Text

August 2002 Exam 50-400/2002-301 Final Written SRO Exam
	ML022670157
Person / Time
Site:	Harris
Issue date:	03/22/2002
From:	Ernstes M; Division of Reactor Safety II
To:	Scarola J; Carolina Power & Light Co
References
	50-400/02301 50-400/02301
	Download: ML022670157 (136)
	v • d • e

Final Submittal

(Blue Paper)

1. Senior Operator Written Examination

SHEARON HARRIS

EXAM 2002-301

50-400

AUGUST 26 - 29, 2002

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 1

Given the following conditions:

"* The Main Turbine is operating at 1800 rpm in preparation for synchronizing to the

grid.

"* Reactor power is being maintained at approximately 12% using the Condenser Steam

Dumps.

Condenser Vacuum Pump 'A' is under clearance.

Condenser Vacuum Pump 'B' trips.

Assuming NO operator actions, condenser vacuum degrades until ...

a. the turbine and the reactor trip, and condenser steam dump operation is blocked

b. the turbine trips, and condenser steam dump operation is blocked, but the reactor

remains critical

c. condenser steam dump operation is blocked, but vacuum stabilizes above the

turbine trip setpoint

d. the turbine and reactor trip, but vacuum stabilizes above the steam dumps

interlock setpoint

ANSWER:

a. the turbine and the reactor trip, and condenser steam dump operation is blocked

// 5 -,--/. 3,o 1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 2

Given the following conditions:

A reactor trip and safety injection have occurred.

Steam Generator parameters have decreased to the following values:

SG LEVEL PRESSURE

A 32% 870 psig

B 12% 420 psig

C 34% 830 psig

NO operator actions have been taken.

Which of the following components is mispositioned?

a. lFCV-205 IB, MDAFW FCV to B SG, CLOSED

b. 1FCV-2051C, MDAFW FCV to C SG, OPEN

c. 1MS-70, MS B SG to AFW Turbine, CLOSED

d. 1MS-72, MS C SG to AFW Turbine, OPEN

ANSWER:

d. 1MS-72, MS C SG to AFW Turbine, OPEN

O6/ ft3'03

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 3

If a Containment Ventilation Isolation (CVI) signal occurred, which of the following

Containment Ventilation fans would NOT trip directly from the CVI signal, but would

trip as a result of being interlocked with other fans?

a. Normal Purge Supply fans (AH-82 A & B)

b. Pre-Entry Purge Makeup fans (AH-81 A & B)

c. Airborne Radioactivity Removal fans (S-lA & B)

d. CNMT Pre-entry Purge Exhaust fans (E-5 A & B)

ANSWER:

b. Pre-Entry Purge Makeup fans (AH-81 A & B)

,/Z/ )t%;V3-O

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 4

Hydrogen concentration in the Waste Gas System, downstream of the catalytic

recombiners, is limited to 4% to ...

a. maintain levels below flammability limits.

b. ensure proper operation of the recombiner.

c. limit the volume of waste gas generated.

d. minimize the radioactive content of the waste gas decay tanks.

ANSWER:

a. maintain levels below flammability limits.

) 7/;7,-,)-5

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 5

Given the following conditions:

A large break LOCA has occurred.

Containment pressure peaked at 15 psig and has decreased to 6 psig.

Actions are being taken to place the plant in cold leg recirculation in accordance with

EPP-010, "Transfer to Cold Leg Recirculation."

Two (2) CSIPs, two (2) RHR Pumps, and two (2) Containment Spray Pumps are

running.

The crew has just completed alignment of Safety Injection for recirculation and is in

the process of verifying Containment Spray alignment when the Reactor Operator

notes Containment Sump level is 25%.

Which of the following actions should be taken?

a. e Stop both trains of Containment Spray

Maintain both trains of RHR Pumps and CSIPs operating

b. 9 Stop both trains of Containment Spray

Stop one (1) train of RHR Pumps and CSIPs

c. * Stop one (1) train of Containment Spray

Stop one (1) train of RHR Pumps and CSIPs

d. * Stop both trains of Containment Spray

Stop both trains of RHR Pumps and CSIPs

ANSWER:

d. * Stop both trains of Containment Spray

Stop both trains of RHR Pumps and CSIPs

b/Ao

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 6

Given the following conditions:

The plant is operating at 93% power.

Condensate Pump 1B trips on motor overcurrent.

Condensate Booster Pump 1B trips as a result of the trip of Condensate Pump lB.

Which of the following describes the effect of these events on the Main Feed Pumps

AND the required operator action?

a. * Main Feed Pumps 1A and 1B remain running

0 Trip the reactor and go to PATH-I

b. e Main Feed Pumps IA and 1B remain running

Verify a turbine runback occurs

c. * Main Feed Pump 1B trips

0 Trip the reactor and go to PATH- 1

d. * Main Feed Pump 1B trips

Verify a turbine runback occurs

ANSWER:

c. e Main Feed Pump 1B trips

Trip the reactor and go to PATH-1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 7

Given the following conditions:

The plant is solid in Mode 5 with one (1) RCP in operation.

RHR Pump A-SA is providing letdown flow with PK-145.1, LTDN PRESSURE

1CS-38, in MAN.

CSIP A-SA is providing RCS makeup and seal injection.

If instrument air is lost to 1CS-38 (PCV-145), the operator should ...

a. trip CSIP A-SA.

b. trip RHR Pump A-SA.

c. maintain letdown flow using HC-142.1, RHR Letdown lCS-28.

d. open one PRZ PORV.

ANSWER:

a. trip CSIP A-SA.

Ac/Wj q &ý-- /, X"; ,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 8

Given the following conditions:

"* An I&C technician reports that both of the Control Room Normal Outside Air Intake

Isolation radiation monitors have failed detectors.

"* It will take somewhere between four (4) and eight (8) hours to replace the detectors.

Which of the following states the action which must be taken within one (1) hour, in

accordance with Technical Specification 3.3.3.1 ?

a. Establish operation of the Control Room Emergency Filtration System in the

Recirculation Mode of Operation

b. Initiate the preplanned alternate method of radiation monitoring

c. Return the monitors to service, or be in Hot Standby within the next six (6) hours

d. Perform a surveillance test on the Control Room Emergency Filtration System, or

be in Hot Standby within the next six (6) hours

ANSWER:

a. Establish operation of the Control Room Emergency Filtration System in the

Recirculation Mode of Operation

9.K5;-11

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 9

Given the following conditions:

A reactor trip occurred from 75% power approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago.

The operating crew is attempting to close the Reactor Trip Breakers.

All controls and switches are in their normal alignment for plant conditions.

Assuming all other conditions are met for closing the Reactor Trip Breakers, which of the

following sets of conditions would physically allow the breakers to close when the

REACTOR TRIP BREAKERS TRAINS A&B switch is taken to the CLOSE position?

a. * SG'A'levelis 18%

IR channel N-36 is failed high

b. * SG 'A' level is 18%

RCP 'A' is secured

c. * IR channel N-36 is failed high

PRZ pressure is 1920 psig

d. e PRZ pressure is 1920 psig

RCP 'A' is secured

ANSWER:

d. & PRZ pressure is 1920 psig

RCP 'A' is secured

Doi /&/-/ //

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 10

The plant is in Mode 1.

When entering the Personnel Air Lock, how is the inside door checked closed and what

would be the consequences of attempting to enter with the inside door open?

a. * The outside door contains a visual indication (red/green light) of the inside

door's position

Technical Specifications would be violated

b. 9 The equalizing valve will NOT open if the inside door is open

Technical Specifications would be violated

c. * The outside door contains a visual indication (red/green light) of the inside

door's position

An interlock will prevent entry if the inside door is open

d. * The equalizing valve will NOT open if the inside door is open

An interlock will prevent entry if the inside door is open

ANSWER:

c. * The outside door contains a visual indication (red/green light) of the inside

door's position

An interlock will prevent entry if the inside door is open

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 11

Given the following conditions:

Containment Pressure Channel I, PT-950A, is in TEST for surveillance testing

purposes.

Containment Pressure Channel III, PT-952A, is failed low.

A large break LOCA occurs and actual Containment Pressure reaches 21 psig.

Which of the following describes the response of the Containment Spray system?

a. NEITHER train of Containment Spray will automatically actuate

b. ONLY Train 'A' of Containment Spray will automatically actuate

c. ONLY Train 'B' of Containment Spray will automatically actuate

d. BOTH trains of Containment Spray will automatically actuate

ANSWER:

d. BOTH trains of Containment Spray will automatically actuate

/9 01

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 12

Given the following conditions:

"* Several Fuel Handling Building (FHB) area radiation monitors on both trains have

reached the high alarm setpoint.

"* AOP-005, "Radiation Monitoring System," has directed the operator to verify that the

FEB ventilation has shifted to the emergency exhaust lineup.

Both FHB Emergency Exhaust Fans, E-12 and E-13, are RUNNING.

FHB Emergency Exhaust Fan Inlets, 1FV-2 SA and 1FV-4 SB, are OPEN

Which of the following additional alignments is expected?

a. * FHB Operating Floor Supply Fans (AE-56, AH-57, AH-58, AH-59)

SECURED

FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)

OPEN

b. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)

RUNNING

FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)

OPEN

c. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)

RUNNING

FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)

SHUT

d. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)

SECURED

FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)

SHUT

ANSWER:

d. * FHB Operating Floor Supply Fans (AH-56, AH-57, AH-58, AH-59)

SECURED

FHB Normal Exhaust Isolation Dampers (FL-D4, FL-D5, FL-D21, FL-D22)

SHUT

(/4 Ct'1//

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 13

Given the following conditions:

A loss of offsite power has occurred with the plant at 100% power.

The operating crew is performing the actions of EOP-EPP-001, "Loss of AC Power to

1A-SA and lB-SB Buses."

A SGTR has been identified in SG 'C'.

SGs 'A' and 'B' are being depressurized to 180 psig.

Which of the following describes the method used AND the bases for depressurizing SGs

'A' and 'B' to 180 psig?

a. * Method - Operate the SG PORVs 'A' and 'B' from the MCB

Bases - Lower RCS pressure below ruptured SG pressure to backfill from SG

'C' to the RCS

b. * Method - Operate the SG PORVs 'A' and 'B' locally

Bases - Lower RCS pressure below ruptured SG pressure to backfill from SG

'C' to the RCS

c. * Method - Operate the SG PORVs 'A' and 'B' from the MCB

0 Bases - Minimize RCP seal damage and RCS inventory loss

d. * Method - Operate the SG PORVs 'A' and 'B' locally

Bases - Minimize RCP seal damage and RCS inventory loss

ANSWER:

d. * Method - Operate the SG PORVs 'A' and 'B' locally

Bases - Minimize RCP seal damage and RCS inventory loss

73/4- Dj oc. c//

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 14

Chemistry reports that the RCS Dose Equivalent Iodine (DEI- 13 1) activity has exceeded

the limit and a shutdown is required.

The plant is to be placed in Hot Standby with T-avg less than 500'F to ...

a. enhance the ability of the mixed bed demineralizers to remove fission products in

the event of a small break LOCA.

b. minimize the deposition of fission products and activation products on the core

surfaces in the event of a large break LOCA.

c. prevent additional fuel cladding oxidation from occurring in the event of a large

break LOCA.

d. prevent the release of radioactivity to the environment in the event of a SGTR.

ANSWER:

d. prevent the release of radioactivity to the environment in the event of a SGTR.

S 07 /ý

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 15

Given the following conditions:

The plant is operating at 50% power.

Bank 'D' Control Rods are at 140 steps.

All control systems are in automatic and at program values.

The Median Select AT Circuit output has failed high.

Which of the following will occur?

a. ALB-020-2-1, TURBINE AUTOMATIC LOADING STOP, alarms

b. ALB-013-8-3, BANK LO-LO INSERTION LIMIT, alarms

c. Bank 'D' Control Rods step inward

d. Charging flow increases

ANSWER:

b. ALB-013-8-3, BANK LO-LO INSERTION LIMIT, alarms

Y64 6k /62

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 16

Which one of the following statements describes the reason why some selected 480-V

MCC loads have two supply breakers in series?

a. The loads are safety-related, requiring redundant train protection

b. The loads are in Containment, requiring redundant overcurrent protection for the

penetration

c. The loads are safety-related, requiring redundant protection with different

overcurrent trip setpoints

d. The loads are capable of being operated from the ACP, requiring redundant

control functions

ANSWER:

b. The loads are in Containment, requiring redundant overcurrent protection for the

penetration

XIA/ D rt.'

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 17

Given the following conditions:

RCS boron concentration is 1900 ppm.

Boric Acid Tank concentration is 7100 ppm.

Which of the following RWMU Flow Controller potentiometer settings will result in the

HIGHEST ACCEPTABLE total automatic Primary Makeup System flow rate for these

conditions?

a. 5.63

b. 6.25

C. 6.88

d. 7.50

ANSWER:

C. 6.88

/I AS

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 18

Given the following conditions:

The site has experienced a loss of offsite power.

EDG 'A' has started and sequenced all loads.

A valve misalignment has isolated ESW cooling to EDG 'A'.

How long can the EDG operate at full load under these conditions with NO adverse

effects?

a. One (1) minute

b. Five (5) minutes

c. Until Jacket Water Cooler Outlet temperature exceeds 185°F

d. Until Lube Oil Cooler Outlet temperature exceeds 185'F

ANSWER:

a. One (1) minute

1i% Lo 6

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 19

The plant is operating at 100% power with the following conditions:

Time Ambient Temp CT Basin Temp

1200 35 OF 64 OF

1600 20 OF 60 OF

2000 10 OF 58 OF

Which of the following describes the correct CT Deicing Gate Valve alignment for these

conditions?

1600 2000

a. Full Open Full Open

b. Full Open Half Open

c. Half Open Full Open

d. Half Open Half Open

ANSWER:

b. Full Open Half Open

W1,79 Olfo?

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 20

Given the following conditions:

A fire has occurred in cable spread Room A - RAB 286 which requires a plant

shutdown.

'A' SG pressure is 1000 psig.

'A' SG wide range level is 78%.

'A' SG narrow range level is unavailable.

AFW flow is being supplied to 'A' SG.

Which of the following actions should be taken?

a. Decrease AFW flow to lower 'A' SG wide range level to < 75%

b. Decrease AFW flow to lower 'A' SG wide range level to < 57%

c. Increase AFW flow to raise 'A' SG wide range level to > 57%

d. Increase AFW flow to raise 'A' SG wide range level to > 75%

ANSWER:

a. Decrease AFW flow to lower 'A' SG wide range level to < 75%

WA0,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 21

Given the following conditions:

The plant is operating at 30% power.

All control systems are in automatic.

T-ref fails low.

Which of the following describes the INITIAL response of the rod control system?

a. Step in at 8 steps per minute to reduce Tavg to 553 0 F

b. Step in at 8 steps per minute to reduce Tavg to 557fF

c. Step in at 72 steps per minute to reduce Tavg to 553 0 F

d. Step in at 72 steps per minute to reduce Tavg to 557°F

ANSWER:

d. Step in at 72 steps per minute to reduce Tavg to 557'F

o,06/ /"O"-)

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 22

While establishing a bubble in the PRZ per GP-002, "Normal Plant Heatup From Cold

Solid to Hot Subcritical MODE 5 to MODE 3," letdown pressure control valve 1CS-38

(PK-145.1), Low Pressure Letdown Pressure Controller, opens.

Which of the following describes why PK-145.1 opens?

a. Thermal expansion of liquid in the pressurizer

b. Change in CCW heat load

c. Spray valves are shut while drawing a bubble

d. Switchover of letdown to orifices from RHR-CVCS cross-connect

ANSWER:

a. Thermal expansion of liquid in the pressurizer

I DID K 1. o

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 23

Given the following conditions:

Feed water flow is being transferred from the Main Feed Regulating Bypass Valves

to the Main Feed Regulating Valves.

All six (6) valves are in MANUAL control and are open.

A reactor trip occurs and RCS Tavg stabilizes at no-load conditions.

Which of the following describes the expected status of the Main Feed Regulating Valves

and the Main Feed Regulating Bypass Valves?

a. Main Feed Regulating Valves OPEN

Main Feed Regulating Bypass Valves OPEN

b. e Main Feed Regulating Valves OPEN

Main Feed Regulating Bypass Valves CLOSED

C. Main Feed Regulating Valves CLOSED

Main Feed Regulating Bypass Valves OPEN

d. * Main Feed Regulating Valves CLOSED

Main Feed Regulating Bypass Valves CLOSED

ANSWER:

C. Main Feed Regulating Valves CLOSED

Main Feed Regulating Bypass Valves OPEN

9 0/ X t 0,2

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 24

Given the following conditions:

The plant is being heated up with RCS temperature at 280TF.

Containment pressure is indicating (-) 0.8 inches WG.

ICB-2 & CB-D51 SA, Vacuum Relief 1CB-2 & CB-D51 SA, is in AUTO.

1CB-6 & CB-D52 SB, Vacuum Relief 1CB-6 & CB-D52 SB, is in AUTO.

Assuming NO operator actions, which of the following will automatically occur?

a. 1CB-2 & CB-D51 SA will open when Containment pressure decreases to (-) 1.0

inches WG; 1CB-6 & CB-D52 SB will open if Containment pressure continues to

decrease to (-) 2.25 inches WG

b. 1CB-6 & CB-D52 SB will open when Containment pressure decreases to (-) 1.0

inches WG; 1CB-2 & CB-D51 SA will open if Containment pressure continues to

decrease to (-) 2.25 inches WG

c. 1CB-2 &CB-D51 SA and 1CB-6 & CB-D52 SB will both open when

Containment pressure decreases to (-) 1.0 inches WG

d. 1CB-2 & CB-D51 SA and 1CB-6 & CB-D52 SB will both open when

Containment pressure decreases to (-) 2.25 inches WG

ANSWER:

d. 1CB-2 & CB-D51 SA and 1CB-6 & CB-D52 SB will both open when

Containment pressure decreases to (-) 2.25 inches WG

K14 7Kg V/\/Y CAz

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 25

A loss of 125 VDC bus DP-1B-SB has just occurred.

Which of the following AFW Pumps, if any, are considered inoperable?

a. NO AFW pumps are inoperable

b. ONLY MDAFW Pump lB-SB is inoperable

c. ONLY the TDAFW Pump is inoperable

d. BOTH MDAFW Pump lB-SB and the TDAFW Pump are inoperable

ANSWER:

d. BOTH MDAFW Pump lB-SB and the TDAFW Pump are inoperable

Y// 66 3 l<Z2. OI

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 26

Given the following conditions:

The plant is being maintained at 1900 psig.

RCS temperature is 5000 F and stable.

xccss lztdo'ur and normal letdownarz both-itr ýerie.

cPw'V- CA( oAqrr.-Q /WUt6w&. $i, j-/

The following indications are noted:

"* Normal letdown is 67 gpm

"* RCP 1A seal injection flow is 9 gpm

"* RCP 1B seal injection flow is 7 gpm

"* RCP 1C seal injection flow is 8 gpm

"* RCP 1A seal leakoff flow is 2.5 gpm

"* RCP 1B seal leakoff flow is 2.0 gpm

"* RCP IC seal leakoff flow is 2.5 gpm

In order to maintain pressurizer level constant, charging flow should be adjusted to

indicate ...

a. 36 gpm.

b. 43 gpm.

c. 50 gpm.

d. 74 gpm.

ANSWER:

c. 50 gpm.

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 27

Which of the following describes the start sequence of the Fire Pumps?

a. The Motor Driven Fire Pump will only start after a 30 second time delay if the

Diesel Driven Fire Pump has received a start signal and is not maintaining Ž 100

psig.

b. The Motor Driven Fire Pump will start at * 93 psig and the Diesel Driven Fire

Pump will start at * 83 psig.

c. The Diesel Driven Fire Pump will start at * 93 psig and the Motor Driven Fire

Pump will start at * 83 psig.

d. The Diesel Driven Fire Pump will only start after a 30 second time delay if the

Motor Driven Fire Pump has received a start signal and is not maintaining Ž 100

psig.

ANSWER:

b. The Motor Driven Fire Pump will start at * 93 psig and the Diesel Driven Fire

Pump will start at * 83 psig.

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 28

Given the following conditions:

An operator is required to complete a valve lineup in an area where the radiation level

is 50 mrem/hour.

The operator's current annual Total Effective Dose Equivalent (TEDE) is 1450 mrem.

All of the operator's dose has been received while working at Harris Nuclear Plant.

What is the MAXIMUM time that the operator may work in this area and still remain

within CP&L's Annual Administrative Dose Limit?

a. One (1) hour

b. Eleven (11) hours

c. Fifty-one (51) hours

d. Seventy-one (71) hours

ANSWER:

b. Eleven (11) hours

4 a ,< ,. O

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 29

Given the following:

The unit is at 45% power.

RCP 'B' trips.

All SG level controllers are in AUTO.

NO operator action is taken.

Which of the following describes the response of SG 'B' level?

a. Increases to approximately 70% and stabilizes without any significant decrease in

level during the transient

b. Decreases to approximately 30% and stabilizes without any significant increase in

level during the transient

c. Increases to approximately 70% and then decreases to approximately 30% before

stabilizing

d. Decreases to approximately 30% and then increases to approximately 70% before

stabilizing

ANSWER:

d. Decreases to approximately 30% and then increases to approximately 70% before

stabilizing

49 )/Sc

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 30

Given the following conditions:

Shortly following a loss of offsite power, the following indications are noted on Train

'A' Emergency Safeguards Sequencer (ESS) light box:

CNMT FAN CNMT FAN CNMT FAN CNMT FAN SW BSTR PUMP

HIGH AH-2A HIGH AH-2B LOW AH-2A LOW AH-2B START A

LIT OFF OFF OFF LIT

Prior to AUTO ACT COMPLETE MAN LOAD PERMITTED (Load Block 9)

lighting, a steam break occurs inside Containment, causing a Safety Injection.

Following completion of the sequencer, which of the following indications would be

expected on the Train 'A' ESS light box?

CNMT FAN CNMT FAN CNMT FAN CNMT FAN SW BSTR PUMP

HIGH AH-2A HIGH AH-2B LOW AH-2A LOW AH-2B START A

a. LIT LIT OFF OFF LIT

b. LIT LIT OFF OFF OFF

c. OFF OFF LIT OFF LIT

d. OFF OFF LIT OFF OFF

ANSWER:

c. OFF OFF LIT OFF LIT

,y/V/,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 31

Given the following conditions:

A reactor trip and safety injection occurred several minutes ago.

A loss of offsite power has just occurred.

Both 6.9 KV buses 1A-SA and lB-SB are being supplied by the diesel generators.

Which of the following components has NO power available?

a. Containment Fan Cooler AH- 1

b. Containment Fan Coil Unit AH-37A

c. Primary Shield Cooling Fan S-2A

d. Reactor Support Cooling Fan S-4A

ANSWER:

b. Containment Fan Coil Unit AH-37A

Yv 022/U2OI /

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 32

Given the following plant conditions:

The plant is operating at 100% power.

IlCS-7, 45 GPM Letdown Orifice A, and 1CS-8, 60 GPM Letdown Orifice B, are

closed.

1CS-9, 60 GPM Letdown Orifice C, is open.

The Reactor Makeup System is setup properly and is in AUTO.

VCT level transmitter, LT- 112, fails high.

Assuming NO operator action, which of the following describes the plant response?

a. Charging Pump suction is eventually lost as VCT level decreases

b. 1CS-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the VCT and NO

automatic makeup will occur

c. 1CS-120 (LCV-115A), Letdown VCT/Hold Up Tank, aligns to the HUT and a

CONTINUOUS makeup to the VCT will occur

d. 1CS-120 (LCV-1 15A), Letdown VCT/Hold Up Tank, aligns to the HUT and

INTERMITTENT makeups at normal setpoints will occur

ANSWER:

d. 1CS-120 (LCV-115A), Letdown VCT/Hold Up Tank, aligns to the HUT and

INTERMITTENT makeups at normal setpoints will occur

Doo !A/L 66

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 33

Given the following conditions:

CCW Pump 'A' needs to be removed from service for motor replacement.

CCW Pump 'C' is being aligned to replace CCW Pump 'A'.

Which of the following design features is associated with this evolution AND what is the

basis for this design feature?

a. A key-operated interlock is used to prevent aligning CCW Pumps 'A' and 'C' to

6.9 KV Bus 1A-SA simultaneously

b. A key-operated interlock is used to prevent aligning CCW Pump 'C' to 6.9 KV

Buses 1A-SA and 1B-SB simultaneously

c. A common breaker is used to prevent aligning CCW Pumps 'A' and 'C' to 6.9 KV

Bus 1A-SA simultaneously

d. A common breaker is used to prevent aligning CCW Pump 'C' to 6.9 KV Buses

IA-SA and lB-SB simultaneously

ANSWER:

a. A key-operated interlock is used to prevent aligning CCW Pumps 'A' and 'C' to

6.9 KV Bus tA-SA simultaneously

K-/A o D7

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 34

Given the following conditions:

The plant is currently operating at 30% power.

RCS boron concentration is 900 ppm.

Core burnup is 300 EFPD.

Control Bank 'D' rods are inadvertently withdrawn from 135 steps to 155 steps.

BEFORE RCS temperature increases in response to the rod withdrawal, reactor power

will increase from 30% to approximately ...

a. 32%.

b. 36%.

c. 40%.

d. 44%.

ANSWER:

b. 36%.

DO / t'/<'/,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 35

Given the following conditions:

The plant is operating at 68% power.

0 Control Bank D, Group 1, step counter indicates 187 steps.

S Control Bank D, Group 2, step counter indicates 187 steps.

S Control Bank D rod heights are as follows:

Group 1 Rod Steps

H2 186

B8 186

H14 192

P8 180

Group 2 Rod Steps

F6 186

F10 198

K10 186

K6 180

Which of the following describes the Technical Specification action, if any, that must be

taken within one (1) hour for these conditions?

a. NO actions are required

b. Realign rods F10 and K6 within 12 steps of each other

c. Reduce power below 50%

d. Determine the position of the rods using the movable incore detectors

ANSWER:

a. NO actions are required

Kl7 6 7 L ,-

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 36

Given the following conditions:

The AutoLog is NOT functioning.

The Reactor Operator is maintaining a manual log.

The following log entries have been made:

0956 B-SB CSIP trip

1005 Started A-SA CSIP per AOP-018

1011 Established normal letdown

At 1030, the Reactor Operator realizes he forgot to make a 0957 entry that letdown had

been isolated.

Which of the following entries would be a proper entry in accordance with OMM-0 16,

Operator Logs?

a. * 1030 Isolated normal letdown

b. L.E. 1030 Isolated normal letdown

c. *0957 Isolated normal letdown

d. L.E. 0957 Isolated normal letdown

ANSWER:

d. L.E. 0957 Isolated normal letdown

111-4

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 37

Given the following conditions:

"* Following a large break LOCA, a transition has been made from EPP PATH- I to

EPP-010, "Transfer to Cold Leg Recirculation."

"* The operator attempts to open 1RH-25, RHR A to Charging Pump Suction Valve, and

lRH-63, RHR B to Charging Pump Suction Valve.

"* 1RH-25 opens, but 1RH-63 fails to open.

Which of the following describes a condition that prevents 1RH-63 from opening AND

the actions that should be taken?

a. * 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.

Maintain RHR Train 'B' aligned for Cold Leg Injection until RWST level

decreases to 3% and then secure RHR Train 'B'.

b. * 1SI-301, CNMT Sump to RHR Pump 'B' Suction, failed to open.

Maintain RHR Train 'B' aligned for Cold Leg Injection until RWST level

decreases to 3% and then secure RHR Train 'B'.

c. * 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.

Close 1CS-753, CSIP 'B' Alternate Miniflow Isolation, and open lRH-63,

RHR B to Charging Pump Suction Valve.

d. * 1SI-301, CNMT Sump to RHR Pump 'B' Suction, failed to open.

Open 1SI-311, CNMT Sump to RHR Pump 'B' Suction, and open 1RH-63,

RHR B to Charging Pump Suction Valve.

ANSWER:

c. * 1CS-752, CSIP 'B' Alternate Miniflow, failed to close.

Close 1CS-753, CSIP 'B' Alternate Miniflow Isolation, and open 1RH-63,

RHR B to Charging Pump Suction Valve.

1/4 6 5/-4/,0

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 38

Given the following conditions:

The plant is operating at 100% power.

Spent fuel is being moved in Spent Fuel Pool 'B'.

The suction pipe from Spent Fuel Pool 'B' to the Spent Fuel Pool Cooling Pump

completely severs.

Level in the Spent Fuel Pool will decrease and stabilize at ...

a. 18 feet above the fuel assemblies. Makeup should be initiated using AOP-013,

"Fuel Handling Accident."

b. 18 feet above the fuel assemblies. Makeup should be initiated using OP-1 16,

"Fuel Pool Cooling System."

c. 21 feet above the fuel assemblies. Makeup should be initiated using AOP-013,

"Fuel Handling Accident."

d. 21 feet above the fuel assemblies. Makeup should be initiated using OP-1 16,

"Fuel Pool Cooling System."

ANSWER:

b. 18 feet above the fuel assemblies. Makeup should be initiated using OP- 116,

"Fuel Pool Cooling System."

/9/4 -33A,ý2.o3

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 39

Given the following conditions:

The plant has tripped from 100% power due to a trip of 'B'RCP.

W and 'C' RCPs are running.

'A'

Which of the following is the expected RVLIS Dynamic Head indication?

a. 36%

b. 41%

c. 63%

d. 100%

ANSWER:

c. 63%

t/*/9* 0&Dc,/A*-t3

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 40

Given the following conditions:

The plant is operating at 40% power.

AOP-005, "Radiation Monitoring System," has been entered.

REM-1WC-3544, WPB CCW HX Inlet Monitor, is in HIGH alarm.

As a result of the high alarm, which of the following will automatically close?

a. LCC-252, RCP Thermal Barrier Flow Control Valve

b. 3WC-4, WPB CCW Surge Tank Overflow Valve

c. 1CC-304, CCW to Gross Failed Fuel Detector

d. 3WC-7, WPB CCW Surge Tank Drain Valve

ANSWER:

b. 3WC-4, WPB CCW Surge Tank Overflow Valve

Vl 057AU .05

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 41

The following post-SGTR cooldown procedures all cooldown and depressurize the RCS

to RHR conditions:

EPP-017, "Post SGTR Cooldown Using Backfill"

EPP-018, "Post SGTR Cooldown Using Blowdown"

EPP-019, "Post SGTR Cooldown Using Steam Dump"

Which of the following describe how the depressurization and cooldown in EPP-017

differs from that in EPP-018 and EPP-019?

a. * EPP-017 maintains RCS pressure above the ruptured SG pressure

EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG

pressure

b. * EPP-017 maintains RCS pressure below the ruptured SG pressure

EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG

pressure

c. * EPP-017 maintains RCS pressure below the ruptured SG pressure

EPP-018 and EPP-019 maintain RCS pressure above the ruptured SG pressure

d. * EPP-017 maintains RCS pressure the same as the ruptured SG pressure

EPP-018 and EPP-019 maintain RCS pressure below the ruptured SG pressure

ANSWER:

b. * EPP-017 maintains RCS pressure below the ruptured SG pressure

EPP-018 and EPP-019 maintain RCS pressure the same as the ruptured SG

pressure

/(/4 ~ ýtic,7

0

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 42

Given the following conditions:

A Control Bank 'D' rod has dropped into the core while operating at 100% power.

The operating crew has reduced power to 74%.

Three (3) hours later, they are attempting to withdraw the dropped rod.

In accordance with AOP-001, "Malfunction of Rod Control and Indication System," to

maintain programmed Tavg while recovering the dropped rod ...

a. raise turbine load.

b. reduce turbine load.

c. borate the RCS.

d. dilute the RCS.

ANSWER:

a. raise turbine load.

/?ý 0653c).V4

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 43

The plant is in Mode 1.

VCT pressure has decreased to 8 psig.

Which of the following is the effect on the plant?

a. VCT water flashes to steam

b. Insufficient cooling is available to the No. 2 RCP seals

c. Insufficient seal injection is available to the RCPs

d. CSIPs begin cavitating due to gas binding

ANSWER:

b. Insufficient cooling is available to the No. 2 RCP seals

oo0F

D 0ci4ý

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 44

Given the following conditions:

0 A plant startup is being performed per GP-005, "Power Operation (MODE 2 to

MODE 1)."

The SG PORVs controllers are set at 87%.

The Steam Dump Controller has been incorrectly set at 89%.

While preparing to latch the Main Turbine, RCS temperature will be maintained at

approximately ...

a. 553 0 F.

b. 557fF.

c. 5620 F.

d. 564 0F.

ANSWER:

c. 5620 F.

V 3 7/9A 65

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 45

Given the following conditions:

"* The plant is operating at 100% power when a high radiation condition occurs inside

containment.

"* RC-3561A, Containment Ventilation Isolation radiation monitor (Train A), goes into

high (RED) alarm.

"* RC-3561B, Containment Ventilation Isolation radiation monitor (Train B), is out-of

service for testing.

"* RC-3561C, Containment Ventilation Isolation radiation monitor (Train A), does

NOT respond to the high radiation condition.

"* RC-3561D, Containment Ventilation Isolation radiation monitor (Train B), goes into

high (RED) alarm.

Which train(s) of Containment Ventilation Isolation will actuate, if any?

a. NEITHER Train 'A' NOR 'B'

b. Train 'A' ONLY

c. Train 'B' ONLY

d. BOTH Train 'A' AND 'B'

ANSWER:

d. BOTH Train 'A' AND 'B'

qk 07VA/ 3,6

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 46

Given the following conditions:

The unit is in Mode 4, performing a cooldown on RHR.

Both trains of CCW are in service.

NSW Pump 'A' is operating.

NSW Pump 'B' is in standby.

Both ESW Pumps are available, but are NOT running.

NSW Pump 'A' experiences a sheared shaft.

Which of the following automatically occurs AND what is the effect on the plant

cooldown?

a. * ESW aligns on a low flow signal to cool Train 'A' CCW ONLY

Train 'B' RHR and CCW must be secured.

b. * ESW aligns on a low flow signal to cool BOTH trains of CCW.

Neither train of RHR and CCW must be secured.

c. * ESW aligns on a low pressure signal to cool Train 'A' CCW ONLY.

Train 'B' RHR and CCW must be secured.

d. e ESW aligns on a low pressure signal to cool BOTH trains of CCW.

Neither train of RHR and CCW must be secured.

ANSWER:

d. * ESW aligns on a low pressure signal to cool BOTH trains of CCW.

Neither train of RHR and CCW must be secured.

1/i7// D,ý ,. /

IqK

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 47

Which of the following conditions would permit securing Containment Spray per EOP

PATH-I Guide?

a. & Actuation caused by a LOCA

Time since LOCA occurred is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

Containment pressure is 9 psig

b. * Actuation caused by a LOCA

Time since LOCA occurred is 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

Containment pressure is 5 psig

c. * Actuation caused by a Steam Line Break

Time since Steam Line Break occurred is 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

o Containment pressure is 5 psig

d. * Actuation caused by a Steam Line Break

Time since Steam Line Break occurred is 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

Containment pressure is 9 psig

ANSWER:

c. * Actuation caused by a Steam Line Break

"* Time since Steam Line Break occurred is 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

"* Containment pressure is 5 psig

A6 ' 69 ,D

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 48

Given the following conditions:

The plant is in Mode 3 with Tavg at 557TF.

All systems are in their normal alignment.

Safety Injection is manually actuated inadvertently.

Which of the following describes the impact on Instrument Air inside Containment?

a. * IA-819, Containment Instrument Air, closes

SI and Phase A must BOTH be reset to allow opening IA-819

b. * IA-819, Containment Instrument Air, closes

ONLY SI must be reset to allow opening IA-819

c. * IA-819, Containment Instrument Air, closes

ONLY Phase A must be reset to allow opening IA-819

d. * IA-819, Containment Instrument Air, remains open

NO actions are required to be taken to restore IA to Containment

ANSWER:

c. * IA-819, Containment Instrument Air, closes

ONLY Phase A must be reset to allow opening IA-819

A i 1 0oJ

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 49

Given the following conditions:

The unit is operating at 100% power.

A turbine trip signal is received.

All automatic actions occur, EXCEPT one (1) Throttle Valve fails to close.

Assuming NO operator actions, which of the following describes the expected FINAL

CONDITION of SG pressure and Turbine First Stage Impulse Pressure as compared to

the 100% power conditions?

a. * SG pressure INCREASES

Turbine First Stage Impulse Pressure INCREASES

b. * SG pressure INCREASES

Turbine First Stage Impulse Pressure DECREASES

c. * SG pressure DECREASES

Turbine First Stage Impulse Pressure INCREASES

d. * SG pressure DECREASES

0 Turbine First Stage Impulse Pressure DECREASES

ANSWER:

b. * SG pressure INCREASES

Turbine First Stage Impulse Pressure DECREASES

DýýIql,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 50

Given the following conditions:

A reactor trip occurred due to a loss of offsite power.

The plant is being cooled down on RHR per EPP-006, "Natural Circulation

Cooldown with Steam Void in Vessel with RVLIS."

RCS cold leg temperatures are 190TF.

Steam generator pressures are 50 psig.

RVLIS upper range indicates greater than 100%.

Three CRDM fans have been running during the entire cooldown.

Steam should be dumped from all SGs to ensure ...

a. boron concentration is equalized throughout the RCS prior to taking a sample to

verify cold shutdown boron conditions.

b. all inactive portions of the RCS are below 200TF prior to complete RCS

depressurization.

c. RCS and SG temperatures are equalized prior to any subsequent RCP restart.

d. RCS temperatures do not increase during the required 29 hour3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months vessel soak period.

ANSWER:

b. all inactive portions of the RCS are below 200'F prior to complete RCS

depressurization.

d,/4 vJVo9§x<'7,oK

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 51

Given the following conditions:

During a reactor startup, power has been stabilized at 108 amps.

Main Feed Pump 'A' is operating and maintaining SG levels at program level.

Main Feed Pump 'B' is secured.

Subsequently, SG 'B' level increases to 85%.

Which of the following is the expected status of the following pumps?

a. * Main Feed Pump 'A' RUNNING

Motor Driven AFW Pumps OFF

Turbine Driven AFW Pump OFF

b. * Main Feed Pump 'A' OFF

Motor Driven AFW Pumps RUNNING

Turbine Driven AFW Pump OFF

c. e Main Feed Pump 'A' OFF

Motor Driven AFW Pumps OFF

Turbine Driven AFW Pump RUNNING

d. * Main Feed Pump 'A' OFF

0 Motor Driven AFW Pumps RUNNING

0 Turbine Driven AFW Pump RUNNING

ANSWER:

b. * Main Feed Pump 'A' OFF

"* Motor Driven AFW Pumps RUNNING

"* Turbine Driven AFW Pump OFF

4 0 5 KA/,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 52

Given the following conditions:

S A loss of offsite power has occurred.

S Both Emergency Diesel Generators are loaded.

ALB-024-3-2, DIESEL GENERATOR A TROUBLE, alarms.

S An operator is sent to investigate and reports the following conditions:

Turbo Oil Press 28 psig and stable

Lube Oil Press 30 psig and stable

Fuel Oil Press 1.5 psig and stable

Day Tank Level 56% and slowly decreasing

Starting Air Pressure 227 psig and slowly decreasing

Jacket Water Pressure 17 psig and stable

Control Air Pressure 53 psig and stable

Which of the following components should have automatically started based on these

conditions?

a. Lube Oil Circulating Pump

b. Auxiliary Lube Oil Pump

c. Fuel Oil Transfer Pump

d. Starting Air Compressor

ANSWER:

b. Auxiliary Lube Oil Pump

A0 05614Az

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 53

Given the following conditions:

PRZ pressure is 1685 psig.

PRT pressure is 15 psig.

Which of the following indications support a diagnosis that a PRZ PORV is stuck open?

TEMP

DOWNSTREAM

PRZ LEVEL OF PORV

a. Increasing 613 0F

b. Increasing 250°F

c. Decreasing 613 0F

d. Decreasing 2500 F

ANSWER:

b. Increasing 250°F

1)0 Ftq 14<a ,,_

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 54

Given the following conditions:

A Reactor Startup is being performed.

Initial Source Range Count Rate was 200 count per second (cps).

2500 pcm has been inserted into the core by withdrawing control rods and Source

Range Count Rate has increased to 400 cps.

Rod withdrawal is continued, and an additional 1250 pcm is added to the core.

Which of the following identifies the approximate condition of the core?

a. The reactor is subcritical with a stable count rate of 500 cps

b. The reactor is subcritical with a stable count rate of 600 cps

c. The reactor is subcritical with a stable count rate of 800 cps

d. The reactor is critical with an increasing count rate

ANSWER:

c The reactor is subcritical with a stable count rate of 800 cps

K14A 151S1-S.a06

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 55

During a plant cooldown and depressurization in preparation for a refueling, the SIS

Accumulators are depressurized and then drained.

The normal drain path for the SIS Accumulators is through the Reactor Coolant Drain

Tank ...

a. to the Recycle Holdup Tank.

b. to the Waste Holdup Tank.

c. via the Spent Fuel Pool Cooling System to the Refueling Water Storage Tank.

d. via the Spent Fuel Pool Cooling System to the Transfer Canal.

ANSWER:

a. to the Recycle Holdup Tank.

/-1, ot2/CU/07

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 56

Given the following conditions:

The plant is in Hot Standby.

Letdown flow is 105 gpm.

CSIP 'B' is operating.

A loss of 125 VDC Emergency Bus DP-1B-SB occurs.

With NO operator actions, which of the following is the response of the plant?

a. Seal injection will be lost

b. Charging pump suction will shift to the RWST

c. Letdown line flashing will occur

d. RCS inventory will be lost

ANSWER:

d. RCS inventory will be lost

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 57

Which of the following sets of conditions would require that the Reactor Coolant Pumps

be secured?

a. * RCS is currently at 5250F during a plant heatup

Operating CSIP has tripped

CCW Heat Exchanger outlet temperature is 95°F

ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is NOT alarming

b. * RCS is currently at 375'F during a plant heatup

Operating CSIP has tripped

CCW Heat Exchanger outlet temperature is 11 20F

ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming

c. e RCS is currently at 525F during a plant heatup

CSIP 'A' is operating

CCW Heat Exchanger outlet temperature is 108'F

ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is NOT alarming

d. e RCS is currently at 3750 F during a plant heatup

CSIP 'A' is operating

CCW Heat Exchanger outlet temperature is 122$F

ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming

ANSWER:

b. o RCS is currently at 375$F during a plant heatup

"* Operating CSIP has tripped

"* CCW Heat Exchanger outlet temperature is 112F

"* ALB-5-1-2B, RCP THERM BAR HDR LOW FLOW, is alarming

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 58

Given the following conditions:

A loss of offsite power has occurred.

SG levels are being maintained constant using AFW in manual control.

ERFIS is out-of-service.

SG pressures are at 885 psig and decreasing slowly.

RCS pressure is 1935 psig and stable.

Core exit thermocouples are 6240 F and stable.

RCS hot leg temperatures are 605'F and stable.

RCS cold leg temperatures are 5320 F and decreasing slowly.

The operator is verifying natural circulation flow in EPP-004, "Reactor Trip Response."

Which of the following describes the status of natural circulation flow criteria per EPP

004?

a. The natural circulation criteria of EPP-004 has been met

b. RCS cold leg temperature criteria has NOT been met

c. RCS hot leg temperature criteria has NOT been met

d. RCS subcooling criteria has NOT been met

ANSWER:

d. RCS subcooling criteria has NOT been met

/0A 119 3, o/

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 59

Which of the following would require that Independent Verification be performed in

accordance with OPS-NGGC-1303, "Independent Verification?"

a. During Mode 5, a valve in the Containment Spray system is being repositioned for

testing and the OP lineup will be completed prior to Mode 4 entry

b. During Mode 1, a valve in the Main Steam system is being placed under clearance

and the valve is only accessible with a manlift

c. During Mode 4, a valve in CVCS inside containment is being positioned for

draining and the valve is located in an area where the temperature is 134°F

d. During Mode 3, a valve in CVCS is being placed under clearance and the valve is

located in a radiation field of 175 mRem/hr with an estimated verification time of

6 minutes

ANSWER:

b. During Mode 1, a valve in the Main Steam system is being placed under clearance

and is only accessible with a manlift

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 60

Given the following conditions:

Train 'A' RHR has just been placed in service in accordance with GP-007, "Normal

Plant Cooldown MODE 3 to MODE 5."

Train 'B' RHR is still aligned for ECCS Mode.

Interlock P-12 has been bypassed and the Condenser Steam Dumps are in operation.

Train 'A' equipment is in operation.

Both CSIPs are still available.

RCP 'C' has been secured for the cooldown.

A loss of 6.9 KV Bus 1A-SA occurs and EDG 1A-SA fails to start.

Which of the following describes the impact of the loss of Bus 1A-SA on the plant?

a. TDAFW Pump becomes inoperable

b. RCPs 'A' and 'B' must be secured

c. RHR cooling capability is temporarily lost

d. Condenser steam dump capability is lost

ANSWER:

c. RHR cooling capability is temporarily lost

,0 6 Vl ,

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 61

Given the following conditions:

FRP-P. 1, "Response to Imminent Pressurized Thermal Shock," is being performed.

Safety Injection CANNOT be terminated due to inadequate RCS subcooling.

However, RCS subcooling is adequate to start an RCP.

Which of the following describes the bases for RCP operation under these conditions?

a. Provide additional RCS subcooling

b. Provide mixing of injection water and reactor coolant

c. Supply additional heat input into the RCS

d. Provides normal sprays for the depressurization

ANSWER:

b. Provide mixing of injection water and reactor coolant

/1//i WOE et YK.20 7.->

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 62

Given the following conditions:

REM-3502A, Containment RCS Leak Detection Radiation monitor, is in service.

REM-3502B, Containment Pre-Entry Purge Radiation monitor, is in service.

Which of the following describes the effect on these monitors if a Containment Isolation

Phase 'A' actuation occurs?

a. * REM-3502A remains in service

REM-3502B remains in service

b. e REM-3502A remains in service

a REM-3502B is isolated

c. * REM-3502A is isolated

REM-3502B remains in service

d. * REM-3502A is isolated

REM-3502B is isolated

ANSWER:

c. * REM-3502A is isolated

REM-3502B remains in service

$64 DZ73AdV, X Z

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 63

Given the following conditions:

"* A LOCA has occurred inside Containment, resulting in a reactor trip and a safety

injection.

"* A transition has just been made from EPP PATH- I to FRP-P. 1, "Response to

Imminent Pressurized Thermal Shock."

Containment pressure is 7 psig and increasing slowly.

All RCPs have been secured.

Pressurizer level is off-scale low.

RVLIS Full Range indicates 88%.

Core exit thermocouples are 2400 F and decreasing

RCS cold leg temperatures are 230'F and decreasing.

RCS pressure is 285 psig and stable.

ERFIS indicates subcooling is 177fF.

RHR HX header flows are both 0 gpm.

SG levels are as follows:

SG LEVEL

A 32%

B 10%

C 26%

Which of the following actions should be taken in accordance with FRP-P. 1, "Response

to Imminent Pressurized Thermal Shock?"

a. Maintain total AFW flow > 210 KPPH until at least one (1) SG is >40% level

b. Secure AFW flow to all SGs

c. Maintain cold leg injection flow, but secure one (1) CSIP

d. Return to EOP-PATH-1

ANSWER:

a. Maintain total AFW flow > 210 KPPH until at least one (1) SG is >40% level

%//14

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 64

Given the following conditions:

A loss of secondary heat sink has occurred.

Attempts are made to restore main feedwater using FRP-H. 1, "Response to Loss of

Secondary Heat Sink."

All RCPs are stopped.

SG level wide range levels are all below 5%.

Core exit thermocouple temperatures are increasing.

PRZ pressure is 2180 psig and increasing rapidly.

Which of the following describes the sequence of actions to be taken?

a. * Actuate Safety Injection

Verify all PRZ PORVs automatically open when pressure increases

b. * Actuate Safety Injection

Open all PRZ PORVs after verifying Safety Injection flowpath

c. e Open all PRZ PORVs

Verify Safety Injection automatically actuates when pressure decreases

d. * Open all PRZ PORVs

0 Actuate Safety Injection after verifying the PRZ PORVs are open

ANSWER:

b. & Actuate Safety Injection

Open all PRZ PORVs after verifying Safety Injection flowpath

414 VVc 6Jti/20&/

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 65

Given the following conditions:

"* Reactor power is 8%.

"* The turbine is at 1800 rpm, in preparations for synchronizing to the grid.

"* A reactor trip occurs.

Which of the following describes why the Main Turbine must be tripped under these

conditions?

a. Prevent an uncontrolled RCS cooldown

b. Generate an additional reactor trip signal

c. Minimize the depletion of SG inventory

d. Minimize the pressure increase in the RCS

ANSWER:

a. Prevent an uncontrolled RCS cooldown

3/411 9 D7 /&/ D3

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 66

Given the following conditions:

"* PRZ pressure is being controlled in automatic at 2235 psig.

"* Pressure transmitter PT-444 fails high.

"* Approximately 10 seconds after the failure, the operator places PK-444A in

MANUAL.

Which of the following actions is the operator required to take to restore PRZ pressure to

2235 psig?

a. Raise controller output to cause heaters to energize and spray valves to close

b. Raise controller output to cause spray valves to open and heaters to deenergize

c. Lower controller output to cause heaters to energize and spray valves to close

d. Lower controller output to cause spray valves to open and heaters to deenergize

ANSWER:

c. Lower controller output to cause heaters to energize and spray valves to close

' 7A41.2,,/S

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 67

The plant is in Mode 3 with the Shutdown Banks withdrawn when the following events

occur:

The reactor trip breakers open.

ALB-15-2-2, PIC 1-2-3-4-9-10-13-14 POWER FAILURE, alarms.

ALB-15-4-3, PIC 17-18 POWER FAILURE, alarms.

Most lights in the top row of Trip Status Light Boxes are energized.

Several lights in each of the other rows of Trip Status Light Boxes are energized.

ALB-15-1-4, 60 KVA UPS TROUBLE, remains clear.

ALB-15-1-5, 7.5 KVA UPS TROUBLE, remains clear.

ALB-15-3-2, PIC 5-6-7-8-11-12-15-16 POWER FAILURE, remains clear.

ALB-15-5-3, PIC 19 POWER FAILURE, remains clear.

Which of the following buses have been lost?

a. Instrument Bus S-I

b. Instrument Bus S-II

c. UPS Bus UPP-1A

d. UPS Bus UPP-1B

ANSWER:

a. Instrument Bus S-I

d 0 57,q -/S

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 68

Given the following conditions:

"* The crew diagnosed a SG tube leak.

"* REM-1BD-3527, Steam Generator Blowdown, went into high (RED) alarm.

"* In response to the alarm on REM-1BD-3527, the crew performed the required actions

of AOP-016, "Excessive Primary Plant Leakage," Attachment 1, "Primary-To

Secondary Leak."

Which of the following describes the expected indicated trend on REM- IBD-3527 after

the completion of Attachment 1?

a. Stabilizes and then decreases

b. Stabilizes and remains constant

c. Increases and stabilizes at full scale

d. Stabilizes and then increases

ANSWER:

a. Stabilizes and then decreases

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 69

Given the following conditions:

FRP-C. 1, "Response to Inadequate Core Cooling," is being performed following a

small break LOCA.

Containment pressure is 8.5 psig.

Core exit thermocouples are >1400TF.

All efforts to establish SI flow have failed.

The crew has started RCP 'C' in an attempt to lower core exit temperatures, but

temperatures have remained above 1300TF.

SG 'C' level is 55%.

SGs 'A' and 'B' are off-scale low.

Which of the following actions should be taken?

a. Open the PRZ PORVs and RCS vent valves

b. Start RCPs 'A' and 'B' one at a time

c. Close any open PRZ PORVs and RCS vent valves

d. Refill and repressurize the SI Accumulators for continued injection

ANSWER:

a. Open the PRZ PORVs and RCS vent valves

//-/d 071 v6 65

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 70

Given the following conditions:

The unit is in the Source Range during a reactor startup.

Power is lost to Instrument Bus S-III.

A reactor trip occurs.

Which of the following signals caused the reactor trip?

a. Source Range High Count Rate

b. Intermediate Range High Flux

c. Power Range Neutron Flux (Low Setpoint)

d. Turbine Trip

ANSWER:

d. Turbine Trip

/04 a, ,V I;?

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 71

After plant control is completely shifted to the Auxiliary Control Panel in accordance

with AOP-004, "Remote Shutdown", which of the following actions will the operators

have to manually perform?

a. Align CSIP suction to the RWST

b. Transfer control of the EDGs to the local control panels

c. Open the reactor trip breakers

d. Block SIAS to the Emergency Sequencers

ANSWER:

a. Align CSIP suction to the RWST

/A ogA,1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 72

Given the following conditions:

"* During a plant startup, Main Feed Water is aligned to the SGs through the Feed Reg

Valve Bypass FCVs.

"* The controller for FCV-479, SG 'A' Feed Reg Valve Bypass FCV (FK-479.1), has

just been placed in AUTO.

"* The controller for FCV-489, SG 'B' Feed Reg Valve Bypass FCV (FK-489.1), is still

in MANUAL.

"* The controller for FCV-499, SG 'C' Feed Reg Valve Bypass FCV (FK-499.1), is still

in MANUAL.

"* FCV-479 begins going open.

Which of the following failures could have caused the response of FCV-479?

a. SG 'A' Feed Flow Channel FT-475 failing low

b. SG 'A' Steam Flow Channel FT-476 failing high

c. SG 'A' Level Channel LT-476 failing high

d. Power Range Channel N-44 failing high

ANSWER:

d. Power Range Channel N-44 failing high

J0 0351q 3 1 0 /

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 73

Which of the following describes why RCP trip criteria is included in PATH-2?

a. Protect against operator misdiagnosis since RCS pressure should not decrease to

the trip criteria during a SGTR

b. Decrease leakage from the RCS since the total leakage for the duration of the

SGTR is less than it would have been with the RCPs in service

c. Prevent heatup of the RCS since a heatup of the RCS due to the RCPs being in

service increases leakage to the ruptured SG

d. Protect the RCPs from operating with inadequate AP across the number one RCP

seal as a result of the RCS depressurization from the SGTR

ANSWER:

a. Protect against operator misdiagnosis since RCS pressure should not decrease to

the trip criteria during a SGTR

,CIIAE/K/ t o2

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 74

Which of the following describes how the Emergency Sequencer is reset following a loss

of AC power to 6.9 KV Bus 1A-SA which results in actuation of the Sequencer UV

Program?

a. The operator resets the program by turning the SI Reset switch to RESET at least

2.5 minutes after Load Block 9 is completed

b. The operator resets the program by placing both Reactor Trip Breaker A-SA and

Reactor Trip Breaker B-SB to the closed position momentarily after all actuation

signals have been cleared

c. The program automatically resets when Auxiliary Bus D To Emergency Bus A

SA Breaker 104 and Emergency Bus A-SA To Aux Bus D Tie Breaker 105 SA are

closed during the restoration of offsite power

d. The program automatically resets when Diesel Generator A-SA Breaker 106 SA is

opened during the restoration of offsite power

ANSWER:

c. The program automatically resets when Auxiliary Bus D To Emergency Bus A

SA Breaker 104 and Emergency Bus A-SA To Aux Bus D Tie Breaker 105 SA are

closed during the restoration of offsite power

D0D6 ft9 Df

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 75

Given the following conditions:

"* FRP-S. 1, "Response to Nuclear Power Generation / ATWS," is being performed.

"* The operating crew is about to exit FRP-S. 1.

Boration should continue even after exiting FRP-S.1 to ensure ...

a. adequate shutdown margin is established since the criteria for exiting FRP-S. 1 is

only that the reactor be subcritical.

b. the reactor becomes subcritical since the criteria for exiting FRP-S. 1 is only that

the power range channels indicate < 5%.

c. cold shutdown boron concentration is achieved since additional boron, beyond that

needed to make the reactor subcritical, is required to compensate for the cooldown

portion of the recovery.

d. refueling boron concentration is achieved since additional boron, beyond that

needed to make the reactor subcritical, is required to allow for core offloading to

inspect for fuel damage.

ANSWER:

a. adequate shutdown margin is established since the criteria for exiting FRP-S. 1 is

only that the reactor be subcritical.

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 76

Given the following conditions:

At 0645, both RHR pumps were declared inoperable.

At 0700 today, during repair efforts, a Maintenance person exited the area after

receiving a Total Effective Dose Equivalent of 5800 mRem.

At 0730 today, a plant shutdown was commenced due to both RHR pumps being

inoperable.

When are the notifications to the NRC required to be completed by for these events?

a. * 0745 today for the plant shutdown

0800 today for the over-exposure

b. * 0745 today for the plant shutdown

0700 tomorrow for the over-exposure

c. * 1130 today for the plant shutdown

0800 today for the over-exposure

d. 9 1130 today for the plant shutdown

0 0700 tomorrow for the over-exposure

ANSWER:

d. * 1130 today for the plant shutdown

0700 tomorrow for the over-exposure

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 77

Given the attached form from OST-1093 (next page) and the following conditions:

"* Maintenance has been performed on 1CS-752 SB, Charging/SI Pump B-SB Alternate

Miniflow.

"* A full flow test of the valve has been performed in accordance with OST-1093,

"CVCS/SI System Operability Train B."

"* Stroke time in open direction was 5.06 seconds.

"* Stroke time in closed direction was 8.02 seconds.

Which of the following conditions apply to the results of the test?

a. e Declare the valve operable

No additional paperwork is required

b. * Retest the valve if no mechanical failures are known to exist

0 If the valve is within limits on retest, declare the valve operable

No additional paperwork is required

c. * Retest the valve if no mechanical failures are known to exist

If the valve is within limits on retest, declare the valve operable

Initiate a Condition Report identifying the test results

d. * Declare the valve inoperable

Initiate a Condition Report identifying the test results

ANSWER:

c. * Retest the valve if no mechanical failures are known to exist

"* If the valve is within limits on retest, declare the valve operable

"* Initiate a Condition Report identifying the test results

,12

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

ENSURE OST-1093 ATTACHMENT INSERTED HERE

(REMOVE THIS PAGE WHEN INSERTED)

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 78

Given the following conditions:

"* A LOCA occurred several hours ago, resulting in a start of both Containment Spray

Pumps.

"* Only one (1) Containment Spray Pump is currently running due to actions taken in

EPP-012, "Loss of Emergency Coolant Recirculation."

"* A transition has just been made to FRP-J. 1, "Response to High Containment

Pressure."

"* Containment Pressure is 14 psig.

Which of the following actions should be taken?

a. Restart the second Containment Spray Pump if Containment pressure does NOT

decrease below 10 psig before exiting FRP-J.1.

b. Restart the second Containment Spray Pump since pressure is above 10 psig.

c. Continue operation with one Containment Spray Pump.

d. Continue operation with one Containment Spray Pump unless Containment

pressure begins increasing, then start the second pump.

ANSWER:

c. Continue operation with one Containment Spray Pump.

///W/1/43,S/L

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 79

In addition to Radiation Levels inside containment, which of the following parameters

are used to determine whether an entry is required to be made into EPP-FRP-J.3,

"Response to Containment High Radiation Level?"

a. Containment Sump Levels and Containment Ventilation Isolation status

b. Containment Pressure and Containment Sump Levels

c. Containment Pressure and Containment Ventilation Isolation status

d. Containment Sump Levels and Containment Hydrogen Concentration

ANSWER:

b. Containment Pressure and Containment Sump Levels

/6 F/3/4to/0

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 80

Given the following conditions:

"* Just prior to shift change, the oncoming Reactor Operator calls in sick.

"* The shift schedule shows the oncoming crew at minimum complement with the

Reactor Operator, but there is a Licensed Operator (CO) scheduled for the RAB.

The S-SO should

a. use the RAB CO in the control room and replace the RAB whenever possible.

b. use the RAB CO in the control room and call in a replacement RAB within two

hours.

c. hold the off-going CO until the S-SO can ensure a replacement will arrive within

two hours.

d. hold the off-going CO until a replacement can relieve the off-going CO.

ANSWER:

d. hold the off-going CO until a replacement can relieve the off-going CO.

/9V * 6z1 5/

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 81

Given the following conditions:

On May 1, at 0100, a plant shutdown was initiated from 100% in preparations for

conducting a refueling.

The reactor was shutdown at 1100 on the same date.

CCW heat exchanger outlet temperature is currently 86.80 F.

When is the EARLIEST that fuel movement in the reactor vessel is allowed to begin?

a. May 5th at 0500

b. May 5th at 1500

c. May6thatO100

d. May6that 1100

ANSWER:

d. May6that 1100

6K62 §b?

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 82

Which of the following satisfies the Technical Specification bases requirement for off

site power distribution?

a. The requirement can only be satisfied by the off-site transmission lines that feed

the SUTs directly (Cary Regency Park and Cape Fear North)

b. The requirement can only be satisfied by the off-site transmission lines that do not

feed the respective north or south switchyard bus through a jumper

c. The requirement is satisfied as long as the switchyard alignment is such that power

is available from the off-site transmission network to both SUTs regardless of the

number of transmission lines available

d. The requirement is satisfied as long as there are two separate off-site transmission

lines that can power the SUTs (either through the switchyard or directly)

ANSWER:

d. The requirement is satisfied as long as there are two separate off-site transmission

lines that can power the SUTs (either through the switchyard or directly)

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 83

Given the following conditions:

During a refueling outage, the SRO-Fuel Handling reports that the crew is having

difficulties loading several fuel assemblies in the vicinity of the hot legs due to the

flow through the piping.

He has requested that the RHR system be secured to allow loading the assemblies.

He estimates that it will take up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to load the assemblies in the vicinity of the

hot legs.

Which of the following identifies the MAXIMUM number of consecutive hours the

RHR system may be secured under these conditions?

a. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

b. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

c. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

d. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

ANSWER:

a. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

dA ý,') ) (

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 84

A group of armed intruders is attempting to enter the Protected Area. Security has

notified the Control Room that a deviation from the Security Plan is necessary to combat

the intruders.

Which of the following is required, according to PRO-NGGC-0200, "Procedure Use and

Adherence?"

a. The deviation shall be approved by the Manager - Operations prior to performing

the action

b. The deviation shall be approved by the Superintendent - Shift Operations prior to

performing the action

c. The state and counties must be notified as soon as possible after performing the

action and within 60 minutes in all cases

d. The NRC must be notified prior to performing the action

ANSWER:

b. The deviation shall be approved by the Superintendent - Shift Operations prior to

performing the action

ý1(74C--,),),)

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 85

Given the following conditions:

Following a reactor trip and safety injection concurrent with a loss of offsite power, a

transition has been made to EPP-015, "Uncontrolled Depressurization Of All Steam

Generators."

Emergency Diesel Generator 1B-SB has tripped and cannot be restarted.

The TD AFW pump has tripped on overspeed and cannot be reset.

MDAFW pump 1A-SA is tagged out.

SG 'A' narrow level is 15%.

SG 'B' and 'C' narrow range levels are off-scale low.

Core exit thermocouple temperatures are all between 705'F and 720F.

Which of the following actions should be taken?

a. Continue in EPP-015, "Uncontrolled Depressurization Of All Steam Generators"

b. Transition to EPP-001, "Loss of AC Power to 1A-SA and lB-SB Buses"

c. Transition to EPP-FRP-C.1, "Response to Inadequate Core Cooling"

d. Transition to EPP-FRP-H. 1, "Response to Loss of Secondary Heat Sink"

ANSWER:

d. Transition to EPP-FRP-H.1, "Response to Loss of Secondary Heat Sink"

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 86

Given the following conditions:

The unit is in Mode 5 with the RCS filled.

RHR Train 'A' is in operation.

RHR Train 'B' is operable, but not in operation.

SG wide range levels are:

SG LEVEL

A 81%

B 68%

C 63%

Maintenance requests that RHR Pump 'B' be removed from operable status for

several hours for minor maintenance.

Which of the following describes the acceptability of removing RHR Pump 'B' from

service under these conditions?

a. It may NOT be done because the SGs are not an adequate heat sink under these

conditions.

b. It may NOT be done because two RHR trains are required at all times for Mode 5.

c. It may be done as long as the RCS remains filled.

d. It may be done as long as RCS temperature remains below 200TF.

ANSWER:

a. It may NOT be done because the SGs are not an adequate heat sink under these

conditions.

A/, 33

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 87

Given the following conditions:

Safety injection is being terminated in accordance with EPP-008, "SI Termination."

The operator reports 1SI-3, BIT Outlet, is closed as directed, but 1SI-4, BIT Outlet,

will NOT close.

An operator unsuccessfully attempts to locally close 1SI-4.

Which of the following actions should be taken?

a. * Unlock and close 1SI-2, BIT Inlet, ONLY

0 Establish normal charging flow while waiting for 1SI-2 to be closed

b. * Unlock and close 1SI-2, BIT Inlet, ONLY

Wait until 1SI-2 is closed before establishing normal charging flow

c. & Unlock and close BOTH 1S-1, BIT Inlet, and ISI-2, BIT Inlet

Establish normal charging flow while waiting for 1S-1 and 1SI-2 to be closed

d. a Unlock and close BOTH 1S-1, BIT Inlet, and 1SI-2, BIT Inlet

Wait until 1SI-1 and 1SI-2 are closed before establishing normal charging flow

ANSWER:

d. * Unlock and close BOTH 1S-1, BIT Inlet, and 1SI-2, BIT Inlet

Wait until 1S-1 and 1SI-2 are closed before establishing normal charging flow

)V4 IA,-E AA-.0Z

v§A

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 88

Given the following conditions:

At 0530, RCS temperature was being maintained at 550 0 F.

A small break LOCA occurred.

At 0545, the crew is ready to commence a cooldown to cold shutdown in accordance

with EPP-009, "Post LOCA Cooldown and Depressurization."

RCS temperature at 0545 is 490'F.

Which of the following identifies the lowest allowable temperature of the RCS at 0630 if

the crew begins the MAXIMUM permissible cooldown rate AND the basis for this

temperature limit?

a. 450'F to ensure that a transition is NOT required to be made to FRP-P. 1,

"Response to Imminent Pressurized Thermal Shock"

b. 450'F to ensure that Technical Specification cooldown limits are NOT exceeded

c. 415'F to ensure that a transition is NOT required to be made to FRP-P. 1,

"Response to Imminent Pressurized Thermal Shock"

d. 415TF to ensure that Technical Specification cooldown limits are NOT exceeded

ANSWER:

b. 450'F to ensure that Technical Specification cooldown limits are NOT exceeded

W$ 6 3/ Y ,2

41

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 89

Given the following conditions:

"* Following a loss of offsite power, EPP-001, "Loss of AC Power to 1A-SA and lB-SB

Buses," is being performed.

"* Safety Injection has been actuated and reset.

"* Attachment 5, "6.9 KV Emergency Bus Breakers," has been completed and all

breakers have been verified open.

"* The SGs are being depressurized.

"* Several minutes later, Emergency Diesel Generator 1A-SA is started.

"* SG pressures are stabilized.

"* ESW Pump 1A-SA is started and the valve alignment for Header 'A' has been

verified.

Plant conditions are now:

"* EDG 1A-SA is running.

"* ESW Pump 1A-SA is running.

"* NO other pumps are running.

"* NO SI valves have repositioned from their "at power" position.

"* RCS pressure is 1400 psig.

"* RCS temperature is 4920 F.

"* RCS subcooling is 96 0 F.

"* PRZ level is 6%.

Which of the following identifies the procedure(s) to be used for recovery from this

condition?

a. EPP-002, "Loss Of All AC Power Recovery Without SI Required"

b. EPP-003, "Loss Of All AC Power Recovery With SI Required"

c. EOP-PATH-1 and AOP-025, "Loss of One Emergency AC Bus or One

Emergency DC Bus," performed concurrently

d. EOP-PATH-1 and FRP-I.2, "Response to Low Pressurizer Level," performed

concurrently

ANSWER:

b. EPP-003, "Loss Of All AC Power Recovery With SI Required"

9, ,/ S,,._-5,5

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 90

Conditions meeting the Emergency Classification criteria for a Notification of Unusual

Event have been determined to have existed, but no longer exist.

As the Site Emergency Coordinator you should ...

a. declare and terminate the event in a single notification message.

b. declare the event in a notification message and terminate the event in a followup

message.

c. notify the NRC of the conditions, but NO notifications to the state and county

would be performed.

d. notify Licensing of the need to generate an LER, but no other notifications would

be performed.

ANSWER:

a. declare and terminate the event in a single notification message.

g'ý2'V q6

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 91

A reactor startup is being performed following a mid-cycle outage per GP-004, "Reactor

Startup (Mode 3 to Mode 2)."

Estimated Critical Conditions are as follows:

TIME 1830

BORON CONC. 1215 ppm

CONT BANK 'C' POSTION 218 steps

CONT BANK 'D' POSTION 90 steps

ECC - 500 PCM POSITION 45 steps on Bank 'D'

ECC + 500 PCM POSITION 197 steps on Bank 'D'

ROD INSERTION LIMIT 0 steps on Bank 'D'

The Actual Critical Conditions are as follows:

TIME 1836

BORON CONC. 1198 ppm

CONT BANK 'C' POSTION 110 steps

CONT BANK 'D' POSTION 0 steps

Which of the following actions must be taken?

a. Shut down the reactor using OP-104, "Rod Control System," AND borate, as

needed, to increase RCS boron concentration to 1215 ppm

b. Maintain critical conditions AND borate, as needed, to increase RCS boron

concentration to 1215 ppm

c. Shut down the reactor using OP-104, "Rod Control System," AND initiate

Emergency Boration per AOP-002, "Emergency Boration"

d. Trip the reactor AND initiate Emergency Boration per AOP-002, "Emergency

Boration"

ANSWER:

c. Shut down the reactor using OP-104, "Rod Control System," AND initiate

Emergency Boration per AOP-002, "Emergency Boration"

t/14 0 ý_ 4 ,I/

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 92

Given the following conditions:

A small break LOCA has occurred.

Entry has been made into FRP-C. 1, "Response to Inadequate Core Cooling."

Core exit thermocouples are all indicating between 740 'F and 760 TF and rising

slowly.

RCS pressure has stabilized at 805 psig.

PZR level is off-scale low.

RVLIS Full Range is indicating 32% and lowering slowly.

NO CSIPs are available.

SG narrow range levels are all off-scale low.

Total AFW flow to the SGs is 240 KPPH.

Which of the following actions should be taken?

a. Dump steam to cooldown and depressurize the RCS to cause the SI accumulators

to dump

b. Open the RCS Head Vent valves to depressurize the RCS to cause the SI

accumulators to dump

c. Start an RCP immediately to provide forced cooling flow

d. Open the PRZ PORVs to depressurize the RCS to cause the SI accumulators to

dump

ANSWER:

a. Dump steam to cooldown and depressurize the RCS to cause the SI accumulators

to dump

I/ld

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 93

Given the following conditions:

A large break LOCA has occurred.

EPP-012, "Loss of Emergency Coolant Recirculation," is being performed.

One (1) CSIP is operating with a flow rate of 520 gpm.

One (1) RHR pump is operating with a flow rate of 3350 gpm.

Time after trip and SI is 73 minutes.

SI CANNOT be terminated due to insufficient subcooling.

Which of the following actions should be taken to MINIMIZE SI flow while still

maintaining the minimum required flow for decay heat removal?

a. Stop the CSIP

b. Start the standby CSIP

c. Manually throttle high head SI flow

d. Stop the RHR pump

ANSWER:

d. Stop the RHR pump

7*4z

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 94

Given the following conditions:

A reactor startup is in progress.

Power level is stable at 10-8 amps.

Electrical Maintenance reports there is a potential problem with the inverter for

Instrument Bus IDP-1A-SI and recommends placing the bus on the alternate power

supply (PP- 1A21 1-SA).

Which of the following describes the effect of permitting this re-alignment?

a. NO reactor trip occurs, but the reactor startup is delayed due to C-1, Intermediate

Range Rod Stop

b. NO reactor trip occurs, but the reactor startup is delayed due to C-2, Power Range

Overpower Rod Stop

c. Reactor trip on Intermediate Range High Flux

d. Reactor trip on Power Range High Flux Low Setpoint

ANSWER:

c. Reactor trip on Intermediate Range High Flux

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 95

Given the following conditions:

A reactor trip and safety injection occurred.

During the performance of PATH-1, an ORANGE path was noted on the Core

Cooling status tree and a transition was made to the appropriate procedure.

Which of the following describes how the CSF status trees should be monitored at this

point?

a. Suspend monitoring until actions have been completed for the ORANGE path

condition

b. Monitor for information only until actions have been completed for the ORANGE

path condition

c. Monitor every 10 to 20 minutes

d. Monitor continuously

ANSWER:

d. Monitor continuously

A/ 6102,/ /1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 96

Given the following conditions:

The plant is operating at 50% power

Train 'A' safety equipment is in service

ALB 24-1-2, 6.9kV EMER BUS A-SA TROUBLE, in alarm

ALB 25-1-2, 6.9kV EMER BUS B-SB TROUBLE, in alarm

AEP-2-8, DEGRADED VOLTAGE, in alarm

AEP-2-9, DEGRADED VOLTAGE, in alarm

Emergency 6.9 kV Buses IA-SA and lB-SB both indicating approximately 6500

volts

Emergency 480V Buses all indicating approximately 450 volts

Which of the following Emergency Buses will be first to be supplied by its EDG AND

which procedure will be used to direct this action?

a. * Emergency Bus A-SA

AOP-028, "Grid Instability"

b. * Emergency Bus A-SA

OP-155, ""Diesel Generator Emergency Power System"

c. * Emergency Bus B-SB

AOP-028, "Grid Instability"

d. * Emergency Bus B-SB

OP-155, ""Diesel Generator Emergency Power System"

ANSWER:

c. * Emergency Bus B-SB

AOP-028, "Grid Instability"

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 97

Given the following conditions:

"* A LOCA outside containment has resulted in unsafe radiological conditions in the

RAB.

"* The crew has taken all the actions of EPP-013, "LOCA Outside Containment," to

isolate the break.

Which of the following is the PRIMARY indication used in EPP-013 that the actions

taken have been successful AND which procedure should be transitioned to when the

isolation is successful?

a. * RAB sump level alarms clearing

Transition to PATH- 1

b. * RCS pressure increasing

Transition to PATH-1

c. * RAB sump level alarms clearing

Transition to EPP-008, "SI Termination"

d. * RCS pressure increasing

Transition to EPP-008, "SI Termination"

ANSWER:

b. * RCS pressure increasing

Transition to PATH-1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 98

Given the following conditions:

A reactor trip occurred from 23% power.

Shutdown Bank 'B' Rod L-5 is indicating 228 steps.

Control Bank 'C' Rod K-8 is indicating 6 steps.

All other rods have the Rod Bottom Lights lit.

RCS boron concentration at the time of the trip was 845 ppm.

The plant is to be maintained at no-load Tavg.

Which of the following actions should be taken AND what is the MINIMUM RCS boron

concentration that must be achieved?

a. Emergency Borate to raise RCS boron concentration to 1307 ppm

b. Emergency Borate to raise RCS boron concentration to 2282 ppm

c. Normal Borate to raise RCS boron concentration to 1307 ppm

d. Normal Borate to raise RCS boron concentration to 2282 ppm

ANSWER:

a. Emergency Borate to raise RCS boron concentration to 1307 ppm

3/43

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 99

Given the following conditions:

The plant is operating at 40% power.

A fire alarm has been received.

Which of the following conditions would require that a plant shutdown be required at the

earliest time?

a. * RHR Pump 1A-SA has been out-of-service for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for maintenance

The fire requires de-energizing Emergency Bus 1A-SA

b. * RHR Pump 1A-SA has been out-of-service for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for maintenance

The fire is contained in the CSIP 1A-SA pump room

c. * Containment Spray Pump lB-SB has been out-of-service for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for

maintenance

The fire requires de-energizing Aux Bus B

d. & Containment Spray Pump lB-SB has been out-of-service for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for

maintenance

The fire is contained in the CSIP 1A-SA pump room

ANSWER:

c. * Containment Spray Pump 1A-SA has been out-of-service for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months for

maintenance

The fire requires de-energizing Aux Bus B

D61 99$?-1

Harris Nuclear Plant

August 2002 - SRO Exam

ANSWER KEY

QUESTION: 100

Given the following plant conditions:

A small break LOCA has occurred.

A transition has been made to EPP-009, "Post LOCA Cooldown and

Depressurization."

Containment pressure is 6.1 psig.

RCS subcooling is 55 0F by ERFIS.

PRZ level is 31%.

Both CSIPs are injecting through the BIT.

Both RHR pumps are secured.

The operators are depressurizing the RCS to refill the pressurizer to > 40% when

subcooling is noted to decrease to 350 F.

Which of the following actions should be taken?

a. Continue the depressurization in EPP-009

b. Stop the depressurization and continue in EPP-009

c. Stop the depressurization and transition to PATH-I

d. Reinitiate SI and transition to PATH-1

ANSWER:

a. Continue the depressurization in EPP-009

Harris Nuclear Plant

SRO Written Reference

SRO SUPPLIED REFERENCES

AOP-036, Attachment 6 ............ SG Wide Range Level vs. SG Pressure

AP-617, Attachment 1 ............... Immediate Notification Requirements

EPP-012, Attachment 1 ............ Minimum SI Flow Rate vs Time After Reactor Trip

OP-107, Attachment 19 ............. Makeup Concentration Limits

OP- 141, Attachment 5 ............... Cooling Tower Cold Weather Operation

PLP-114, Attachment 2 ............ Refueling Operations

Curves A-I 1-6 through -1 ......... Differential and Integral Rod Worth Curves

Curves C- Il-I through -3 ........... Power Defect Curves

Steam Tables

SAFE SHUTDOWN FOLLOWING A FIRE

I

Attachment 6

Sheet 1 of 1

SG Wide Range Level vs. SG Pressure

100

DO

4,

.4

r

it

Ig 70

rA

55

2(13 4w1 roe (AD OM am InM RD 11w 1200

InD 2(11

SG Pressure (PSIG)

- END ATTACHMENT 6 -

Rev. 19

Attachment I

Sheet I of 8

IMMEDIATE NOTIFICATION REQUIREMENTS

The following tables are divided into sections based upon the time allowed for reporting

the respective events as follows:

I One Hour Notifications

II Four Hour Notifications

III Eight Hour Notifications

IV Twenty-four Hour Notifications

NOTE: The events listed in this attachment may be concurrent with conditions that

result in a declared emergency. In the case of a declared emergency, the

notification made under the Emergency Plan and implementing procedures

satisfies the notifications required by this procedure. Written reports will be

based on §50.73 and Technical Specifications regardless of whether the initial

notification is made under the Emergency Plan or this procedure.

I. ONE HOUR NOTIFICATIONS

I.A. OPERATIONAL EVENTS -10 CFR 50.72 (b) (1)

1. Technical Specification Deviations (10 CFR 50.54x)

2. Safety Limit Violation (TS 6.7.1)

I.B. RADIOLOGICAL EVENTS

1. Radioactive Shipments (Note 1)

2. Loss or Theft of Licensed Material/Radiological Sabotage (Note 2)

3. Exposure to Individuals or Releases (Note 3)

4. Accidental Criticality in the Fuel Handling Building (Note 4)

I.C. SECURITY THREAT (Note 10)

1. Adversary Threat

2. Security Program Vulnerabilities

3. International Atomic Energy Agency (IAEA) Representative

I.D. FITNESS FOR DUTY (Note 11)

1. FFD - NRC Employee

AP-617 Rev. 18 Page 11 of 37

Attachment 1

Sheet 2 of 8

IMMEDIATE NOTIFICATION REQUIREMENTS

II. FOUR HOUR NOTIFICATIONS

OPERATIONAL EVENTS 10 CFR 50.72 (b) (2)

1. Initiation of any Nuclear Plant Shutdown required by Technical

Specifications.

2. Unplanned Actuation of the reactor protection system (scram) when the

reactor was critical and any event that results or should have resulted in

ECCS discharge into the RCS.

3. Off-Site Notification Has Been or Will Be Made

III. EIGHT HOUR NOTIFICATIONS

1. Degraded or Unanalyzed Condition

2. Loss of Emergency Response Capability (Note 5)

3. Unplanned Actuation of selected ESF Systems

Refer to NUREG 1022 System Actuation to identify applicable

system actuations.

4. Loss of a Safety Function

5. Transport of Contaminated Individual

IV. TWENTY-FOUR HOUR NOTIFICATIONS

1. EXPOSURE TO INDIVIDUALS OR RELEASES

a. Radiological Exposure/Release (Note 6)

b. Other Releases (Note 7)

2. VIOLATION OF OPERATING LICENSE CONDITIONS (Note 8)

3. FITNESS FOR DUTY PROGRAM EVENTS (Note 9)

AP-617 Rev. 18 Page 12 of 37

Attachment 1

Sheet 3 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

NOTIFICATION REFERENCE WRITTEN FOLLOW-UP

RADIOACTIVE SHIPMENTS

§20.1906(d)(1) NRC Notification Also

Removable contamination from a received package Required per

§71.87(i)

containing radioactive material in excess of the limits §20.1906(d)(1)

specified in §71.87(i)

§20.1906(d)(2) NRC Notification Also

Radiation levels from a received package of Required per

radioactive material in excess of the limits specified in §71.47

§71.47. §20.1906(d)(2)

§20.1906(d)(1)

The involved H.P. Supervisor shall immediately notify

the final delivery carrier. Follow-up NRC notification

shall be made by Regulatory Affairs §73.71(a)(5)

Security threats or theft of licensed material shall be

reported to site Security personnel. After initial §73.71(b)(2)

notification or after submission of 30 day report,

additional information shall be reported to NRC as it is

available and within 30 days of discovering additional

information. Per §73.71 (a)(5) and §73.71 (b)(2),

significant supplemental information which becomes

available after the initial telephonic notification or after

the submission of the written report must be

telephonically reported to the NRC Operations Center

and also submitted in a revised written report. (Written

reports will be submitted on USNRC Form 366 and will

be provided a number unique to Safeguards Events.

These reports will not be a part of the AEOD tracking

program for LERs.)

2. LOSS OR THEFT OF LICENSED MATERIAL/ RADIOLOGICAL SABOTAGE

Any loss or theft or attempted theft of:

a) Licensed material in an aggregatequantity equal §20.2201(a)(1)(i) 30 Day WrittenperReport

§20.2201(d) also required

to or greater than 1,000 times the quantity §70.52(b) §20.2201(b)

specified in Appendix C to §20.1000-§20.2401

under such circumstances that it appears that an

exposure could result to persons in unrestricted

areas,

§73.71 (a) (loss/theft 30 Day Written Report

b) Any Special Nuclear Material or spent fuel, only) also required per

§74.11 §73.71(a)

§150.16(b) 15 Day Written Report

may also be required per

§150

15 Day Written Report

c) Greater than 10 curies of tritium at any one time or §30.55(c) also required

100 curies in one calendar year, or

§40.64(c) 15 Day Written Report

d) More than 15 pounds of uranium or thorium at any also required

more than 150 pounds in one calendar §150.17(c)

one time or

year.

§73.71(a)

Recovery of or accounting for loss of any shipment of

Special Nuclear Material or spent fuel

AP-617 Rev. 18 Page 13 of 37

... s.. .. O.&4*flAtL-.t. . .? . -. * - fl . .* . *. ',d.'.. - in. q1* .Yt:a

Attachment 1

Sheet 4 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

REFERENCE WRITTEN FOLLOW-UP

NOTIFICATION

§73.71(a)(5)

Security threats or theft of licensed material shall be

reported to site Security personnel. After initial §73.71(b)(2)

notification or after submission of 30 day report,

additional information shall be reported to NRC as it is

available and within 30 days of discovering additional

information. Per §73.71(a)(5) and §73.71(b)(2),

significant supplemental information which becomes

available after the initial telephonic notification or after

the submission of the written report must be

telephonically reported to the NRC Operations Center

and also submitted in a revised written report. (Written

reports will be submitted on USNRC Form 366 and will

be provided a number unique to Safeguards Events.

These reports will not be a part of the AEOD tracking

program for LERs.)

3. EXPOSURE TO INDIVIDUALS OR RELEASES

Any event involving by-product, source or Special

Nuclear Material that may have caused or threatens to

cause:

a) An individual to receive: §20.2202(a)(1) LER required by

§50.73(a)(2)(viii),

1) A total effective dose equivalent of>_25 Rem (a)(2)(ix) and §20.2203

2) An eye dose equivalent of _75 Rem

3) A shallow-dose equivalent to the skin or

extremities of Ž250 Rad

4) An intake of 5 ALl in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

b) Release of radioactive material in excess of §20.2202(a)(2) LER required by

§50.73(a)(2)(viii),

§50.72(b)(2)(iv)

Technical Specification Instantaneous Limits shall (a)(2)(ix) and §20.2203

be declared an emergency in accordance with

PEP-310. The reporting requirements of PEP

310 shall take precedence over the less

restrictive times for reporting requirements of

§20.2202 and §50.72(b)(2) for releases.

4. ACCIDENTAL CRITICALITY IN FUEL HANDLING

BUILDING

Accidental criticality of special nuclear material. §70.52(a) None

AP-617 Rev. 18 Page 14 of 37

'-k

Attachment 1

Sheet 5 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

NOTIFICATION REFERENCE WRITTEN FOLLOW-UP

5. LOSS OF EMERGENCY RESPONSE CAPABILITY

Any event that results in a major loss of assessment

capability, offsite response capability, or

communications capability (e.g., significant portion of

Control Room indication, Emergency

Telecommunication System, or offsite notification

system).

This includes loss of any of the following:

a) All dedicated Emergency Telecommunication

System phone links to the NRC, as determined by

the Emergency Planning Organization.

b) Offsite siren capability for greater than one hour

as follows:

i) Greater than 16 of the 81 sirens (20% of

system) reported as out of service, or

ii) All sirens in a single county out of service.

The Customer Service Center or on-call ERO SEC

or EP Advisor will notify the Control Room of a

siren problem by telephone.

c) Selective Signaling System phones from the

Control Room, ACP, or EOF to local, county, and

state warning points. Reporting is required only if

these communication links camot be

compensated for by other readily available off-site

communication systems.

d) National Weather Service primaryand back-up

NOAA Weather Radio transmitters at Fayetteville

or primary and back-up NOAA Weather Radio

transmitters at Durham. The National Weather

Service will contact the Control Room if either of

these two conditions exists.

AP-617 Rev. 18 Page 15 of 37

Attachment 1

Sheet 6 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

The following Warning Sirens will not operate when power is lost to the identified transformers. The

Control Room staff is to use this table to determine Reportability.

TRANSFRMR -t-R-EN COUNTY TRANSFRMR SIREN COUNTm(

TRANSFRMR SIREN COUNTY 8185 25 Wake

Z885AC Chatham

1078K 17 Chatham 13 B8728H 38 Wake

1447K 20 Chatham J047BH 70 Harnett

U B924SH A Wake

1598K 1 Chatham J445 Harnett

G754 600 K684BH 67A Wake

1774K 3 Chatham Harnett 67 Wake

L424BH 57 X462AC

A250AF 8 Chatham Harnett 63 Wake

M049 ý66 J681BH

BQ63AF 6A Chatham Harnett 66 Wake

0180 71 Chatham M086 599 J7026H

69 Hlarnelt

Hameet L166BH 62 Wake

53 Chatham P268

CB36AC L869BH 48A Wake

16 Chatham EMC 42 Lee

CCO90 M408 65 Wake

24 Chatham EMC 45 Lee

GD97AC N909 71 Wake

Chatham EMC 46 Lee

CR13AF 6 N991 51 Wake

EMC 48 Lee

CS05AF 7 Chatham 32 Wake

EMC 58 Lee S087BH

CZ64AF 5 Chatham C Wake

V959AC 39. Lee $275BH

D697AC 14 Chatham $551 34 Wake

EMC 9 Chatham 921 28 Wake

40 Wake S716BH 37 Wake

EMC 12 Chatham 12688

1099K 31 Wake SOLAR E01 Wake

L780AF Z Chatham SOLAR E02 Wake

49 Chatham 1187K 26 Wake

M556AC SOLAR E03 Wake

27 Chatham 1394K 21 Wake

M580AC SOLAR E04 Wake

2324K 4 Wake

N218AF 54 Chatham SOLAR 705 Wake

55 Chatham 584K 22 Wake

N279BH V SOLAR E06 Wake

689K Wake

X392AC 41 Chatham SOLAR E07 Wake

G918BH 0 Chatham 7162K

8371 K 72 SOLAR 208 Wake

X595AC 15 Chatham

44 Chatham 951K 19 Wake SOLAR E09 Wake

ZO9OAC

APEX CITY 29 E10

2278AC 52 Chatham SOLAR Wake

36 Wake

10 Chatham APEX CITY 30 U343BH

Z561AC

Note: If power is lost to Siren 39 and the Electric Membership Corporation cannot be contacted, it

should be conservatively assumed that power has been lost to all sirens in Lee County.

NOTIFICATION REFERENCE WRITTEN FOLLOW-UP

(§20.2202(b) 30 Day Written Report

6. RADIOLOGICAL EXPOSURE/RELEASE Also Required per

§20.2203

Any event involving licensed material possessed by the

licensee that may have caused or threatens to cause

an individual to receive, in a period of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

a) A total effective dose equivalent > 5 Rem; or

b) An eye dose equivalent > 15 Rem; or

c) A shallow-dose equivalent to the skin or

extremities> 50 Rem; or

d) An intake of > 1 ALl.

7. OTHER RELEASES

Any Unusual or Important Environmental Events Env. Prot. Plan 30 Day Written Report

Section 4.1, also required

PLP-500

AP-617 Rev. 18 Page 16 of 37

Attachment 1

Sheet 7 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

8. VIOLATION OF OPERATING LICENSE CONDITIONS

1) Any event resulting in the plant operating in a OL Section 2.G LER required per OL

manner which violates the SHNPP Facility Section 2.G

Operating License, Section 2.C:

a) Reactor Core Thermal Power Level exceeds OL Section 2.C.1

2900 MWt.

Average thermal power level for any eight

hour period exceeding 2900 MWt.

Instantaneous thermal power level exceeding

2958 MWt (102%) or average thermal power

levels equivalent to 2958MWt (102%) for a

15-minute period, 2929 MWt (101%) for a 30

minute period, 2914 MWt. (100.5%) for a 60

minute period, shall be used for determination

of Reportability

2) A failure to comply with the following administrative LER required per OL

requirements (See Note 1): Section 2.G

a) Deviation from the requirements of the OL Section 2.C.2

Environmental Protection Plan;

b) Failure to comply with ani-trust conditions of OL Section 2.C.3

Appendix C to OL;

c) Failure to comply with new fuel storage OL Section 2.C.10

requirements.

9. FITNESS FOR DUTY PROGRAM EVENTS

1. Sale, use, or possession of illegal drugs within the §26.73(a)(1) None

protected area.

2. Any acts by any person licensed under §55, or by §26.73(a)(2) None

any supervisory personnel assigned to perform

duties within the scope of §26

a) Involving the sale, use, or possession of a

controlled substance,

b) Resulting in a confirmed positive test on such

persons,

c) Involving use of alcohol within the protected

area, or

d) Resulting in a determination of unfitness for

scheduled work due to the consumption of

alcohol.

3. False positive error on a blind performance test App. A to Part 26 None

specimen when error is determined to be B.2.8(e)(5)

administrative.

AP-617 Rev. 18 Page 17 of 37

Attachment 1

Sheet 8 of 8

NOTES

IMMEDIATE NOTIFICATION REQUIREMENTS

10. ADVERSARY (SECURITY) THREAT(I.C.1) 30 Day Written Report

When specified by Security based on applicable §73.71(b) also Required per

§73 App. G

Security Plan Procedure. §73.71(d)

Security threats or theft of licensed material shall be

§73.71 (a)(5)

reported to site Security personnel. After initial §73.71 (b)(2)

notification or after submission of 30 day report,

additional information shall be reported to NRC as it is

available and within 30 days of discovering additional

information. Per §73.71(a)(5) and §73.71(b)(2),

significant supplemental information which becomes

available after the initial telephonic notification or after

the submission of the written report must be

telephonically reported to the NRC Operations Center

and also submitted in a revised written report. (Written

reports will be submitted on USNRC Form 366 and will

be provided a number unique to Safeguards Events.

These reports will not be a part of the AEOD tracking

program for LERs.)

SECURITY PROGRAM VULNERABILITIES(I.C.2)

§73.71(b) 30 Day Written Report

When specified by Security based on applicable §73 App. G also Required per

Security Plan Procedure. §73.71(d)

§73.71(a)(5)

Security threats or theft of licensed material shall

§73.71(b)(2)

be reported to site Security personnel. After initial

notification or after submission of 30 day report,

additional information shall be reported to NRC

as it is available and within 30 days of discovering

additional information. Per §73.71(a)(5) and

§73.71 (b)(2), significant supplemental information

which becomes available after the initial

telephonic notification or after the submission of

the written report must be telephonically reported

to the NRC Operations Center and also submitted

in a revised written report. (Written reports will be

submitted on USNRC Form 366 and will be

provided a number unique to Safeguards Events.

These reports will not be a part of the AEOD

tracking program for LERs.)

INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA)

REPRESENTATIVE (1.C.3)

Individual claiming to be an IAEA representative who is §75.7 None

not accompanied by an NRC employee and has no

prior confirmation of credentials in writing.

§75.6 and §75.7

Notification is to Director, Office of Nuclear

Reactor Regulation

11. FITNESS FOR DUTY - NRC EMPLOYEE

§26.27(d)

Notification of NRC employee's unfitness for duty. None

Per §26.27(d), the appropriate Regional

Administrator must be notified immediately by

telephone. During other than normal working

hours, the NRC Operations Center must be

notified.

AP-617 Rev. 18 Page 18 of 37

I LOSS OF EMERGENCY COOLANT RECIRCULATION

Attachment I

Sheet 1 of 1

MINIMUM SI FLOW RATE VERSUS TIME AFTER REACTOR TRIP

TIME AFTER REACTOR TRIP MINIMUM SI FLOW (GPM)

10 TO 15 MINUTES 500

15 TO 20 MINUTES 450

20 TO 25 MINUTES 425

25 TO 30 MINUTES 400

30 TO 40 MINUTES 3/5

40 TO 50 MINUTES 350

50 TO 60 MINUTES 325

1 TO 1.5 HOURS 300

1.5 TO 2 HOURS 275

2 TO 3 HOURS 250

3 TO 4 HOURS 225

GREATER THAN 4 HOURS 200

- END

EOP-EPP-012 Rev. 16 Page 61 of 65

is # II 1i.1

Attachment 19

Sheet 1 of 9

Makeup Concentration Limits

These tables were derived per calculation HNP-I/INST-1056 using the equations of

Attachment 3 and provide a means to select an appropriate RWMU Total

Makeup Flow Rate (Q) which will yield a desired blended flow boron concentration when

matched to the BAT concentration and the Boric Acid Flow Rate span of I to 30 gpm.

It is necessary to select lower RWMU Total Makeup Flow Rates when high boron

concentrations are required because the Boric Acid Flow is limited by system configuration to

a maximum of 33 gpm. This maximum Boric Acid Flow capability however is not used as the

basis for these tables because it is necessary to allow some margin for possible system

performance degradation. Therefore a maximum Boric Acid Flow of 30 gpm is used as the

basis for the following tables.

Each sheet of the tables is applicable to a specific RWMU Total Makeup Flow Rate (Q).

The maximum ppm results have been rounded down to the nearest whole number and the

minimum ppm results have been rounded up to the nearest whole number.

OP-107 Rev. 42 Page 245 of 256

Attachment 19

Sheet 2 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 120qpm

1 . To determine the maximum boron concentration for which makeup will be

reliable at 120 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 120 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 120 GPM MINIMUM PPM FOR 120 GPM

MAKEUP MAKEUP

CONCENTRATION (PPM) (1 GPM BA FLOW)

(30 *PM BA FLOW)

7000 1750 59

7050 1762 59

7100 1775 60

7150 1787 60

7200 1800 60

7250 1812 61

7300 1825 61

7350 1837 62

7400 1850 62

7450 1862 63

7500 1875 63

7550 1887 63

7600 1900 64

7650 1912 64

7700 1925 65

7750 1937 65

I OP-107 Rev. 42 Page 246 of 256

Attachment 19

Sheet 3 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 110.qpm

1 . determ ine the m axim um boron concentration for w hich m akeup w ill be

To

reliable at 110 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 110 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 110 GPM MINIMUM PPM FOR 110 GPM

MAKEUP MAKEUP

CONCENTRATION (PPM) (1 GPM BA FLOW)

(30 GPM BA FLOW)

7000 1909 64

7rNN 1922 65

7100 1936 65

7150 1950 65

7200 1963 66

7250 1977 66

7300 1990 67

7350 2004 67

7400 2018 68

7450 2031 68

7500 2045 69

7550 2059 69

7600 2072 70

7650 2086 70

7700 2100 70

7750 2113 71

Rev. 42 Page 247 of 256 1

I OP-107 R 4a 2 5

Attachment 19

Sheet 4 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 100gpm

1. To determine the maximum boron concentration for which makeup will be

reliable at 100 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 100 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 100 GPM MINIMUM PPM FOR 100 GPM

CONCENTRATION (PPM) MAKEUP MAKEUP

(30 GPM BA FLOW) (1 GPM BA FLOW)

7000 2100 70

7050 2115 71

7100 2130 71

7150 2145 72

7200 2160 72

7250 2175 73

7300 2190 73

7350 2205 74

7400 2220 74

7450 2235 75

7500 2250 75

7550 2265 76

7600 2280 76

7650 2295 77

7700 2310 77

7750 2325 78

OP-107 Rev. 42 Page 248 of 256

Attachment 19

Sheet 5 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 90gpm

1. To determine the maximum boron concentration for which makeup will be

reliable at 90 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 90 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 90 GPM MINIMUM PPM FOR 90 GPM

CONCENTRATION (PPM) MAKEUP MAKEUP

(30 GPM BA FLOW) (1 GPM BA FLOW)

7000 2333 78

7050 2350 79

7100 2366 79

7150 2383 80

7200 2400 80

7250 2416 81

7300 2433 82

7350 2450 82

7400 2466 83

7450 2483 83

7500 2500 84

7550 2516 84

7600 2533 85

7650 2550 85

7700 2566 86

7750 2583 87

I OP-107 Rev. 42 Page 249 of 256

Attachment 19

Sheet 6 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 80qpm

1. To determine the maximum boron concentration for which makeup will be

reliable at 80 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 80 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BOR( MAXIMUM PPM FOR 80 GPM MINIMUM PPM FOR 80 GPM

MAKEUP MAKEUP

(30 GPM BA FLOW) (1 GPM BA FLOW)

U

7000 2625 88

7050 2643 89

7100 2662 89

7150 2681 90

7200 2700 90

7250 2718 91

7300 2737 92

7350 2756 92

7400 2775 93

7450 2793 94

7500 2812 94

7550 2831 95

7600 2850 95

7650 2868 96

7700 2887 97

7750 2906 97

OP-107 Rev. 42 Page 250 of 256

Attachment 19

Sheet 7 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 70gpm

1. To determine the maximum boron concentration for which makeup will be

reliable at 70 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 70 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 70 GPM MINIMUM PPM FOR 70 GPM

CONCENTRATION (PPM) MAKEUP MAKEUP

(30 GPM BA FLOW) (I GPM BA FLOW)

7000 3000 100

7050 3021 101

7100 3042 102

7150 3064 103

7200 3085 103

7250 3107 104

7300 3128 105

7350 3150 105

7400 3171 106

7450 3192 107

7500 3214 108

7550 3235 108

7600 3257 109

7650 3278 110

7700 3300 110

7750 3321 111

OP-107 Rev. 42 Page 251 of 256

Attachment 19

Sheet 8 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 60pm

1. To determine the maximum boron concentration for which makeup will be

reliable at 60 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 60 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 60 GPM MINIMUM PPM FOR 60 GPM

MAKEUP MAKEUP

CONCENTRATION (PPM) (1 GPM BA FLOW)

(30 GPM BA FLOW)

7*nn 3500 117

7050 3525 118

7100 3550 119

7lgn 3575 120

7200 3600 120

7250 3625 121

7300 3650 122

7350 3675 123

7400 3700 124

125

3725

7450

7500 3750 125

7550 3775 126

7600 3800 127

7650 3825 128

7700 3850 129

7750 3875 130

I OP-107 Rev. 42 Page 252 of 256

Attachment 19

Sheet 9 of 9

Makeup Concentration Limits for: RWMU Total Makeup Flow Rate Q = 50gpm

1. To determine the maximum boron concentration for which makeup will be

reliable at 50 gpm total flow, select the BAT boron concentration which is equal

to or lower than current BAT boron concentration.

2. To determine the minimum boron concentration for which makeup will be reliable

at 50 gpm total flow, select the BAT boron concentration which is equal to or

higher than current BAT boron concentration.

BAT BORON MAXIMUM PPM FOR 50 GPM MINIMUM PPM FOR 50 GPM

CONCENTRATION (PPM) MAKEUP MAKEUP

(30 GPM BA FLOW) (1 GPM BA FLOW)

7000 4200 140

7050 4230 141

7100 4260 142

7150 4290 143

7200 4320 144

7250 4350 145

7300 4380 146

7350 4410 147

7400 4440 148

7450 4470 149

7500 4500 150

7550 4530 151

7600 4560 152

7650 4590 153

7700 4620 154

7750 4650 155

OP-107 Rev. 42 Page 253 of 256

Attachment 5

Sheet 1 of 1

Coolino Tower Cold Weather Operation

90

L

W

Li 80

W

Li

70

a:

C

z

W

60

0

I-

0 50

09

0

Co

0

40

32

30

-30 -20 -10 0 10 20 30 32 40

AMBIENT AIR TEMPERATURE (DEGREES F)

NORMAL OPERATION - COOLING TOWER DEICING GATE VALVES OPEN

AS IS - COOLING TOWER DEICING GATE VALVES REMAIN AS IS

HALF OPEN - COOLING TOWER DEICING GATE VALVES HALF OPEN

P*/ ABNORMAL OPERATION-"NO CONDENSER HEAT LOAD" AREA. IN THIS AREA PERFORM SECTION 8.6.

'A A

CCý vlý

ýCýýJý_ (ii 0

pAý MAY~P6~p

OP-141 Rev. 17 ,age 38 of 40

Attachment 2

Sheet 1 of 3

Refueling Operations

1.0 OPERATIONAL REQUIREMENTS - DECAY TIME

1.1 The reactor shall be subcritical for a minimum period of time as

determined by Table A.

APPLICABILITY: During movement of irradiated fuel in the reactor vessl.

ACTION:

With the reactor subcritical for a time less than determined by Table A,

suspend all operations involving movement of irradiated fuel in the

reactor vessel. Fuel movement in the reactor vessel may continue

provided the minimum decay time is greater than the time shown on.'

Table A.

2.0 SURVEILLANCE REQUIREMENTS 2.1 The reactor shall be determined to have been subcritical for a minimum

period of time as determined using Table A by verification of the date

and time of subcriticality prior to movement of irradiated fuel in the

reactor vessel.

2.2 CCW temperature shall be monitored every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the movement of

fuel in the reactor vessel to ensure the temperature used to determine

decay time is not exceeded.

Table A

Time from Reactor Subcritical (Hours) Effective CCW Temperature (oF)

100 88.9

120 91.8

144 94.3

168 96.2

192 97.9

216 99.1

240 100.2

NOTE 1: - Linear interpolation between listed points is acceptable.

NOTE 2: - These delay times are applicable to end of cycle full core off-loads only. A

mid-cycle core off-load assumes two CCW and Fuel Pool Cooling trains

available and does NOT require compliance with these limits.

NOTE 3: - Effective CCW temperature refers to actual CCW heat exchanger outlet

temperature plus 5oF.

NOTE 4: - The table assumes the core off-load duration is 39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months or greater.

PLP-114 Rev. 14 Page 8 of 26

Attachment 2

Sheet 2 of 3

Refueling Operations

3.0 OPERATION REQUIREMENTS - COMMUNICATIONS

3.1 Direct communications shall be maintained between the control room and

personnel at the refueling station in containment.

APPLICABILITY: During CORE ALTERATIONS.

ACTION:

When direct communications between the control room and personnel at the

refueling station cannot be maintained, suspend all CORE ALTERATIONS.

4.0 SURVEILLANCE REQUIREMENTS:

4.1 Direct communications between the control room and personnel at the

refueling station in containment shall be demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE

ALTERATIONS.

5.0 OPERATIONAL REQUIREMENTS - REFUELING MACHINE

5.1 The refueling machine and auxiliary hoist shall be used for movement of

drive rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine, used for movement of fuel assemblies, having:

1. A minimum capacity of 4000 pounds, and

2. An automatic overload cutoff limit less than or equal to 2700

pounds.

The auxiliary hoist, used for latching and unlatching drive rods, having:

b.

1. A minimum capacity of 3000 pounds, and

2. A 0 - 2000 pound digital load indicator that shall be used to

monitor loads to prevent lifting more than 600 pounds.

APPLICABILITY: During movement of drive rods or fuel assemblies within the

reactor vessel.

ACTION:

With the requirements for the refueling machine and/or auxiliary hoist

OPERABILITY not satisfied, suspend use of any inoperable refueling

machine and/or auxiliary hoist from operations involving the movement of

drive rods and fuel assemblies within the reactor vessel.

6.0 SURVEILLANCE REQUIREMENTS 6.1 The refueling machine used for movement of fuel assemblies within the

reactor vessel shall be demonstrated OPERABLE, within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to

the start of such operations, by performing a load test of at least 4000

pounds and demonstrating an automatic load cutoff at less than or equal

to 2700 pounds.

PLP-114 Rev. 14 Page 9 of 26

Attachment 2

Sheet 3 of 3

Refueling Operations

6.0 SURVEILLANCE REQUIREMENTS (continued)

of

6.2 The auxiliary hoist and associated load indicator used for movement

OPERABLE

drive rods within the reactor vessel shall be demonstrated

within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a

load test of at least 900 pounds.

7.0 OPERATIONAL REQUIREMENTS - CRANE TRAVEL / FUEL HANDLING BUILDING

7.1 Loads in excess of 2300 pounds shall be prohibited from travel over fuel

assemblies in the storage pool.

APPLICABILITY: With irradiated fuel assemblies in the storage pool.

ACTION:

place the

With the requirements of the above specification not satisfied,

crane load in a safe condition.

8.0 SURVEILLANCE REQUIREMENTS

loads

8.1 Crane interlocks and physical stops which prevent crane travel with

in excess of 2300 pounds over fuel assemblies shall be demonstrated

days

OPERABLE within 7 days prior to crane use and at least once per 7

thereafter during crane operation.

9.0 OPERATIONAL REQUIREMENTS

the

9.1 Spent Fuel Pool loads used for plant operations scenarios assume that

is no

refueling outage duration (reactor shutdown to re-synchronization)

shorter than 20 days.

10.0 SURVEILLANCE REQUIREMENTS

be

10.1 Prior to Entry into MODE 1 following a refueling outage, it must

confirmed that the duration of a refueling outage is greater than

20 days.

PLP-114 Rev. 14 Page 10 of 26

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

OIL (05 EFPDF 161), HZP, WITH NO XENON

-1800 Tý -18

1_=

piiý Hf tfIL. -17

-1700 -

-1600 ti~ -,--

SJ31

L4-- - I- I-F-I-k - I -t--

I- -I4-9--

--4--4--

-4--

.-4-.-J--4.-

I-h t - I -

-U

A-

I r1 Fl - - I - T TrFI

V

4-

-16

I

1 m

1

-4- T ttz

-15

I

-1500

-4-Vt- -14

-1400 L

I--- H -t- I

-13 c

-1300

-12 'nt

m

E -1200 m

0C.)

S -11

-1100

I

-10 r

0 -1000

-9 c

o -900

0 -8 0

j, -800 -7 -

c5 -700 -5

Lu

I-_

z -600

HIII:

-500

-400

I -H- t-3

-300

-2

X

-200

-100 7I t. t -4-II I I I ;-i--j p -1

25 PCM/iIV ii "2ZItIiLZV'1

1tbLv4ici1/2" ___ -----if__'JN1Ei

0

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

0 103 231

a1 BANK 0

231

128

BANK C

CURVE NO. A-11-6 REV NO. 0

ORIGINATOR F',*---, DATE 01,2 SD/0

SUPERVISOR * DATE

SUPERINTENDENT- /t(04oile

SHIFT OPERATIONS ,,DATE

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

MOL (161 < EFPD - 334), HZP, WITH NO XENON

-1800 -18

-1700

-1600

-1500

-1400

t

f ISI

S

4\ '1 VtI I1IVJ1*TVI

EIII }IIII!

i ill'* S I

I

-17

-16

-15

-14

-13 a

-1300 =5

-12

-12 rn

m

-1200 rn

E m

-11 z

-1100

I-L -10

0 -1000

0

-9 0:

0 -900

I0 -8 0

-800 -7

cc

o -700

w

I-. -6

Z -600

-5

-500

T-

-4

-400

-3

-300

-2

-200

-1

-100

25 PCMIDIV 0

0

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

J 231

0 10

BANK D

128 231

BANK C

CURVE NO. A-11-7 REV NO. 0

ORIGINATOR - DATE i0/gsc1

SUPERVISOR DATE /o -_ -o\

SUPERINTENDENT

SHIFT OPERATIONS DATE ZLZY- & I

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

-

EOL (334 < EFPD < 507), HZP, WITH NO XENON

-2000 1 1 j i I IiI I IjZ II IlI.IiII I J - - -20

1 J.-I-L 4-4 -19

-1900 .

-1800 -18

,ti -17

.. 700 - -fr-H-I-I-i-H--i--pt-

-1600 -16

-1500

- - -4-4- - - 4-

-15

0

- HH F-1 H -14%

!

-1400

TmTgffý =: m

'-I

-13 mr

E -1300 It ---

in m

-1200 4-1 t t4 !4 -C

--- -12-

z

I0

n' -11 ::

-1100 Em

-H -o 0

0 -1000 -10

0

t-t LA

IF- H-F t-H-F- 'U -I -9 0

n

cr. -900

a -7

-8 'r

n-' -800 I I I LA -C

-7 *

I -700

z

S RiiI il I -6 'o

-600

'-I- 4 F-tH-F Ii

- H-P -5

-500 Em ---------------------------T

-4

-400

-300 Thtw-tt+/- - -3

I i I I. H . . . ..

I I F H F I F

-2

-200 IL I R 1 1 444444-A+++4-Y4- I-T- -4 1 1

44 -1

-100

25 PCM/DIV 0

0 -

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

0 103 231

BANK D

128 231

BANK C

CURVE NO. A-11-8 REV NO. 0

ORIGINATOR C a27ZJ DATE 00/-25-/0

SUPERVISOR * DATE 41"./,t

SUPERINTENDENT

SHIFT OPERATIONS L. DATE 1 e

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

BOL (0 < EFPD < 161), HFP, EQUILIBRIUM XENON

-1700 -17

-16

-1600

-15

-1500

-14

-1400

-1300

Ij--44 -13

-12

-1200 a

m

-11 =I

-1100 m

z

C)

-10

I- -1000 r

0CL -9 O

-900 0

0 -8

-800 0

-J -7

-700

I- -6

Z -600 -4-r

-5

-500

WWp

~~~~ iM l I -4

-400

-3

-300

-2

-200

-1

_1 Nil

25 PCM/DIV

0 Tff-- . . -t-rKm*, _[ 0

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

0 103 231

BANK D

128 231

BANK C

CURVE NO. A-11-9 REV NO. 0

ORIGINATOR "DATE /0/--5/0D

SUPERVISOR DATEE

SUPERINTENDENT

SHIFT OPERATIONS DATE /o0 /

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

MOL (161 < EFPD < 334), HFP, EQUILIBRIUM XENON

-1800 -18

ifrtw:LItfl): :1: :hi 1'JZIz I:j44 : iir :z 4ZzI 14 FHZIZ

I I I I -I

I II II II I I I 1 I I IJ I ~ I- I

JJ I q

-- 17

-1700 1 t-t7týl I I 1 1 1ý

-1600

J-I- - 1 1 111 1+ý l 11+ -16

-*tft1 jArtzLIM IIt rI T

-1500 I!BFEHIIIHIHTFFHTh'A H

-15

-1400

I~tIHt 'tIit--HLtLtztc Mt H-

111111ý11itil",

-iV *-14

1ýI I: :1-:1 I N: I: I 1 1:): I: "-13o

-1300 :125122: I2iltttT2

I I I II I I II II -n

-1200 -12 m

E Ile 'I m

a..

C., -1-t-JA i -1 1 z

-1100

0 F- -10

-1000 _L L II: II

0

-900 L- .-A I I I I I4-1

0

. -800

g##ItHEIW3 -8

0

0C, :nu-r1:lL:FFfls --7

-700 I-1

I-

z

-600

-4

I i-H

-6

-5

-500 pq 1i 11 1 11~ 11 H- IF::11, ... N -

44F FFYllIIVEE

FF4 H-!-*

-400 -4

-3

-300

-200

- ~~U~h~43 :4i~~f: UIT c1 -2

-100 ~jjp

1 44-A- -1

25 PCMfDIV

0 0

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

0 103 231

BANK D

128 231

BANK C

CURVE NO. A-11-10 REV NO. 0

ORIGINATOR 4S/-l DATE /O/Z-?*/o,

SUPERVISOR DATE /ýZt?/.24

SUPERINTENDENT- (L/S/

SHIFT OPERATIONS DATE /

HARRIS UNIT 1 CYCLE 11

DIFFERENTIAL AND INTEGRAL

ROD WORTH CONTROL BANKS D and C

MOVING WITH 103 STEP OVERLAP

EOL (334 < EFPD < 507), HFP, EQUILIBRIUM XENON

-2100 -l--l--rrr))Lt+/-441111 I J112111441J11i1jJ1i112J7T -21

-2000

-1900

1 [-ý-p --- 4-4-+/-+/-q - -20

-19

-18

-1800 4-H t-H,-fr-H

-4 * -17

-1700 "H+-H-1 -VI- r I I I

-4- * -16

-1600

-15

-1500

I"11

E -1400

qq 4-tHH-i- - -13 .z

C-, -1300

0a r

-1200 -I

0 - -11 a

-1100

-- 10 m

cc -1000

-9 0

oH3

-J ----------

w -900 wJttl.ZtCI I

cc

0 H- H- --8 >

GZ -800 0

--7

-700 -H- A I -I I - * I-F- I i I I I I

PF44A--X44-4-J:HF4-H-H 1 -H- -6

-600j

-500 flit- ~ l ~ 11

-ý,

'I

r -5

-400

+/-:A

F- -1,--

1:1

1-

I: 1

-4- qxtt H- 4=

--

-444-

-300 22:irnlr

-- 2

-200

ttr 44Th1t t F@ -1

-100

25 PCM/DIV

0-

!

....f..

I.H. . I

-tI1

fI

I

-0

0 20 40 60 80 100 120 140 160 180 200 220 240

5 STEPS/DIV

0 103 231

BANK D

128 231

BANK C

CURVE NO. A-11-11 REV NO. 0

ORIGINATOR 6L- - DATE /0/2r5/0(

V SUPERVISOR DATE N

SUPERINTENDENT

SHIFT OPERATIONS DATE 1/0L2dL

HARRIS UNIT 1 CYCLE 11

POWER DEFECT vs. POWER LEVEL

for VARIOUS BORON CONCENTRATIONS

BOL (0O5 EFPDS 161)

0

-200

-400

-600

-800

C,

0U. -1000

w

cr

w

C

-1200

w

0

-1400

0~

b-

-1600 2100 ppm

1800 ppm

-1800 1500 ppm

1200 ppm

-2000 900 ppm

-2200

20 PCM/OIV

-2400

100

0 20 40 60 80

1 -//DIV

POWER LEVEL (PERCENT)

CURVE NO. C-11-1 REV NO. 0

ORIGINATOR ./DATE //.5/oi

SUPERVISOR DATE /,/Z2/',,

SUPERINTENDENT

SHIFT OPERATIONS DATE /6 /

HARRIS UNIT 1 CYCLE 11

POWER DEFECT vs. POWER LEVEL

for VARIOUS BORON CONCENTRATIONS

MOL (161 <EFPDP 334)

.0

-200

-400

-600

-800

E

L)

w -1000

1u

H_ -1200

C)

O'

I--

0.

-1400

-1600

-1800

1800 ppm

1500 ppm

-2000

1200 ppm

-2200 900 ppm

20 PCM/D!V

600 ppm

-2400

0 20 40 60 80 100

1 /./DIV

POWER LEVEL (PERCENT)

CURVE NO. C-11-2 REV NO. 0

ORIGINATOR /4- DATE /01,27/0,

SUPERVISOR DATE

SUPERINTENDENT

SHIFT OPERATIONS- DATE

HARRIS UNIT 1 CYCLE 11

POWER DEFECT vs. POWER LEVEL

for VARIOUS BORON CONCENTRATIONS

EOL (334 < EFPD _ 507)

0

-200

-400

-600

-800

-1000

E -1200

I -1400

Lu

LI

-1600

0

U.I -1800

I-

-2000

0 -2200

-2400

-2600

1200 ppm

-2800 900 ppm

600 ppm

-3000 300 ppm

-3200 0 ppm

20 PCM/DIV

-3400

0 20 40 60 80 100

1 */0 D0IV

POWER LEVEL (PERCENT)

CURVE NO. C-11-3 REV NO. 0

ORIGINATOR A dte/k-( DATE /O/1

01-/-1

SUPERVISOR DATE ,/ttz /Sk

SUPERINTENDENT

SHIFT OPERATIONS DATE Zo_/(10/

Attachment 5

Sheet 1 of 1

Coolina Tower Cold Weather Ooeration

90

80

70

60

50

40

32

30

-30 -20 -10 0 10 20 30 32 40

AMBIENT AIR TEMPERATURE (DEGREES F)

rq NORMAL OPERATION - COOLING TOWER DEICING GATE VALVES OPEN

U AS IS - COOLING TOWER DEICING GATE VALVES REMAIN AS IS

[] HALF OPEN - COOLING TOWER DEICING GATE VALVES HALF OPEN

ABNORMAL OPERATION-"NO CONDENSER HEAT LOAD" AREA. IN THIS AREA PERFORM SECTION 8.6.

ý, P-C

4LCýj ýVuz, f#I

f410ýý

OP-141 Rev. 17 Page 38 of 40

ML022670157

Text

Navigation menu

Search