ML15097A267

301 Final SRO Written Exam

ML15097A267
Person / Time
Site:	Saint Lucie
Issue date:	04/07/2015
From:	NRC/RGN-II
To:	Florida Power & Light Co
References
50-335/15-301, 50-389/15-301 50-335/OL-15, 50-389/OL-15
Download: ML15097A267 (149)

Text

HLC 22 SRO NRC EXAM Question 1 Unit 1 is at 45% power steady state. A leak on the upper instrument tap (reference leg) associated with the SELECTED Pressurizer (Pzr) Level Transmitter, LT-1110X occurs.

Assuming NO Operator actions, which ONE of the following would be the response of the Pressurizer Pressure Level Control System?

Indicated Pzr Level Actual Pzr Level A. lowers rises B. rises rises C. rises lowers D. lowers lowers Page 1 of 100

HLC 22 SRO NRC EXAM Question 2 Given the following conditions on Unit 1:

• The reactor has been manually tripped and the crew is implementing 1-EOP-01, Standard Post Trip Actions

• The Core Heat Removal Safety Function has JUST been completed.

• Pressurizer pressure is 1590 psia

• Containment pressure is 5.5 psig Which ONE of the following completes the statements below:

A Safety Injection Actuation signal __(1)__ occurred.

The required status of the Reactor Coolant pumps at this time is __(2)__ should be secured.

A. 1) HAS

2) ONE RCP in each loop B. 1) HAS

2) ALL four RCPs C. 1) HAS NOT

2) ONE RCP in each loop D. 1) HAS NOT

2) ALL four RCPs Page 2 of 100

HLC 22 SRO NRC EXAM Question 3 Given the following conditions on Unit 2:

• A Loss of Coolant Accident (LOCA) has occurred

• Reactor Coolant System pressure is 75 psia Which ONE of the following is monitored or performed with regards to Short and Long Term Core Cooling given the above conditions?

Short Term Cooling Long Term Cooling A. Low Pressure Safety Injection operation/ Align Shutdown Cooling Safety Injection Tanks discharge B. Low Pressure Safety Injection operation/ Align High Pressure Safety Injection Safety Injection Tanks discharge hot and cold leg injection with a Recirculation Actuation Signal C. High Pressure Safety Injection operation/ Align Shutdown cooling Steam Generator steaming D. High Pressure Safety Injection operation/ Align High Pressure Safety Injection Steam Generator steaming hot and cold leg injection with a Recirculation Actuation Signal Page 3 of 100

HLC 22 SRO NRC EXAM Question 4 Unit 2 is at 100% power Given the following indications on the 2A1 Reactor Coolant Pump (RCP):

• Controlled Bleedoff (CBO) temperature is 225oF

• CBO flow is 2.5 gpm

• Vapor Seal Cavity pressure is 80 psia

• Upper Seal Cavity pressure is 2235 psia

• Middle Seal Cavity pressure is 2235 psia Which ONE of the following describes the status of the RCP seals and required action?

A. ONLY One RCP seal has failed; perform a controlled shutdown then secure the 2A1 RCP.

B. ONLY Two RCP seals have failed; trip the reactor and secure the 2A1 RCP.

C. ONLY Two RCP seals have failed; perform a controlled shutdown then secure the 2A1 RCP.

D. Three RCP seals have failed; trip the reactor and secure the 2A1 RCP.

Page 4 of 100

HLC 22 SRO NRC EXAM Question 5 Unit 1 is operating at 100% power.

The SNPO reports that a break in the Charging header has occurred.

The following indications noted on the RTGB:

• Charging Header pressure is 0 psig

• Charging Header flow is 0 gpm

• Reactor Cavity leakage is 0.1 gpm and steady Which ONE of the following completes the statements below:

Annunciator __(1)__ should be in alarm for this condition.

Due to the leak location, __(2)___ Charging pump(s) can be used when aligning per 1-AOP-02.03, Alternate Charging Flow Path Through the Auxiliary HPSI Header.

(Reference Provided)

A. 1) M-7, Regenerative Heat Exchanger Letdown P High

2) either the 1A, 1B OR 1C B. 1) M-28, Regenerative Heat Exchanger Outlet Temp High

2) either the 1A, 1B OR 1C C. 1) M-7, Regenerative Heat Exchanger Letdown P High

2) ONLY the 1A D. 1) M-28, Regenerative Heat Exchanger Outlet Temp High

2) ONLY the 1A Page 5 of 100

HLC 22 SRO NRC EXAM Question 6 Given the following conditions on Unit 1:

• The unit is currently shutdown for a refueling outage

• The Reactor Head is being removed

• The RCS level 35 feet

• Time since reactor shutdown is 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />

• Reactor Coolant System (RCS) temperature is 95º F Which ONE of the following choices completes the statement below:

If all SDC was lost, with the conditions given above, the Time to Boil is approximately____ minutes.

(References Provided)

A. 28 B. 30 C. 37 D. 40 Page 6 of 100

HLC 22 SRO NRC EXAM Question 7 Given the following conditions on Unit 1:

• The 1A and 1B Component Cooling Water (CCW) pumps are running

• The 1C CCW pump is in standby in accordance with 1-NOP-14.02, CCW System Operation, and is aligned to the B train electrically and mechanically

• A Loss of Offsite Power (LOOP) has just occurred

• The 1B CCW pump failed to start following the LOOP

• 1 minute later, a Loss of Coolant Accident (LOCA) with SIAS and CIAS actuation occurred Which ONE of the following describes the configuration of the CCW system at this time?

(Assume no operator actions)

A. ONLY 1 CCW pump supplying all CCW loads due to the failure of 1B CCW pump to start.

B. Two CCW pumps supplying all CCW loads due to SIAS start of the 1C CCW pump.

C. ONLY 1 CCW pump supplying its own safety header due to ESFAS isolation of the Non-Essential header.

D. Two CCW pumps supplying all CCW loads due to SIAS start of the 1C CCW pump until the Non-Essential header isolates on a loss of instrument air.

Page 7 of 100

HLC 22 SRO NRC EXAM Question 8 Given the following conditions on Unit 1:

• The unit is operating at 100% power

• PIC-1110X, Pressurizer (Pzr) pressure controller is selected

• Pzr pressure is being maintained at 2250 psia with Pzr Backup Heaters B1, B2, and B5 ON with a 10% output to the Pzr Proportional Heaters

• PT-1110Y is out of service for I&C calibration PT-1110X begins to slowly fail high.

Which ONE of the following choices completes the statement below:

IAW 1-AOP-01.10 Pressurizer Pressure and Level, the operator is directed to take manual control of HIC-1100, Spray Valve controller and __(1)__ its output in order to

__(2)__ ACTUAL Pzr Pressure.

A. 1) lower

2) lower B. 1) lower

2) raise C. 1) raise

2) raise D. 1) raise

2) lower Page 8 of 100

HLC 22 SRO NRC EXAM Question 9 Given the following conditions on Unit 1:

• Reactor power is 100%

• A total loss of feedwater event occurred

• The Reactor Protection System FAILED TO AUTOMATICALLY TRIP the reactor when trip conditions were present

• No operator actions have been taken Which ONE of the following describes the expected reactivity change AND the impact on the Reactor Coolant System (RCS) Pressure from this event?

Over the next 5 minutes, reactor power will:

A. Stay the same; an RCS over pressure condition will be prevented due to the event being within the design bases of the Pressurizer Safety valves.

B. Stay the same; an RCS pressure excursion that could potentially challenge the RCS pressure boundary integrity will be prevented by the Diverse Scram System.

C. Lower; an RCS over pressure condition will be prevented due to the event being within the design bases of the Pressurizer Safety valves.

D. Lower; an RCS pressure excursion that could potentially challenge the RCS pressure boundary integrity will be prevented by the Diverse Scram System.

Page 9 of 100

HLC 22 SRO NRC EXAM Question 10 Given the following conditions on Unit 1:

• The unit was operating at 100% power

• A Steam Generator Tube Rupture (SGTR) occurs

• 1-EOP-04, SGTR is being implemented

• SIAS actuation has occurred

• A Reactor Coolant System (RCS) cool down and depressurization is in progress The crew is evaluating RCS Void Elimination per 1-EOP-99, Appendix K, RCS Fill and Drain Method of Void Elimination.

Which ONE of the following is an indication that voids are present in the Reactor Vessel Head?

A. RVLMS indicates segments voided.

B. Pressurizer Level lowering while spraying.

C. Pressurizer level rising while charging to the RCS.

D. The saturation temperature of the RCS is higher than the temperature in the upper Rx Head region.

Page 10 of 100

HLC 22 SRO NRC EXAM Question 11 Given the following:

• The Unit 1 reactor tripped from 100% power 15 minutes ago

• The 1A and 1B Auxiliary Feedwater (AFW) pumps header flow control valves have been throttled to the desired flow rate for BOTH SGs

• The 1C AFW pump header flow control valves were closed

• The 1A and 1B Steam Generator (SG) Narrow Range levels are 10% and slowly rising

• A Loss of Offsite Power (LOOP) has just occurred and BOTH Emergency Diesel Generators (EDGs) are supplying power to their respective 4.16kVAC busses To maintain Reactor Coolant System temperature stable, which ONE of the following describes the manual operator action that must be taken on the AFW header flow control valves in response to the LOOP?

The 1A and 1B AFW pump header flow control valves __(1)___.

The 1C AFW pump header flow control valves __(2)__.

A. 1) require throttling closed

2) require throttling closed B. 1) remain throttled

2) require throttling closed C. 1) require throttling closed

2) remain closed D. 1) remain throttled

2) remain closed Page 11 of 100

HLC 22 SRO NRC EXAM Question 12 Given the following conditions on Unit 1:

• A Station Blackout has occurred and 1-EOP-10, Station Blackout, is being implemented

• The crew is evaluating Single Phase Natural Circulation criteria

• The BRCO reports the T between REP CET and T hot is 18oF and increasing Which ONE of the following choices describes:

1. The contingency action, in 1-EOP-10, Station Blackout, that is to be taken if Single Phase Natural Circulation criteria is BEING challenged?

AND

2. The operational implications if the temperature difference between T hot and REP CET exceeds the Single Phase Natural Circulation criteria limits?

A. 1) Verify Two Phase Natural Circulation CRITERIA.

2) The Hot Legs are becoming thermodynamically uncoupled from the core.

B. 1) Verify Two Phase Natural Circulation CRITERIA.

2) The Reactor Coolant System is approaching saturated conditions.

C. 1) Ensure proper control of Steam Generator feeding and steaming.

2) The Hot Legs are becoming thermodynamically uncoupled from the core.

D. 1) Ensure proper control of Steam Generator feeding and steaming.

2) The Reactor Coolant System is approaching saturated conditions.

Page 12 of 100

HLC 22 SRO NRC EXAM Question 13 Given the following conditions for Unit 2:

• The unit is operating at 100% power

• The 2AB 4.16kVAC bus is aligned to the 2B3 4.16kVAC bus

• HVA/ACC-3C is the running air conditioning unit

• HVA/ACC-3A and HVA/ACC-3B control switches are aligned as required per 2-NOP-25.07, Control Room Ventilation System

• A Loss of Offsite Power (LOOP) has occurred Which ONE of the following describes the final configuration of HVA/ACC-3A and HVA/ACC-3B Air Conditioning units AFTER the EDG loading sequence is complete?

HVA/ACC-3A is __(1)__.

HVA/ACC-3B is __(2)__.

A. 1) running

2) not running B. 1) not running

2) not running C. 1) not running

2) running D. 1) running

2) running Page 13 of 100

HLC 22 SRO NRC EXAM Question 14 Given the following conditions on Unit 1:

• The unit is operating at 100% power

• A loss of the 1MB 120VAC Vital Instrument Inverter occurs Which ONE of the following choices completes the statements below:

The B Channel Steam Generator (SG) Pressure and Level Safety channel indication will be affected on the __(1)__ .

The Reactor Protection System coincidence for a reactor trip will now be __(2)__ on Low SG Pressure or Level.

A. 1) A and B SGs

2) 1 out of 3 B. 1) A and B SGs

2) 2 out of 3 C. 1) B SG ONLY

2) 1 out of 3 D. 1) B SG ONLY

2) 2 out of 3 Page 14 of 100

HLC 22 SRO NRC EXAM Question 15 Given the following conditions on Unit 1

• Unit 1 was operating at 100% power

• The 1AB busses are aligned to the A train

• The unit tripped and the crew is carrying out actions of EOP-01, Standard Post Trip Actions

• During the Maintenance of Vital Auxiliaries Safety Function, the DRCO reports that the 1AB DC bus is de-energized Which ONE of the following actions are required in accordance with EOP-01 AND also satisfies the design basis:

Within ______ , re-energize the 1AB DC bus from the ________.

A. 10 minutes; 1AB Battery Charger B. 10 minutes; 1B DC bus C. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; 1AB Battery Charger D. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; 1B DC bus Page 15 of 100

HLC 22 SRO NRC EXAM Question 16 Given the following:

• Unit 2 is operating at 100% power.

• The 2B3 4.16kVAC bus has been de-energized due to a differential current lockout

• The 2AB 4.16kVAC bus is aligned to the A train

• The crew was UNSUCCESSFUL in resetting the differential current lockout relay

• The US directed that the 2B Intake Cooling Water (ICW) pump control switch be placed in Pull-to-Lock AND the 2C ICW pump MECHANICALLY aligned to the B train

• The US also directed that the 2B Component Cooling Water pump (CCW) pump control switch be placed in Pull-to-Lock AND the 2C CCW pump MECHANICALLY aligned to the B train

• The 2C ICW pump and the 2C CCW pump were then started as directed by 2-AOP-47.01B, Loss of a Safety Related AC Bus-Train B Which ONE of the following statements describes ALL of the consequences associated with the ICW system AND the CCW system being in this configuration per Ops Policy 503?

A. Offsite Power Operability for two ICW pumps running on the same electrical bus; Offsite Power Operability for two CCW pumps running on the same electrical bus.

B. Offsite Power Operability for two ICW pumps running on the same electrical bus.

C. Offsite Power Operability for two ICW pumps running on the same electrical bus; Offsite Power Operability AND Emergency Diesel Generator Operability for two CCW pumps running on the same electrical bus.

D. Offsite Power Operability AND Emergency Diesel Generator Operability for two CCW pumps running on the same electrical bus.

Page 16 of 100

HLC 22 SRO NRC EXAM Question 17 Given the following:

• An uncomplicated trip from 100% power has occurred on Unit 2

• 2-EOP-01, Standard Post Trip Actions is being implemented

• Reactor power is 1.0E-6% and lowering as indicated on Wide Range Instrumentation Which ONE of the following describes the status of the Nuclear Startup Channels?

A. Energized. They automatically energized at 1.0 E-5% power.

B. NOT energized. They will be manually energized while implementing 2-EOP-01, Standard Post Trip Actions.

C. Energized. They automatically energized on the reactor trip signal.

D. NOT energized. They will be manually energized after 2-EOP-02, Reactor Trip Recovery is entered.

Page 17 of 100

HLC 22 SRO NRC EXAM Question 18 Given the following:

• Unit 1 Reactor is at a stable power level of 2% in preparation for plant start up

• An ESDE on the 1A Steam Generator (SG) inside containment occurred

• Reactor Coolant System (RCS) T avg is currently 515°F and lowering

• RCS pressure is 2220 psia and lowering

• Pressurizer level is 25% and lowering

• Both Atmospheric Dump Valves (ADVs) are in service operating in Auto After the reactor is tripped, which ONE of the following describes:

1) The expected plant response.

AND

2) After the 1A SG dries out, operate the 1B ADV at a saturation pressure for A. 1) BOTH SGs will depressurize at the same rate until MSIS.

2) RCS T avg of 525oF to 535oF B. 1) ONLY the 1A SG will depressurize.

2) the lowest RCS T COLD C. 1) BOTH SGs will depressurize at the same rate until MSIS.

2) the lowest RCS T COLD D. 1) ONLY the 1A SG will depressurize.

2) RCS T avg of 525oF to 535oF Page 18 of 100

HLC 22 SRO NRC EXAM Question 19 Unit 2 is in Mode 2 with the following:

• A reactor startup is in progress using 2-GOP-302, Reactor Plant Startup - Mode 3 to Mode 2

• Reactor power is 1.0E-5%

• CEA group 5 is at 52 inches and being withdrawn in manual sequential to a planned stopping height of 60 inches

• CEA #59 does not move with the rest of the Group 5 CEAs If the CEAs were continued to be withdrawn, which ONE of the following alarms are expected during this evolution?

A. K-29, CEA PDIL (ADS).

B. K-18, Auto Withdraw Prohibit.

C. K-11, CEA Motion Inhibit.

D. K-26, CEA Withdrawal Prohibit.

Page 19 of 100

HLC 22 SRO NRC EXAM Question 20 Given the following:

• Unit 2 tripped from 100% due to a Loss of Offsite Power (LOOP)

• All AB electrical busses are aligned to the B side

• The 2B Charging pump is out of service with its control switch in stop Following the trip:

• The 2A Emergency Diesel Generator (EDG) tripped on overspeed Which ONE of the following choices completes the statement below:

If a condition requiring the crew to emergency borate existed, operators would be required to start the 2C Chg pump and emergency borate via the ___(1)____ to establish a flowpath.

If a SIAS were to subsequently occur, the 2C Chg pump breaker would __(2)__.

A. 1) 2B Boric Acid Makeup pump and Emergency Borate valve (V2514)

2) open then reclose on the load block sequencer B. 1) Gravity Feed valves (V2508 and V2509)

2) remain closed C. 1) 2B Boric Acid Makeup pump and Emergency Borate valve (V2514)

2) remain closed D. 1) Gravity Feed valves (V2508 and V2509)

2) open then reclose on the load block sequencer Page 20 of 100

HLC 22 SRO NRC EXAM Question 21 Given the following conditions on Unit 2:

• The reactor is operating at 100% power

• Channel X for Pressurizer (Pzr) Level control is SELECTED

• Channel Y for Pzr Level control, LT-1110Y, failed LOW Which ONE of the following states:

1) the CURRENT status of the Pzr Heaters AND

2) after selecting the LEVEL position on the Back Up Interlock Bypass Key Switch (B/U INTLK BP Key), which Pzr heaters are available?

A. 1) All Pzr heaters off

2) All B. 1) All Pzr heaters off

2) A side Pzr heaters ONLY C. 1) ONLY B side Pzr heaters off

2) A side Pzr heaters D. 1) ONLY B side Pzr heaters off

2) All Page 21 of 100

HLC 22 SRO NRC EXAM Question 22 Given the following:

• Core Reload is in progress on Unit 2

• 32 Fuel Assemblies have been loaded into the core

• During insertion of the 33rd assembly, the BRCO reports that count rate has increased from 10 CPS to 24 CPS Which ONE of the following describes the action required in accordance with 0-NOP-67.05, Refueling Operation, for the stated conditions?

A. Direct Reactor Engineering to assess the nuclear instrument readings used for the 1/M plot and verify that they do NOT extrapolate to 0 when fuel movement resumes.

B. Direct Reactor Engineering to validate the method being used to calculate the 1/M and renormalize the 1/M prior to inserting the next fuel assembly.

C. Stop inserting the fuel assembly and verify Shutdown Margin while maintaining the fuel assembly at the current position.

D. Stop inserting the fuel assembly and withdraw it into the Refueling Machine Hoist/Mast then verify Shutdown Margin.

Page 22 of 100

HLC 22 SRO NRC EXAM Question 23 Unit 1 is performing a Liquid Release of the 1A Waste Monitor Tank when the following alarm is received:

LIQUID WASTE RAD HIGH N-37 LIQUID RELEASE FLOW CONTROL VALVE, FCV-6627X, indicates open and will not close from the RTGB.

IAW 1-AOP-06.02, Uncontrolled Release Of Radioactive Liquids, which ONE of the following is a required SUBSEQUENT Operator action?

A. At the 1A Waste Monitor Storage Tank, stop the 1A Waste Monitor pump from the local control box AND at the CCW platform, lock closed V21462, Waste Monitor Pumps Discharge to Discharge Canal Isolation valve.

B. At the 1A Waste Monitor Storage Tank, close FCV-6627X AND at the CCW platform, lock closed V21462, Waste Monitor Pumps Discharge to Discharge Canal Isolation valve.

C. From the Liquid Waste Control panel ONLY, stop the 1A Waste Monitor pump AND close V21462, Waste Monitor Pumps Discharge to Discharge Canal Isolation valve.

D. From the Liquid Waste Control panel, close FCV-6627X AND at the Waste Monitor Storage Tanks, stop the 1A Waste Monitor pump from the local control box.

Page 23 of 100

HLC 22 SRO NRC EXAM Question 24 Given the following on Unit 1:

• Several fire detectors have alarmed in the Cable Spreading Room Which ONE of the following describes the expected actuation?

A. Halon will charge the header, but only discharge into the room when the header fuseable link melts.

B. Halon will charge the header and then automatically discharge into the room.

C. Pre-action dry pipe header will be charged with water, and then automatically discharge into the room.

D. Pre-action dry pipe header will be charged with water, but only discharge into the room when the header fuseable link melts.

Page 24 of 100

HLC 22 SRO NRC EXAM Question 25 Given the following conditions on Unit 1:

• The reactor is operating at 100% power

• A full B train INADVERTENT CIAS occurred

• 1-AOP-69.01 Inadvertent ESFAS Actuation was entered

• Section 4.2.2 Recovery from Inadvertent CIAS is being performed for B train CIAS components

• B train valve V6555 Waste Gas Containment Isolation valve (downstream) is indicating OPEN on the RTGB

• A train valve V6554 Waste Gas Containment Isolation valve (upstream) is indicating OPEN on the RTGB

• The affected Containment Penetration is #31 Which ONE of the following MEETS the required action, if any, per 3.6.3.1, Containment Isolation Valves? (consider each selection independently)

A. NONE since V6554 is OPERABLE.

B. V6554 must be CLOSED ONLY.

C. V6555 must be CLOSED AND DE-ENERGIZED.

D. V6555 must be CLOSED ONLY.

Page 25 of 100

HLC 22 SRO NRC EXAM Question 26 Given the following conditions on Unit 1:

• The reactor is at 100% power

• Increased Reactor Coolant System (RCS) activity has caused entry into 1-AOP-01.06, Excessive RCS Activity

• The specific activity of the RCS has stabilized at 0.15 uCi/gram Dose Equivalent (DEQ) Iodine 131

• DEQ XE-133 is at normal levels IAW 1-AOP-01.06, which ONE of the following is a required action based on the current level of DEQ I-131 and the reason for the action?

A. Isolate letdown to prevent high gamma radiation areas in the RCA.

B. Position V6307, Flash Tank Divert Valve, to the Flash Tank to strip fission product gasses.

C. Commence a plant shutdown due to RCS chemistry limits being exceeded.

D. Place all available CVCS Ion Exchangers in service to provide additional mechanical filtration to reduce RCS activity.

Page 26 of 100

HLC 22 SRO NRC EXAM Question 27 Given the following:

• Unit 2 is operating at 100% power

• The 2A Charging (Chg) pump is running

• The 2B Chg pump switch is in AUTO

• The Chg pump selector switch is in the 2A-2B position A Small Break Loss of Coolant Accident then occurs with the following:

• The Reactor is manually tripped

• SIAS has actuated

• Pressurizer level is 25% and slowly rising

• All Charging pumps are now running Complete the following statements:

(consider each statement independently)

The 2C CHG pump ___(1)____ following the SIAS signal.

Later in the event, if Pressurizer Level rises above Chg pump cut off set point, the 2B CHG pump (back-up) will __(2)___.

A. 1) was manually started

2) remain running B. 1) was manually started

2) auto stop C. 1) automatically started

2) remain running D. 1) remains running

2) auto stop Page 27 of 100

HLC 22 SRO NRC EXAM Question 28 The following conditions exist on Unit 1:

• The Reactor Coolant System (RCS) is solid, preparing for Reactor Coolant Pump sweeps IAW 1-NOP-01.05, Filling and Venting the RCS

• Shutdown Cooling has been secured

• RCS temperature is 130°F

• RCS pressure is 310 psia

• Secondary side of the Steam Generator (SG) is 165°F Based upon the given conditions, which ONE of the following results could occur if an RCP is started?

A. Excessive core uplift.

B. No adverse results should be expected.

C. An RCS pressure transient outside the LTOP system capacity.

D. Allowable brittle fracture stress limits for the SG could be exceeded.

Page 28 of 100

HLC 22 SRO NRC EXAM Question 29 Given the following:

• The Unit 1 Reactor Coolant System (RCS) is operating on solid pressure control

• One train of Shutdown Cooling (SDC) is in service

• BOTH LCV-2210P & Q, Letdown Level Control Valves, are open with their controller in MANUAL

• BOTH PCV-2201P & Q, Letdown Pressure Control Valves, are throttled with their controller in AUTO set at 100 psig Which ONE of the following choices completes the statement below:

If HCV-3657, SDC Temp Control Valve, is throttled OPEN, what is the effect on RCS temperature AND how would the Letdown Pressure Control valves respond to maintain letdown pressure on setpoint?

RCS temperature will:

A. rise causing PCV-2201P&Q to throttle open.

B. lower causing PCV-2201P&Q to throttle closed.

C. rise causing PCV-2201P&Q to throttle closed.

D. lower causing PCV-2201P&Q to throttle open.

Page 29 of 100

HLC 22 SRO NRC EXAM Question 30 Given the following:

• Unit 2 is in Mode 5 on Shutdown Cooling (SDC)

• The B SDC train is in service

• The A SDC train is in standby

• The following annunciator is received:

4.16KV SWGR 2B2 CURRENT TRIP A-13 Which ONE of the following describes the response of the 2B Low Pressure Safety Injection pump?

A. Continues in operation.

B. Its control breaker opens, but can be manually closed.

C. Its control breaker opens, but recloses following a time delay.

D. Stops due to loss of AC power and its breaker cannot be reclosed.

Page 30 of 100

HLC 22 SRO NRC EXAM Question 31 Given the following conditions:

• Unit 1 is in MODE 4

• 1A Shutdown Cooling (SDC) loop is in service for Reactor Coolant System (RCS) cooldown

• 1B1 and 1B2 Reactor Coolant Pumps (RCP) are running

• SDC flow rate returning to the core is 3400 gpm as indicated on FI-3306

• RCS temperature is 295°F and being held constant for shift turnover

• HCV-3657 SDC Temp Control Valve, is open and FCV-3306 SDC Return Flow Valve, is closed

• 2 of the 4 LPSI header injection valves are open

• The air supply line breaks on HCV-3657 Which ONE of the following choices completes the statement below:

As a result of the loss of air, HCV-3657 will __(1)__and SDC flow rate returning to the core will __(2)___.

A. 1) close

2) lower to approximately 0 gpm B. 1) fail as is

2) remain unchanged C. 1) close

2) lower then return to 3400 gpm D. 1) open

2) rise to approximately 4600 gpm Page 31 of 100

HLC 22 SRO NRC EXAM Question 32 Unit 2 is performing a cooldown per 2-GOP-305, Reactor Plant Cooldown - Hot Standby To Cold Shutdown, for a refueling outage with the following conditions:

Time: 11:30

• RCS pressure is 1750 psia

• RCS temperature is 504°F

• 2A & 2B S/Gs are 740 psia Time: 11:32

• 2A Steam Line ruptures outside Containment and just upstream of the MSIV

• RCS pressure is 1630 psia and rapidly lowering

• SG pressures are 580 psia and rapidly lowering Which ONE of the following states:

1) What ESFAS actuation(s) must be MANUALLY actuated?

2) Which procedure will be implemented to mitigate the event?

A. 1) BOTH MSIS, and SIAS.

2) 2-EOP-05, ESDE.

B. 1) ONLY SIAS.

2) 2-EOP-05, ESDE.

C. 1) BOTH MSIS, and SIAS.

2) 2-ONP-01.01 Plant Condition 1 Steam Generator Heat Removal LTOP Not in Effect.

D. 1) ONLY SIAS.

2) 2-ONP-01.01 Plant Condition 1 Steam Generator Heat Removal LTOP Not in Effect.

Page 32 of 100

HLC 22 SRO NRC EXAM Question 33 Given the following on Unit 1:

• The unit has just completed drawing a bubble from a solid plant condition

• Quench Tank pressure was noted to be 1 psig

• Pressurizer (Pzr) is being maintained at 200 psia by cycling backup Pzr heaters

• RCS temperature is 205°F If a PORV is venting fluid to the Quench Tank (QT), which ONE of the following would apply to both the expected PORV tailpipe temperature AND the Technical Specification Limiting Condition for Operation leakage limit for this type of leakage?

(Assume 100% quality. References provided)

A. ~310ºF; 1 gpm B. ~310ºF; 10 gpm C. ~380ºF; 1 gpm D. ~380ºF; 10 gpm Page 33 of 100

HLC 22 SRO NRC EXAM Question 34 Given the following on Unit 1:

• Unit 1 is operating at 100% power

• LS-14-1A (A side CCW Surge Tank Level Switch) on the Component Cooling Water (CCW) surge tank fails low causing the following annunciator CCW SURGE TANK LEVEL HIGH/

COMPARTMENT A LEVEL LOW S-6 Which ONE of the following describes the impact on CCW operation and Reactor Coolant Pump (RCP) Cooling:

A. No Non-Essential header valves isolate, A and B CCW header supplies flow to the RCPs.

B. A side Non-Essential header valves auto close; B CCW header supplies flow to the RCPs.

C. A side Non-Essential header valves should be manually closed; All RCPs should be secured.

D. All four Non-Essential header valves auto close; All RCPs should be secured.

Page 34 of 100

HLC 22 SRO NRC EXAM Question 35 On Unit 1, which ONE of the following DIRECTLY feeds the Pressurizer PORV OPEN signal when the PORV selector switch is in:

1) NORMAL RANGE

2) LOW RANGE DIRECTLY from the:

A. 1) RPS high pressurizer pressure trip bistables

2) Pressurizer pressure instruments PT-1103, 1104 B. 1) Pressurizer pressure instruments PT-1102 A through D

2) Pressurizer pressure instruments PT-1102 A through D C. 1) RPS high pressurizer pressure trip bistables

2) Pressurizer pressure instruments PT-1102 A through D D. 1) Pressurizer pressure instruments PT-1102 A through D

2) Pressurizer pressure instruments PT-1103, 1104 Page 35 of 100

HLC 22 SRO NRC EXAM Question 36 Given the following conditions on Unit 1:

The unit is operating at 100% power.

Reactor Protection System (RPS) Channel A Tcold RTD (TE-1112CA) has been inadvertently set to the MAXIMUM value of the Tcold range.

Which ONE of the following describes the effect of this adjustment on the given parameters?

Channel A : TM/LP Trip Setpoint Delta-T Power Indication A. Lowers Lowers B. Lowers Raises C. Raises Raises D. Raises Lowers Page 36 of 100

HLC 22 SRO NRC EXAM Question 37 Given the following conditions on Unit 2:

• Unit 2 is at 18% power

• The Main Generator is on-line

• The crew is raising turbine load at 1 MW/min IAW 2-GOP-201, Reactor Plant Startup - Mode 2 to Mode 1 The Main Transformer 2A Differential Current Alarm has annunciated resulting in:

• Turbine tripping on a generator lockout Which ONE of the following describes the expected plant response and subsequent actions?

The Reactor should _________.

A. remain CRITICAL; Verify SBCS quick opens to remove reactor heat and enter 2-AOP-53.03, Main Generator B. remain CRITICAL; Ensure SBCS responds in pressure control mode to remove reactor heat and enter 2-AOP-53.03, Main Generator C. have automatically TRIPPED; Manually trip the reactor and enter EOP-01, Standard Post Trip Actions D. have automatically TRIPPED; Manually trip the reactor then enter EOP-15, Functional Recovery, due to the Reactivity Control Safety Function not being met Page 37 of 100

HLC 22 SRO NRC EXAM Question 38 Given the following conditions on Unit 2:

The crew has entered 2-EOP-03, LOCA with the following:

• Containment Spray Actuation Signal (CSAS) has automatically actuated

• 2-HVS-1A and 2-HVS-1C, Containment Fan Coolers, are running

• 2A Containment Spray (CS) pump running with a flow rate of 2600 gpm

• 2B CS pump failed to auto start

• Containment Pressure is 7 psig and rising

• Containment Temperature is 220°F and rising

• Hydrogen Analyzers have just been placed in service Which ONE of the following describes the status of the Containment Temperature and Pressure Safety Function and required actions (if any) per 2-EOP-03, LOCA?

A. Met; based on current Containment Temperature and Pressure.

B. Not Met; but will be if the HVS-1B Containment Fan Cooler is started.

C. Not Met; but will be if the 2B CS pump is started with a flow rate of 2700 gpm.

D. Met; based on ONLY 2 Containment Fan Coolers running and the 2A CS pump at its current flow rate.

Page 38 of 100

HLC 22 SRO NRC EXAM Question 39 Given the following events and conditions:

• Unit 2 is operating at 100%

• The 2C Auxiliary Feedwater (AFW) pump is out of service Time: 0200:

• A loss of off-site power (LOOP) occurred

• The 2A Emergency Diesel Generator (EDG) automatically loaded on the bus

• The 2B EDG started but its output breaker DID NOT auto close

• 2A and 2B Steam Generator (S)G levels are 10% NR lowering Time: 0204:

A train of Auxiliary Feedwater (AFW) is feeding the 2A SG Time: 0205:

The 2B EDG output breaker was manually closed Which ONE of the following statements correctly describes the B Train AFW status at Time 0206, and any subsequent actions (if any)?

The 2B AFW pump:

A. is running and feeding the 2B SG. No further action is required.

B. is running but NOT feeding the 2B SG. Open the 2B AFW flow control valves to establish the desired flow rate.

C. is NOT running. Manually initiate AFAS-2 using all 4 Manual Initiation switches on RTGB-202.

D. is NOT running. Manually start the B train of AFW in accordance with the operator hard card.

Page 39 of 100

HLC 22 SRO NRC EXAM Question 40 Given the following condition on Unit 2:

Time: 0200

• A Small Break Loss of Coolant Accident has occurred

• A loss of offsite power (LOOP) occurs

• 2A and 2B Emergency Diesel Generators start and load as expected Time: 0220

• Safety Injection Signal Actuates (SIAS)

Which ONE of the following describes the expected condition of the Containment Fan Coolers following the LOOP, then following the SIAS?

__(1)__ Containment Cooling Fans running in ___(2)___ speed.

0200 0220 A. 1) Three Four

2) Fast Fast B. 1) Three Four

2) Fast Slow C. 1) Four Four

2) Fast Fast D. 1) Four Four

2) Fast Slow Page 40 of 100

HLC 22 SRO NRC EXAM Question 41 Given the following conditions on Unit 2:

• A Large Break LOCA has occurred

• The B Containment Spray Actuation signal (CSAS) failed to automatically or manually actuate

• The 2B Containment Spray pump was manually started by the RCO Which ONE of the following identifies the status of the B Train Iodine Removal System at this time?

2B Hydrazine Pump SE-07-3B, 2B Hydrazine Pump Discharge Valve A. OFF CLOSED B. ON CLOSED C. ON OPEN D. OFF OPEN Page 41 of 100

HLC 22 SRO NRC EXAM Question 42 The REGULARLY scheduled QUARTERLY Code Run Test has just been performed on the 2A Containment Spray (CS) pump IAW 2-OSP-07.04A, 2A Containment Spray and 2A Hydrazine Pump Code Run.

Which ONE of the following describes the specific parameter, from this particular surveillance, that must be satisfied in order to be in compliance with the Unit 2 CONTAINMENT SPRAY and COOLING SYSTEM Technical Specification LCO, 4.6.2.1?

A. 2A Hydrazine Pump flow as indicated on RTGB-206 B. 2A CS Pump developed head by a locally installed pressure gauge C. 2A CS Pump vibration by local vibrometer D. 2A CS Pump discharge flow as indicated on RTGB 206 Page 42 of 100

HLC 22 SRO NRC EXAM Question 43 Which ONE of the following is the function of the 103% Turbine Overspeed Protection Circuit (OPC)?

A. Reheat stop valves and extraction non-return valves are maintained closed to prevent LP turbine overspeed following a turbine trip.

B. Reheat stop valves and extraction non-return valves momentarily close to prevent LP turbine overspeed following a turbine trip.

C. Governor valves and intercept valves are maintained closed to prevent turbine overspeed following a complete loss of electrical load.

D. Governor valves and intercept valves momentarily close to prevent turbine overspeed following a complete loss of electrical load.

Page 43 of 100

HLC 22 SRO NRC EXAM Question 44 Given the following:

• Unit 1 reactor is at 65% power

• The 1A Main Feedwater pump trips Which ONE of the following describes the expected plant response BEFORE the automatic Reactor and Turbine Trip on Low Steam Generator (SG) Level?

(Assume no operator action)

Reactor Coolant System (RCS) Pressure will:

A. increase because the RCS delta T power increases.

B. increase because the RCS temperature increases due to increased SG temperatures.

C. decrease because the increased boiling rate in the SG tube bundle region decreases Tavg.

D. decrease because the SG level initially increases due to swell, causing a contraction of the RCS inventory.

Page 44 of 100

HLC 22 SRO NRC EXAM Question 45 Given the following conditions on Unit 2:

• The unit has just tripped from 100% power due to a loss of the 2A 125VDC bus

• The 2AB 125VDC bus is aligned to the B train

• 2-EOP-01, Standard Post Trip Actions is being implemented

• No contingency actions have been taken Which ONE of the following completes the statements below with regards to the configuration of the 2B and 2C Auxiliary Feedwater pumps (AFW) immediately following AFAS actuation?

The 2B AFW pump is __(1)__. The 2C AFW pump is running feeding __(2)__.

A. 1) running feeding the 2B Steam Generator

2) ONLY the 2A Steam Generator B. 1) locked out due to a Feedwater Header Rupture ID;

2) ONLY the 2A Steam Generator.

C. 1) running feeding the 2B Steam Generator;

2) BOTH the 2A and 2B Steam Generator.

D. 1) locked out due to a Feedwater Header Rupture ID;

2) BOTH the 2A and 2B Steam Generator.

Page 45 of 100

HLC 22 SRO NRC EXAM Question 46 Given the following on Unit 2:

• The reactor is operating at 100% power

• The 2A3 4.16kVAC bus has been de-energized due to an inadvertent trip of the 2A3-2A2 bus tie breaker, 4.16 KV BUS TIE 2A3-2A2

• The 2A Emergency Diesel Generator (EDG) has started and its output breaker is closed

• 2-AOP-47.01A, Loss of a Safety Related AC Bus-Train-A, is being implemented

• The SNPO reports that there is NO apparent damage found on the 2A3-2A2 bus tie breaker For the stated conditions, which ONE of the following completes the statements below:

IAW 2-AOP-47.01A Attachment 7, Restoration of Offsite Power with 2A EDG in Operation, the 2A EDG synchroscope should be slowly rotating in the __(1)__

direction.

If NON-EMERGENCY EDG trips were locked in AT THIS TIME, they would actuate when the __(2)__.

A. 1) clockwise (fast)

2) 2A3 to 2A2, 4.16kVAC bus tie breaker is closed B. 1) clockwise (fast)

2) Synch Plug switch is placed into the TIE 2A3/2B3 position C. 1) counterclockwise (slow)

2) 2A3 to 2A2, 4.16kVAC bus tie breaker is closed D. 1) counterclockwise (slow)

2) Synch Plug switch is placed into the TIE 2A3/2B3 position Page 46 of 100

HLC 22 SRO NRC EXAM Question 47 Given the following conditions on Unit 1:

The unit is at 100%

An automatic reactor trip occurs Which ONE of the following completes the statement below:

The __(1)___ signal DIRECTLY generates the __(2)__ bus automatic transfer from the auxiliary transformers to the Start-up transformers.

A. 1) TDM-1 and TDM-2

2) Live B. 1) TDM-1 and TDM-2

2) Dead C. 1) Generator Lockout

2) Live D. 1) Generator Lockout

2) Dead Page 47 of 100

HLC 22 SRO NRC EXAM Question 48 Given the following conditions on Unit 1:

• Unit 1 is at 100%

• The 1AA Battery Charger is out of service

• Annunciator B-20, "125V DC Bus/ 1A Batt Chgr/ Batt Rm Fan Trouble" has just alarmed

• The SNPO has been dispatched to verify the status of the 1A battery charger Which ONE of the following describes an indication(s), FROM THE CONTROL ROOM, that can be used to determine if the 1A vital DC bus is powered from the Battery ONLY?

A. ONLY the 1A DC Bus voltmeter on RTGB-101 indicates 130VDC slowly lowering.

B. ONLY the 1A DC Bus ammeter on RTGB-101 indicates a discharge rate.

C. Both an ammeter on RTGB-101 indicates a discharge rate AND the voltmeter on RTGB-101 indicates 130VDC slowly lowering.

D. A white light above the 1A DC Bus voltmeter on RTGB-101 is lit signifying potential is being supplied from a battery charger.

Page 48 of 100

HLC 22 SRO NRC EXAM Question 49 The Unit 2 SNPO has reported the following local alarm on the 2A Emergency Diesel Generator (EDG)

STARTING AIR PRESS LOW 5-1 The SNPO reports a relief valve lifted and has reseated on the two Air Receiver Tanks with the following pressures being reported:

• 2A1: 45 psig, 2A2: 50 psig, 2A3: 152 psig, 2A4: 155 psig Which ONE of the following describes the effect of this condition on the 2A EDG OPERABILITY AND if a start signal occurs prior to any operator action?

A. INOPERABLE; 2A EDG would start.

B. OPERABLE; 2A EDG would NOT start.

C. OPERABLE; 2A EDG would start.

D. INOPERABLE; 2A EDG would NOT start.

Page 49 of 100

HLC 22 SRO NRC EXAM Question 50 Given the following:

• Unit 1 is at 100% power

• The SNPO reports that the 1A Diesel Fuel Oil Transfer (DFOT) pump has tripped

• The Diesel Oil Storage Tanks (DOST) are above the Technical Specification minimum levels For the stated conditions, which ONE of the following completes the statements below:

Technical Specification Limiting Condition for Operation 3.8.1.1, Electrical Power Systems - AC Sources, is __(1)__.

AND If a Loss of Offsite Power were to occur, the expected system response would be

__(2)__.

(Assume no operator action)

A. 1) MET

2) The 1A EDG would run for ~47 minutes before running out of fuel B. 1) MET

2) The 1A EDG would run with make-up via gravity feed C. 1) NOT MET

2) The 1A EDG would run for ~47 minutes before running out of fuel D. 1) NOT MET

2) The 1A EDG would run with make-up via gravity feed Page 50 of 100

HLC 22 SRO NRC EXAM Question 51 Given the following on Unit 1:

Unit 1 is at 100% power Unit 2 is at 100% power The Unit 1 in-service waste gas decay tank has ruptured with a gas release in progress Which ONE of the following completes the statements below:

A MINIMUM of___(1)_____ Unit 1 Control Room Outside Air Intake Rad Monitor(s) in HIGH alarm, causes ___(2)_____ control room(s) to go into recirc mode?

A. 1) One

2) ONLY the Unit 1 B. 1) Two

2) ONLY the Unit 1 C. 1) One

2) BOTH Unit 1 and Unit 2 D. 1) Two

2) BOTH Unit 1 and Unit 2 Page 51 of 100

HLC 22 SRO NRC EXAM Question 52 Which ONE of the following describes the function of the TCV-13-2A & 2B, 1A & 1B Turbine Cooling Water Heat Exchanger Outlet Valves, AND MV-21-2 & 3, Intake Cooling Water A & B Train to Turbine Cooling Water Heat Exchanger Valves?

TCV-13-2A & 2B controls __(1)__flow through the Turbine Cooling Water heat exchanger.

MV-21-2 & 3 close, to isolate TCW, on receipt of a SIAS signal to __(2)__.

A. 1) Intake Cooling Water

2) provide backpressure for Intake Cooling Water pumps to prevent run-out flow during accident conditions B. 1) Turbine Cooling Water

2) provide backpressure for Intake Cooling Water pumps to prevent run-out flow during accident conditions C. 1) Turbine Cooling Water

2) ensure sufficient cooling flow is available to the Component Cooling Water Heat Exchangers D. 1) Intake Cooling Water

2) ensure sufficient cooling flow is available to the Component Cooling Water Heat Exchangers Page 52 of 100

HLC 22 SRO NRC EXAM Question 53 Given the following:

Unit 1 is at 100% power The following Control Room annunciators and indications are received:

INSTR AIR INSTR AIR PRESS COMPR HIGH/LOW AUTO START F-5 F-21

• Instrument Air Pressure indication on RTGB-102, Pl-18-9, indicates 97 psig, lowering slowly The NPO reports that the1C and 1D Instrument Air Compressors are running.

For the stated conditions, which ONE of the following completes the statements below:

IAW 1-AOP-18.01, Instrument Air Malfunction:

If Instrument Air header pressure continued to lower, the unit to unit cross-tie from Unit 2 would open at __(1)__.

If Instrument Air header pressure continued to degrade, the crew should evaluate the need to perform a controlled shutdown when Instrument Air header pressure reaches a MINIMUM of __(2)__.

A. 1) 95 psig

2) 75 psig B. 1) 85 psig

2) 75 psig C. 1) 95 psig

2) 60 psig D. 1) 85 psig

2) 60 psig Page 53 of 100

HLC 22 SRO NRC EXAM Question 54 Given the following conditions on Unit 1:

• A loss of offsite power (LOOP) has occurred on Unit 1

• The crew is performing actions of 1-EOP-09, Loss of Off Site Power / Loss of Forced Circulation

• The SM directs the crew to perform a controlled cooldown to 350°F in accordance with 1-AOP-01.13, Natural Circulation Cooldown Which ONE of the following local (field) operator action(s) must be completed to DIRECTLY support the cooldown from the CONTROL ROOM?

A. Reset Non-Essential Load Breakers IAW 1-EOP-99, Appendix P, Restoration of Components Actuated by ESFAS.

B. Crosstie Unit 1 Condensate Storage Tank with Unit 2 Condensate Storage Tank IAW 1-AOP-09.02, Auxiliary Feedwater.

C. Perform 1-EOP-99, Appendix X, Secondary Plant Post Trip Actions, Section 2.

D. Perform 1-EOP-99, Appendix H, Operation of the 1A and 1B Instrument Air Compressors.

Page 54 of 100

HLC 22 SRO NRC EXAM Question 55 Which ONE of the following completes the statements below:

IAW 2-NOP-68.01, Containment Building Access Hatches-Operation, interlocks that are installed to prevent both airlock doors from being open at the same time must be functional __(1)__. A control room annunciator alarms to alert the control room operators if __(2)__ door(s) is(are) open.

A. 1) in Modes 1 thru 4 AND Mode 6

2) EITHER the inner OR outer B. 1) in Modes 1 thru 4 AND Mode 6

2) BOTH the inner AND outer C. 1) ONLY in Modes 1 thru 4

2) EITHER the inner OR outer D. 1) ONLY in Modes 1 thru 4

2) BOTH the inner AND outer Page 55 of 100

HLC 22 SRO NRC EXAM Question 56 Given the following on Unit 1:

• A plant start-up is in progress with power stable at 15%

• The Main Generator is on line

• CEAs are at 110 inches on group 7

• One CEA in group 7 is 3 inches below the remainder of CEAs

• The CEDS control is placed in Manual Individual mode to realign the CEA

• When the RCO initially withdraws the CEA, the CEA moves outward but it continues to WITHDRAW when the control switch is released due to a control switch failure As a result, which ONE of the following will occur?

(Assume NO Operator action)

As power rises:

A. BOTH T avg and T ref will increase at the same rate and the CEA withdrawal should stop when a group deviation occurs.

B. BOTH T avg and T ref will increase at the same rate and the CEA withdrawal should stop when an Automatic CEA Withdrawal Prohibit occurs.

C. ONLY T avg will increase and the CEA withdrawal should stop when a group deviation occurs.

D. ONLY T avg will increase and the CEA withdrawal should stop when an Automatic CEA Withdrawal Prohibit occurs.

Page 56 of 100

HLC 22 SRO NRC EXAM Question 57 Which ONE of the following completes the statements below:

If the external leads of a Core Exit Thermocouple short together, a __(1)___

temperature than actual would be indicated.

In accordance with Ops Policy-503, Technical Specification Guidance, a TOTAL of

__(2)___ Core Exit Thermocouples per core quadrant are required to be OPERABLE for Mode 1 - 3 operation.

A. 1) higher

2) 2 B. 1) lower

2) 2 C. 1) lower

2) 4 D. 1) higher

2) 4 Page 57 of 100

HLC 22 SRO NRC EXAM Question 58 Given the following conditions on Unit 2:

• The plant is currently in Mode 3

• Radiation Protection has requested that a containment purge be performed to reduce particulate activity in Containment Which ONE of the following describes the required method of purging the Containment atmosphere under these conditions?

Operation of:

A. HVE-1&2, airborne radioactivity removal system IAW 2-NOP-25.05, Containment Ventilation Systems.

B. HVE-7A/B through the Continuous Purge (mini-purge) mode filter train IAW 2-NOP-25.02, Continuous Containment / Hydrogen Purge System Operation.

C. HVE-6A/B through the Shield Building Ventilation filter train IAW 2-NOP-25.01, Shield Building Ventilation System Operation.

D. HVE-8A/B through the normal Containment Purge (main-purge) System filter train IAW 2-NOP-06.20, Controlled Gaseous Batch Release to Atmosphere.

Page 58 of 100

HLC 22 SRO NRC EXAM Question 59 Given the following on Unit 1:

• A small break LOCA has occurred

• 1-EOP-03, LOCA, was entered and the direction to place both Hydrogen Analyzers in service was given Which ONE of the following completes the statements below:

IAW with 1-EOP-99, Appendix L, Placing a Hydrogen Analyzer in Service, aligning the sampling point for the Hydrogen Analyzer requires positioning valves at (in) the __(1)__.

AFTER 1-EOP-99, Appendix L, Placing a Hydrogen Analyzer in Service, had been performed, a Loss of Offsite (LOOP) occurred resulting in BOTH Emergency Diesel Generators starting and energizing the 1A3 and 1B3 4.16 kv busses.

Without any operator action, Hydrogen Analyzer panel operation IN THE CONTROL ROOM __(2)__ be affected.

A. 1) Hydrogen Analyzer Cubicle on the 43 foot elevation in the Fan Room

2) WOULD B. 1) Control Room

2) WOULD C. 1) Hydrogen Analyzer Cubicle on the 43 foot elevation in the Fan Room

2) WOULD NOT D. 1) Control Room

2) WOULD NOT Page 59 of 100

HLC 22 SRO NRC EXAM Question 60 IAW 1-NOP-67.04, Refueling Machine Operation, which ONE of the following would be expected while operating the Refueling Machine (RFM) during a core onload?

A. A Bridge and Trolley lockout interlock is present any time the RFM Hoist is being lowered or raised, while on index at a core location, to prevent RFM movement.

B. If fuel movement had to be suspended, the RFM operator should ensure that the fuel transfer carriage is placed in its stored position on the Spent Fuel Pool side.

C. When lowering the RFM Hoist over the core with a fuel assembly grappled on, hoist motion is automatically stopped when the Lower Grapple operate zone is reached.

D. Operation with the Programmable Logic Controller (PLC) switch in the OVRD position is permitted to override an OVERLOAD condition on a fuel assembly that is being raised out of the core.

Page 60 of 100

HLC 22 SRO NRC EXAM Question 61 Given the following conditions on Unit 1:

• The Unit is in Mode 3

• Control Element Assembly testing is planned

• 1-GOP-302, Reactor Plant Startup - Mode 3 to Mode 2 Which ONE of the following Reactor Protection System RPS trips is NOT bypassed during this testing?

A. Low Steam Generator Level B. Thermal Margin / Low Pressure C. Low Steam Generator Pressure D. Low Reactor Coolant System Flow Page 61 of 100

HLC 22 SRO NRC EXAM Question 62 Given the following Unit 1:

• Unit 1 is at 12% power with Tavg at 535°F and Steam Generator pressure at 900 psia

• The Main Generator has just been synchronized to the grid

• The Main Generator picked up 40 MWe and then TRIPPED 5 seconds later Which ONE of the following states the expected plant response?

A. Reactor power indicates 12% with Steam Bypass Control System valves opening to control steam generator pressure at 940 psia for 4 minutes then ramp to 900 psia over the next 5 minutes.

B. Reactor power indicates 12% with Steam Bypass Control System valves opening to control steam generator pressure at 900 psia ONLY.

C. Reactor trips with Steam Bypass Control System closing to control steam generator pressure at 940 psia for 4 minutes then ramp to 900 psia over the next 5 minutes.

D. Reactor trips with both Atmospheric Dump Valves opening to control steam generator pressure at 900 psia.

Page 62 of 100

HLC 22 SRO NRC EXAM Question 63 Given the following conditions on Unit 2:

• Annunciator N-34 Gas Analyzer Trouble is in alarm

• The gas analyzer local panel oxygen monitor and Chemistry samples on the in-service gas decay tank indicate 2.2%

IAW 2-AOP-06.03, Waste Gas System, which ONE of the following actions should be taken?

Suspend all additions of waste gases and _______.

A. ensure the Gas Surge header inlet to the Gas SURGE Tank auto closes and the nitrogen purge supply to the Gas SURGE Tank auto opens B. manually align nitrogen to the Gas SURGE Tank C. ensure the Gas Surge header inlet to the Gas DECAY Tank auto closes and the nitrogen purge supply to the Gas DECAY Tank auto opens D. manually align nitrogen to the in-service Gas DECAY Tank Page 63 of 100

HLC 22 SRO NRC EXAM Question 64 Given the following conditions on Unit 2:

TIME 1200: The unit had to be manually tripped due to a severe DEH leak on Governor Valve #1 1202: A 350 gpm tube rupture developed in the 2A Steam Generator (SG) 1205: SIAS & CIAS actuations occurred on low Reactor Coolant System (RCS) pressure 1210: 2-EOP-01, SPTAs are in progress and the crew has just begun to evaluate the Containment Conditions safety function Which ONE of the following rad monitor alarms and trend indications would be AVAILABLE AT THIS TIME to determine the status of the Containment Conditions safety function (Secondary Activity)?

(ASSUME NO OPERATOR ACTIONS HAVE BEEN TAKEN and no other complications exist post trip)

A. Main Steam Line Rad Monitor AND SG Blowdown Rad Monitor B. Steam Jet Air Ejector AND SG Blowdown Rad Monitor C. ONLY Main Steam Line Rad Monitor D. ONLY Steam Jet Air Ejector Page 64 of 100

HLC 22 SRO NRC EXAM Question 65 Unit 1 is at 100% power, steady state.

• The 1B Intake Cooling Water (ICW) pump has just tripped unexpectedly

• It is desired to START the 1C ICW pump to replace the 1B ICW pump In accordance with 1-AOP-21-03B, 1B INTAKE COOLING WATER SYSTEM HEADER, which ONE of the following answers the questions below:

1) What is the MINIMUM electrical alignment to start the 1C ICW pump for the given plant conditions?

2) Prior to starting the 1C ICW pump, if Main Generator Average Cold Gas Temperature were to rise to 47.7°C, what is the required action the crew must take at that point in time?

A. 1) The 1AB 4.16 kVAC bus, 1AB 480 V bus AND the 1AB DC bus aligned to the B side.

2) Reduce MVARS to minimum.

B. 1) The 1AB 4.16 kVAC bus, 1AB 480 V bus AND the 1AB DC bus aligned to the B side.

2) Reduce power as necessary to stabilize main generator cold gas temperature.

C. 1) ONLY the 1AB 4.16 kVAC bus aligned to the B 4.16 KV side.

2) Reduce MVARS to minimum.

D. 1) ONLY the 1AB 4.16 kVAC bus aligned to the B 4.16 KV side

2) Reduce power as necessary to stabilize main generator cold gas temperature.

Page 65 of 100

HLC 22 SRO NRC EXAM Question 66 Given the following on Unit 2:

• A Core Reload is in progress

• Both BF 3 Log Startup Channels are in service with the B Channel selected for audible count rate monitoring (control room and containment)

• While Reactor Engineering was monitoring the indicated count rates for a 1/M plot, it was determined that the B Log Startup Channel had failed low IAW 0-NOP-67.05, Refueling Operation, which ONE of following describes the impact on the fuel movement?

A. With the B Log Startup Channel inoperable, the core reload MUST STOP until it is returned to service.

B. With the A Log Startup Channel operable, the core reload can continue as long as audible is selected for that channel.

C. The core reload can continue by substituting a Wide Range Log Safety Channel and selecting the audible count rate to that channel.

D. The core reload can continue by substituting an Appendix R Wide Range (Excore) and selecting the audible count rate to channel A.

Page 66 of 100

HLC 22 SRO NRC EXAM Question 67 Unit 1 is operating at 100% when the following annunciator alarmed due to a loss of power to the MA Instrument Inverter:

120V AC INSTR BUS MA INVERTER TROUBLE B-43 Which ONE of the following Control Room alarms would NOT be expected to annunciate for this event?

A. Q-19, 1B S/G PRESS MSIS CHANNEL TRIP B. Q-16, CNTMT RAD HIGH CIS CHANNEL TRIP C. R-10, ENGINEERED SAFEGUARDS ATI FAULT D. R-11, CNTMT PRESS HIGH CSAS CHANNEL TRIP Page 67 of 100

HLC 22 SRO NRC EXAM Question 68 Given the following on Unit 2:

• Reactor start-up is in progress

• Mode 2 has just been declared Reactor Engineering predicts that the reactor will become critical approximately 600 pcm earlier than the ECC calculation.

Which ONE of the following describes the correct course of action IAW 2-GOP-302, Reactor Plant Startup - Mode 3 to Mode 2?

Stop CEA withdrawal, _________.

A. and immediately initiate emergency boration B. and ensure CEAs are inserted to a position equivalent to (-) 500 pcm from the apparent critical position C. maintain the present rod height, and verify that the 1/M predicted rod position is greater than minimum rod height for criticality D. raise the Reactor Coolant System boron concentration until criticality can be predicted within the proper +/-pcm band. Recalculate the ECC prior to continuing CEA withdrawal Page 68 of 100

HLC 22 SRO NRC EXAM Question 69 Given the following:

• Unit 1 is at 100% power

• An equipment clearance order (ECO) is being developed for a system with the following parameters: pressure - 100 psia and temperature - 130°F

• Due to an equipment deficiency, only a SINGLE CHECK VALVE is available as an isolation boundary IAW OP-AA-101-1000, Clearance and Tagging, at a minimum, which ONE of the following describes the Energy classification and whose approval is required for the clearance to allow use of a SINGLE CHECK VALVE as an isolation boundary?

A. High Energy; Nuclear Joint Safety Committee B. Low Energy; Nuclear Joint Safety Committee C. Low Energy; Shift Manager D. High Energy; Shift Manager Page 69 of 100

HLC 22 SRO NRC EXAM Question 70 Given the following conditions on Unit 1:

Due to equipment unavailability, the Containment Fan Cooler flow weekly surveillance check must be delayed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

To ensure compliance with the Technical Specification Limiting Condition of Operation (LCO), which ONE of the following describes how the surveillance must be tracked?

A. With an Action Request (AR) ONLY.

B. the Equipment Out Of Service program AND an Action Request (AR).

C. 1-OP-0010125A, Data Sheet 29, Surveillance Tracker (Data Sheet Pink).

D. 1-OP-0010125A, Data Sheet 30, Unscheduled Surveillances and Evolution Tracking (Data Sheet Green).

Page 70 of 100

HLC 22 SRO NRC EXAM Question 71 Given the following conditions on Unit 1:

• 1-GOP-302, Reactor Plant Startup - Mode 3 to Mode 2 is being implemented

• Reactor power is at 1.0E-2% and slowly increasing

• An oil leak on the 1A2 Reactor Coolant Pump (RCP) is suspected

• A containment entry is required to check the RCP Oil Collection Tank level

• No Management Exemptions have been authorized for this entry IAW ADM- 09.05, Containment Entries Mode 1-4, which ONE of the following conditions apply?

A Containment entry:

A. cannot occur because reactor power level is beyond the administrative limit for entry inside the biological shield wall.

B. can occur ONLY IF reactor power is lowered to <1.0E-3%.

C. can occur ONLY IF ALL Regulating Group CEAs are inserted.

D. can occur IF power level is stabilized at the current value.

Page 71 of 100

HLC 22 SRO NRC EXAM Question 72 Given the following:

• The Unit 2 reactor is operating 100%

• The Radiation Monitoring Computer System (RMCS) is magenta for the B Main Steam Line (MSL) radiation monitor

• There is a suspected tube leak in the B Steam Generator (SG)

• The Unit Supervisor directs use of the Portable RM-23P at the B MSL line RM-80 Microprocessor When the RM-23P is placed in service, which ONE of the following describes:

1) the Control Room (CR) Alarm Function AND

2) the location where the B MSL radiation levels can be read CR Alarm function B MSL radiation reading A. None ONLY Local at RM-23P B. RMCS Local at RM-23P AND RM-23 at PACB-2 C. None Local at RM-23P AND RM-23 at PACB-2 D RMCS RMCS AND Local at RM-23P Page 72 of 100

HLC 22 SRO NRC EXAM Question 73 Given the following:

• Unit 1 tripped from 100% power

• The A Steam Generator (SG) has experienced an Excess Steam Demand event due to a failed open Main Steam Safety Valve (MSSV)

• The B SG has a sheared U-tube

• The crew has entered 1-EOP-15, Functional Recovery Which ONE of the following completes the following statement:

The MOST AFFECTED SG is the __(1)__ SG. It will remain the MOST AFFECTED SG until __(2)__.

A. 1) A

2) the failed open MSSV on the A SG is gagged B. 1) A

2) 1-EOP-99, Appendix R, Steam Generator Isolation, is complete on the A SG C. 1) B

2) 1-EOP-99, Appendix R, Steam Generator Isolation, is complete on the B SG D. 1) B

2) Reactor Coolant System pressure is reduced within 50 psia of the B SG pressure.

Page 73 of 100

HLC 22 SRO NRC EXAM Question 74 Given the following on Unit 1:

Time 0010:

• The Unit is at 100% power

• Pressurizer level is 66%

• Pressurizer pressure is 2250 psia

• 2 Charging pumps are running

• Letdown flow is 84 gpm

• The 1A Steam Generator (SG) has a known tube leak of 15 gallons per day

• A Main Steam Line radiation monitor is reading 0.5 mr/hr

• A containment particulate radiation monitor is reading 210 cpm Time 0020:

• Pressurizer level is at 64%

• Pressurizer pressure is 2235 psia

• 2 Charging pumps are running

• Letdown flow is 50 gpm

• A Main Steam Line radiation monitor is reading 0.5 mr/hr

• A containment particulate radiation monitor is reading 400 cpm Which ONE of the following completes the following statements:

1) The type of event that is occurring at time 0020 is a __(1)__

2) For the purposes of Emergency Plan classification, the increased RCS leakage at time 0020 is considered __(2)__ leakage.

A. 1) Steam Generator Tube Leak

2) IDENTIFIED B. 1) Steam Generator Tube Leak

2) UNIDENTIFIED C. 1) Reactor Coolant System Leak

2) IDENTIFIED D. 1) Reactor Coolant System Leak

2) UNIDENTIFIED Page 74 of 100

HLC 22 SRO NRC EXAM Question 75 Given the following conditions on Unit 1:

• 1-EOP-03, LOCA is being implemented

• Pressurizer Pressure 900 psia and slowly lowering

• T HOT is 508°F and slowly lowering

• REP CET 515°F and slowly lowering

• Pressurizer Level is 55% and slowly rising

• Both Steam Generators are 25% Narrow Range and rising with total AFW flow of 350 gpm

• Both Steam Generator pressures are 660 psia slowly lowering

• ECCS flow is 650 gpm

• Containment Temperature is 185°F and slowly lowering Which ONE of the following states the strategy that should be implemented AT THIS TIME? (Consider each selection independently)

(References Provided)

Cooldown and ___________.

A. depressurize to maximize ECCS flow B. ensure 1-EOP-99, Figure 2 is being maintained C. allow Pressurizer level to rise to 68% then throttle ECCS flow and secure Charging D. throttle ECCS flow and cycle a Charging pump to allow Pressurizer level to stabilize and be maintained between 30% to 68%

Page 75 of 100

HLC 22 SRO NRC EXAM Question 76 Given the following:

• Unit 1 is implementing 1-EOP-02, Reactor Trip Recovery following an inadvertent A train SIAS

• The 1C Charging pump was out of service prior to occurrence of the event

• The 1A Charging pump did not start on the SIAS and could not be manually started

• BRCO reports that for the last five (5) minutes the Quench Tank level has risen approximately 20%

• Reactor Coolant System (RCS) pressure is 1950 psia and slowly lowering

• Minimum subcooling is not being met Which ONE of the following states:

1) The required action to take AT THIS TIME AND

2) The required Reactor Coolant System Leakage Technical Specification Action Statement that applies for the current conditions?

(References Provided)

A. 1) Re-diagnose the event and enter 1-EOP-03, LOCA.

2) Reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

B. 1) Enter 1-AOP-01.08, RCS Leakage Abnormal, and perform the Safety Function Status Checks of 1-ONP-01.01, Plant Condition 1 Steam Generator Heat Removal LTOP Not in Effect.

2) Reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

C. 1) Re-diagnose the event and enter 1-EOP-03, LOCA.

2) Be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> ONLY.

D. 1) Enter 1-AOP-01.08, RCS Leakage Abnormal, and perform the Safety Function Status Checks of 1-ONP-01.01, Plant Condition 1 Steam Generator Heat Removal LTOP Not in Effect.

2) Be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> ONLY.

Page 76 of 100

HLC 22 SRO NRC EXAM Question 77 Given the following conditions on Unit 1:

• A core RELOAD is in progress and is 95% complete

• The 1A Low Pressure Safety Injection pump (LPSI) is operating and aligned to provide Shutdown Cooling (SDC) to the Reactor Coolant System (RCS)

• 1B LPSI pump is out of service for pump maintenance

• The 1A LPSI pump has just tripped due to a motor fault What is the required Tech Spec Action that applies and what is (are) the bases?

The core on load ___(1)____ all operations that would cause an introduction into the RCS, coolant with a boron concentration less than the Tech Spec refueling boron concentration are suspended AND all containment penetrations are CLOSED within the next four hours.

At least one Shutdown Cooling loop in operation ensures __(2)__.

A. 1) may continue for up to one hour provided that

2) adequate mixing B. 1) may continue for up to one hour provided that

2) sufficient decay heat removal capability AND minimizes the effect of a boron dilution event C. 1) must be suspended and additionally,

2) adequate mixing D. 1) must be suspended and additionally,

2) sufficient decay heat removal capability AND minimizes the effect of a boron dilution event Page 77 of 100

HLC 22 SRO NRC EXAM Question 78 Given that Unit 2 is in Mode 1 with the following conditions:

• The crew entered 2-AOP-14.01,Component Cooling Water Abnormal Operations due to a leak in the 2B CCW Heat Exchanger

• The Unit Supervisor has determined that the 2B Component Cooling Water (CCW) Heat Exchanger is to be isolated

• The 2C CCW pump is in its normal standby alignment

• The required MECHANICAL alignment of the 2C CCW pump has been performed Complete the following statements :

In accordance with 2-AOP-14.01, ELECTRICALLY align the 2C CCW pump to the

__(1)__.

For the given alignment, CCW loads must be isolated to ensure the CCW Heat Exchanger design bases CCW flow limit will not exceed a MAXIMUM flow of __(2)__.

A. 1) B train

2) 18,500 gpm B. 1) A train

2) 14,600 gpm C. 1) B train

2) 14,600 gpm D. 1) A train

2) 18,500 gpm Page 78 of 100

HLC 22 SRO NRC EXAM Question 79 Given the following:

At time 0210, a Station Blackout occurred on Unit 2 At time 0225, the crew entered 2-EOP-10, Station Blackout with the following conditions:

• Reactor Coolant System (RCS) pressure is ~1820 psia and constant

• Both SG water levels are being restored with Auxiliary Feedwater (AFW) at 200 gpm from the 2C AFW pump

• Loop Tcold temperatures are 540°F and lowering

• Loop Thot temperatures are 575°F and constant

• REP CET is 587°F and constant Complete the following statements:

IAW the Unit 2 FSAR Station Blackout Analysis, operator control of the Atmospheric Dump Valves is credited to occur NO LATER THAN _(1)__.

For the given conditions, the Atmospheric Dump Valves must be throttled more in the__(2)__ direction.

A. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) OPEN B. 1) 30 minutes

2) OPEN C. 1) 30 minutes

2) CLOSED D. 1) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

2) CLOSED Page 79 of 100

HLC 22 SRO NRC EXAM Question 80 A Loss of Offsite Power occurred with the following:

Time 0220

• Unit 1 was at 100% power and both Emergency Diesel Generators (EDGs) started

• Unit 2 is in a refueling outage, DEFUELED and (1) one EDG Generator started and the other EDG is Out of Service Time 0240

• On Unit 1, ONE of the running EDGs tripped on Overspeed

• On Unit 2, the ONLY running EDG tripped on Differential Current For the given times, which ONE of the following identifies the Emergency Classifications that must be reported to the State Watch Office and NRC?

(References Provided)

A. Time 0220, Alert - notifications required.

Time 0240, No change in classification - No additional notification required.

B. Time 0220, Unusual Event - notifications required.

Time 0240, Site Area Emergency - notifications are required.

C. Time 0220, Unusual Event - notifications required.

Time 0240, Alert - notifications required.

D. Time 0220, Alert - notifications required.

Time 0240, Site Area Emergency - notifications are required.

Page 80 of 100

HLC 22 SRO NRC EXAM Question 81 Given the following conditions:

• Unit 1 is in Mode 3

• Switchyard Voltage is 231kV in support of system distribution work

• The Contingency Analysis program predicts no change in switchyard voltage

• Maintenance personnel are troubleshooting the control circuit for 1-HVS-1B, Containment Cooling Fan

• Maintenance requests to stop the 1B Containment Cooling Fan and leave the control switch in AUTO and the breaker left ON while troubleshooting continues

• All other Containment Cooling Fans are in operation In accordance with Ops Policy 503, Technical Specification Guidance, which ONE of the following describes the administrative requirement for this configuration?

Declare:

A. BOTH the 1B Offsite Power Circuit inoperable AND the 1B Emergency Diesel Generator inoperable.

B. BOTH the 1A Offsite Power Circuit inoperable AND the 1A Emergency Diesel Generator inoperable.

C. ONLY the 1A Offsite Power Circuit inoperable.

D. ONLY the 1B Offsite Power Circuit inoperable.

Page 81 of 100

HLC 22 SRO NRC EXAM Question 82 Given the following:

• Unit 1 is at 100% power

• All CEAs were at 136 inches when CEA 56 dropped with the rod bottom light lit on the core mimic display

• Efforts to re-align CEA 56 have been in progress for 60 minutes, but CEA 56 has not been re-aligned to the proper height IAW 1-AOP-66.01, Dropped or Misaligned CEA Abnormal Operations Which ONE of the following states:

1) The CEA 56 position indication on the Distributed Control System (DCS) prior to CEA recovery.

2) The procedure requirements of 1-AOP-66.01, Dropped or Misaligned CEA Abnormal Operations AT THIS TIME?

A. 1) DCS would indicate 0 inches.

2) Continue efforts to re-align CEA 56 and concurrently reduce power to 70% IAW 1-AOP-22.01, Rapid Downpower.

B. 1) DCS would indicate 0 inches.

2) Suspend efforts to re-align CEA 56 and reduce power to 70% IAW 1-AOP-22.01, Rapid Downpower.

C. 1) DCS would indicate 136 inches.

2) Continue efforts to re-align CEA 56 and concurrently reduce power to 70% IAW 1-AOP-22.01, Rapid Downpower.

D. 1) DCS would indicate 136 inches.

2) Suspend efforts to re-align CEA 56 and reduce power to 70% IAW 1-AOP-22.01, Rapid Downpower.

Page 82 of 100

HLC 22 SRO NRC EXAM Question 83 Given the following:

• The Unit 2 operating crew has evacuated the Control Room due to the presence of toxic fumes

• 2-ONP-100.02 Control Room Inaccessibility is being implemented

• All Operator actions in the Control Room were performed prior to evacuation

• RCO A is maintaining Hot Standby conditions at the Remote Shutdown Panel Complete the following statements:

IAW 2-ONP-100.02 Control Room Inaccessibility, Figure 1, Subcooled margin is determined using the values of Reactor Coolant System ____(1)____ and PRESSURIZER PRESSURE.

If the PRESSURIZER PRESSURE instrument used to calculate subcooling margin failed low, the Remote Shutdown Tech Spec Action that is applicable is:

Restore the inoperable channel to operable status within __(2)__ days.

(Reference Provided)

A. 1) Thot

2) 7 B. 1) Thot

2) 30 C. 1) Tcold + 50°F

2) 7 D. 1) Tcold + 50°F

2) 30 Page 83 of 100

HLC 22 SRO NRC EXAM Question 84 The following annunciator has just been received on Unit 1:

THERMAL POWER CHANGE EXCEEDS 15% PER HOUR L-6 Per Technical Specification Table 4.4-4, which ONE of the following describes the required surveillance AND the bases for that action?

The Specific Activity of the Reactor Coolant System shall be determined to be within LCO limits for the stated time interval by performing an isotopic analysis for:

A. Iodine once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to establish a trend to determine that Dose Equivalent Iodine-131 is less than the LCO limit over the duration of the power change.

B. One sample for Iodine between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due to the expected Iodine peak during the period.

C. Xenon once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to determine that Dose Equivalent Xenon-133 is less than the LCO limit over the duration of the power change.

D. One sample for Xenon between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> due to the expected Xenon peak during the period.

Page 84 of 100

HLC 22 SRO NRC EXAM Question 85 Unit 2 is performing a Natural Circulation cooldown per 2-AOP-01.13, Natural Circulation Cooldown.

At time 0220 the following conditions were noted:

• Pressurizer pressure is 1620 psia

• SIAS has been blocked

• Reactor Coolant System (RCS) temperature is 490°F and lowering

• Boric Acid Makeup Tanks and Refueling Water Tank are NOT available for makeup to the RCS

• Pressurizer level is 26% and slowly lowering, based on the Pzr Level Accuracy vs. Temperature curve and Pzr Level Cold Cal instrumentation (LI-1103)

Based on the conditions above, Unit 2 must be in HOT SHUTDOWN within the following

__(1)___ in accordance with Technical Specifications 3.4.3, Pressurizer.

In accordance with 2-AOP-01.13, Natural Circulation Cooldown, the Unit Supervisor should direct RCS make up alignment FROM the Safety Injection tanks DIRECTLY to the __(2)___.

A. 1) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

2) suction of the Charging Pumps B. 1) 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

2) Volume Control Tank C. 1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

2) suction of the Charging Pumps D. 1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

2) Volume Control Tank Page 85 of 100

HLC 22 SRO NRC EXAM Question 86 Given the following:

• Unit 1 is in Mode 3

• The Reactor Coolant System is at normal operating temperature and pressure

• A FAILED SOLENOID on the air supply line to 1-HCV-14-1, CCW to the Reactor Coolant Pump (RCPs), has caused a loss of instrument air to 1-HCV-14-1

• The Unit Supervisor directs the opening of 1-HCV-14-1 by installing a nitrogen jumper bypassing the failed solenoid in accordance with 1-NOP-01.02, RCP Operation, section 5.2, Local Restoration of CCW to RCPs Which ONE of the following completes the statements below:

1) What is the impact of the loss of instrument air on RCP operation?

All RCPs _(1)_.

2) Following the above system alteration, 1-HCV-14-1__(2)__ per Technical Specification 3.6.3.1, Containment Isolation Valves A. 1) must be secured due to a total loss of CCW to the RCPs

2) is considered OPERABLE B. 1) may continue to run, but CCW flow to the RCPs will be reduced

2) is considered OPERABLE C. 1) may continue to run, but CCW flow to the RCPs will be reduced

2) must be considered INOPERABLE D. 1) must be secured due to a total loss of CCW to the RCPs

2) must be considered INOPERABLE Page 86 of 100

HLC 22 SRO NRC EXAM Question 87 Unit 2 was operating at 100% power. Given the following events and conditions on 8/20 at 1200:

• Proportional heater bank P1 breaker failed

• Surveillance 2-OSP-100.02 Schedule of Periodic Tests, Checks and Calibrations Week 2 step 4.4.2.1 (Thursday) was conducted in order to verify adequate Pressurizer Heater Capacity

• The Distributed Control System (DCS) direct reading point for Backup Heater B1 kW indicates 149 kW Which ONE of the following statements correctly describes the required maintenance actions to allow continued operation at 100%?

Schedule maintenance to repair and have the:

A. backup heater group B1 made operable NO later than 8/20 at 1800.

B. backup heater group B1 made operable NO later than 8/23 at 1200.

C. proportional heater group made operable NO later than 8/20 at 1800.

D. proportional heater group made operable NO later than 8/23 at 1200.

Page 87 of 100

HLC 22 SRO NRC EXAM Question 88 Unit 2 is at 100% power.

• LIS-07-2A, Refueling Water Tank (RWT) level instrument, failed low and has been placed in BYPASS within one hour

1) If OPERABILITY was NOT RESTORED within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the required action IAW Tech Spec 3.3.2 ESFAS Instrumentation would be to __(1)__.

AND

2) ONE of the Bases for the above action is that if __(2)__?

A. 1) place it in the tripped condition; power operation can continue

2) an additional non-associated 120V Vital AC Instrument bus was lost, concurrent with a LOCA, RAS could fail to respond B. 1) place it in the tripped condition; power operation can continue

2) a second RWT level channel failed, RAS actuation could occur prematurely and align ECCS pumps to an inadequate suction source during accident conditions C. 1) perform a plant shutdown

2) an additional non-associated 120V Vital AC Instrument bus was lost, concurrent with a LOCA, RAS could fail to respond D. 1) perform a plant shutdown

2) a second RWT level channel failed, RAS actuation could occur prematurely and align ECCS pumps to an inadequate suction source during accident conditions Page 88 of 100

HLC 22 SRO NRC EXAM Question 89 Given the following on Unit 2:

• The unit is at 100% power

• The 2A & 2B Component Cooling Water (CCW) Hx outlet temperatures rose from 88°F to 98°F over the last hour due to high DP conditions in CCW HX strainers and high ocean injection temperatures

• The following annunciators have alarmed:

CNTMT FAN CLR CNTMT FAN CLR HVS-1A/1B HVS-1C/1D TEMP HIGH T-9 TEMP HIGH T-15

• The standby Containment Fan Cooler had been started IAW 2-NOP-25.04, Containment Fan Cooler Operations, but has just tripped

• 2-AOP-25.01, Loss of RCB Cooling Fans, has been entered

• Containment temperature is 121°F and slowly rising Which ONE of the following identifies:

1) The action that should be directed per 2-AOP-25.01 "Loss of RCB Cooling Fans" after the reactor is tripped?

Commence a cooldown to:

2) The Bases for the action?

To limit containment air temperature so that:

A. 1) COLD SHUTDOWN ONLY.

2) Containment Vessel Temperature does not exceed its design temperature of 264°F during steam line break and LOCA conditions.

B. 1) COLD SHUTDOWN ONLY.

2) maintain the Reactor Vessel Support Structure within its design basis.

C. 1) AT LEAST HOT SHUTDOWN.

2) Containment Vessel Temperature does not exceed its design temperature of 264°F during steam line break conditions and LOCA conditions.

D. 1) AT LEAST HOT SHUTDOWN.

2) maintain the Reactor Vessel Support Structure within its design basis.

Page 89 of 100

HLC 22 SRO NRC EXAM Question 90 Unit 1 is at 100% power. The crew has entered 1-AOP-08.02, Steam Generator Tube Leak and has been logging the following secondary radiation monitor readings every 15 minutes.

Time: 0105 0120 0135 Blowdown: A S/G 980 cpm A S/G 1090 cpm A S/G 1620 cpm B S/G 120 cpm B S/G 125 cpm B S/G 125 cpm Air Ejector: 420 cpm 720 cpm 920 cpm Based on the above readings, which ONE of the following states THE EARLIEST TIME for the following actions to be directed IAW1-AOP-08.02, Steam Generator Tube Leak

1) Performing a Rapid Downpower AND

2) Entering Mode 3 (Reference Provided)

A. 1) 0135

2) 0735 B. 1) 0120

2) 0735 C. 1) 0120

2) 0420 D. 1) 0135

2) 0420 Page 90 of 100

HLC 22 SRO NRC EXAM Question 91 Given the following on Unit 2:

• The unit is at 3% power

• Main Feedwater and the Steam Bypass Control System are in service

• Power is being raised to 12% to place the main generator on line IAW 2-GOP-201, Reactor Plant Start Up - Mode 2 to Mode 1

• The following annunciators have alarmed:

REACTOR LOCAL POWER TM/LP POWER HIGH DENSITY CHANNEL CHANNEL TRIP CHANNEL TRIP TRIP L-9 L-22 L-36

• The Board RCO reports A Channel Axial Shape Index is off-scale HIGH POSITIVE

• The ONE hour Technical Specification required actions have been completed Complete the following statements:

The __(1)__ on the affected linear range nuclear instrument.

The Unit __(2)__.

A. 1) Lower detector has failed high

2) Up Power can continue to 12%

B. 1) Upper detector has failed high

2) Up Power can continue to 12%

C. 1) Lower detector has failed high

2) must be stabilized until the instrument is repaired D. 1) Upper detector has failed high

2) must be stabilized until the instrument is repaired Page 91 of 100

HLC 22 SRO NRC EXAM Question 92 Given the following conditions on Unit 1:

• The unit is in Mode 3 and stable

• The Containment Atmosphere PARTICULATE radioactivity monitor has been declared inoperable 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> ago.

1) If a loss of instrument air occurred, due to a ruptured airline in Containment, the Containment Atmosphere GASEOUS Radioactivity monitor __(1)__ be AFFECTED?

2) Based on that determination, an inventory balance __(2)__ a containment atmosphere grab sample, must be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to satisfy the RCS Leakage Detection system LCO 3.4.6.1, RCS Leakage Detection Systems per Ops Policy 503, Technical Specification Guidance?

(Reference Provided)

A. 1) WOULD NOT

2) AND B. 1) WOULD NOT

2) OR C. 1) WOULD

2) AND D. 1) WOULD

2) OR Page 92 of 100

HLC 22 SRO NRC EXAM Question 93 Given the following information:

• Unit 2 is at 100% power

• A team has entered the Unit 2 Containment for a valve inspection

• A blown fuse results in a loss of the power to RIS-26-3, A Channel Containment (CIS) Radiation Monitor

• The crew has entered 2-AOP-26.02, Area Radiation Monitors

• The crew has bypassed the failed ESFAS channel Which ONE of the following completes the statements below:

The containment evacuation alarm __(1)__ automatically upon the fuse failure.

Per the Unit 2 Technical Specifications, if RIS-26-3 is not restored to an operable status WITHIN 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the A CIS Radiation bistable __(2)__; restore to operability by the next Cold Shutdown.

A. 1) sounded

2) MAY be maintained in bypass B. 1) did not sound

2) MAY be maintained in bypass C. 1) sounded

2) MUST be placed in trip D. 1) did not sound

2) MUST be placed in trip Page 93 of 100

HLC 22 SRO NRC EXAM Question 94 The design bases event for Limiting Condition of Operation 3.1.1.1, Shutdown Margin, is:

A. Positive reactivity addition resulting from a Rod Ejection event at beginning of core life from 0% power conditions.

B. Positive reactivity addition resulting from a Rod Ejection event at end of core life from 100% power conditions.

C. Excessive cooldown resulting from a Main Steam Break at beginning of core life from 100% power conditions.

D. Excessive cooldown resulting from a Main Steam Break at end of core life from 0%

power conditions.

Page 94 of 100

HLC 22 SRO NRC EXAM Question 95 In accordance with 2-GOP-365 Refueling Operations, to perform Core Alterations on Unit 2, minimum Reactor Vessel water level must be:

1) 23 feet above the ___(1)____.

2) The Technical Specification basis for this minimum water level is to ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of ___(2)____ fuel assembly.

Which ONE of the following completes the statements above:

A. 1) reactor vessel flange

2) an irradiated B. 1) reactor vessel flange

2) ONLY a RECENTLY irradiated C. 1) fuel

2) ONLY a RECENTLY irradiated D. 1) fuel

2) an irradiated Page 95 of 100

HLC 22 SRO NRC EXAM Question 96 Given the following information:

• Unit 1 is at 100% power

• There is a planned cable repair on Containment Fan Cooler 1-HVS-1A

• The planned repair duration is 4 days of the allowed 7 day Limiting Condition for Operation Which ONE of the following completes the statement below:

In accordance with WM-AA-100-1000, Work Activity Risk Management, the work control process would classify this repair activity as ________, a Risk Management Plan ________ mandatory.

A. High Risk; IS B. High Risk; IS NOT C. Medium Risk; IS D. Medium Risk; IS NOT Page 96 of 100

HLC 22 SRO NRC EXAM Question 97 Given the following:

• Unit 1 is in Mode 4 making preparations to enter Mode 3 following a 40 day refueling outage In accordance with TS 3.7.1.2, Auxiliary Feedwater System and Ops Policy 503, Technical Specification Guidance complete the following:

Mode 3 is allowed to be entered providing 1-OSP-09.01C, 1C AFW Pump Code Run is performed within __(1)__.

• Once the 1C AFW Pump surveillance was commenced, the following annunciator comes in and REMAINS lit:

1C AFW Pump Turbine Failure/Trip/

SS Isol G-46 RTGB-202 indication for MV-08-3, 1C AFW Pump Throttle/Trip indicates DUAL position.

The cause of the above alarm is due to a(n) __(2)__.

Which ONE of the following completes the statements above:

A. 1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

2) Electrical overspeed trip B. 1) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

2) Mechanical overspeed trip C. 1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

2) Electrical overspeed trip D. 1) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

2) Mechanical overspeed trip Page 97 of 100

HLC 22 SRO NRC EXAM Question 98 Given the following conditions on Unit 1:

TIME 2000, March 16th

• A Work Execution Coordinator (WEC) SRO is performing an observation of a task in a contaminated work area

• The workers performing the task reported that the WEC SRO is acting erratically and believe he may not be fit for duty

• The WEC SRO falls on the floor and suffers an injury that requires immediate transportation to an offsite medical facility

• The fall caused the individuals Protective Clothing to tear exposing his skin

• During transport, contamination surveys of the WEC SROs exposed skin indicated a value of 2500 DPM (beta-gamma)

TIME 2300, March 16th

• The hospital reported to the Shift Manager that the injured individuals BAC was 0.05% as determined from a confirmatory blood test drawn at 2100

• Investigation revealed that the individuals consumption of alcohol did NOT occur at work that night Given the conditions above, which ONE of the following is (are) the required NRC notification (s), if any?

(Reference Provided)

A. A contaminated injured person being transported offsite ONLY.

B. A Fitness for Duty violation ONLY.

C. A contaminated injured person being transported offsite AND a Fitness for Duty violation.

D. No NRC notifications are required.

Page 98 of 100

HLC 22 SRO NRC EXAM Question 99 Unit 1 has entered EOP-06, Total Loss of Feedwater, with the following indications:

• 1A Steam Generator (SG) level is 21% wide range

• 1B Steam Generator (SG) level is 13% wide range

• With the Atmospheric Dump Valves (ADVs) 100% open, Tcold has increased from 535°F to 541°F over the past few minutes Which ONE of the following describes the course of action that the crew should implement AND the reason?

A. Based on SG levels, initiate once-through cooling by referring to EOP-15, Functional Recovery Success Path HR-3 while remaining in EOP-06.

B. Based on Tcold rise, exit EOP-06, go to EOP-15, Functional Recovery Step 1 and initiate once-through cooling when the Success Path HR-3 step is reached.

C. Based on Tcold rise, initiate once-through cooling by referring to EOP-15, Functional Recovery Success Path HR-3 then exiting EOP-06 and entering EOP-15 at step1.

D. Based on SG levels, initiate once-through cooling by referring to EOP-15, Functional Recovery Success Path HR-3 then exiting EOP-06 and entering EOP-15 at step1.

Page 99 of 100

HLC 22 SRO NRC EXAM Question 100

• A General Emergency has been declared by the Emergency Coordinator for Unit 1

• A Steam Generator Tube Rupture has occurred on the 1A Steam Generator (SG) along with other complications

• Off-Site Dose Calculations are being performed

• The following meteorological data has been collected:

o 10 meter wind direction is 170° o 57.9 meter wind direction is 168°

• The Control Room actions for 1-EOP-99 Appendix R, Steam Generator Isolation have been completed with the exception that MV-08-13, SG 1A to Auxiliary FeedWater (AFW) Pump 1C DID NOT CLOSE

• The NPO has NOT performed ANY local actions of 1-EOP-99 Appendix R for the 1A SG

• The 1C AFW Pump is operating to provide AFW flow to the 1B SG Which ONE of the choices below identifies:

1) the affected sectors for the protective action recommendations?

2) whether a release is occurring?

Wind Sectors Wind Sectors Wind Sectors From Affected From Affected From Affected 348-11 HLK 123-146 PQR 236-258 CDE 11-33 JKL 146-168 QRA 258-281 DEF 33-56 KLM 168-191 RAB 281-303 EFG 56-78 LMN 191-213 ABC 303-326 FGH 78-101 MNP 213-236 BCD 326-348 GHU 101-123 NPQ There is no 0 sector There is no I sector A. 1) QRAB

2) A release IS occurring B. 1) RAB

2) A release IS occurring C 1) QRAB

2) A release IS NOT occurring D. 1) RAB

2) A release IS NOT occurring Page 100 of 100

St. Lucie NRC Exam HLC-22 ROISRO Answer Key Question Answer

1. C

2. B

3. B

4. C

5. D

6. B

7. C

8. B

9. D

10. A

11. C

12. C

13. A

14. A

15. B

16. A

17. D

18. B

19. C

20. B

21. B

22. D

23. A

24. B

25. C

26. B

27. A

28. C

29. B

30. B

31. A 32.

33. B

34. A

35. A

36. D

37. C

38. C 5jei 1 tes+i-.. U 3 cJ..&-jcd 1 i

c C LCrriz*S.

Rev 13/23/20158:13:17 AM

St. Lucie NRC Exam HLC-22 Question Answer New PSL Modified Comp Memory Bank PSL Bank

39. C

40. D

41. B

42. B

43. D

44. B

45. A

46. C

47. D

48. A

49. C

50. D

51. A

52. D

53. B

54. D

55. C

56. C

57. C

58. B

59. B

60. A

61. A

62. B

63. A

64. D

65. C

66. D

67. D

68. B

69. B

70. C

71. D

72. A

73. A

74. D

75. B SRO Exam Rev 13/23/20158:40:24 AM

St. Lucie NRC Exam HLC-22 Question Answer

80. C

81. C

82. B

83. D

84. B

85. B

86. D

87. B

88. D

89. D

90. C

91. A

92. D

93. A

94. D

95. A

96. A

97. B

98. C

99. C 100. B Rev 1 3/23/20158:14:32 AM

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REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 19 of 31 PROCEDURE NO.:

1-AOP-03.02 ST. LUCIE UNIT 1 ATTACHMENT I Estimated Time to Core Boiling (Page 1 of 5)

DETERMINE estimated time to core boiling as follows (Section 6.2 Commitment 2):

A. RECORD the following information at the time SDC was lost:

Time Shutdown Cooling Lost from Section 4.2 Step 2 RCS Temperature from Section 4.2 Step 2 °F Vessel / Cavity Level ft Time since shutdown (hours)

B. RECORD time to boil at mid-loop per Table 1:

(minutes)

C. RECORD additional time to boil for a 100 ft 3 volume per Table 2:

(minutes /100 ft ).

3 D. RECORD additional 100 ft 3 volumes above mid-loop per Table 3:

(100 ft

)

3 E. CALCULATE Time to Boil as follows:

+ x = *min Time to boil Additional time to boil Additionall00 3 ft volumes Time to Boil Stepl.B Stepl.C Stepl.D

2. IF core shuffle or reload has NOT been competed, THEN CALCULATE time boiling will occur as follows:

+ =

Time SDC Lost Time to Boil Time Boiling Will Occur Stepl.A Stepl.E

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 20 of 31 PROCEDURE NO.:

1-AOP-03.02 ST. LUCIE UNIT 1 ATTACHMENT I Estimated Time to Core Boiling (Page 2 of 5)

3. IF core shuffle or reload has been completed, THEN CALCULATE time boiling will occur as follows (Section 6.1 .3 Management Directive 1):

x(1.35) =

Time SDC Lost Time Boiling Will Occur Stepl.A

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 21 of 31 PROCEDURE NO.:

1-AOP-03.02 ST. LUCIE UNIT 1 ATTACHMENT I Estimated Time to Core Boiling (Page 3of5)

NOTE

. Temperatures between columns, select higher temperature.

• Time since shutdown between rows, select the lesser time.

Table 1 Time to Boil (minutes) at Mid-Loop RCS Temperature °F)

Shutdown Time 60 80 90 100 110 120 130 140 200 (Hours) 24 13.1 11.4 10.5 9.64 8.77 7.90 7.03 6.17 1.08 48 16.0 13.9 12.8 11.8 10.7 9.66 8.60 7.54 1.32 60 17.3 15.0 13.9 12.7 11.6 10.4 9.30 8.16 1.43 72 18.6 16.1 14.9 137 12.4 11.2 9.98 8.76 1.54 90 20.4 17.7 16.4 15.0 13.7 12.3 11.0 9.62 1.69 102 21.6 18.7 17.3 15.9 14.4 13.0 11.6 10.2 1.78 120 23.2 20.1 18.6 17.0 15.5 14.0 12.4 10.9 1.91 132 24.2 21.0 19.4 17.8 16.2 14.6 13.0 11.4 2.00 144 25.1 21.8 20.1 185 16.8 15.1 13.5 11.8 2.07 240 31.0 26.9 24.8 22.8 20.7 18.7 16.6 14.6 2.56 300 33.8 29.3 27.0 24.8 22.6 20.3 18.1 15.9 2.79 360 36.3 31.5 29.0 26.6 24.2 21.8 19.4 17.1 2.99 480 40.9 35.5 32.8 30.1 27.3 24.6 21.9 19.3 3.38 600 45.4 39.4 36.4 33.3 30.4 27.3 24.3 21.4 3.75 840 54.2 47.0 43.4 39.8 36.2 32.6 29.0 25.5 4.47 1440 73.9 64.1 59.2 54.3 49.4 44.5 39.6 34.8 6.10

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 22 of 31 PROCEDURE NO.:

1-AOP-03.02 ST. LUCIE UNIT 1 ATTACHMENT I Estimated Time to Core Boiling (Page 4 of 5)

NOTE Temperatures between columns, select higher temperature.

Time since shutdown between rows, select the lesser time.

Table 2 Additional Time To Boil (minutes) for Each 100 cubic feet of RCS Volume (Section 6.1.3 Management Directive 2)

RCS Temperature (°F Shutdown Time 60 80 90 100 110 120 130 140 200 (Hours) 24 0.913 0.792 0.731 0.671 0.610 0.550 0.490 0.430 0.0754 48 1.12 0.968 0.893 0.819 0.746 0.672 0.598 0.525 0.0921 60 1.21 1.05 0.967 0.887 0.807 0.727 0.647 0.568 0.0996 72 1.30 1.12 1.04 0.952 0.866 0.781 0.695 0.610 0.107 90 1.42 1.23 1.14 1.04 0.951 0.857 0.763 0.669 0.117 102 1.50 1.30 1.20 1.10 1.00 0.905 0.805 0.707 0.124 120 1.61 1.40 1.29 1.19 1.08 0.972 0.865 0.760 0.133 132 1.68 1.46 1.35 1.24 1.13 1.01 0.902 0.792 0.139 144 1.75 1.52 1.40 1.28 1.17 1.05 0.937 0.823 0.144 240 2.16 1.87 1.73 1.58 1.44 1.30 1.16 1.01 0.178 300 2.35 2.04 1.88 1.73 1.57 1.42 1.26 1.11 0.194 360 2.52 2.19 2.02 1.85 1.69 1.52 1.35 1.19 0.208 480 2.85 2.47 2.28 2.09 1.90 1.71 1.53 1.34 0.235 600 3.16 2.74 2.53 2.32 2.11 1.90 1.69 1.49 0.261 840 3.77 3.27 3.02 2.77 2.52 2.27 2.02 1.77 0.311 1440 5.14 4.46 4.12 3.78 3.44 3.10 2.76 2.42 0.424

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 23 of 31 PROCEDURE NO.:

1-AOP-03.02 ST. LUCIE UNIT 1 ATTACHMENT I Estimated Time to Core Boiling (Page 5of5)

NOTE If elevation is between rows, select the lesser value.

Table 3 Additional Volume (100 cubic feet) Above Mid-Loop (Section 6.1.3 Management Directive 2)

Vessel Head Vessel Head Vessel Head Installed or Not Installed Not Installed Not Installed Elevation Additional Elevation Additional Elevation Additional Volume Volume Volume Feet 100 ft3 Feet 100 ft3 Feet 100 ft3 29.5 0.00 36.5 14.20 48.5 229.00 30.0 0.40 37.0 23.15 49.0 237.95 30.5 0.81 37.5 32.10 49.5 246.90 31.0 1.21 38.0 41.05 50.0 255.85 31.5 1.61 38.5 50.00 50.5 264.80 32.0 2.02 39.0 58.95 51.0 273.75 32.5 2.42 39.5 67.90 51.5 282.70 33.0 2.82 40.0 76.85 52.0 291.65 33.5 3.23 40.5 85.80 52.5 300.60 34.0 3.63 41.0 94.75 53.0 309.55 34.5 4.04 41.5 103.70 53.5 318.50 35.0 4.44 42.0 112.65 54.0 327.45 35.5 4.84 42.5 121.60 54.5 336.40 36.0 5.25 43.0 130.55 55.0 345.35 43.5 139.50 55.5 354.30 44.0 148.45 56.0 363.25 44.5 157.40 56.5 372.70 45.0 166.35 57.0 381.15 45.5 175.30 57.5 390.10 46.0 184.25 58.0 399.05 46.5 193.20 58.5 408.00 47.0 202.15 59.0 416.95 47.5 211.10 59.5 425.90 48.0 220.05 60.0 434.85

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 18 of 37 PROCEDURE NO.:

2-AOP-03.02 ST. LUCIE UNIT 2 ATTACHMENT I Estimated Time to Core Boiling (Page 1 of 4)

I. DETERMINE estimated time to core boiling as follows (Section 6.2 Commitment 3):

A. RECORD the following information at the time SDC was lost:

Time Shutdown Cooling Lost from Section 4.2 Step 2 RCS Temperature from Section 4.2 Step 2 °F Vessel / Cavity Level ft Time since shutdown (hours)

B. RECORD time to boil at mid-loop per Table 1:

(minutes)

C. RECORD additional time to boil for a 100 ft 3 volume per Table 2:

(minutes /100 ft ).

3 D. RECORD additional 100 3 ft volumes above mid-loop per Table 3:

(100 ft

)

3 E. CALCULATE Time to Boil as follows:

+ .. . . x mm Time to boil Additional time to boil Additional 100 ft 3 volumes Time to Boil Step 1 .B Step! .C Step 1 .D

2. IF core shuffle or reload has NOT been competed, THEN CALCULATE time boiling will occur as follows:

+ =

Time SDC Lost Time Boiling Will Occur Stepl.A

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 19 of 37 PROCEDURE NO.:

2-AOP-03.02 ST. LUCIE UNIT 2 ATTACHMENT I Estimated Time to Core Boiling (Page 2 of 4)

3. IF core shuffle or reload has been completed, THEN CALCULATE time boiling will occur as follows (Section 6.1 .3 Management Directive 1):

+ x(l.35) =

Time SDC Lost Time to Boil Time Boiling Will Occur Stepl.A Step i.E NOTE Temperatures between columns, select higher temperature.

Time since shutdown between rows, select the lesser time.

Table 1 Time to Boil (minutes) at Mid-Loop RCS Temperature (°F)

Shutdown Time (Hours) 60 80 90 100 110 120 130 140 200 24 11.9 10.4 9.57 8.77 7.99 7.20 6.40 5.62 0.986 48 14.6 12.7 11.7 10.7 9.76 8.79 7.83 6.87 1.20 60 15.8 13.7 12.6 11.6 10.6 9.51 8.47 7.43 1.30 72 17.0 14.7 13.6 12.5 11.3 10.2 9.09 7.98 1.40 90 18.6 16.1 14.9 13.7 12.4 11.2 9.98 8.76 1.54 102 19.7 17.0 15.7 14.4 13.1 11.8 10.5 9.25 1.62 120 21.1 18.3 16.9 15.5 14.1 12.7 11.3 9.94 1.74 132 22.0 19.1 17.6 16.2 14.7 13.3 11.8 10.4 1.82 144 22.9 19.8 18.3 16.8 15.3 13.8 12.3 10.8 1.89 240 28.2 24.5 22.6 20.7 18.9 17.0 15.1 13.3 2.33 300 30.7 26.7 24.6 22.6 20.6 18.5 16.5 14.5 2.54 360 33.0 28.6 26.4 24.3 22.1 19.9 17.7 15.5 2.73 480 37.2 32.3 29.8 27.4 24.9 22.4 20.0 17.5 3.07 600 41.3 35.9 33.1 30.4 27.6 24.9 22.2 19.5 3.41 840 49.3 42.8 39.5 36.2 33.0 29.7 26.4 23.2 4.07 1440 67.3 58.4 53.9 49.4 45.0 40.5 36.1 31.7 5.55

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 20 Of 37 PROCEDURE NO.:

2-AOP-03.02 ST. LUCIE UNIT 2 ATTACHMENT I Estimated Time to Core Boiling (Page 3 of 4)

NOTE

• Temperatures between columns, select higher temperature.

• Time since shutdown between rows, select the lesser time.

Table 2 Additional Time To Boil (minutes) for Each 100 cuc feet of RCS Volume RCS Temperature (°F Shutdown Time (hours) 60 80 90 100 110 120 130 140 200 24 0.913 0.792 0.731 0.671 0.610 0.550 0.490 0.430 0.0754 48 1.12 0.968 0.893 0.819 0.746 0.672 0.598 0.525 0.0921 60 1.21 1.05 0.967 0.887 0.807 0.727 0.647 0.568 0.0996 72 1.30 1.12 1.04 0.952 0.866 0.781 0.695 0.610 0.107 90 1.42 1.23 1.14 1.04 0.951 0.857 0.763 0.669 0.117 102 1.50 1.30 1.20 1.10 1.00 0.905 0.805 0.707 0.124 120 1.61 1.40 1.29 1.19 1.08 0.972 0.865 0.760 0.133 132 1.68 1.46 1.35 1.24 1.13 1.01 0.902 0.792 0.139 144 1.75 1.52 1.40 1.28 1.17 1.05 0.937 0.823 0.144 240 2.16 1.87 1.73 1.58 1.44 1.30 1.16 1.01 0.178 300 2.35 2.04 1.88 1.73 1.57 1.42 1.26 1.11 0.194 360 2.52 2.19 2.02 1.85 1.69 1.52 1.35 1.19 0.208 480 2.85 2.47 2.28 2.09 1.90 1.71 1.53 1.34 0.235 600 3.16 2.74 2.53 2.32 2.11 1.90 1.69 1.49 0.261 840 3.77 3.27 3.02 2.77 2.52 2.27 2.02 1.77 0.311 1440 5.14 4.46 4.12 3.78 3.44 3.10 2.76 2.42 0.424

REVISION NO.: PROCEDURE TITLE: PAGE:

SHUTDOWN COOLING ABNORMAL OPERATIONS 21 of 37 PROCEDURE NO.:

2-AOP-03.02 ST. LUCIE UNIT 2 ATTACHMENT I Estimated Time to Core Boiling (Page 4of4)

NOTE If elevation is between rows, select the lesser value.

Table 3 Additional Volume (100 cubic feet) Above Mid-Loop Vessel Head Installed Vessel Head Not Vessel Head Not or Not Installed Installed Installed Additional Additional Additional Elevation Elevation .

Elevation Volume Volume Volume Feet 100 ft3 Feet 100 ft3 Feet 100 ft3 29.5 0.00 36.5 14.20 48.5 229.00 30 0.40 37 23.15 49 237.95 30.5 0.81 37.5 32.10 49.5 246.90 31 1.21 38 41.05 50 255.85 31.5 1.61 38.5 50.00 50.5 264.80 32 2.02 39 58.95 51 273.75 32.5 2.42 39.5 67.90 51.5 282.70 33 2.82 40 76.85 52 291.65 33.5 3.23 40.5 85.80 52.5 300.60 34 3.63 41 94.75 53 309.55 34.5 4.04 41.5 103.70 53.5 318.50 35 4.44 42 112.65 54 327.45 35.5 4.84 42.5 121.60 54.5 336.40 36 5.25 43 130.55 55 345.35 43.5 139.50 55.5 354.30 44 148.45 56 363.25 44.5 157.40 56.5 372.70 45 166.35 57 381.15 45.5 175.30 57.5 390.10 46 184.25 58 399.05 46.5 193.20 58.5 408.00 47 202.15 59 416.95 47.5 211.10 59.5 425.90 48 220.05 60 434.85

a C

3 0

C C

a aI in I

REVISION NO.: PROCEDURE TITLE: PAGE:

53 PPENDICES I FIGURES I TABLES I DATA SHEETS PROCEDURE NO.: 132 of 176 1-EOP-99 ST. LUCIE UNIT 1 FIGURE IA RCS PRESSURE TEMPERATURE (Page 1 of 1)

(Containment Temperature Less Than or Equal to 200°F)

CAUTION The RCP NPSH curve assumes instrumentation should be monito pp

[dance with tEOP99, Table 13.

2200 . An Operator Aid has been placed in the Unit I Control Room RCO desk (Any revision to this section of the procedure shall verify the validity 2000 of the Operator Aid and, if changes are necessary, notify the EOP Coordinator).

1800 1600 .

Acceptable Range z 140n 0.

. 1200 -

I I

Maximum Minimum Subcooled° Subcooled 0.

- 1000

• 0

.E 800-

/ 1/

600 400 200 Shut

°Includcs Instrument Uncertainties 0

0 100 I I 4 200 300 400 500 600 700 800 Indicated RCS Temperature (°F)

RCS Pressure Range Required QSPDS Subcooled Margin Reading (Rep CET) 2250 psia to 1000 psia 40 to 180°F 1000 psia to 500 psia 50 to 170°F Less than 500 psia 80 to 160°F

REVISION NO.: PROCEDURE TITLE:

53 PPENDICES I FIGURES / TABLES I DATA SHEETSI PROCEDURE NO.: I 133 of 176 1-EOP-99 ST. LUCIE UNIT 1 FIGURE lB RCS PRESSURE TEMPERATURE (Page 1 of 1)

(Containment Temperature Greater Than 200°F)

CAUTION The RCP NPSH curve assumes one pump is operating in each loop. RCP instrumentation should be monitored for seal and pump performance in accordance with 1-EOP-99, Table 13.

U U,

a z

U, I

a, 0

a, N

U, U,

a, 0

a, U

I 0 100 200 300 400 500 600 700 800 Indicated RCS Temperature (°F) .001S SQl S0.,a OR:. S

REVISION NO.: PROCEDURE TITLE: PAGE:

53 PPEND ICES / FIGURES I TABLES I DATA SHEETS PROCEDURE NO.: 134 of 176 1-EOP-99 ST. LUCIE UNIT 1 FIGURE 2 SAFETY INJECTION FLOW VS. RCS PRESSURE (Page 1 of 1) 1300 1200 1100 1000 900 VI U 800 VI VI U

I 0

700 U

N I 600 VI VI w 500 0

400 300 200 100 0

0 1000 2000 3000 4000 5000 6000 Total Safety Injection Flow (gpm) P/FQP/-FOP 9FlJ2,.lncV

REVISION NO.: PROCEDURE TITLE: PAGE:

50 APPENDICES / FIGURES I TABLES / DATA PROCEDURE NO.: 126 of 171 SHEETS 2-EOP-99 ST. LUCIE UNIT 2 FIGURE IA RCS PRESSURE TEMPERATURE (Page 1 ofi)

(Containment Temperature Less Than or Equal to 200°F)

CAUTION The RCP NPSH cue assumes instrumentation should be monit

[dance with 2EOP, Tab 13.

NUtS 2400 An Operator Ard ha, been placed or hp Un) 2 Control Ronnr RCO dock (Any rcpeion 1° tb, section piPe procedure shall perrfy the validry of the Operator Aid and, if chaegeo are necencary. noldy foe EOP Coordirurtor, 2200 2000 1800 Maximum Subcooled 1600 a

a) 1400 a

C) 1200 a) a

• 0 a) 1000 0

Minimum Subcooled 800 600 400 lpciudes Instrument Uncertarettes 200 0

0 100 200 300 400 500 600 700 800 nica1ed RCS Temperature (F) 1 prCpSQ.cc.p.har CRC RCS Pressure Range Required QSPDS Subcooled Margin Reading (Rep CET) 2250 psia to 1000 psia 40 to 180°F 1000 psia to 500 psia 50 to 170°F Less than 500 psia 80 to 160°F

REVISION NO.: PROCEDURE TITLE: PAGE:

50 APPENDICES / FIGURES I TABLES I DATA 127 of 171 PROCEDURE NO.: SHEETS 2-EOP-99 ST. LUCIE UNIT 2 FIGURE lB RCS PRESSURE TEMPERATURE (Page 1 of 1)

(Containment Temperature Greater Than 200°F)

CAUTION The RCP NPSH curve assumes one pump is operating in each loop. RCP instrumentation should be monitored for seal and pump performance in accordance with 2-EOP-99, Table 13.

2400 NOTE An Operator Aid has been placed in the Unit 2 Control Room RCO desk. (Any revoron to this section of the procedure shall verify the vakdity of the Operator Aid arid, if changes are necessary, notify I EOP Coordinator.(

2200 2000 1800

• 1.

1600 U) ci.

0) ci 0)

U) 11) 1400 0

0)

N ci U.)

1200 U) 0)

ci

• 0 0)

CD 1000 C.) Minimum

-o RCP NPSH Subcooled*

C 800 600 400 1ncludes Instrument Uncertainties 200 0

0 100 200 300 400 500 600 700 800 Indicated RCS Temperature (F) P.CPS EQP90g 8,Re2xf

REVISION NO.: PROCEDURE TITLE: PAGE:

50 APPENDICES I FIGURES / TABLES / DATA 128 of 171 PROCEDURE NO.: SHEETS 2-EOP-99 ST. LUCIE UNIT 2 FIGURE 2 SAFETY INJECTION FLOW VS. RCS PRESSURE (Page 1 of 1) 1200 NOTE 1100 This curve represents minimum expected SI Flow. If measured flow 1000 is less than this figure, then SI System lineup should be verified.

900 l1 a

w 800 z

a, I 700 0

I a,

N 600 Ill a, 500 a.

400 2 Full Trains

.-ln Operation 300 200 100 1 FullTrain :7 in Operation 0

0 1000 2000 3000 4000 5000 6000 Total Safety Injection Flow (gpm)

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for each Reactor Coolant System Pressure Isolation Valve identified in Table 3.4.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit in 3.4.6.2.e above reactor operation may continue provided that at least two valves, including check valves, in each high pressure line having a non-functional valve are in and remain in the mode corresponding to the isolated con dition. Motor operated valves shall be placed in the closed posi tion, and power supplies deenergized. (Note, however, that this may lead to ACTION requirements for systems involved.) Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ST. LUCIE - UNIT 1 3/4 4-14 Order dated 1/20/81 Amendment No. 4-1-8, 200

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,

c. *Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> except when operating in the shutdown cooling mode,

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

e. Verifying each Reactor Coolant System Pressure Isolation Valve leakage (Table 3.4.6-1) to be within limits:

1. Prior to entering MODE 2 after refueling,

2. Prior to entering MODE 2, whenever the plant has been in COLD SHUTDOWN for 7 days or more if leakage testing has not been performed in the previous 9 months,

3. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

4. The provision of Specification 4.0.4 is not applicable forentry into MODE 3or4.

f. Whenever integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in each high pressure line having a leaking valve shall be recorded daily; and

g. Primary-to-secondary leakage shall be verified 150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.**

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Not applicable to primary-to-secondary leakage.

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

ST. LUCIE - UNIT 1 3/4 4-14a Amendment No. 4-n, 200

TABLE 3.4 6-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Check Valve No.

V3227 V31 23 V3217 V3 113 V3237 V3133 V3247 V3143 V3 124 V31 14 V31 34 V3 144 NOTES (a) Maximum Allowable Leakage (each valve):

1. Leakage rates less than or equal to 1.0 gpm are acceptable.

2. Leakage rates greater than 1 .0 gpm but less than or equal to 5.0 gpm are acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount the reduces the margin between previous measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1 .0 gpm but less than or equal to 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are unacceptable.

(b) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

(c) Minimum test differential pressure shall not be less than 150 psid.

ST. LUCIE - UNIT 1 3/4 4-14b Order dated 4/20/81

501 Prolonged Loss of All OS-site and All On-Site AC Power 551 Less of All OS-site and All On-Site AC Power to Emergency SA5 AC Power Capability To Emergency Russes Reduced To A Single Power In Emergency Busses.

SU1 Loss of All OS-nile AC Power to Emergency Busses forth Busnos for fS misules or longer. Source Per f h Minutes or Longer Such ThaI Any Additional Single Failore Minutes or Losger.

ruor root Would Result hr Stafron Blacknal.

Operalieg Mode ApplicabIlity: 1, 2, 3, 4 Operalieg Mode Applicabdily: 2, 3, 4 r Operaling Mode Applluabilily: 1, 2, 3, 4 Dperalieg Mode Applicablldy: 1, 2, 3, 4 EAL Values: EAL Valaeu:

EAL Value: EAL Valaes:

Role Rare Role The Emergency Courdinafor shorrld not umd unhl lbs apphcable The Emergency Coordinator should not wart uoht the applicable hme yg 0 The Emergency Ceerdinufyr should net wail uoht fhe applicable yme has elapsed, hot ohoold declare fye ecevl 400000 and is elapsed, hot should declare the eceet as 500045 d is detemrincd that the rime has elapsed, but ohoold declare fhn ecenf as 5000 as d is defenrrirred that the csnddinn has ecceeded, or wry hkety exceed, condAon has exceeded. or wdt likely exceed Ihe applicable lime. deterrrriscd that the cooddion has ecceeded, or wE hkety esceed LU lye apphcable lime the applicable time

1. Loss of all Off-site AND all On-site AC power to A3 416KV 1. Less al all Off-srle AND all Oe-sile AC Power le A3 4 16 KV AND 0 AND 63 4.16 KVbusses. 1. AC power capability to A3 4.16KV AND 83 4.16KV busses reduced led single fLoss of as 06-site AC power to A3 4 16KV AND 634.16KV for 0*

634.16KV busses fur lh minutes or longer poascr soame tsr lb minules or lnngnr. 15 mrnules or longer. C, AND AND a- EIThER of Ac following:

(1) Restoration of at tenor one emergency bus is less than 4 a. ANY uddilional sivgle power source fuilure nell resuh lOd SI oboe Blackout.

hours is NOT likely OR (2) RCS and Core Heat Removal Safely function is NOT mel.

C,)

SG2 Aulornalic Tyy and All Macoar Actions Fail to Shaldowe 552 Automatic Triy Fails to Shuldows rho Reader AND Manual SA2 Autowulic Trip Fails In Shuldawu Ihe Reactor AND the Manual Actions Taken z Ihe Reactor AND ledicalion of an Eat rome Challenge to SUB Inudnedenr Crdicalfp.

Actions Takes horn Ike Reactor Turbine Onnerator Board from the Reuslor Turbine Generutur Rnurd )RTOR) are Ssccesoful in Shulheg 0 Ihe Abililylo Coot the Core Points. )RTGR) urn NOT Successlul in Shutting Down the Reactor. Down the Reuclor I 0

Operalieg Made Applluabilily: 1, 2 Dperutiug Made Applicability: 1, 2 Dperatieg Made Appliuabilitp: z 1, 2 Operutieg Mode Applicability: 3, 4 D z Ll PAL Values: EAL Vuloes: 0 PAL Values: EAL Values: -J I

1. AutomatIc trip failed Is sf uldown Ike reactor. Ac uulomolic Irip fulled to shutdown Ihe reactor 0 lAo Aolomulic Sty fulled to okuldown Ike reactor I. UNPLANNED 050luisod pnnihse stud-up rule ekoemed no LU AND nuclear lnslrumnnluhon. I AND AND 0 lii

a. ALL Manaal aclioss failed to shaldown roe reactor as indicated a. Manual actions taken at the Reactor Turbine Deneralor Board I

by:

a. Manual acfiees taken at Iha Reader Turbine Ouceralor Board )RTGR) 0

)RTOB) DO NOT shatdown the reactor as indicated by saccessfutly shutdown the reactor as indicated by ALL of the tnlluwing XC, >-

• Reactor power is NOT droypiog 10 less than b% power

• Reactor power is NOT dropping to less than 5% power 0

• Reactor power is dropping lo less than 5% power

• All CEAs am NOT inseded

• ALL full strength CERn are NOT inseded

• Negation slay-up rate

• All CEAs are insedod or boralion in progress l.t2 0 AND LU

6. EITHER of the following eoisl or have eccuved duels coolinued power generalios: D

)f) Core Heal Remnnal Safely Function NOT mel. -J DR U-(2) RCS Heat Rewesal Safely Funclien NOT mel.

553 Loss of All Vilol DC Power for IS Minutes or Longer.

Operalieg Made Appllcubililp: 1, 2, 3, 4 PAL Values:

UNPLANNED A parameter change or an enrol thur is nor the LU On-sire AC power maybe prodded by the nlher Unils Rote Emnrgency Diesel Generator )EDG) by successlul S-Ire Is The Emergency Coordinator should net wail ostil the apphcable resuk of an inlended onnlulion and requires correclino or egkerlhn A3 or 634.16KV bus. hme San elapsed, but should declare lye ecent as soon os a is wdiguline uslisns. 0 determined fhuf ISa eondrtion has eoceeded or wid hknly exceed, a

the apphcahle time. 0 0

1. Less than 112 VDC on t)2)A, 1(2)6 AND t)2)AR Vrul DC busses fortS minulro or longer.

SHOTCONDITIONS IS EAL-I-IOT BASIS PAGE REVISION I ST LUCIE PLANT CLASSIFICATION TOOL 5 HOTCONDITIONS IS

CG1 Loss of RCS Inventory Affecting Fuel Clad lntegtity with CSI Loss of RCS Insentonp Ahecting Core Decay Heat Remonal CAl Loss ut RCS tncenlorp. Cill RCS Leakage.

Containment Challenged. Capability.

mel Operating Mode Applicability: 5, 6 Operating Made Applinabitdp: 5, 6 Operating Made Applicability: 5, 6 Operating Mode Apptinabilitp: 5 EAL Values: hAL Values: EAL Values:

EAL Values:

Note Note ReM The Emergency Coordinator should nut waif volit the applicable The Enrergenny Coordinator should nut wed onht the applicable tiwn The Emengnucp Coord:natar should not wad onht the apyhcable Swn han Note hone has elapsed, but should ductore the count as soon as it is has elapsed but should declare the anent ao soon an it d determined - etapond, bat should dectare the event an soon and is determined that the The Emergency Caordiualur shnuld net wad until tIre apphca bIn detennnined that the cundihan wrIt hkety eucued the apphcahte that the naeddiae win Oheln auoeed ba aeohcabte time. I canddion idg hbetn exceed the aoohoable lime. lime has elapsed but should declare the event as noon and is hme determined that the csndthnn wrIt Phety eaoeed the applicable time.

1. Cnetainment challenge indicated Sb ANY 01 the following: 1. Wth CONTAINMENT CLOSURE NOT established RCS lend less t. RCS tend less than:

• CONTAINMENT CLOSURE NOT established than: 1. RCS leakage resohs in the inahitily to maintain or restore PZR 23.2 inches on LI-i it 7 OR Ll-t 1 t 7-1

• UNPLANNEO rise in containment pressure

• tO inches ou LI-it 17 OR LI-it 17-i level 2b% to 35% tot 15 minutes or tonoer

• Coctainmcnl Hydrogen grealerthan 4% OR OR 0 AND 0 z

2. RCS lend cannot he monitored fur t S minutes nn longer I 2, Wth CONTAINMENT CLOSURE established RCS lend lens than: CU2 UNPLANNED Loss of RCS Inventory. z 0 ANY one of the tnllowing applies:

• inches on Ll-t t t7 OR LI-it i7-t AND LU I Operating Made Applicability: 6

> ci

a. Cure onconnry for 30 minutes or lunger as indicated bp: OR a. Loss of RCS rnnnntnrp as indicated by uneuptained lend nse in ANY ol the z z totlnwtng: D RCS level less than: Note LL 3, RCS lend cannot be monitored tot 30 minnles or longer with a Ions ot

• Containment bump LU -J

• 5 inches on Lt-11t7 OR LI-1il7-t Reader Pressure Vessel innentorp as indicated by ANY of the The Emnrgeecy Cuordrnator should nut wad onht the apptica Ste

• Reactor Canitp bump hme has elapsed, bat should declare the eaent as 500rl usa is CD Pnttowing:

• Safeguards bump OR determiond that the oonditiun unIt hbety enceed the apphcahte

• Coclainment High Range Radiation Monitor reading grealer

• Holdup Tanks time.

than 1.3 End R.inr

• Refueling Waler Tank

b. RCS tenet cannot be monitored with cure unconery indicated by

• Erratic source range monitor Indicatise Ui LU ANY of the tnttowieg for 30 minutes or longer: I. UNPLANNED RCS lend drop au indicated bp EITHER of the -J I-.

• Unenptained teed rise in ANY of the tultowing 0

• Containment High Range Radiation Monitor reading iollawing: U)

• Containment bump >-

greater than 4.0 boA Rthn

• RCS water level drop below 00 inches on LI-lit y-i tori C.) 0

• Reactor Canitp bump

• Erratic source range monitor indication minutes or longer when the RCS tenel band is established Sateguards bump 0

• UNPLANNED tenet rise in ANY of the totluwing: aboon the Reactor Pressore Vessel flange.

• Holdup Tankn z

• Cnetainmnnt bump

• Refueling Water Tack

• RCS water Inoel drop Setous the RCS tenel Sand for iS minutes or lunger when the RCS lend band is established -J

• Reactor Candy bump LU

• Sateguardn bump below the Reactor Pressure Vessel flange.

D

• Holdup Tanbs LL

• Refueling Water Tank ON LU

2. RCS lenel cannot be monitored with a loss of RCS incentory as indicated Span oneoplained tenet dse in ANY of the following: z

• Containment bump

• Reactor Canilp Somp

• Safeguards bump 0

0

• HoMup Tanks

• Refueling Water Tank I

0 C43 Loss nt alt 00-site and all On-site AC Power In Emergency Bosses Fur th CU3 AC Power Capabrlrlp To Emergency Busses Reduced to a minutes nr Longer. 0 Single Power Source tort b Minolns or longer such that -J CONTAINMENT CLOSURE The procedurally deltned actions lust tsar Aep Addgionat Single Failure Would Result in Station Planned enolalinos to test, mantpalate, repair, yedarm maintenance 0 taken Is secure cnnlainment and its associated structures. or modifratines to sgstems and egoipment that resoti in an SAL value Blackoal.

spstems. and cemynnents as a t000lional barrier Ia fssinn being meter eocoeded are nut subject to classigcalian and acliuatinn product release under eoislrng plant cnsdrhsns. requrremenls as tong as Ihe esotutlan proceeds as planned and is Operating Mode Applinabititp: 5, 6, Defugled Operatieg Mode Applicabititp: 5, 6 1

Ci within the operational limtaliens imposed bp the specific operating UNPLANNED A parameter change on an enenl that is not the license. (St. Lode Tanhnical Basis, Section 3.BJ EAL Values:

SAL Values:

resuh stan intended evolution and regutres corrective or mitigalise actinns. Nate LU Nate The Emergency Coordinator should cot wail until the apphcabte hme has The Emergency Coordinator should not wart avId tIre apphcabte etapnerh but should declare the event an noon and io determined that the condition wilt hkety exceed the apphcabte hwe.

hme has elapsed. but oheald declare the enact as nuns and is 0 Fewer source meaen edher the A3 ar B3 4.16KV bus is belog ted determined that the canddioo wilt hhety eaceed the applicable 0 from aug eigble source. C) 1, Lens stall Ott-see AND all On-site AC Power to A3 dub KV AND 63 4.t6 KV bosses tot lb minutes or losget.

1. AC power capability to A3 atE KV AND 534.16EV reducad to a single power source tnr lb minutes or longer.

On-site AC puwer meg be prnsrded kg Ihe other Units Emengency AND Diesel Gennralur (EDS) bp successful X-tie to edherthe A3 or 63 4.16KV bus. a. ANY additional single power source taitare will result in Slatinn Blackout.

CCOLD CONDITIONS 12 EPIP-Ol F03 PAGE REVISION I ST LLICIE PLANT CLASSIFICATION TOOL C COLDCONDITIONS 12

INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown system transfer switches, control and instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE remote shutdown channels less than the Required Number of Channels shown in Table 3.3-9, either restore the inoperable channel to OPERABLE status within 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With the number of OPERABLE remote shutdown channels less than the Minimum Channels OPERABLE requirements of Table 3.3-9, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6.

4.3.3.5.2 Each remote shutdown system instrumentation transfer switch and control circuit shall be demonstrated OPERABLE by verifying its capability to perform its intended function(s) at least once per 18 months.

ST. LUCIE - UNIT 2 3/4 3-38

DAILY CHEMISTRY REPORT ST. LUCIE UNIT # I PRIMARY CHEMISTRY DATE: 10/3/2014 REACTOR COOLANT SYSTEM Parameter iJ Hydrogen j Surge Boron jts ppm I vorn I 0Db ppb oob uc,YrnI uci/mi I I I CC/kQ Dorn ILimits ee COP-05 04 for limits Result 534 2.35 1.18 0.62 < 5 1.67E-01 41.80 536 Date 101312014 101312014 101312014 101312014 101312014 101312014 101312014 10I112014 10I3I2014 913012014 RCS Boron Sample Time: 0:05 -131 1.20E-04

= -134 = 2.63E-03 Iodine Sample Date: 10/3/2014 Zinc Pump Stroke: 2.00 Iodine Ratio 1.24E-01

= Xe Ratio = 0.78 Chg Pmp 1 Xe Ratio >1.0 should be investigated for potential fuel defect RWST ---

SFP bAMF Parameter Sirs Boron Silica Boron Silica A Boron B Boron 1AI 1A2 IB1 1B2 Units ppm ppb ppm ppb ppm ppm ppm ppm ppm ppm Limits See COP-05.04 for limits Result 2130 20558 2181 32400 5512 5816 2072 2085 2085 2081 Date 912812014 91112014 1012/2014 101212014 912912014 912912014 101212014 1012/2014 101212014 10/2)2014 Time 7:12PM I 9:30 AM 9:30 AM I 9:30 AM I 9:30 AM SECONDARY CHEMISTRY I AM (ENERATQRS Parameter Chloride Fluoride Sulfate Cation Conductivity Sodium Gross Act WL) Rate Units ppb ppb ppb umhos/cm ppb uci/mi GPM Limits See COP-05.04 for limits Date 10/ZI2014 I 101212014 101212014 1013/2014 I 10/3/2014 I 10/312014 10/312014 lB Result 4b.31 Date 10/2l2014 10/2/2014 101212014 101312014 I 101312014 1013I2014 CONDENSATE lEEU Parameter Cation Conductivity Diss. 02 pH Ammonia Iron Copper Units umhosicm ppb ppm ppb ppb Limits See 0P-05.Q4 for limits Result 1.21 0.020 Date 10/312014 10/3/2014 10/3/2014 I 10)2/2014 9/30/2014 9130/2014 Unit I Projected Steam Generator Leak Rate Calculations(Info Only)

Using the Air Ejector Monitor Using the Blow Down Monitors Leakrate Projected Reading Leakrate Projected_Reading 5gpd= 53 cprn .

A SIG B SIG 3ogpd= 216 cpm 5 gpd = 241 294 cprn 75 gpd = 509 cprn 30 gpd = 478 820 cpm 100 gpd= 672 cpm 75gpd= 904 1767 cpm 150 gpd= 997 cprn 100 gpd= 1140 2293 cpm 1 GPM = 9400 cpm l5Ogpd= 1614 3345 cpm Current = <1 gpd 1 GPM 13826 30484 cprn Current = <1 ( <1 gpd Air Ejector Set Point Basis: Blow Down Set Point Basis:

Alert: 2X Average Background Alert: 2X Average Background High: 3X Average Background

j ST. LUCIE PLANT OPS-503 OPERATIONS DEPARTMENT POLICY REVISION 63 TECHNICAL SPECIFICATION GUIDANCE r?.E. Page 13of40 Operational Guidance for Section 314.4 3/4.4.1 Reactor Coolant Loops and Coolant Circulation

1. Cold Shutdown Loops Filled: The Reactor Coolant System loops are considered filled while in a Cold Shutdown condition when the following conditions are met:

A. The RCS must have been filled and vented in accordance with 1-NOP-Ol .05 /

2-NOP-Ol .05, RCS Fill and Vent, (If the Unit was just removed from service, the fill and vent was performed during the preceding startup).

B. The RCS is at an inventory level of greater than or equal to 30% pressurizer level as indicated on Ll-1103.

C. The RCS is capable of being pressurized to greater than or equal to 70 psia (55 psig) within the allowed time from loss of cooling to initiation of core boiling as per Attachment 1 in 1-AOP-03.02 / 2-AOP-03.02 as indicated on PlC / P1 -1103 or PlC / P1 -1104.

D. Both SG narrow range levels greater than 10%.

E. The RCS has not been drained below 63 as indicated on the refueling level indicator LI-i 117-i since the last RCS fill and vent procedure was performed.

3/4.4.6.1 Leakage Detection Systems The Reactor Cavity Flow Monitoring System consists of the Weir Pit, the flowmeter (FR-07-3), and the associated alarms (N-35 (Unit 1) from FR-07-3 and N-46 (both Units) from LS-07-12). The function of the system is to detect a 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2. The Reactor Cavity Flow Monitoring System is considered fully operable when both an alarm function works and the flow indication is functional. If either the alarm(s) or flow indication is out of service, the system is operable, but degraded subjected to the following compensatory actions:

A. If the Rx Cavity Leakage High Alarm is out of service, initiate a DS 30 to ensure leakage flow rates per FR-07-3 are recorded hourly. Unit 1 would require both N-35 and N-46 to be declared OOS to require this action.

B. If FR-07-3 is out of service, no exceptional contingency actions are required.

\*.I(, ST. LUCIE PLANT OPS-503 OPERATIONS DEPARTMENT POLICY REVISION 63 TECHNICAL SPECIFICATION GUIDANCE Page 14 of 40 Operational Guidance for Section 314.4 (continued) 3/4.4.6.1 Leakage Detection Systems (continued)

3. With the Reactor Cavity Flow Monitoring system inoperable (neither the flowmeter nor alarms are functional), operation may continue provided the following conditions are met:

A. For up to 30 days provided the following occurs:

1. The Containment Atmosphere Radioactivity Monitor (particulate) remains operable.

2. RCS Inventory balance is performed every 24 Hours.

B. For up to 7 days provided the following occurs

1. The Containment Atmosphere Radioactivity Monitor (gaseous) remains operable.

2. RCS Inventory balance is performed every 24 Hours.

3. Containment atmosphere grab samples are analyzed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4. If the Containment Atmosphere Radioactivity Monitor (particulate) is restored, then apply the time in the step3.A.

Plant management expectation is that an RCS inventory balance shall be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the Reactor Cavity Flow Monitoring system is inoperable. Refer to Technical Specification 3.4.61.

4. With the Containment Atmosphere Radioactivity Monitors (gaseous and particulate) inoperable, operation may continue for up to 30 days provided at least one of the following occurs:

A. RCS inventory balance performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Containment atmosphere grab samples are analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Plant management expectation is that an RCS inventory balance or grab samples shall be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the Containment Atmosphere Radioactivity Monitors are inoperable. Refer to TS 3.4.6.1.

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REGULATORY REPORTING 15 of 97 PROCEDURE NO..

LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 1 of 8)

Declaration of an Emergency Class (See NUREG 1022 Section 3.1.1) 1 Hour Report § 50.72(a)(1)(i) The declaration of any of the Emergency Classes specified in the licensees approved Emergency Plan.

Plant Shutdown Required by Technical Specifications (See NUREG 1022 Section 3.2.1) 4 Hour Report § 50.72(b)(2)(i) The initiation of any 60 Day LER § 50.73(a)(2)(i)(A) The completion of nuclear plant shutdown required by the plants any nuclear plant shutdown required by the plants Technical Specifications. Technical Specifications.

Operation or Condition Prohibited by Technical Specifications (See NUREG 1022 Section 3.2.2) 60 Day LER § 50.73(a)(2)(i)(B) Any operation or condition which was prohibited by the plants Technical Specifications except when:

(1) The Technical Specification is administrative in nature; (2) The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or (3) The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of discovery of the event.

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5 REGULATORY REPORTING 16 Of 97 PROCEDURE NO.:

Ll-AP,-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 2 of8)

Deviation from Technical Specifications Authorized under § 50.54(x)

(See NUREG 1022 Section 3.2.3) 1 Hour Reports 50.72(b)(1) ... any deviation from the 60 Day LER § 50.73(a)(2)(i)(C) Any deviation from the plants Technical Specifications authorized pursuant to plants Technical Specifications authorized pursuant to

§ 50.54(x) of this part. § 50.54(x) of this part.

Degraded or Unanalyzed Condition (See NUREG 1022 Section 3.2.4) 8 Hour Report § 50.72(b)(3)(ii) Any event or condition 60 Day LER 50.73(a)(2)(ii) Any event or condition that that results in: resulted in:

(A) The condition of the nuclear power plant, (A) The condition of the nuclear power plant, including its principal safety barriers, being including its principal safety barriers, being seriously degraded; or seriously degraded; or (B) The nuclear power plant being in an (B) The nuclear power plant being in an unanalyzed condition that significantly unanalyzed condition that significantly degraded plant safety.

degrades plant safety.

External Threat or Hampering (See NUREG 1022 Section 3.2.5) 60 Day LER § 50.73(a)(2)(iii) Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.

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REGULATORY REPORTING 17 of 97 PROCEDURE NO.

Ll-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 3 of 8)

System Actuation (See NUREG 1022 Section 32.6) 4 Hour Report § 50.72(b)(2)(iv)(A) Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre planned sequence during testing or reactor operation.

4 Hour Report § 50.72(b)(2)(iv)(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

8 Hour Report § 50.72(b)(3)(iv)(A) Any event or 60 Day LER § 50.73(a)(2)(iv)(A) Any event or condition that results in valid actuation of any of the condition that resulted in manual or automatic systems listed in paragraph (b)(3)(iv)(B) of this actuation of any of the systems listed in paragraph section, except when the actuation results from and is (a)(2)(iv)(B) of this section, except when:

part of a pre-planned sequence during testing or reactor operation. (1) The actuation resulted from and was part of a pre-planned sequence during testing or reactor operation; or (2) The actuation was invalid and; (i) Occurred while the system was properly removed from service; or (ii) Occurred after the safety function had been already completed.

As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv)(A) other than actuation of the RPS when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.

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REGULATORY REPORTING 18 of 97 PROCEDURE NO:

Ll-AA-1 02-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 4 of 8) 8 Hour Report § 50.72(b)(3)(iv)(B) The systems to 50.73(a)(2)(iv)(B) The systems to which the which the requirements of paragraph (b)(3)(iv)(A) of equirements of paragraph (a)(2)(iv)(A) of this section this section apply are: ipply are:

(1) Reactor protection system (RPS) including: 1) Reactor protection system (RPS) including:

reactor scram and reactor trip. reactor scram or reactor trip.

(2) General containment isolation signals affecting :2) General containment isolation signals affecting containment isolation valves in more than one containment isolation valves in more than one system or multiple main steam isolation valves system or multiple main steam isolation valves (MSIVs). (MSIVs).

(3) Emergency core cooling systems (ECCS) for 3) Emergency core cooling systems (ECCS) for pressurized water reactors (PWR5) including: high- pressurized water reactors (PWR5) including: high-head, intermediate-head, and low-head injection head, intermediate-head, and low-head injection systems and the low pressure injection function of systems and the low pressure injection function of residual (decay) heat removal systems. residual (decay) heat removal systems.

(4) LCCS for boiling water reactors (BWRs) including: 4) ECCS for boiling water reactors (BWRs) including:

high-pressure and low-pressure core spray high-pressure and low-pressure core spray systems: high-pressure coolant injection system: systems; high-pressure coolant injection system; low pressure injection function of the residual heat low pressure injection function of the residual heat removal system. removal system.

(5) BWR reactor core isolation cooling system: 5) BWR reactor core isolation cooling system:

isolation condenser system: and feedwater isolation condenser system: and feedwater coolant injection system. coolant injection system.

(6) PWR auxiliary or emergency feedwater :6) PWR auxiliary or emergency feedwater system. system.

(7) Containment heat removal and (7) Containment heat removal and depressurization depressurization systems, including systems, including containment spray and fan containment spray and fan cooler systems cooler systems.

(8) Emergency ac electrical power systems, including: (8) Emergency ac electrical power systems, emergency diesel generators (EDGs): hydroelectric including: emergency diesel generators (EDG5):

facilities used in lieu of EDGs at the Oconee hydroelectric facilities used in lieu of EDGs at the Station: and BWR dedicated Division 3 EDGs. Oconee Station; and BWR dedicated Division 3 EDGs.

Actuation of the RPS when the reactor is critical is (9) Emergency service water systems that do not eportable under § 50.72(b)(2)(iv)(B) normally run and that serve as ultimate heat sinks.

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REGULATORYREPORTING 19of97 PROCEDURE NO.:

LI-AA-1 02-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 5 of 8)

Event or Condition that Could Have Prevented Fulfillment of a Safety Function (See NUREG 1022 Section 3.2.7) 8 Hour Report § 50.72(b)(3)(v) Any event or 60 Day LER § 50.73(a)(2)(v) Any event or condition condition that at the time of discovery could have that could have prevented the fulfillment of the safety prevented the fulfillment of the safety function of function of structures or systems that are needed to:

structures or systems that are needed to:

(A) Shut down the reactor and maintain it in a safe (A) Shut down the reactor and maintain it in a safe shutdown condition; shutdown condition; (B) Remove residual heat; (B) Remove residual heat; (C) Control the release of radioactive material; or (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

(D) Mitigate the consequences of an accident.

8 Hour Report § 50.72(b)(3)(vi) Events covered in 50.73(a)(2)(vi) Events covered in paragraph (b)(3)(v) of this section may Include one or paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual and/or procedural inadequacies. However, individual component failures need not be reported pursuant to component failures need not be reported pursuant to paragraph (b)(3)(v) of this section if redundant paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and equipment in the same system was operable and available to perform the required safety function. available to perform the required safety function.

Common Cause Inoperability of Independent Trains or Channels (See NUREG 1022 Section 3.2.8) 60 Day LER § 50.73(a)(2)(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:

(A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

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Ll-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 6 of 8)

Radioactive Release (See NUREG 1022 Section 3.2.9) 60 Day LER § 50.73(a)(2)(viii)(A) Any airborne radioactive release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, resulted in airborne radionuclide concentrations in an unrestricted area that exceeded 20 times the applicable concentration limits specified in appendix B to part 20, table 2, column 1.

60 Day LER § 50.73(a)(2)(viii)(B) Any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in appendix B to part 20, table 2, column 2, at the point of entry into the receiving waters (i.e.,

unrestricted area) for all radionuclides except tritium and dissolved noble gases.

Internal Threat or Hampering (See NUREG 1022 Section 3.2.10) 60 Day LER § 50.73(a)(2)(x) Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

Transport of a Contaminated Person Offsite (See NUREG 1022 Section 3.2.11) 8 Hour Report § 50.72(b)(3)(xii) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

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REGULATORY REPORTING 21 of 97 PROCEDURE NO Ll-AA-1 02-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 7 of 8)

News Release or Notification of Other Government Agency (See NUREG 1022 Section 3.2.12) 4 Hour Report § 50.72(b)(2)(xi) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

Loss of Emergency Preparedness Capabilities (See NUREG 1022 Section 3.2.13) 8 Hour Report § 50.72(b)(3)(xiii) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, emergency notification system, or offsite notification system).

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REGULATORY REPORTING 22 of 97 PROCEDURE NO.:

Ll-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT I REPORTABLE EVENTS (Page 8 of 8)

Single Cause that Could Have Prevented Fulfillment of the Safety Functions of Trains or Channels in Different Systems (See NUREG 1022 Section 3.214) 60 Day LER § 50.73(a)(2)(ix)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems that are needed to:

(1) Shut down the reactor and maintain it in a safe shutdown condition; (2) Remove residual heat; (3) Control the release of radioactive material; or (4) Mitigate the consequences of an accident.

§ 50.73(a)(2)(ix)(B) Events covered in paragraph (ix)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (ix)(A) of this section if the event results from:

(1) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (2) Normal and expected wear or degradation.

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REGULATORY REPORTING 23 of 97 PROCEDURE NO.:

LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 1 of 12)

GENERAL CONDITION REFERENCE AID This table is provided as an aid to locate pertinent subject areas within Attachment 2. Topics have been abbreviated and may refer to multiple entries in Attachment 2 which may differ depending upon the characteristics of the event, the nature of the threat, and the level of compensation. Additionally, more than one line number may apply to a given situation.

Reportability determinations shall be made only after referral to the pertinent sections of .

Line #s Topic Report oggable Significant FFD 1 2, 3 Bomb / Extortion / Vandalism Threat or Conspiracy 1 hr X 4, 5 Safeguards Information Compromise 1 hr X 6 Offsite Communications Loss 1 hr 7 Civil Disturbance 1 hr 8 Assault on Power Reactor (Actual or Imminent) 15 mm; 9 Unauthorized personnel gain access to site PANA 1 hr 1 hr 10 Tampering Confirmed 1 hr 1 1 12 Safeguards Suspension 1 hr 13,14 (Uncompensated) Fire/ 1 hr Explosion 15 Badge Falsification 1 hr 16 Uncompensated Key or Badge Loss; X 1 hr Compensated Badge Loss) X 1 hr 17 Less Than Minimum Security Force / Strike 18, 19 Security System Power Loss 1 hr X 20 Intrusion Detection Loss 1 hr X 21 22 Vital Area Door Alarms and Locks 1 hr X 23, 24 Contraband in PA/VA (actual or 1 hr X 25 attempted) 1 hr x 26, 27 Weapon Loss 1 hr x CCTV Alarm Loss / Fail 28, 29 Security Computer Fail X 30 Alarm Station Alarm Loss X 31 Card Reader Fail VA (compensated) x

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REGULATORY REPORTING 24 of 97 PROCEDURE NO.:

LI-AA-1 02-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 2 of 12) 32 Alarm Failures (compensated) X 33 Lighting Zones (compensated) X 35 Tailgating into an Authorized Access Required 1 hr X 36 Area lhr X Escort Separation x

Search Equipment Fail 38 Vehicle Barrier System X 39 Compensatory Measure Fail 40 PA/VA Security Design Flaw (compensated) 41 False I Nuisance Alarms 1 hr X 42 Pre-Employment Screening Incomplete X 43 Safeguards System Fail / Degrade X 44 Safeguard Effectiveness Reduction 24 hr X 45 Illegal drugs or alcohol in PA 24 hr 46 Controlled Substance / Alcohol Licensed Operator or Supervisor 47 Licensed Operator or Supervisor FFD 24 hr 48 FFD Programmatic Failures 24 hr 49 FFD False Positive or Negative 24 hr 50 Cyber Security Event (Voluntary SIDS Entry) 51 Category 1 or Category 2 - Radioactive Material Immediate LLEA 4 hr 30 day written report 52 Confirmed Suspicious Activity related to possible LLEA theft, sabotage, diversion, radioactive Cat. 1 or 2 4 hr NRC materials

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REGULATORY REPORTING 25 of 97 PROCEDURE NO.

LI-AA-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 3 of 12)

EVENT REPORT DISCUSSION Credible Bomb or Extortion Threat One Hour In addition to the INITIAL telephone report, a telephone report of the RG 5.62 C.2.2.1 Report results of a bomb search shall be made within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the completion of Paragraph l(a)(1), (2), or (3) of Appendix G the search. Implementation of compensatory measures does not affect or alter reporting requirements.

2 Non-Credible bomb or extortion threat. Loggable VERIFY the bomb or extortion threat is not one of a pattern of harassing RG 5.62 C.2.4.13 threats and does not involve a specific organization or group claiming Paragraph 11(b) of Appendix G responsibility, that the search result is negative, and that no evidence is available other than the threat message.

3 Discovery of acts involving felonious acts, conspiracy to One Hour Due to the seriousness of such event, this event shall be reported within bomb the facility, sabotage or vandalism of vital equipment. Report 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> even IF the individuals unescorted access authorization is RG 5.62 C 2.2.2 & 3 cancelled.

Paragraph I(a)(2) or (3) of Appendix G 4 Compromise (Including loss or theft of documents) of One Hour This is considered a credible threat to the facility. Implementation of Safeguards documents posing a threat Report compensatory measures does not affect or alter reporting requirements.

RG 5.62 C 2.2.17 Paragraph 1(a) of Appendix G 5 Compromise (including theft or loss of documents) of Loggable DETERMINE IF the theft, loss, or compromise would assist an individual Safeguards Information, in gaining undetected or unauthorized access, or would not significantly GL 91-03, Enclosure I assist an individual in an act of sabotage.

Paragraph 11(a) of Appendix G 6 Complete loss of offsite communications. One Hour This shall be reported immediately after restoration of any viable RG 5.62 C.2.2.28 Report communications with the NRC.

Paragraph I (a)(2) of Appendix G

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5 REGULATORY REPORTING 26 of 97 PROCEDURE NO.

Ll-AA-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 4 of 12) 7 Mass demonstration or civil disturbance that may pose a One Hour Implementation of compensatory measures does not affect or alter threat to the facility Report reporting requirements.

RG 5.62, C.2.2.9 & 10 Paragraph I(a)(2) or (3) of Appendix G 8 An assault on a power reactor regardless of whether 15 minutes INITIAL report to be within 15 minutes to comply with Emergency Plan perimeter penetration is achieved. Initial Report, reporting requirements (see Section 4.1). NRC may request an open RG 5.62 C.2.2 12 One Hour line. Since the INITIAL 15 minute report may be abbreviated, FOLLOW Paragraph I(a)(1), (2), or (3) of Appendix G Follow-up up with a 1 one-hour report IF additional required information must be report communicated. Due to the seriousness of this type of event, implementation of compensatory measures or contingency procedures does not affect or alter reporting requirements 9 An actual entry of an unauthorized person into a Protected One Hour Since this would be after the fact of serious safeguards degradation, Area, or Vital Area Report implementation of compensatory measures does not affect or alter RG 5.62 C.2.2.13 GL 91-03, Enclosure I, 9th example reporting requirements.

Paragraph 1(b) of Appendix G 10 Confirmed tampering of suspicious origin with safety or One Hour Since this would be after the fact, implementation of compensatory security equipment. Report measures does not affect or alter reporting requirements.

RG 5.62 C.2.2.11 Paragraph l(a)(1), (2), or (3) of Appendix G 11 Uncompensated suspension of safeguards controls during One Hour Regulatory Guide 5.65 describes safeguards measures that maybe either radiological or non-radiological emergencies that could Report suspended during non-radiological emergencies.

allow undetected or unauthorized access.

RG 5.62 C.2.2.14 Paragraph I (c) of Appendix G

REVISION NO.: PROCEDURE TITLE PAGE REGULATORY REPORTING 27 of 97 PROCEDURE NO.:

Ll-FA-1O2-1O01 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 5 of 12) 12 Suspension of Safeguards due to severe weather under the One Hour Implementation of compensatory measures does not affect or alter provisions of 10CFR5O.54(x) and (y) Report reporting requirements.

RG 5.62 C.2 2.14.

Paragraph 1(c) of Appendix G 13 A lire / explosion that is of suspicious or unknown origin. One Hour A fire I explosion within the PA or Isolation Zone that is of unknown or RG 5.62, C.2.2.5 Report suspicious origin (e.g., it may have been intended to affect the security Paragraph (a)(1), (2) & (3) & 1(d) of Appendix G response, or be used as a diversion) shall be reported within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Para l(a)(2), l(a)(3) and 1(c). IF the fire I explosion results in significant damage it may be reportable under 10CFR5O.72 or 50.73, and IF so, a duplicate report under 10CFR73.71 is not required.

14 A fire / explosion originally thought to be of suspicious or Loggable IF this determination of non-suspicious origin is made within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of unknown origin is subsequently determined, more than one under 73.71 discovery, the fire / explosion need neither be logged nor reported under hour after discovery, to be non-suspicious (e.g., no 10CFR73.71 unless the non-suspicious fire / explosion resulted in deception, diversion or malevolent intent was involved), significant damage, in which case the event shall be logged.

[RG 5.62, C.2.2.5, C.2.5.4J Paragraph I (a) or (d) of Appendix G (IF the lire / explosion results in significant damage it may still be reportable under 10CFR5O.72 or 50.73.)

15 Discovery of intentionally falsified identification badges or One Hour This is considered a safeguards threat and shall always be reported key cards. Report within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of discovery.

RG 5.62 C.2.2 15 Paragraph 1(a) of Appendix G 16 Discovery of uncompensated and unaccounted for, lost or One Hour Events need not be reported IF compensatory measures are taken in stolen key cards, ID. card blanks, keys, or any access Report / accordance with the Physical Security Plan to prevent the use of the lost device that could allow unauthorized or undetected access Loggable or stolen device and verified the lost or stolen device was not used in an to Protected Areas, controlled access areas, or vital areas. unauthorized manner.

RG 5.62 C.2 2.16 Paragraph 1(c) of Appendix G

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5 REGULATORY REPORTING 28 of 97 PROCEDURE NO.

LI-AA-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 6 of 12) 17 Unavailability of a minimum number of security personnel or One Hour IF minimum manning can be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> THEN; LOG ImmInent strike by the security force. Report / During a strike / walkout, minimum manning is not supplemented within 2 RG 5.62 .2.2,19 Loggable hours THEN; a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report is required.

Paragraph 1(c) of Appendix G 18 Uncompensated loss of all power to security systems that One Hour IF the security system is maintained by stand-by power and system could allow undetected or unauthorized to the Protected Report verification is conducted THEN the event is considered properly Area, controlled access area, or vital areas. compensated and need only be logged. IF stand-by power fails prior to RG 5.62 C.2.2.20 restoration of power, this shall be reported within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of loss of stand Paragraph 1(c) of Appendix G by power.

19 Properly compensated loss of the power supply for the entire Loggable / Proper compensation means the availability of immediate back-up intrusion detection system that, if uncompensated, would One Hour emergency power through an uninterruptible power source or the posting allow unauthorized or undetected access. Reportable of security IF back-up power is not available. The posting of security is RG 5.62 C.2.4.7 not considered adequate IF back-up power fails to PROVIDE an Paragraph 11(a) of Appendix G immediate response or fails during the event and shall be reported within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

20 Uncompensated loss of ability to detect within a single One Hour IF properly compensated THEN; LOG intrusion detection zone or all zones. Report / Note: SOCA loss is not required to be reported or logged, as it is RG 5.62 C.2.2.21 Loggable considered to be an early warning system and is not credited for Paragraph 1(c) of Appendix G initiating contingency response.

21 Uncompensated loss of alarm capability to a vital area One Hour IF proper compensatory measures are instituted and a thorough search access door. Report / of the area has been completed as soon as practical, THEN : LOG RG 5.62 C.2.2.21 Loggable Paragraph 1(c) of Appendix G 22 Uncompensated loss of a locking mechanism on a vital area One Hour IF proper compensatory measures are instituted and a thorough search door. Report / of the area has been completed as soon as practical, THEN : LOG RG 5.62 C.2.2.21 Loggable Paragraph 1(c) of Appendix G

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REGULATORY REPORTING 29 of 97 PROCEDURE NO.

LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 7 of 12) 23 Discovery of the actual or attempted introduction into or One Hour Implementation of compensatory measures does not affect or alter possession within the protected area or vital area of Report reporting requirements.

contraband (i.e. weapons, explosives or incendiary devices) that constitutes a significant threat.

RG 5.62 C.2.2.23; GL 91-03; 10CFR73 Appendix G Paragraph 1(d) 24 Discovery of contraband inside the protected area that is not Loggable VERIFY contraband would not constitute a threat.

a significant threat.

RG 5.62 C.2.2.23; GL 91-03, Enclosure 1, 3rd example; 10CFR73 Appendix G Paragraph 11(b) 25 Loss of a security weapon at the site. One Hour Implementation of compensatory measures does not affect or alter RG 5.62 C.2.2.24 Report! reporting requirements. IF weapon is retrieved within one hour of Paragraph I(a)(3) of Appendix G Loggable discovery of its loss THEN LOG 26 Total loss of ability to remotely assess alarms by CCTV One Hour IF proper compensatory measures are instituted THEN; LOG RG 5.62 C.2.2.18 Report!

Paragraph 1(c) of Appendix G Loggable 27 Properly compensated CCTV failure in a single zone while Loggable Proper compensation means providing other assessment capabilities of the intrusion detection system remains operational. equal value or posting security.

RG 5.62 C.2.4.4 Paragraph 11(a) of Appendix G 28 Properly compensated security computer failures Loggable Proper compensation means IF back-up power restores systems to RG 5.62 C.2.4.1 operation or security is posted to offer an equivalent level of protection Paragraph 11(a) of Appendix G and thoroughly searching all affected areas.

REVISION NO. PROCEDURE TITLE: PAGE REGULATORY REPORTING 30 of 97 PROCEDURE NO. I Ll-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 8 of 12) 29 Properly compensated computer failures that may not Loggable Proper compensation means verification has been performed to enable unauthorized or undetected access, determine unauthorized or undetected access did not or could not occur RG 5.62 C.2.4.9 Paragraph 11(b) of Appendix G 30 Properly compensated loss of the capability of a single alarm Loggable Proper compensation means measures were immediately taken to offer station to monitor or remotely assess alarms but monitoring an equivalent level of protection.

or assessment capability remains in the other station.

RG 5.62 C.2.4.10 Paragraph 11(b) of Appendix G 31 Properly compensated vital area card reader failure. Loggable Proper compensation means posting of security and conducting a RG 5.62 C.2.4.2 thorough search of the area as soon as practical.

Paragraph 11(a) of Appendix G 32 Properly compensated alarm failures. Loggable Proper compensation means USE of back-up equipment or posting RG 5.62 C.2,4.3 security and conducting a thorough search of the affected area(s).

Paragraph II (a) of Appendix G 33 Properly compensated failure or degradation of single Loggable Proper compensation means the USE of stand-by power, or using low-perimeter lighting zone if the intrusion detection system level light devices, or using portable lighting systems, or posting security.

remains operational.

RG 5.62 C.2.4.5 Paragraph 11(a) of Appendix G 34 Properly compensated loss of a badge by an individual. Loggable Proper compensation means canceling the badge and verifying the RG 5.62 C 2.4.6 badge was not used in unauthorized manner.

Paragraph 11(a) of Appendix G

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LI-AA-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 9 of 12) 35 Tailgating by an employee of the licensee or contractor to Loggable I One hour Report IF either (1) Malevolent intent existed to enter the area gain access to an area to which authorized access is One Hour undetected, or (2) Tailgating was committed by person without authorized required. Reportable access which can not be satisfactorily explained or where the individuals RG 5.62 C.2.4.11 GL 91-03 Enclosure 1 9th example level of screening would not qualify them for unescorted access; Paragraph 11(b) of Appendix G Otherwise Loggable.

36 A visitor becomes separated from their escort. Loggable I VERIFY the escort (or other authorized person) immediately (within a few RG 5.62 C.2.2.13 GL 91-03 Enclosure 1, 8th example One Hour minutes) recognizes the situation and corrects it. IF not corrected quickly Paragraph 11(b) of appendix G Reportable THEN; One Hour Report 37 Undetected search equipment failure allowing unsearched Loggable IF search equipment failure is discovered before anyone gains individuals into the Protected Area. unsearched entry into the Protected Area and individuals are searched GL 91-03, Enclosure 1, 7th example with back-up equipment, no report or LOG is necessary Paragraph 11(b) of Appendix G 38 Proper compensation of a failure or degradation of the Loggable Proper compensation means measures were taken to offer an equivalent Vehicle Barrier System occurs that would decrease level of protection effectiveness.

Paragraph 1(c) of Appendix G 39 A failed compensatory measure, such as an inattentive Loggable TAKE immediate action for the failed equipment or inattentive security security officer or failed equipment that fails after being officer.

successfully established for a degraded security system.

RG 5.62 C.2.2.22 GL 91-03, Enclosure 1, 2nd example Paragraph 11(a) of Appendix G 40 Properly compensated design flaw or vulnerability in a Loggable Proper compensation means measures were immediately taken to offer protected area or vital area safeguards barrier, an equivalent level of protection RG 5.62 C.2.3.1; GL 91-03, Enclosure 1, 1st example.

Paragraph 11(a) of Appendix G

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LI-M-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 10 of 12) 41 False or nuisance alarms for which a pattern of such alarms Loggable Proper compensation means measures were taken to offer an equivalent emerges or that are so frequent that the effectiveness of the level of protection alarm system is degraded.

RG 5.62 C.2.2.3 Paragraph II (a) of Appendix G 42 Incomplete pre-employment screening records (to include Loggable I Proper compensation means the unescorted access is cancelled until falsification of a minor nature), inadequate administration of, One Hour resolved. IF it is determined that unescorted access would not have been control or evaluation of psychological tests. Reportable granted based on the new or developed information, THEN a 1 Hour GL 91-03, Enclosure 1, 13th example Report is required.

43 Any previous undiscovered or unreported failure or Loggable VERIFY adequate compensatory measures were instituted in accordance degradation or vulnerability of a safeguards system that with the Physical Security Plan requirements and offer an equivalent level could have allowed unauthorized or undetected access to of protection.

the protected area or vital area if compensatory measures had not been established.

RG 5.62 C.2.3.1 Paragraph 11(a) of Appendix G 44 Any other threatened, attempted, or committed act not Loggabie VERIFY actions or compensatory measures taken offer an equivalent defined in Appendix G that has the potential for reducing the level of protection.

effectiveness of the safeguards system below that committed in the physical security or contingency plan or the actual condition of such reduction in effectiveness.

RG 5.62 C.2.3.2 Paragraph 11(b) of Appendix G 45 The sale, use, or distribution, possession, or presence of 24 Hour Implementation of compensatory measures does not affect or alter illegal drugs or alcohol occurring within the Protected Area, Report reporting requirements.

is a Significant Fitness-for-Duty Event. Notification 10 CFR 26.719

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REGULATORY REPORTING 33 of 97 PROCEDURE NO..

Ll-AA-102-1 001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 11 of 12) 46 The sale, use, or possession of controlled substance, 24 Hour The occurrence includes licensees or other entitys FFD policy (including confirmed positive test, use of alcohol within the protected Report subversion as defined in § 26.5). Implementation of compensatory area, or unfit for scheduled work due to the consumption of Notification measures does not affect or alter reporting requirements.

alcohol by a any person licensed under 10 CFR Part 55 to operate a power reactor, FFD program personnel or any supervisory personnel then, is a Significant Fitness-for-Duty Event.

10 CFR 26.719 47 Should any Fitness-for-Duty failure occur for a licensed 24 Hour Implementation of compensatory measures does not affect or alter operator or supervisory personnel then classify the event as Report reporting requirements.

a Significant Fitness-for-Duty Event. Notification 10 CFR 26.719 48 Any intentional act that casts doubt on the integrity of the 24 Hour Implementation of compensatory measures does not affect or alter FFD program; and any programmatic failure, degradation, or Report reporting requirements.

discovered vulnerability of the FFD program that may permit Notification undetected drug or alcohol use or abuse by individuals within a Protected Area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program.

1 OCFR26.71 9 49 If a false positive error occurs on a blind performance test 24 Hour Implementation of compensatory measures does not affect or alter sample submitted to an HHS-certified laboratory, or a false Report reporting requirements.

negative error occurs on a quality assurance check of Notification validity screening tests, as required in § 26.137(b), the licensee or other entity shall notify the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after discovery of the error.

1 OCFR26.71 9

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LI-AA-102-1001 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 2 SAFEGUARDS EVENT CLASSIFICATION (Page 12 of 12) 50 Suspicious Cyber Security Incidents are requested to be SID Entry NRC NSIR Information Assessment Team Advisory lA-i 3-01 UCriteria for reported via SID. Reporting of Suspicious Activity Associated with Cyber Security Incidents, provides updates to previous notices discusses NSIR recommendations for licensees to voluntarily reporting incidents via NRC Security Incident Database (SID). IA-13-01, IA-04-08, IA-05-01, IA-00-03 and IA-12-02 contain additional information regarding the types of activities for which NSIR has requested voluntary reports via SID.

51 a) Unauthorized entry resulting in an actual or attempted 4 Hour Since this would be after the fact, implementation of compensatory theft, sabotage, or diversion of Category 1 or Category 2 Report / measures does not affect or alter reporting requirements.

quantities of radioactive material. 30-day 10CFR37.57 Written Immediately notify Local Law Enforcement Agency (LLEA) after Report determination that event occurred.

Notify NRC Operations Center within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of discovery. NRC notification should not be at the expense of causing a delay or interfering with the LLEA response.

Initial telephonic notification shall be followed by a written report to NRC by an appropriate method list in 10CFR37.7 providing sufficient information for NRC analysis and evaluation, including identification of any necessary corrective actions to prevent future instances.

52 Any suspicious activity related to possible theft, sabotage, or 4 Hour Assess activity and notify LLEA, as appropriate. If LLEA notified, notify diversion of Category 1 or Category 2 quantities of Report if NRCs Operation Center as soon as possible, but not later than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> radioactive material. LLEA notified after notifying LLEA.