Immediate Actions.

The Main Turbine shaft has STOPPED.

Operators began the AOP 693 Follow Up Actions.

As the CRS, review the status of alarms on 1C07A (provided) and determine which of the following directions are appropriate for the given plant conditions.

Direct operators to

a.

make one attempt to reset the overload trip on the EBOP.

b.

restart the Lift Pumps (1P90A-F), TGOP, and Turning Gear Drive Motor.

c.

reduce Main Generator gas pressure and then secure the Hydrogen Seal Oil System.

d.

initiate a transfer of lube oil from the Clean Lube Oil Tank, 1T-2B, to the Turbine Lube Oil Tank, 1T-1.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 25 Written Exam

SENIOR REACTOR OPERATOR 2005 NRC EXAM INDEX OF ATTACHMENTS Integrated Plant Operating Instructions IPOI-8 Outage and Refueling Operations Attachments o Appendix 2 DAEC Decay Heat Curve o Attachment 3 Time-To-Boil Calculation o Attachment 4 Time-To-150°F Calculation Abnormal Operating Procedures AOP 149 Loss of Decay Heat Removal Attachments o Appendix 1 Heatup Rate Curve - RPV Flooded o Appendix 2 1 Heatup Rate Curve - RPV Level at 200 o Appendix 3 Loss of Fuel Pool Cooling Heatup Rate Curve AOP 913 Fire Emergency Operating Procedures Graph 1 RPV Saturation Temperature Graph 4 Heat Capacity Limit Graph 5 Pressure Suppression Pressure Graph 7 Drywell Spray Initiation Limit EOP-3 Secondary Containment Control Technical Specifications 3.5.1 ECCS Operating 3.6.3.2 Primary Containment Oxygen Concentration 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air Emergency Plan Implementing Procedures EAL-04 System Malfunction Table Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm_Rev1.doc Page 26 Written Exam

Figure for Question # 2 Senior Reacto rator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 2A_Rev1.doc Page 2A Written Exam r Ope

Figures for Question # 6 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 6A_Rev1.doc Page 6A Written Exam

Figures for Question # 11 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 11A_Rev1.doc Page 11A Written Exam

Figures for Question # 11 6

Rev.

RC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 11A_Rev1.doc Page 11B Written Exam Figures for Question # 11 6

Rev.

RC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 11A_Rev1.doc Page 11B Written Exam Senior Reactor Operator, 6000 1

Topic 05-ILC-SRO-N Manual Transfer Switch 1Y36 480 VAC MCC 1B32 1B3211 1B3216A 1Y1A01 Regulating Transformer 1Y1A 1Y1A02 Dist Panel 1Y16 1Y16-10 EPA C1 EPA C2 1Y2A01 Regulating Transformer 1Y2A 1Y2A02 Manual Bypass Switch Panel 1Y25 Dist Panel 1Y26 1Y26-10 Manual Bypass Switch Panel 1Y15 RPS MG Set 1G51 EPA A1 EPA A2 EPA B1 EPA B2 480 VAC MCC 1B42 1B4203A 1B4216 RPS MG Set 1G61 1C15 C71B-S1A MG A ALT 1C17 C71B-S1B ALT MG B C71B-CB1 (1Y30A00) 1Y30 Bus "A" RPS Electrical Distribution Panel 1Y30 Bus "B" 1Y30 W

W W

W RPS Electrical Lineup Date Updated:_________________

Senior Reactor Operator, 6000 1

Topic 05-ILC-SRO-N Manual Transfer Switch 1Y36 480 VAC MCC 1B32 1B3211 1B3216A 1Y1A01 Regulating Transformer 1Y1A 1Y1A02 Dist Panel 1Y16 1Y16-10 EPA C1 EPA C2 1Y2A01 Regulating Transformer 1Y2A 1Y2A02 Manual Bypass Switch Panel 1Y25 Dist Panel 1Y26 1Y26-10 Manual Bypass Switch Panel 1Y15 RPS MG Set 1G51 EPA A1 EPA A2 EPA B1 EPA B2 480 VAC MCC 1B42 1B4203A 1B4216 RPS MG Set 1G61 1C15 C71B-S1A MG A ALT 1C17 C71B-S1B ALT MG B C71B-CB1 (1Y30A00) 1Y30 Bus "A" RPS Electrical Distribution Panel 1Y30 Bus "B" 1Y30 W

W W

W RPS Electrical Lineup Date Updated:_________________

Figure for Question # 14 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 14A_Rev1.docPage 14A Written Exam

Figures for Question # 21 West East North Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 21A_Rev1.doc Page 21A Written Exam

Figure for Question # 25 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Attchmnt 25_Rev1.doc Page 25A Written Exam ALARM WINDOW ENGRAVINGS AND GRID LAYOUT 1C07A 1

2 3

4 5

6 7

8 9

10 11 A

EHC 24 VDC POWER FAILURE MAIN TURBINE TRIP EHC PUMPS 1P-97A/B HS NOT IN AUTO EHC PUMPS 1P-97A/B MOTOR OVERLOAD EHC FLUID FILTER PUMP 1P-95 TRIP TURBINE LUBE OIL CONDITIONER 1T-39 OR CENTRIFUGE 1T-3 TROUBLE TURBINE LUBE OIL BEARING HEADER LO PRESSURE MECHANICAL TRIP LOCKOUT TURNING GEAR OIL PUMP 1P-38 HS NOT IN AUTO LIFT PUMPS 1P-90A - F HS NOT IN AUTO HVAC PANEL 1C-23 TROUBLE B

MOTOR SUCTION PUMP 1P-96 HS NOT IN AUTO TURNING GEAR DRIVE MOTOR HS NOT IN AUTO EHC FLUID RESERVOIR 1T-33 HI LEVEL EHC FLUID RESERVOIR 1T-33 LO LEVEL BYPASS VALVE NO. 1 OPEN TURBINE LUBE OIL TANK 1T-1 HI LEVEL EMERGENCY BEARING OIL PUMP 1P-40 RUNNING EMERGENCY BEARING OIL PUMP 1P-40 HS NOT IN AUTO TURNING GEAR OIL PUMP 1P-38 RUNNING LIFT PUMPS 1P-90A - F MOTOR OVERLOAD SBGT PANEL 1C-24 TROUBLE C

MOTOR SUCTION PUMP 1P-96 RUNNING RX FEED PUMP SEAL WATER DRAIN TANK 1T-135 HI/LO LEVEL EHC FLUID HI/LO TEMP EHC PUMPS 1P-97A/B BOTH RUNNING TURBINE LUBE OIL VAPOR EXTRACTOR 1K-7 TRIP TURBINE LUBE OIL TANK 1T-1 HI-HI LEVEL CLEAN LUBE OIL TANK 1T-2B HI LEVEL CIRC WATER VALVE HO-4201/HO-4202 HYDRAULIC OIL LO PRESSURE TURNING GEAR OIL PUMP 1P-38 MOTOR OVERLOAD TURBINE THRUST BEARING WEAR DETECTOR (1C479)

TROUBLE DRYWELL COOLING PANEL 1C-25 TROUBLE D

MOTOR SUCTION PUMP 1P-96 MOTOR OVERLOAD TURNING GEAR ENGAGED EHC FLUID LO PRESSURE EHC FLUID RESERVOIR 1T-33 HI-HI LEVEL HWC PANEL 1C-22 TROUBLE TURBINE LUBE OIL TANK 1T-1 LO LEVEL DIRTY LUBE OIL TANK 1T-2A HI LEVEL EMERGENCY BEARING OIL PUMP 1P-40 MOTOR OVERLOAD MECHANICAL VACUUM PUMP 1P-32 TRIP OFFGAS PANEL 1C-34 TROUBLE CONTROL BLDG HVAC PANEL 1C-26 TROUBLE

IPOI-8 APPENDIX 2 DAEC DECAY HEAT CURVE Page 1 of 2 RFO-18 Decay Heat Curve (Days After Shutdown) 0.00E+00 5.00E+06 1.00E+07 1.50E+07 2.00E+07 2.50E+07 3.00E+07 3.50E+07 1

6 11 16 21 26 31 36 41 46 51 56 61 66 71 76 81 86 91 96 Days After Shutdown Decay Heat BTUs/hr.

Note: The decay heat curves experience negligible changes each cycle. Therefore, the curves in this procedure are not updated unless it is determined that the changes are significant with regard to decay heat & Time-To-Boil determination. Cycle-specific curves are provided to the Operations Department prior to scheduled refueling outages and may be utilized in lieu of the curves provided in this procedure.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 1_Rev1.doc Written Exam Page 1 of 2

IPOI-8 APPENDIX 2 DAEC DECAY HEAT CURVE Page 2 of 2 RFO 18 Decay Heat Curve (Hours After Shutdown) 0.00E+00 1.00E+07 2.00E+07 3.00E+07 4.00E+07 5.00E+07 6.00E+07 7.00E+07 8.00E+07 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Hours After Shutdown Decay Heat BTUs/hr.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 1_Rev1.doc Written Exam Page 2 of 2

ATTACHMENT 3 TIME-TO-BOIL CALCULATION The following formula may be used to calculate Time-To-Boil in either the Reactor or Fuel Pool:

(212°F - Actual Temp)

Time To Boil = -----------------------------------------------------------------------

(Actual Decay Heat - Heat Removal Capacity) / (K)

Actual Temp (°F) - Check the appropriate box for the input source.

Reactor Temperature Fuel Pool Temperature Actual Decay Heat (BTU/Hr) - Check the appropriate box for the input source.

Typical curves in Appendix 2.

Optimized curves verified by Systems Engineering.

Heat Removal Capacity (BTU/Hr) - Check the appropriate box for the input source.

Typical data in Appendix 3.

Optimized data verified by Systems Engineering.

Assume NONE for the shiftly Time-To-Boil Calc.

Constant K - Constant equivalent to the value (m cp). Select the appropriate value for current plant conditions:

Fuel in Vessel, RPV Level at 200" K = 668,600 BTU/°F Fuel in Vessel, RPV Flooded K = 3.99 x 106 BTU/°F Fuel in Fuel Pool, Pool Full K = 2.05 x 106 BTU/°F Hr

=

Boil To Time F/Hr)

(______

F)

(

=

Boil To Time F)

BTU/

(______

BTU/Hr /

(______

F)

(

=

Boil To Time F)

BTU/

(______

/

BTU/Hr)

BTU/Hr

(______

F)

- F (212

=

Boil To Time

°

° Prepared By: ____________________________________Date: ________________ Time: ___________

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 2_Rev1.doc Written Exam

ATTACHMENT 4 TIME-TO-150°F CALCULATION The following formula may be used to calculate Time-To-150°F in either the Reactor or Fuel Pool:

(150°F - Actual Temp)

Time To 150°F = -----------------------------------------------------------------------

(Actual Decay Heat - Heat Removal Capacity) / (K)

Actual Temp (°F) - Check the appropriate box for the input source.

Reactor Temperature Fuel Pool Temperature Actual Decay Heat (BTU/Hr) - Check the appropriate box for the input source.

Typical curves in Appendix 2.

Optimized curves verified by Systems Engineering.

Heat Removal Capacity (BTU/Hr) - Check the appropriate box for the input source.

Typical data in Appendix 3.

Optimized data verified by Systems Engineering.

Assume NONE for the shiftly Time-To-Boil Calc.

Constant K - Constant equivalent to the value (m cp). Select the appropriate value for current plant conditions:

Fuel in Vessel, RPV Level at 200" K = 668,600 BTU/°F Fuel in Vessel, RPV Flooded K = 3.99 x 106 BTU/°F Fuel in Fuel Pool, Pool Full K = 2.05 x 106 BTU/°F Hr

=

150 To Time F/Hr)

(______

F)

(

=

150 To Time F)

BTU/

(______

BTU/Hr /

(______

F)

(

=

150 To Time F)

BTU/

(______

/

BTU/Hr)

BTU/Hr

(______

F)

- F (150

=

150 To Time F

F F

F

°

° Prepared By: ____________________________________Date: ________________ Time: ___________

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 3_Rev1.doc Written Exam

AOP 149 APPENDIX 1 HEATUP RATE CURVE - RPV FLOODED Vessel Water Heatup Rate (Floodup Condition) 0.00 1.00 2.00 3.00 4.00 5.00 6.00 7.00 8.00 T ime Since Shutdown (Days)

CAUTION The initial heatup rate in the vessel may be higher than the calculated value when RHR or Fuel Pool Cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the vessel and upper levels of the cavity or in the spent fuel pool.

In some cases local boiling may occur but bulk boiling will not occur as long as cooling is restored within the calculated time-to-boil period.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 4_Rev1.doc Written Exam

AOP 149 APPENDIX 2 HEATUP RATE CURVE - RPV LEVEL AT 200" Vessel Water Heatup Rate (Water Level =200")

0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 T ime Since Shutdown (Days)

CAUTION The initial heatup rate in the vessel may be higher than the calculated value when cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the fuel area and upper levels of vessel. In some cases local boiling may occur but bulk boiling will not occur as long as cooling is restored within the calculated time-to-boil period.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 5_Rev1.doc Written Exam

AOP 149 APPENDIX 3 LOSS OF FUEL POOL COOLING HEATUP RATE CURVE Fuel Pool Isolated Water Heatup Rate (Full Core Offload and Previous Spent Fuel) 0.00 2.00 4.00 6.00 8.00 10.00 12.00 14.00 16.00 18.00 T ime Since Shutdown (Days)

CAUTION The initial heatup rate in the spent fuel pool may be higher than the calculated value when cooling is removed from service. The calculation used to generate the heatup rate curves assumes instantaneous mixing and heat transport from the fuel area to the remainder of the system volume. In addition, the calculated heatup rates reflect bulk temperatures not local temperatures. Under natural circulation conditions and the resulting time delay in heat transport, considerable differences in temperature may exist between the fuel area and measured temperatures in fuel pool cooling heat exchanger inlets.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 6_Rev1.doc Written Exam

AOP 913 FIRE ABNORMAL OPERATING PROCEDURE AOP 913 FIRE Usage Level Reference Use Enter the following as applicable:

FIRE ALARMS REQUIRING IMMEDIATE FIRE BRIGADE ACTIVATION PAGE 2

FIRE AREAS PAGE 9

SAFE SHUTDOWN PATHS PAGE 15 OFFSITE ASSISTANCE PAGE 93 ATTACHMENT 1, RPV LEVEL CONTROL PAGE 95 ATTACHMENT 2, CONTROL BLDG. MANUAL DAMPER CONTROL PAGE 99 NOTE Refer to EPIP for EAL Assessment NOTE Throughout this procedure fire response actions will require entry into various T.S. LCOs.

AOP 913 Page 1 of 103 Rev. 40 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE ALARMS REQUIRING IMMEDIATE FIRE BRIGADE ACTIVATION AOP 913 Page 2 of 103 Rev. 40

1C40, A-2 DG 1G-31 OR DAY TANK 1T-37A PRE-ACTION SYSTEM NO. 2 INITIATED A-3 DG 1G-21 OR DAY TANK 1T-37A PRE-ACTION SYSTEM NO. 3 INITIATED B-1 DIESEL FIRE PUMP 1P-49 SPRINKLER SYSTEM NO. 7 INITIATED C-1 RCIC ROOM DELUGE NO. 1 INITIATED C-2 HPCI ROOM DELUGE NO. 2 INITIATED

1C40A, A-1 PUMPHOUSE S/R PIPING AREA FIRE SPRINKLER INIT A-3 REACTOR BLDG R.S.P. 1C-388 FIRE A-5 R.B. HT. EXCH./CHILLER EL. 812' FIRE SPRINKLER INIT D-3 R.B. EQUIPMENT HATCH EL. 786' FIRE DELUGE INIT E-1 CONTROL BLDG. RSP FUSE PANEL 1C422A FIRE E-3 REACTOR BLDG. RSP FUSE PANEL 1C422B FIRE

1C40, DETECTION ZONE WHITE INDICATING LIGHTS (RED FRONT PANEL)

DET. ZONE 13 250V BATTERY RM. (MIDDLE)

DET. ZONE 14 125V BATTERY ROM (WEST)

DET. ZONE 15 1A4 SWGR RM. (WEST)

DET. ZONE 16 1A3 SWGR RM. (EAST)

DET. ZONE 17 125V BATTERY RM. (EAST)

DET. ZONE 21 A DIESEL GEN. RM. (EAST)

DET. ZONE 22 DIESEL GEN. RM. (WEST)

1C40, FIUs RAN-36A REMOTE FIRE ANNUNCIATOR (LOCATED ON SIDE)

17 RX. BLDG. N.E. CORNER ROOM DET. ZONE 41

18 RX. BLDG. N.W. CORNER ROOM DET. ZONE 42

19 RX. BLDG. S.E. CORNER ROOM DET. ZONE 43

20 RX. BLDG. S.W. CORNER ROOM DET. ZONE 44

21 PUMP HOUSE DET. ZONE 45

26 INTAKE DET. ZONE 50 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 3 of 103 Rev. 40 IMMEDIATE ACTIONS

1. Determine location of the fire by reviewing 1C40 and 1C40A annunciators, 1C40B alarm messages and zone indicating units (ZIU) alarms.

2. If the fire alarm is the result of a smoke alarm or a trouble alarm, send an operator to the scene to determine the extent of the fire.

3. If the alarm window has a RED Lens and/or a fire is seen, activate the DAEC Fire Brigade by sounding the site fire alarm and making the appropriate announcement over the plant page. Repeat Site Fire alarm and page announcement. If the cause is known not to be a fire, then the Fire Brigade need not be activated.

4. IF the fire is in the Control Room, Cable Spreading Room, Control Bldg.

HVAC Area or Back Panel Area THEN enter AOP 915, Shutdown Outside Control Room and execute concurrently with this procedure.

5. If the fire cannot be extinguished promptly or if fire is outside the protected area, request offsite fire assistance per the OFFSITE ASSISTANCE section and activate the backup fire brigade.

6. As necessary, provide for maximum water to the fire pump suction as follows:

a.

Maximize makeup flow to the stilling basin.

b.

For a fire at the intake structure provide 2000 gpm or more to the discharge canal with blowdown or radwaste dilution for use as a drafting source by offsite fire department. For all other fires secure blowdown and radwaste dilution.

7. Go to the FIRE AREA section to determine the appropriate SAFE SHUTDOWN PATH subsection (ADM, CB1, RB1, etc.) to enter

8. Perform the steps of the appropriate SAFE SHUTDOWN PATH subsection in parallel with the FOLLOW-UP ACTIONS section.

AUTOMATIC ACTIONS

- Electric Fire Pump 1P-48 starts

- Diesel Fire Pump 1P-49 starts

- Affected fire area sprinkler/deluge system activates.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 4 of 103 Rev. 40 FOLLOW-UP ACTIONS

1. IF any of the following systems are affected by the fire Hydrogen Seal Oil Turbine Lube Oil Generator H2 System Main Transformer Main Turbine Main Generator Both Reactor Recirc MG Sets Alterrex EHC Oil System Both Feed Pumps Both Cond. Pumps Both Circ. Water Pumps Both Cooling Towers THEN shut down the reactor as follows:

1) Run recirc flow control to 27 Mlb/hr.

2) Manually scram the reactor

3) Enter IPOI 5, Reactor Scram concurrently with this procedure

4) Deenergize affected equipment

2. IF only one Reactor Recirc MG Set is affected by fire THEN reduce recirc MG Set speed of the affected MG set to minimum AND trip the affected MG set AND Perform ARP 1C04A A-4 [1C04B,A-1]

"A"["B"] RECIRC MG DRIVE MOTOR TRIP OR MOTOR OVERLOAD

3. IF only one of the following components have been affected by the fire:

Condensate Pumps Reactor Feed Pumps Circ Water Pumps Cooling Towers THEN reduce reactor power to less than 50% as rapidly as possible per IPOI 4, Section 6.0, Fast Power Reduction AND secure the affected equipment MANUAL SCRAM Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 5 of 103 Rev. 40 FOLLOW-UP ACTIONS (continued)

NOTE For the determination of EALs, the fifteen (15)-minute clock begins when the Control Room is notified or the control room alarm is verified.

4. IF any fire in a safe shut down equipment area cannot be extinguished within fifteen (15) minutes of detection OR the fire spreads into another safe shutdown equipment area.

THEN refer to EPIP for EAL determination

5. IF the fire is outside the protected area THEN the Operations Shift Manager or Control Room Supervisor shall determine if Fire Brigade response can be made without endangering the power plant.

NOTE There are NO fire related EAL's for the ISFSI.

6. IF the fire is inside the ISFSI area THEN the Operations Shift Manager or Control Room Supervisor shall determine the possible threat to the stored fuel.

AND Determine if Fire Brigade response can be made without endangering the Power Plant.

7. Shut down equipment and electrical distribution affected by the fire.

8. Refer to Technical Specifications and enter appropriate LCOs.

9. As time allows, monitor the operation of the Electric Fire Pump 1P-48 and the Diesel Fire Pump 1P-49, if running.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 6 of 103 Rev. 40 FOLLOW-UP ACTIONS (continued)

10. WHEN the fire has been extinguished THEN complete the following actions:

1) Notify DAEC Fire Marshal.

2) Secure any activated deluge or sprinkler system.

3) Station a reflash watch in the fire area.

4) Return Fire Protection System to standby by securing fire pumps per OI 513, Section 7.0 and return fire pumps to standby readiness per section 3.4 and 3.5.

5) Direct Health Physics personnel to conduct surveys in the fire area.

6) Restore fire fighting equipment to its required locations OR place fire watches until the equipment is declared operable.

7) Restore normal stilling basin makeup and blowdown, as necessary.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 7 of 103 Rev. 40 PROBABLE ANNUNCIATORS Any fire annunciator on Panel 1C-40 or 1C-40A 1C40B Alarm Zone Indicating Unit (ZIU) Alarms PROBABLE INDICATIONS Any Site Fire/Smoke alarm activation Any sprinkler/deluge system activation Any unplanned automatic start of a fire pump Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 8 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE AREAS AOP 913 Page 9 of 103 Rev. 40 FIRE AREAS Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE AREAS AOP 913 Page 10 of 103 Rev. 40 NOTE PALO FIRE DEPARTMENT ---------- 911 CEDAR RAPIDS FIRE DEPARTMENT ---------- 911 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE AREAS AOP 913 Page 11 of 103 Rev. 40 ADMIN/SECURITY IF the fire is in the THEN go to Admin Bldg First Floor Second Floor Third Floor Security Bldg SCP Lobby, Ingress/Egress Offices, Lunch Room CAS,UPS,Repair Room AFP-51 AFP-52 AFP-53 AFP-54 AFP-54 AFP-55 AFP-55 ADM CONTROL BLDG IF the fire is in the THEN go to Cable Spreading Room Control Room Control Room HVAC Room*

AFP-25 AFP-26 AFP-27 AOP 915 Requires Manual Operator Action within 20 minutes*

Battery Room 1D2 Essential Switchgear 1A4 AFP-23 AFP-24 CB2 Battery Room 1D1 Essential Switchgear 1A3 North Turb. Bldg/Rx Bldg Air Lock in overhead (chase above airlock)

AFP-23 AFP-24 NONE CB3 Battery Room Corridor Battery Room 1D4 AFP-23 AFP-23 CB4

A fire in the Control Room HVAC Room also requires the opening of V-33-0220, Sprinkler system #12 Shutoff. (TURB. Bldg., 757' North Open End, East of Feed water Reg. Valves)

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE AREAS AOP 913 Page 12 of 103 Rev. 40 INTAKE STRUCTURE IF the fire is in the THEN go to A RWS Pumps A RWS Screen Wash Area AFP-31 AFP-32 IS1 B RWS Pumps B RWS Screen Wash Area AFP-31 AFP-32 IS2 PUMP HOUSE IF the fire is in the THEN go to B RHR/ESW Service Water Pumps RWS Makeup Valve Area South end Pump House Elev 747 North end Pump House Elev 747' AFP-28 AFP-30 AFP-30 AFP-30 PH1 A RHR/ESW Service Water Pumps AFP-28 PH2 OUTSIDE AREA IF the fire is in the THEN go to Div. 1 Manhole (manholes from near turbine bldg to intake structure)

Div. 2 Manhole (manholes from near turbine bldg to intake structure)

Diesel Fuel Oil Supply A Cooling Tower B Cooling Tower Off Gas Stack Switchyard Standby Transformer 1X4 East Warehouse West Warehouse ISFSI None None None AFP-73 AFP-73 None AFP-74 AFP-70 AFP-67 AFP-68 AFP-79 YARD Startup Transformer 1X3 Auxiliary Transformer 1X2 Main Transformers 1X1 AFP-71 AFP-72 AFP-69 TB1 Requires manual operator action within 30 minutes Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE FIRE AREAS AOP 913 Page 13 of 103 Rev. 40 TURBINE BLDG IF the fire is in the THEN go to Turbine Bldg Startup transformer 1X3 Main Transformers 1X1 Auxiliary Transformer 1X2 AFP 14-22 AFP 71 AFP 69 AFP 72 TB1 Requires manual operator action within 30 minutes.

REACTOR BLDG IF the fire is in the THEN go to Torus Room (Bays 1-5 and 11-16 )

NW Corner Room HPCI ROOM SW CORNER ROOM North CRD AREA South CRD AREA CRD Repair Room RHR Valve Room Steam Tunnel AFP-01 AFP-01 AFP-03 AFP-02 AFP-04 AFP-05 AFP-04 AFP-06 AFP-17 RB1 Requires Manual Operator Action within 20 minutes if all high pressure injection systems are lost and a SRV is stuck open.

Torus Room (Bays 6-10)

S.E. Corner Room RCIC Room Radwaste Tank Room (1T-70)

AFP-01 AFP-02 AFP-03 AFP-03 RB2 Reactor Building 786' Reactor Building 812' Reactor Building 833' Reactor Building 855' North RB Chase North RB Stair # 8 South RB Stair # 6 RB Exhaust Fan Room AFP- 07 & 08 AFP- 09 & 10 AFP-11 & 12 AFP-13 AFP-04 AFP-04 AFP-05 AFP-10 RB3 Requires Manual Operator Action within 20 minutes if all high pressure injection systems are lost and a SRV is stuck open.

N.E. Corner Room AFP-01 RB4 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 14 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 15 of 103 Rev. 40 SAFE SHUTDOWN PATHS Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS ADM AOP 913 Page 16 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS ADM AOP 913 Page 17 of 103 Rev. 40 ADM INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN ADMIN BLDG.

FIRST FLOOR AFP-51 SECOND FLOOR AFP-52 THIRD FLOOR AFP-53 SECURITY BLDG.

SCP AFP-54 LOBBY, INGRESS, EGRESS AFP-54 OFFICES, LUNCH ROOM AFP-55 CAS, UPS, REPAIR ROOM AFP-55 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS ADM AOP 913 Page 18 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray X

A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler X

Control Room Lights X

B RHR/CS Room Cooler Offsite Power X

A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power X

A RHR SW and HVAC X

250 VDC power X

B RHR SW and HVAC A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS ADM AOP 913 Page 19 of 103 Rev. 40 Entry Conditions for ADM FIRE in ADM AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Instrument AC if FUEL ZONE level only indication available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS ADM AOP 913 Page 20 of 103 Rev. 40 MANUAL ACTIONS FOR LEVEL INDICATIONS IF the FUEL ZONE LEVEL THEN 1. OPEN 1Y1118 INDICATORS are the only level

2. OPEN 1Y1141 indicators being used to monitor

3. Place 1Y10 in ALTERNATE level and 120 VAC Instrument Control Power Panel 1Y11 is lost IF the FUEL ZONE LEVEL THEN 1. OPEN 1Y2118 INDICATORS are the only level

2. OPEN 1Y2141 indicators being used to monitor

3. Place 1Y20 in ALTERNATE level and 120 VAC Instrument Control Power Panel 1Y21 is lost LEVEL CONTROL:

Use A Core Spray with normal operating instructions and any systems directed by CRS TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use A RHR with normal operating instructions and any systems directed by CRS.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use A RHR with normal operating instructions and any systems directed by CRS.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 21 of 103 Rev. 40 CB2 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN CONTROL Battery Room 1D2 Essential Switchgear 1A4 AFP-23 AFP-24 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 22 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC A Core Spray X

A Diesel and support systems X

B Core Spray B Diesel and support systems A RHR/CS Room Coolers X

Control Room Lights X

B RHR/CS Room Coolers Offsite Power A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power A RHR SW and HVAC X

250 VDC power B RHR SW and HVAC A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 23 of 103 Rev. 40 NOTE Smoke may enter the Control Room via a common supply and exhaust HVAC duct with CB2. Smoke may be detected by sight, smell, or eye/throat irritation.

The following steps should be taken anytime there is a fire in area CB2.

NOTE When Fire exists in the CB2 fire area, the availability of the Remote Shutdown Panel cannot be guaranteed.

IF Smoke is detected in Control

Room, THEN 1. evacuate unnecessary personnel from the Control Room

2. direct the operating crew to don SCBA's

3. Scram the reactor.

MANUAL SCRAM

1. At 1C26A[B] secure Control Room ventilation as follows:

a.

Verify secured CONTROL RM RETURN FAN 1V-RF-30A(B) using handswitch HS6104A (B).

b.

Verify secured CONTROL ROOM SUPPLY FAN 1V-AC-30A(B) using handswitch HS6113A (B).

c.

At 1C26A verify secured COMPUTER ROOM SUPPLY FAN 1V-AC-33 using handswitch HS6120U.

2. Provide ventilation to the Control Room and Control Building HVAC room as follows:

a.

Block open Control Room access Doors 420, 423 and Admin building roof access door 301A.

b.

Direct installation of a smoke ejector on the Admin building roof outside door 301A blowing air into the corridor to provide positive pressurization of the corridor to improve Control Room habitability.

c.

Block open the intake plenum Doors 415 and 416 to evacuate smoke out the main air intake.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 24 of 103 Rev. 40

3. When the fire is extinguished and the smoke has been cleared, and air quality has been verified in the Control Room and Control Building HVAC Room, restore ventilation as follows:

a.

Remove SCBAs.

b.

At 1C26A start CONTROL ROOM SUPPLY FAN 1V-AC-30A by momentarily placing handswitch HS6113A in the START position.

c.

At 1C26A start CONTROL RM RETURN FAN 1V-RF-30A by momentarily placing handswitch HS6104A in the START position.

d.

At 1C26A start COMPUTER ROOM SUPPLY FAN 1V-AC-33 by momentarily placing handswitch HS6120U in the START position.

4. Control room temperatures may be high (above 90°F) due to loss of damper control. The temperatures may be lowered by taking the following manual actions As Soon As Possible.

a.

Above 1V-AC-30B in the CB HVAC Room, close D61-0011 per AOP-913, ATTACHMENT 2.

5. Close doors 415, 416, 420, 423 and 301A.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 25 of 103 Rev. 40 Entry Conditions for CB2 FIRE in CB2 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. If FUEL ZONE indicators are the only indication being used and 1Y11 is lost.

OR

3. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

OR

4. Loss of 1A3 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 26 of 103 Rev. 40 RPV LEVEL-PRESSURE INDICATION As Soon As Possible:

IF FUEL ZONE indicators are the only indication being used and 1Y11 is lost THEN 4. Monitor RPV level/pressure at 1C56 until Instrument AC is restored

5. Open 1Y1128, which is causing the loss of 1Y11

6. Place 1Y10 in ALTERNATE LEVEL CONTROL Use A Core Spray systems with normal operating instructions and any systems directed by CRS TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use A systems with normal operating instructions and any systems directed by CRS

1. A RHR HX Inlet Temp (TE-1945D) may be the only suppression pool water temperature indication available. This point can be monitored at recorder TRS-1945 at Panel 1C21.

Indication is available after flow has started.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB2 AOP 913 Page 27 of 103 Rev. 40 SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use A RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. At 1D42 open 1D4206(MO-1909 may spuriously operate due to the fire).

2. Manually open MO-1909 from the RHR VALVE ROOM.

CAUTION When a 480 or 4160 breaker cubicle is entered for racking in or out a breaker the personnel protection requirements specified in ACP 1408.25 Electrical Safety shall be adhered to.

3. At 1B44 open and rack out 1B4401.

4. At 1C08 close 1B3401.

1A3 Restoration The fire may have damaged the auto tripping and closing function of supply breakers to 1A3

1. Refer to AOP 301 tab 4, "RESTORING POWER TO ESSENTIAL 4160 BUSES". Ensure that the "Bus Transfer Switch" (HS 143-3 on 1C08) is in the MANUAL position before tripping breaker 1A301 to allow 1G-31 output breaker to be manually closed to restore power to 1A3.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 28 of 103 Rev. 40 CB3 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN CONTROL Battery Room 1D1 AFP-23 Essential switchgear 1A3 AFP-24 North Turb. Bldg/Rx. Bldg None Air lock in overhead (Chase above airlock)

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 29 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray X

B Diesel and support systems X

A RHR/CS Room Coolers Control Room Lights X

B RHR/CS Room Coolers X

Offsite Power A RHR Suppression Pool Cooling 1A3 B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 30 of 103 Rev. 40 NOTE Smoke may enter the Control Room via a common supply and exhaust HVAC duct with CB3. Smoke may be detected by sight, smell, or eye/throat irritation.

The following steps should be taken anytime there is a fire in area CB3.

NOTE When Fire exists in the CB3 fire area, the availability of the Remote Shutdown Panel cannot be guaranteed.

IF Smoke is detected in Control

Room, THEN

1. evacuate unnecessary personnel from the Control Room

2. direct the operating crew to don SCBA's

3. Scram the reactor.

MANUAL SCRAM

1. At 1C26A[B] secure Control Room ventilation as follows:

a.

Verify secured CONTROL RM RETURN FAN 1V-RF-30A(B) using handswitch HS6104A (B).

b.

Verify secured CONTROL ROOM SUPPLY FAN 1V-AC-30A(B) using handswitch HS6113A (B).

c.

At 1C26A verify secured COMPUTER ROOM SUPPLY FAN 1V-AC-33 using handswitch HS6120U.

2. Provide ventilation to the Control Room and Control Building HVAC room as follows:

a.

Block open Control Room access Doors 420, 423 and Admin building roof access door 301A.

b.

Direct installation of a smoke ejector on the Admin building roof outside door 301A blowing air into the corridor to provide positive pressurization of the corridor to improve Control Room habitability.

c.

Block open the intake plenum Doors 415 and 416 to evacuate smoke out the main air intake.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 31 of 103 Rev. 40

3. When the fire is extinguished and the smoke has been cleared, and air quality has been verified in the Control Room and Control Building HVAC Room, restore ventilation as follows:

a.

Remove SCBAs.

b.

At 1C26B start CONTROL ROOM SUPPLY FAN 1V-AC-30B by momentarily placing handswitch HS6113B in the START position.

c.

At 1C26B start CONTROL RM RETURN FAN 1V-RF-30B by momentarily placing handswitch HS6104B in the START position.

4. Control room temperatures may be high (above 90°F) due to loss of damper control. The temperatures may be lowered by taking the following manual actions As Soon As Possible.

a.

Above 1V-AC-30A in the CB HVAC Room, close D61-0017 per AOP-913, ATTACHMENT 2.

5. Close doors 415, 416, 420, 423 and 301A.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 32 of 103 Rev. 40 Entry Conditions for CB3 FIRE in CB3 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems OR

2. If FUEL ZONE indicators are the only indication being used and 1Y21 is lost OR

3. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 33 of 103 Rev. 40 RPV LEVEL CONTROL Use B Core Spray with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

250 VDC (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action required to support MO-1909).

The fire may have caused a loss of 250 VDC.

1. Open breaker 1D40 ckt. 3. (250 VDC Battery Charger 1D43)

2. Open breaker 1D40 ckt. 6. (uninterruptible AC Inverter 1D45 supply)

3. Place 1D44 in service per OI-388.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

NOTE The following steps, if implemented, will override the Group 4 automatic closure of MO1908 and/or MO1909.

1. If MO1909 is closed with an isolation signal present due to a fire, override and open MO1909 as follows:

a.

Obtain 14 AWG jumper from the CRS desk or from the warehouse, stock item 100-4864.

b.

At 1C42, lift and tape lead (either field or panel side) at terminal BB44.

c.

At 1C42, install 14 AWG jumper from terminal BB40 to terminal BB42.

d.

When MO 1909 indicates OPEN at 1C03, remove jumper from terminal BB40 and BB42.

2. If MO 1908 is closed with an isolation signal present due to a fire, override and open MO 1908 as follows:

a.

Obtain 14 AWG jumper from the CRS desk or from warehouse, stock item 100-4864.

b.

At 1C41, lift and tape lead (either field or panel side) at terminal BB44.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB3 AOP 913 Page 34 of 103 Rev. 40

c.

At 1C41, install 14 AWG jumper from terminal BB40 to terminal BB42.

d.

When MO 1908 indicates OPEN at 1C03, remove jumper from terminal BB40 and BB42.

CAUTION Perform Manual Actions 3 to 11 listed below in the order listed. These steps are performed to backfeed 1B34.

3. At 1C388 place keylocked handswitch 52-4401/SS in the "EMERG" position.

4. At 1C388 place keylocked handswitch 43-206 in the "EMERG." position.

5. At 1C388 use handswitch 52-4401E/CS to open breaker 52-4401.

6. At 1C390 place keylocked handswitch 52-303/SS in the "EMERG" position.

7. At 1C390 place keylocked handswitch HS 2011B in the "EMERG" position.

8. At 1C390 use handswitch 52-3401E/CS to open breaker 1B3401.

9. At 1B34 open breaker 1B3400.

10. At 1C390, use handswitch 52-3401E/CS to close breaker 1B3401.

11. At 1C388 use handswitch 52-4401E/CS to close 1B4401.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB4 AOP 913 Page 35 of 103 Rev. 40 CB4 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN CONTROL Battery Room Corridor AFP-23 Battery Room 1D4 AFP-23 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB4 AOP 913 Page 36 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler X

B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC A Core Spray A Diesel and support systems X

B Core Spray B Diesel and support systems A RHR/CS Room Cooler X

Control Room Lights X

B RHR/CS Room Cooler Offsite Power A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power X

A RHR SW and HVAC X

250 VDC power X

B RHR SW and HVAC A or B SBGT and Stack Exh. fans X

NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB4 AOP 913 Page 37 of 103 Rev. 40 Entry Conditions for CB4 FIRE in CB4 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

OR

3. Loss of 1A3 OR

4. Loss of 1A3 switchgear cooling components which will cause high switchgear temperatures.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB4 AOP 913 Page 38 of 103 Rev. 40 Ventilation

1. As soon as possible verify cooling is available in the 1A3 Switchgear Room by observing air flow from the room supply ducts.

NOTE By analysis results, the following steps must be completed in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if cooling is not available.

2. If flow is not available and re-entry into the Battery Room Corridor or cable spreading room is not possible, then open Door 122 between the 1A3 Switchgear room and the Turbine Building and set up smoke ejectors to provide air movement into the switchgear room.

3. If flow is not available and re-entry into the Battery Room Corridor and cable spreading room is possible, then open Door 122 between 1A3 Switchgear Room and the Turbine Building and verify the following dampers are open:

1V-FD-317 Between Battery Room Corridor and 1D1 Battery Room. Access is from the Battery Room.

NOTE CARDOX is required to be tagged out per OI-513 before entry into the cable spreading room.

1V-FD-300 Between Battery Room Corridor and the Cable Spreading Room. Access is from the Cable Spreading Room in the North HVAC chase. (Knife and Prybar may be necessary to access damper. Tools are staged in the EOP Toolbox.)

Ventilation (actions required as soon as possible after the CB4 fire)

1. Following the fire, verify cooling is available to the 1D1 Battery Room.

2. If the Battery Room exhaust is not available, provide an exhaust path to the Turbine building or the Admin building until the HVAC system is restored.

RPV LEVEL CONTROL Use RCIC system with normal operating instructions and any systems directed by CRS.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS CB4 AOP 913 Page 39 of 103 Rev. 40 TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use A RHR systems with normal operating instructions and any systems directed by CRS.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use A RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. At 1D42 open breaker 1D4206 for MO1909.

2. In the RHR Valve Room, manually open MO-1909.

1A3 Restoration The fire may have damaged the auto tripping and closing function of supply breakers to 1A3.

1. Refer to AOP 301 tab 4, "RESTORING POWER TO ESSENTIAL 4160 BUSES".

2. Once 1A3 is repowered and 1B32 is restored, return 1D20 to service with 1D120 as charger. Refer to OI-302, section 6.4.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS IS1 AOP 913 Page 40 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS IS1 AOP 913 Page 41 of 103 Rev. 40 IS1 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN INTAKE A RWS PUMPS AFP - 31 INTAKE A RWS SCREEN WASH AREA AFP - 32 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS IS1 AOP 913 Page 42 of 103 Rev. 40 Intake Structure (A RWS Pumps, A RWS Screen Wash Area)

The Systems credited for a safe shutdown in this tab include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray X

B Diesel and support systems A RHR/CS Room Coolers Control Room Lights X

B RHR/CS Room Coolers X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS IS2 AOP 913 Page 43 of 103 Rev. 40 IS2 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN INTAKE B RWS PUMPS AFP - 31 INTAKE B RWS SCREEN WASH AREA AFP - 32 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS IS2 AOP 913 Page 44 of 103 Rev. 40 Intake Structure (B RWS PUMPS, B RWS Screen Wash Area)

The Systems credited for a safe shutdown in this tab include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray X

A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler X

Control Room Lights X

B RHR/CS Room Cooler Offsite Power X

A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power X

A RHR SW and HVAC X

250 VDC power X

B RHR SW and HVAC A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH1 AOP 913 Page 45 of 103 Rev. 40 PH1 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN PUMPHOUSE B RHR/ESW SERVICE WATER PUMPS RWS MAKEUP VALVE AREA 727 South end of Pumphouse Elevation 747 AFP-28 AFP-30 AFP-30 PUMPHOUSE North end of Pumphouse Elevation 747' AFP-30 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH1 AOP 913 Page 46 of 103 Rev. 40 Pumphouse (B RHR/ESW Service Water Pumps, RWS makeup valve area 727, or South end Pumphouse Elevation 747)

The systems credited for a safe shutdown in this subsection includes as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler X

B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Coolers X

Control Room Lights X

B RHR/CS Room Coolers Offsite Power X

A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 (NO 1B46)

X RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power X

A RHR SW and HVAC X

250 VDC power X

B RHR SW and HVAC A or B SBGT and Stack Exh. fans X

NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH1 AOP 913 Page 47 of 103 Rev. 40 North end of the Pumphouse Elevation 747 The systems credited for safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC B RPV/DW/Torus Instrumentation B ESW and HVAC X

SRVs ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler X

B CB HVAC X

A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans X

NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH1 AOP 913 Page 48 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH2 AOP 913 Page 49 of 103 Rev. 40 PH2 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN PUMPHOUSE A RHR/ESW SERVICE WATER PUMPS AFP - 28 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS PH2 AOP 913 Page 50 of 103 Rev. 40 Pumphouse (A RHR/ESW Service Water Pumps)

The systems credited for safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs ESW/RHRSW Discharge X

HPCI and Room Cooler X

A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans X

NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS YARD AOP 913 Page 51 of 103 Rev. 40 YARD INFORMATION THIS SUBSECTION COVERS THE FOLLOWING FIRE LOCATIONS BUILDING LOCATION AREA FIRE PLAN NONE DIV I MANHOLE None NONE DIV II MANHOLE None NONE DIESEL FUEL OIL SUPPLY None NONE A COOLING TOWER AFP-73 NONE B COOLING TOWER AFP-73 NONE OFF GAS STACK None NONE SWITCH YARD AFP-74 NONE STANDBY TRANSFORMER 1X4 AFP-70 WAREHOUSE EAST WAREHOUSE AFP-67 WAREHOUSE WEST WAREHOUSE AFP-68 Electrical Equipment Building ISFSI AFP-79 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS YARD AOP 913 Page 52 of 103 Rev. 40 Div. 1 Manhole (manholes from near turbine building to intake structure)

The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

ADS A CB HVAC HPCI and Room Cooler B CB HVAC X

RCIC and Room Cooler A Instrument AC X

A LPCI B Instrument AC X

B LPCI X

A Diesel and support systems A Core Spray B Diesel and support systems B Core Spray Control Room Lights X

A RHR/CS Room Cooler Offsite Power X

B RHR/CS Room Cooler X

1A3 X

A RHR Suppression Pool Cooling 1A4 X

B RHR Suppression Pool Cooling X

LPCI Swing Bus X

RHR Drain to Radwaste X

A 125 VDC power X

A RHR SDC B 125 VDC power X

B RHR SDC X

250 VDC power X

A RHR SW and HVAC A or B SBGT and Stack Exh. fans B RHR SW and HVAC X

NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS YARD AOP 913 Page 53 of 103 Rev. 40 Div. 2 Manhole (manholes from near turbine building to intake structure)

The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

ASC A CB HVAC X

HPCI and Room Cooler B CB HVAC RCIC and Room Cooler A Instrument AC X

A LPCI X

B Instrument AC X

B LPCI A Diesel and support systems A Core Spray B Diesel and support systems B Core Spray Control Room Lights X

A RHR/CS Room Cooler X

Offsite Power X

B RHR/CS Room Cooler 1A3 X

A RHR Suppression Pool Cooling X

1A4 X

B RHR Suppression Pool Cooling LPCI Swing Bus X

RHR Drain to Radwaste X

A 125 VDC power X

A RHR SDC X

B 125 VDC power X

B RHR SDC 250 VDC power X

A RHR SW and HVAC X

A or B SBGT and Stack Exh. fans B RHR SW and HVAC NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS YARD AOP 913 Page 54 of 103 Rev. 40 Diesel Fuel Oil Supply The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler B CB HVAC A LPCI X

A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler Offsite Power X

A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power X

A RHR SW and HVAC X

250 VDC power X

B RHR SW and HVAC A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS YARD AOP 913 Page 55 of 103 Rev. 40 A Cooling Tower, B Cooling Tower, Off Gas Stack, Switchyard, Standby Transformer, East Warehouse, and West Warehouse and ISFSI The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

ASC A CB HVAC X

HPCI and Room Cooler B CB HVAC RCIC and Room Cooler A Instrument AC X

A LPCI X

B Instrument AC X

B LPCI A Diesel and support systems X

A Core Spray B Diesel and support systems X

B Core Spray Control Room Lights X

A RHR/CS Room Coolers X

Offsite Power B RHR/CS Room Coolers 1A3 X

A RHR Suppression Pool Cooling X

1A4 X

B RHR Suppression Pool Cooling LPCI Swing Bus X

RHR Drain to Radwaste X

A 125 VDC power X

A RHR SDC X

B 125 VDC power X

B RHR SDC 250 VDC power X

A RHR SW and HVAC X

A or B SBGT and Stack Exh. fans B RHR SW and HVAC NOTE For the determination of EALs, the fifteen (15) minute clock begins when the control room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 56 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 57 of 103 Rev. 40 TB1 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN Turbine Building Turbine Building Interior AFP 14-22 NONE Startup Transformer 1X3 AFP - 71 Main Transformers 1X1 AFP - 69 Auxiliary Transformer 1X2 AFP - 72 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 58 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation X

A ESW and HVAC B RPV/DW/Torus Instrumentation B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI X

B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 59 of 103 Rev. 40 Entry Conditions for TB1 FIRE in TB1 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

OR

3. 1P-99B will not manually start from 1C06 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 60 of 103 Rev. 40 NOTE Primary generator protective relaying may be unavailable due to fire damage.

Emergency Diesel Generators 1G21 and 1G31 may also be unavailable. To prevent grid instability and potential loss of Offsite and Onsite AC power systems and subsequent loss of Safe Shutdown capability due to motoring of the Turbine Generator, it may be necessary to manually trip switchyard "H" and "I" breakers.

LEVEL CONTROL (30 minute action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. At 1B44 open breaker 1B4493 for MO-1905.

2. In RHR Valve Room manually open MO1905.

3. The fire may have damaged the logic for MO-2010. This valve may have to be repositioned by using its handwheel.

4. If B ESW pump will not run, B Core Spray and RHR room cooling is lost.

Following vessel reflood, secure one RHR pump and maintain RPV level using one pump. Following ESW restoration and room cooler startup, both RHR pumps may be operated as necessary.

TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. Due to fire damage, If B ESW Pump 1P-99B will not start from 1C-06, perform the following:

a.

Open 1B4207, 1P-99B ESW Pump.

b.

Cut cables 2B4207-E and 2B4207-P.

c.

Replace the blown control power fuses at 1B4207.

d.

Close 1B4207, 1P-99B ESW Pump.

e.

Start 1P-99B from 1C06.

2. At 1B44 open breaker 1B4434 RHR HX 1E-201B SERVICE WTR OUTLET VLV MO-1947.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 61 of 103 Rev. 40

3. In NWCR manually open MO1947, RHR HX 1E-201B SERVICE WTR OUTLET Isolation.

4. At 1C03 place handswitch HS1903B in the " Manual Override" position and then momentarily place HS1903C in the "Manual" position. This will allow MO1932 and MO1934 to be operated by their handswitches.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 62 of 103 Rev. 40 CONTROL ROOM VENTILATION If control room temps are high (above 90°F) due to loss of ESW flow, the temperatures may be lowered by taking the following manual actions.

1. Block open Admin Bldg door on third floor which connects with the outside area (Admin Bldg roof) by the elevator.

2. Block open Control Room access doors 420 and 423.

3. Block open the intake plenum door 416.

4. Block open the access door on the lower east side of 1VAC30B.

NOTE Expect a battery room Low Flow Alarm

5. Above 1VAC030B in the CB HVAC room open D61-0011 per AOP 913, ATTACHMENT 2.

6. At 1C-26B start 1VAC30B.

7. Verify 1VEF-30B is off.

8. Block open all west switchgear (1A4) room doors.

9. Block open all battery room doors.

10. Block open the door from the battery room corridor to the Administration building.

11. Block open outside double doors in Administration building by elevator.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. If MO-1909 will not operate then:

a.

Open breaker 1D42 ckt 1 (B Recirc M/G Set Emerg. Aux Oil Pump 1P-204B).

b.

Open breaker 1D42 ckt 3 (Steam Line Drain Outbd. Isolation MO-4424).

c.

Open breaker 1D40 ckt 8 (250 VDC MCC 1D41).

d.

Open breaker 1D40 ckt 4 (Emergency Seal Oil Pump 1D-93 Starter).

e.

Open breaker 1D40 ckt 7 (Turbine Emergency Bearing Oil Pump 1P-40 Starter).

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 63 of 103 Rev. 40

f.

Place 1D44 in service per OI-388.

g.

Close/verify closed breaker 1D40 ckt 5 (250 VDC MCC 1D42).

2. Before starting RHR pumps in shutdown cooling mode, block open contact (1-2) of relays E11-K19B and E11-K22B at 1C33.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS TB1 AOP 913 Page 64 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 65 of 103 Rev. 40 RB1 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN REACTOR BUILDING Torus (Bays 1-5 AND 11-16)

AFP - 01 NW Corner Room AFP - 01 HPCI Room AFP - 03 SW Corner Room AFP - 02 North CRD Area AFP - 04 South CRD Area AFP - 05 CRD Repair Boom AFP - 04 RHR Valve Room AFP - 06 Steam Tunnel AFP - 17 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 66 of 103 Rev. 40 The systems credited for safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC X

RPV Isolation (Group Isolations)

X B RWS and HVAC A RPV/DW/Torus Instrumentation X

A ESW and HVAC X

B RPV/DW/Torus Instrumentation B ESW and HVAC SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC X

RCIC and Room Cooler B CB HVAC A LPCI A Instrument AC X

B LPCI B Instrument AC A Core Spray X

A Diesel and support systems X

B Core Spray B Diesel and support systems A RHR/CS Room Cooler X

Control Room Lights X

B RHR/CS Room Cooler Offsite Power A RHR Suppression Pool Cooling X

1A3 X

B RHR Suppression Pool Cooling 1A4 RHR Drain to Radwaste X

LPCI Swing Bus A RHR SDC X

A 125 VDC power X

B RHR SDC B 125 VDC power A RHR SW and HVAC X

250 VDC power B RHR SW and HVAC A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 67 of 103 Rev. 40 Entry Conditions for Subsection RB1 FIRE in RB1 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Instrument AC if FUEL ZONE level only indication available.

OR

3. Indications of a relief valve open. (Torus Level oscillations or Reactor Level decrease.)

OR

4. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG). (No High Drywell Pressure Indication may be available)

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 68 of 103 Rev. 40 RPV/TORUS LEVEL-PRESSURE INDICATION IF the FUEL ZONE LEVEL INDICATORS are being used to monitor level and 120 VAC Instrument Control Power Panel is lost THEN 1. Monitor RPV Level and pressure at 1C56

2. Restore Control Room Level and Pressure indication Open the following breakers 1Y1108 1Y1118 1Y1134 1Y1141 1Y1140 RESTORE POWER to 1Y11 Close 1B3216 Place 1Y10 in ALTERNATE Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 69 of 103 Rev. 40 LEVEL CONTROL (20 minute action)

NOTE The CRS should direct use of any systems available.

Use A Core Spray with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. When Reactor Pressure is less than 450 PSIG, enable the A Core Spray inject valves per AOP 913 Attachment 1. AOP 913, Attachment 1 package is located in the EOP box in the Control Room Backpanel area.

a.

Open MO-2117 using controls at Panel 1C03. Verify MO-2117 is open.

b.

Inject with Core Spray, maintain RPV Level 170" to 211".

NOTE Fire damage may interfere with the closing of MO2238.

2. Isolate the HPCI Steam Supply Isolation as follows:

to close MO2238:

a.

At 1C-32, place HS2229A in the "Override" position.

b.

At 1C-03 close MO2238 using HS2238.

TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use A RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. A RHR HX Inlet Temp (TE-1945D) may be the only suppression pool water temperature indication available. This point can be monitored at Recorder TRS-1945 at Panel 1C21.

Indication is available after flow is started.

2. At 1B34 open 1B3436 breaker for MO2046.

3. At 1B34 open 1B3423 breaker for MO2005.

4. At 1B34 open 1B3425 breaker for MO2007.

5. At Torus Bay 14, close V-19-0048. This valve will be used as an isolation for spuriously operating valves.

6. At Torus Bay 16, manually open/verify open MO-2005.

7. At Torus Bay 16, manually open/verify open MO-2007.

8. In the HPCI room manually open/verify open MO-2046.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 70 of 103 Rev. 40 ISOLATION OF THE SDV (9 HOUR ACTION)

9. Unlock and close V-18-02 located at the Scram Air header (757' South Rx. Bldg)

10. Remove pipe cap and open V-18-1636 (located above the CRD Hatch) or remove pipe cap and open V-18-1638 (located at bay 10 Torus Catwalk Mezz.) to isolate the scram discharge volume.

MANUAL ACTIONS NEEDED FOR SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use A RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. At 1B34 open breaker 1B3423 for MO2005.

2. At Torus Bay 16, manually close/verify closed MO-2005. This valve could spuriously operate

3. At 1B4 Open breaker 1B403 to deenergize 1B44 (1B44 is in the fire area).

4. In the NW Corner room, manually close/verify closed MO-1920. This valve could spuriously operate.

5. In the NW Corner room, manually close/verify closed MO-1912. This valve could spuriously operate.

NOTE Request Maintenance to take the following steps to close MO-4627 and allow shutdown cooling operation. As an option, if accessible a Drywell entry could be performed to close MO4627.

Following steps are one method to ensure that the valve is closed.

6. Actions for closing MO-4627.

a.

Obtain 308 and 914 cables from warehouse, stock items 100-4862 and 100-4863.

b.

At 1B34 remove control power fuse in 1B3491.

c.

After the fire is out and access to RB 757' is restored, determinate cable 1B3491-A at 1B3491 and at 1JX105A port #9.

d.

Determinate cable 1B3491-C at 1B3491 and at 1JX105A port #3.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 71 of 103 Rev. 40

e.

Route new (the length should be minimized) 308 cable for 1B3491-A and new 914 cable for 1B3491-C from 1JX105A to 1B3491.

f.

Replace MO-4627 control power fuses.

g.

If power is not available to MCC 1B34A, perform step 7, otherwise perform step 8.

7. Actions to restore power to MCC 1B34A:

a.

Obtain 14 AWG jumper from the CRS desk or from warehouse, stock item 100-4864.

b.

Isolate MCC 1B34A and MCC 1B44A.

c.

Pull control power fuses for 1B3401.

d.

Place HS 52-3401/CS on 1C08 in TRIP and PULL TO LOCK.

e.

Determinate cable 5B3444-A in 1B34A.

f.

Determinate cable 1B3401-E in 1B34A at TB-1 terminals 4 and 5.

Install a 14 AWG jumper from TB-1 terminal 4 to TB-1 terminal 5.

g.

Replace control Power fuses for 1B3401.

h.

Place HS 52-3401/CS on 1C08 in normal.

8. At 1C04 close MO-4627 using HS4627.

9. At 1C-32 block open contacts (1-2) of relays E11-K19A and E11-K22A.

This will block the spurious actions of suction path trips of the RHR pumps.

The RHR pumps may be started in shutdown cooling mode when needed.

10. At 1D42 open breaker 1D4206 for MO1909.

11. In the RHR Valve Room, manually open MO-1909.

NOTE The following steps if implemented will override the Group 4 automatic closure of MO1908.

12. If MO 1908 is closed with an isolation signal present due to a fire, override and open MO 1908 as follows:

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 72 of 103 Rev. 40

a.

Obtain 14 AWG jumper from the CRS desk or the warehouse, stock item 100-4864.

b.

At 1C41, lift and tape lead (either field or panel side ) at terminal BB44.

c.

At 1C41, install 14 AWG jumper from terminal BB40 to terminal BB42.

d.

When MO 1908 indicates OPEN at 1C03, remove jumper from terminal BB40 and BB42.

13. At 1B34 open breaker 1B3494 for MO2004 and 1B3493 for MO2003.

14. In RHR Valve Room, manually open/verify open MO-2004.

15. In RHR Valve Room, manually open/verify open MO-2003.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB1 AOP 913 Page 73 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 74 of 103 Rev. 40 RB2 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN REACTOR BUILDING Torus Room (Bays 6 - 10)

AFP - 01 SE Corner Room AFP - 02 RCIC Room AFP - 03 Radwaste Tank Room (1T-70)

AFP - 03 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 75 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray X

B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 76 of 103 Rev. 40 Entry Conditions for RB2 FIRE in RB2 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 77 of 103 Rev. 40 Level Control Use B Core Spray with normal operating instructions. The CRS should direct use of any systems available.

TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use B RHR with normal operating instructions. The CRS should direct use of any systems available.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

NOTE MO-2011 and MO-2016 may lose indication. Both valves will be closed and will not be capable of being opened.

CAUTION MO-1909 will open with Reactor Pressure above 135 PSIG and MO-1904 will open with Reactor Pressure above 450 PSIG when the following steps are taken.

1. At transfer station 1C390, Unlock and open the panel access door.

2. Place keylocked handswitch HS 1908A on 1C390 in the "EMERG" position.

Place handswitch HS 1908B on 1C390 in the "OPEN" position.

3. At transfer station 1C388, place keylocked handswitch HS 1909A in the "EMERG" position.

4. Place handswitch HS 1909B on 1C388 in the "OPEN" position.

5. At transfer station 1C388, place keylocked handswitch HS 1904A in the "EMERG" position.

6. Place handswitch HS 1905C on 1C388 in the "OPEN" position.

7. Place handswitch HS 1904B on 1C388 in the "OPEN" position.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 78 of 103 Rev. 40

8. IF power to MO-1909 is

lost, THEN

1) Open 1D40 ckt 8 (250 VDC MCC 1D41) to preserve 250 VDC.

2) Place 1D44 in service per OI-388.

3) Close/ verify closed 1D40 ckt 5 (250 VDC MCC 1D42).

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 79 of 103 Rev. 40 RB3 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN REACTOR BUILDING Elevation 786' AFP-07 & 08 Elevation 812' AFP-09 & 10 Elevation 834' AFP-11 & 12 Elevation 855' AFP-13 North RB Chase AFP-04 North RB Stair #8 AFP-04 South RB Stair #6 AFP-05 RB Exhaust Fan Room AFP-10 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 80 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI X

B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 81 of 103 Rev. 40 Entry Conditions for RB3 FIRE in RB3 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Indications of a relief valve open. Torus Level oscillations maybe the only indication available.

OR

3. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

OR

4. Loss of Instrument AC if FUEL ZONE level only indication available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 82 of 103 Rev. 40 RPV/TORUS LEVEL-PRESSURE INDICATION IF the FUEL ZONE LEVEL INDICATORS are being used to monitor level and 120 VAC Instrument Control Power Panel is lost THEN 1. Monitor RPV Level and pressure at 1C55.

2. Restore Control Room Level and Pressure indication.

Open the following breakers at 1Y21 1Y2112 1Y2116 1Y2122 1Y2126 RESTORE POWER to 1Y21 Place 1Y20 in ALTERNATE LEVEL CONTROL (20 minute action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. At 1C04 close MO2700 using HS2700.

2. At 1C04 close MO2400 using HS2400.

3. At 1B44, open breaker 1B4493 prior to manually operating MO-1905.

4. In the RHR Valve Room, open MO-1905 manually.

5. If B ESW pump will not run, ESW and RHR room cooling is lost. Following vessel reflood, secure one RHR pump and maintain RPV level using one pump. Following ESW restoration and room cooler startup, both RHR pumps may be operated as necessary Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 83 of 103 Rev. 40 CONTROL ROOM VENTILATION If control room temps are high (above 90°F) due to loss of ESW flow, the temperatures may be lowered by taking the following manual actions.

1. Block open Admin Bldg door on third floor which connects with the outside area (Admin Bldg roof) by the elevator.

2. Block open Control Room access doors 420 and 423.

3. Block open the intake plenum door 416.

4. Block open the access door on the lower east side of 1VAC30B.

NOTE Expect a battery room Low Flow Alarm.

5. Above 1VAC-30B in CB HVAC room open D61-0011 per AOP 913, ATTACHMENT 2.

6. At 1C-26 start 1VAC-30B.

7. Verify 1VEF-30B is off.

8. Block open all west switchgear (1A4) room doors.

9. Block open all battery room doors.

10. Block open the door from the battery room corridor to the Administration building.

11. Block open outside doors in Administration building by elevator.

TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available

1. In the Torus Bay 14 close V-19-48. This will be used as an isolation for spuriously operating valves.

2. Start 1P-99B ESW Pump as follows:

a.

At 1C-388 place HS 4928C in the Emergency position.

b.

At 1C-388, place the "Transfer Last" Switch (HS43-206) to Emergency position.

c.

At 1C-388 start ESW pump by using HS-4928D.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 84 of 103 Rev. 40 SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

NOTE The CRS should direct use of any systems available.

Use B RHR with manual actions listed and normal operating instructions.

1. At 1B3, open breaker 1B303.

2. MO-2011 must be manually operated closed.

3. MO-2016 must be manually operated closed.

4. At 1D42, open breakers:

a.

1D42 ckt 1 (B Recirc M/G Set Emerg Aux Oil Pump 1P-204B).

b.

1D42 ckt 2 (A Recirc M/G Set Emerg Aux Oil Pump 1P-204A).

c.

1D42 ckt 4 (RWCU Suction Outbd. Isolation MO-2701).

5. Close/verify closed breaker 1D42 ckt 6 (RHR loop B Shutdown Cooling Outbd. Suction MO-1909).

6. Place 1D44 in service per OI-388.

7. At 1D40 close/verify closed breaker. 1D40 ckt 5 (250 VDC MCC 1D42)

8. If MO-1909 will not operate after 250 VDC restoration, portions of the RHR logic may have been destroyed by the fire. At 1Y11 open breakers associated with all loads and at 1Y10 Place 1Y10 in ALTERNATE. Then at 1Y11 close 1Y1104. This will allow MO-1909 to be operated.

9. Before starting the B and D RHR pumps in Shutdown Cooling mode and prior to closing MO1913 and MO1921, block open contacts 9-10 and 11-12 of relay E11A-K15A. This will allow the pumps to be started.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 85 of 103 Rev. 40 NOTE Request Maintenance to take the following steps to close MO-4628 and allow shutdown cooling operation. As an option, could perform a Drywell entry to close MO4628. Following steps are one method to ensure that the valve is closed.

10. Actions for closing MO-4628.

a.

Open breakers 1B4401 and 1B4491 at 1B44 and 1B44A.

b.

Disconnect the power cable tie between 1B34A and 1B44A.

c.

Place Transfer Switch 52-4401/SS at 1C388 in 'EMERG'.

NOTE Breaker 1B4401 cannot be closed manually or electrically without a supply of 125 VDC power to 1B4401.

d.

Verify/replace control power fuses EF22 and EF23 (10A)at 1C422B.

e.

Close 1B4401 using Handswitch 52-4401E/CS at 1C388.

f.

Verify 1B44A voltage has been established.

g.

Close 1B4491 and close MO4628 from the Main Control room at 1C04.

h. Following valve repositioning, 1B4401, 1B4402, and 1B4491 can be tripped.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB3 AOP 913 Page 86 of 103 Rev. 40 NOTE Request Maintenance to take the following steps to open MO-1908 and allow shutdown cooling operation. As an option, could perform a Drywell entry to open MO1908.

Following steps are one method to ensure that the valve is open.

11. Actions for closing MO1908.

a.

Obtain 308 and 914 cables from warehouse, stock items 100-4862 and 100-4863.

b.

Isolate 1B34 by opening 1B4401, 1B4402, and 1B303 at 1B3 and 1B44.

c.

Pull the new 308 and 914 cable for MO1908 from 1B44 to penetration 1JX105A. This cable does not have to be routed in a raceway.

d.

Open the DARMATT penetration enclosure. Identify MO1908 power and control cables to be spliced. Cut existing damaged power and control cables. Splice the 308 cable to the power cable and the 914 cable to the control cable at the penetration conductors.

e.

Identify a suitable MOV starter to be utilized for MO1908 at 1B44.

Terminate temporary power and control cables for MO1908 at the spare starter. Verify correct thermal overloads.

f.

Close spare breaker and open MO1908 using spare starter controls.

g.

Following valve repositioning, open breaker associated with MO1908.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB4 AOP 913 Page 87 of 103 Rev. 40 RB4 INFORMATION This subsection covers the following fire locations BUILDING LOCATION AREA FIRE PLAN Reactor Building NE Corner Room AFP-01 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB4 AOP 913 Page 88 of 103 Rev. 40 The Systems credited for a safe shutdown in this subsection include as a minimum:

Scram X

A RWS and HVAC RPV Isolation (Group Isolations)

X B RWS and HVAC X

A RPV/DW/Torus Instrumentation A ESW and HVAC B RPV/DW/Torus Instrumentation X

B ESW and HVAC X

SRVs X

ESW/RHRSW Discharge X

HPCI and Room Cooler A CB HVAC RCIC and Room Cooler B CB HVAC X

A LPCI A Instrument AC X

B LPCI X

B Instrument AC X

A Core Spray A Diesel and support systems B Core Spray B Diesel and support systems A RHR/CS Room Cooler Control Room Lights X

B RHR/CS Room Cooler X

Offsite Power X

A RHR Suppression Pool Cooling 1A3 X

B RHR Suppression Pool Cooling X

1A4 X

RHR Drain to Radwaste X

LPCI Swing Bus X

A RHR SDC A 125 VDC power X

B RHR SDC X

B 125 VDC power X

A RHR SW and HVAC 250 VDC power X

B RHR SW and HVAC X

A or B SBGT and Stack Exh. fans NOTE For the determination of EALs, the fifteen (15) minute clock begins when the Control Room is notified or the alarm is verified.

NOTE This procedure identifies the equipment which will be available even after maximum fire damage. All other equipment may be used if available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB4 AOP 913 Page 89 of 103 Rev. 40 Entry Conditions for RB4 FIRE in RB4 AND

1. Unable to maintain Reactor Vessel above 170" with high pressure systems.

OR

2. Loss of Drywell cooling components which will cause high drywell pressure (2 PSIG).

OR

3. B ESW will not start with use of HS-4928B at 1C-06.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB4 AOP 913 Page 90 of 103 Rev. 40 ESW Pump Operation (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Spurious operation of Aux contact CW-K2G24 could prevent auto-start of the B ESW pump. If the fire has only damaged Aux Contact CW-K2G24 the pump can still be started using HS-4928B at 1C06. If the handswitch will not start the pump, request repairs listed in the Torus Cooling section of RB4. These repairs are required to be completed within 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to support suppression pool cooling.

LEVEL CONTROL Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. If B ESW pump will not run using the above manual action, B Core Spray and RHR room cooling is lost. Following vessel reflood, secure one RHR pump and maintain RPV level using one pump. Following ESW restoration and room cooler startup, both RHR pumps may be operated as necessary.

Perform the repairs stated in the Torus Cooling section of RB4.

CONTROL ROOM VENTILATION If control room temps are high (above 90°F) due to loss of ESW flow, the temperatures may be lowered by taking the following manual actions.

1. Block open Admin Bldg door on third floor which connects with the outside area (Admin Bldg roof) by the elevator.

2. Block open Control Room access doors 420 and 423.

3. Block open the intake plenum door 416.

4. Block open the access door on the lower east side of 1VAC30B.

NOTE Expect a battery room Low Flow Alarm.

5. Above 1VAC-30B in CB HVAC room open D61-0011 per AOP 913, ATTACHMENT 2.

6. AT 1C-26B Start 1VAC-30B.

7. Verify 1VEF-30B is off.

8. Block open all west switchgear (1A4) room doors.

9. Block open all battery room doors.

10. Block open the door from the battery room corridor to the Administration building.

11. Block open outside doors in Administration building by elevator.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS RB4 AOP 913 Page 91 of 103 Rev. 40 TORUS COOLING (2.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months action)

Use B RHR with manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

1. Perform the following repairs if the B ESW will not run following Manual Action in "ESW Pump Operation" section. At the breaker cut cable 2B4207-E leads (from 1B4207 to 1C118) and replace control power fuse 6A at 1B4207.

SHUTDOWN COOLING (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action)

Use B RHR and normal operating instructions. The CRS should direct use of any systems available.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE SAFE SHUTDOWN PATHS AOP 913 Page 92 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE OFFSITE ASSISTANCE AOP 913 Page 93 of 103 Rev. 40 OFFSITE ASSISTANCE Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE OFFSITE ASSISTANCE AOP 913 Page 94 of 103 Rev. 40

1. Direct the CAS Operator to contact the Palo and/or Cedar Rapids Fire Departments and request fire fighting assistance.

a.

Request rescue squad assistance if necessary for personnel injuries.

2. Carry out Emergency Plan Implementing Procedures when calling for offsite fire assistance.

3. Notify Security to allow entry of offsite fire fighters and to have the following equipment staged for issue:

Thermoluminescent dosimeters, Self Reading Pocket dosimeters or Electronic dosimeters

4. Send a Health Physics technician with portable radiation and contamination monitoring equipment to meet and guide the offsite fire fighters.

NOTE PALO FIRE DEPARTMENT ---------- 911 CEDAR RAPIDS FIRE DEPARTMENT ---------- 911 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 1 AOP 913 Page 95 of 103 Rev. 40 AOP 913 Page 1 of 3 RPV LEVEL CONTROL FOR A FIRE IN FIRE AREA RB-1 PURPOSE:

To permit RPV level control for a fire in Fire Area RB-1. To defeat the 450 PSIG interlock for the A Core Spray Inject Valve MO-2117. This interlock could be damaged as a result of the RB-1 fire. To maintain the reactor within analyzed conditions, the interlock must be jumpered within 20 minutes of the onset of the fire event.

LOCATION(S):

1C43, 1C03 EQUIPMENT REQUIRED: 14 AWG jumper with alligator clip ends. This jumper is in the AOP 913 Attachment 1 package located in the EOP box in the Control Room Back panel area.

INSTRUCTIONS: Use A Core Spray with the manual actions listed and normal operating instructions. The CRS should direct use of any systems available.

NOTE The relay within Panel 1C43 is identified as E21A-K20A and as wiring device "BF".

A diagram showing the rear view of the relay and terminations is provided on the inside of the panel door.

When reactor pressure is less than 450 PSIG, enable the A Core Spray inject valves. At Panel 1C43, install a jumper across contact E21A-K20A (1 - 7).

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 1 AOP 913 Page 96 of 103 Rev. 40 AOP 913 Page 2 of 3 RPV LEVEL CONTROL FOR A FIRE IN FIRE AREA RB-1 RESTORATION: A Core Spray 450 PSIG interlock:

1. Have Electrical Maintenance verify that the circuit associated with relay E21A-K20A is free from fire damage or repair the circuit so that it is free from fire damage.

2. Verify circuit operability by simulating a 450 PSIG signal and verifying that relay E21A-K20A operates as required.

3. Following circuit verification, remove jumper installed between relay contact E21A-K20A (1 - 7).

REFERENCES

1. AL-E96-020

2. APED-E41-006<1>

3. APED-H11-037

4. APED-H11-077

5. BECH-E121<005>

6. BECH-E840 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 1 AOP 913 Page 97 of 103 Rev. 40 Page 3 of 3 RPV LEVEL CONTROL FOR A FIRE IN FIRE AREA RB-1 13 37 36 27 28 19 30 41 45 32 43 402 38 44 33 42 39 17 16 15 18 11 10 02 21 401 40 14 35 1C03 1C04 1C05 1C06 07 26 23 25 24 34 29 22 20 1

2 Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 98 of 103 Rev. 40 THIS PAGE WAS INTENTIONALLY LEFT BLANK Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 2 AOP 913 Page 99 of 103 Rev. 40 CONTROL BUILDING MANUAL DAMPER CONTROL Page 1 of 3 FOR A FIRE IN VARIOUS FIRE AREAS PURPOSE:

To establish Control building HVAC for a fire in various plant fire areas dampers associated with Control Building HVAC units 1VAC30A and B must be manually blocked open or closed. This is required as a loss of instrument air.

LOCATION(S):

Control Building HVAC Equipment Room above Main Control Room. This work requires climbing on top of the HVAC units. This area is congested. A ladder will be necessary.

EQUIPMENT REQUIRED: Pliers, Tie Wire INSTRUCTIONS: In accordance with the following table, disconnect the damper from the damper operator and reposition the damper to the position indicated for the fire area of concern. The damper may be repositioned by removing the damper linkage pin (pliers and ladder may be necessary), rotating the damper arm to the required position and securing the damper arm with the tie wire supplied with this package.

Reference from TAB Damper Operator Damper Operated Damper Location Required Damper Position Disconnect and Rotate Damper Arm CB2 DO6113B D61-0011 Above 1VAC030B Closed Clockwise CB3 DO6113A D61-0017 Above 1VAC030A Closed Clockwise RB3 DO6113B D61-0011 Above 1VAC030B Open Counter-Clockwise RB4 DO6113B D61-0011 Above 1VAC030B Open Counter-Clockwise TB1 DO6113B D61-0011 Above 1VAC030B Open Counter-Clockwise RESTORATION: Have electrical maintenance verify affected circuits are free from fire damage or have been repaired and tested. Have mechanical maintenance verify instrument tubing is free from fire damage or repaired and tested and that instrument air is available. Following verification, remove previously installed tie wire and reconnect damper arm to damper operator by reinstalling linkage pin previously removed.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 2 AOP 913 Page 100 of 103 Rev. 40 Page 2 of 3 CONTROL BUILDING MANUAL DAMPER CONTROL FOR A FIRE IN VARIOUS FIRE AREAS REFERENCES

1. BECH-M151

2. BECH-M161

3. Calculation CAL-E96-012 (Fire Area CB2)

4. Calculation CAL-E96-013 (Fire Area CB3)

5. Calculation CAL-E96-022 (Fire Area CB3)

6. Calculation CAL-E96-023 (Fire Area RB4

7. Calculation CAL-R96-024 (Fire Area TB1)

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE ATTACHMENT 2 AOP 913 Page 101 of 103 Rev. 40 CONTROL BUILDING MANUAL DAMPER CONTROL Page 3 of 3 FOR A FIRE IN VARIOUS FIRE AREAS 1V-RF-30A 1V-RF-30B Sprinkler 12 Control Room N

Fa2 EL. 800'-4" Ga Hc Hf Ja Stairs 1V-AC-30A 1V-AC-30B DO6113A located between 1V-AC-30A and 1V-RF-30A (D61-0017)

DO6113B located between 1V-AC-30B and 1V-RF-30B (D61-0011)

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 FIRE AOP 913 Page 102 of 103 Rev. 40 References

1. EPIP, EAL Assessment

2. AOP 915, Shutdown Outside Control Room

3. IPOIs 3, 4, 5

4. OI 264, 513

5. EOP 1

6. Commitment AI 7001 (DR 84-364)

7. DAEC Fire Plan

8. DCP 1553

9. DDC 3151

10. DCP 1430

11. ECP 1619, ECP 1630

12. EMA A43417, EMA A43416

13. AR 23595

14. TNI FSAR, Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage system for Irradiated Nuclear Fuel, NUH-003, Latest Revision, NRC Docket No. 72-1004.

15. ACP 118, Conduct of the Duane Arnold Energy Center Interim On-Site Dry Spent Fuel Storage Program.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

AOP 913 Page 103 of 103 Rev. 40 ABNORMAL OPERATING PROCEDURE AOP 913 FIRE Effective Date:

TECHNICAL REVIEW Prepared by:

Date:

Validated by:

Date:

Operations Staff Verified by:

Date:

System Engineer Verified by:

Date:

Fire Protection Reviewed by:

Date:

Operations Committee PROCEDURE APPROVAL I am responsible for the technical content of this procedure.

Approved by Procedure Owner:

Date:

Operations Approved by:

Date:

DAEC Plant Manager Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 7_Rev1.doc Written Exam

200 225 250 275 300 325 350 375 0

25 50 75 100 125 150 RPV PRESSURE (psig)

Graph 1 RPV SATURATION TEMPERATURE 120 212 Action is required 200 225 250 275 300 325 350 375 0

25 50 75 100 125 150 RPV PRESSURE (psig)

Graph 1 RPV SATURATION TEMPERATURE 120 212 Action is required Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 8_Rev1.doc Written Exam

120 130 140 150 160 170 180 190 200 210 220 230 240 0

100 200 300 400 500 600 700 800 900 1000 1100 RPV PRESSURE (psig)

Graph 4 HEAT CAPACITY LIMIT 50 Action is required Bounding curve for torus levels between 8 ft and 13 ft.

120 130 140 150 160 170 180 190 200 210 220 230 240 0

100 200 300 400 500 600 700 800 900 1000 1100 RPV PRESSURE (psig)

Graph 4 HEAT CAPACITY LIMIT 50 Action is required Bounding curve for torus levels between 8 ft and 13 ft.

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 9_Rev1.doc Written Exam

0 5

10 15 20 25 30 6

8 10 12 14 16 TORUS WATER LEVEL (ft)

Graph 5 PRESSURE SUPPRESSION PRESSURE 7.1 13.8 25.9 Action is required Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 10_Rev1.doc Written Exam

0 50 100 150 200 250 300 350 400 0

10 20 30 40 50 6

DRYWELL PRESSURE (psig)

Graph 7 DRYWELL SPRAY INITIATION LIMIT Do NOT initiate drywell sprays in shaded area 3.0 7.2 0

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 11_Rev1.doc Written Exam

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 12_Rev1.doc Written Exam Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 12_Rev1.doc Written Exam

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 1 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 2 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 3 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 4 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 5 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 6 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 13_Rev1.doc Written Exam Page 7 of 7

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 14_Rev1.doc Written Exam

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 15_Rev1.doc Written Exam Page 1 of 2

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 15_Rev1.doc Written Exam Page 2 of 2

Senior Reactor Operator, 60006 Rev. 1 Topic 05-ILC-SRO-NRC 57_05-ILC-SRO-NRC-written_xm-Reference 16_Rev1.doc Written Exam RPS FAILURE None SA2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful Operating Modes: Run, Startup Auto Scram Failure AND Operator actions to reduce power are SUCCESSFUL as indicated by either:

ALL Rods Full-In, OR Reactor Shutdown Under All Conditions Without Boron, OR Reactor power below the APRM Downscale Alarm on ALL valid APRM instruments SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful Operating Modes: Run, Startup In ATWS EOP AND Operator actions to reduce power are UNSUCCESSFUL as indicated by either:

Reactor power above the APRM Downscale Alarm on ANY valid APRM instrument.

OR Boron Injection Initiation Temperature (BIIT) Curve (EOP Graph 6) exceeded.

SG2 Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core Operating Modes: Run, Startup In ATWS EOP AND Loss of adequate core cooling or decay heat Removal capability as indicated by either:

RPV level cannot be maintained above -

25 inches.

OR HCL Curve (EOP Graph 4) exceeded.

SS4 Complete Loss of Heat Removal Capability Operating Modes: Run, Startup, Hot S/D EOP Graph 4 Heat Capacity Limit is exceeded INABILITY TO REACH OR MAINTAIN SHUTDOWN CONDITIONS SU2 Inability to Reach Required Shutdown Within Technical Specification Limits Operating Modes: Run, Startup, Hot S/D Plant NOT brought to required mode within applicable LCO Action Statement Time.

SA3 Inability to Maintain Plant in Cold Shutdown Operating Modes: Cold S/D, Refuel Loss of decay heat removal systems required to maintain Cold Shutdown.

AND Temperature rise that exceeds 212ºF OR Uncontrolled temperature rise approaching 212ºF SS5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel Operating Modes: Cold S/D, Refuel RPV level below 15 inches, indicating that the core is or will be uncovered.

AND Loss of all decay heat removal.

See Fission Barrier Table INSTRUMENTATION /

COMMUNICATION SU3 Unplanned Loss of Most or All Safety System Annunication or Indication in the Control Room for Greater Than 15 Minutes Operating Modes: Run, Startup, Hot S/D Unplanned loss of most or all 1C03, 1C04 and 1C05 annunciators or indicators associated with Critical Safety Functions for greater than 15 minutes.

AND Compensatory non-alarming indications are available.

SA4 Unplanned Loss of Most or All Safety System Annunication or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2)

Compensatory Non-Alarming Indicators Unavailable Operating Modes: Run, Startup, Hot S/D Unplanned loss of most or all 1C03, 1C04 and 1C05 Annuciators or indicators associated with Critical Safety Functions for greater than 15 minutes.

AND Either of the following conditions exist:

A significant plant transient in progress.

OR Loss of compensatory non-alarming indications.

SS6 Inability to Monitor a Significant Transient in Progress Operating Modes: Run, Startup, Hot S/D Significant transient in progress and ALL of the following:

Loss of most or all annunciators on Panels 1C03, 1C04 and 1C05.

AND Loss of compensatory non-alarming indications.

AND Loss of indicators needed to monitor criticality, OR core heat removal, OR Fission Product Barrier status.

See Fission Barrier Table

ML051250432

Text

Navigation menu

Search