Text

2014 Prairie Island Nuclear Generating Plant Initial License Examination Administered Written Examination - Senior Reactor Operator
	ML15126A572
Person / Time
Site:	Prairie Island
Issue date:	08/22/2014
From:	Bielby M; NRC/RGN-III/DRS/OLB
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ML13093A443	List: ML15126A524; ML15126A536; ML15126A543; ML15126A554; ML15126A572;
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SRO APPLICANT EXAMINATION HANDOUTS

Number:

Title:

Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 LEVEL OF USE CONTINUOUS USE Continuous use of procedure required.

Read each step prior to performing.

Mark off steps as they are completed.

Procedure SHALL be at the work location.

PORC REVIEW DATE:

OWNER:

EFFECTIVE DATE:

8/28/13 D. Smith 8/29/13 Page 1of26

N Number:

Title:

Revision Number:

1ECA-0.0 LOSS OF ALL SAFEGUARDS AC POWER REV. 23 A.

PURPOSE This procedure provides actions to respond to a loss of all safeguards AC power.

B.

ENTRY CONDITIONS C.

This procedure is entered from:

1. lE-0, REACTOR TRIP OR SAFETY INJECTION, on the indication that both safeguards buses are deenergized.

2. Anytime power is lost to both safeguards buses after the reactor and turbine have been verified tripped.

ATTACHMENTS:

ATTACHMENT E:

ATTACHMENT G:

ATTACHMENT H:

SG Wide Range Level -

Adverse Conditions Unit 1 Containment Isolation Valve Locations DC Bus Load Shed D.

REFERENCES

1. Commitment 20018117 - Per October 2001 letter to NRC:

Revise Purpose and Entry Conditions page (N).

2. IERLl-11-4, Near-Term Actions to Address the Effects of an Extended Loss of All AC Power in Response to the Fukushima Daiichi Event (M).

Page 2 of 26

Number:

1ECA-0.0 STEP

Title:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE 1------1 RESPONSE NOT OBTAINED i---------

NOTE GSF Status Trees should be monitored for information only.

FR procedures SHALL NOT be implemented.

1 Check If RCS Is Isolated:

2

a. PRZR PORVs -

CLOSED

b. Letdown isolation valves -

CLOSED:

CV-31255

CV-31226

c. Excess letdown isolation valve CV-31330 -

CLOSED Verify AFW Flow-GREATER THAN 200GPM Page 3 of 26

a. IF PRZR pressure less than 2335 psig, THEN manually close PORVs.

b. Manually close valves.

c. Manually close valve.

Perform the following:

a. Verify TD AFW pump running. IF NOT, THEN manually start TD AFW pump.

IF pump can NOT be started, THEN dispatch personnel to locally start pump per C28.1 AOP3, AUX FEEDWATER SYSTEM OPERATION WHEN AC POWER IS LOST.

b. Verify proper safeguards alignment of AFW valves.

IF NOT, THEN locally align valves as necessary.

Number:

Ti LI c,:

Revlsion Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE i--~~~~ RESPONSE NOT OBTAINED i--~~~~~~--.

3 Perform Notifications:

4

Announce "Unit 1 Reactor Trip"

Notify Shift Manager and SEC Check Cooling Water Header Pressures - BOTH GREATER THAN 25 PSIG Page 4 of 26 Verify two cooling water pumps running.

IF pressure(s) NQ_I restored, IHEN perform the following:

a. Stop affected diesel generator(s) until pressure is restored.

b. Initiate C35 AOP2, LOSS OF PUMPING CAPACITY OR SUPPLY HEADER WITHOUT SI.

c. Continue with Step 5.

(

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE....... ----t RESPONSE NOT OBTAINED 1---------

5 Check If. Safeguards Buses Are Available For Sequencer Loading:

a. Check Bus 16 -

AVAILABLE:

Bus 16 green load rejection lights -

LIT

Bus 16 lock out annunciator 47024-0104 -

OFF

b. Check Bus 15 -

AVAILABLE:

Bus 15 green load rejection lights -

LIT

Bus 15 lock out annunciator 47024-0101, -

OFF Page 5 of 26

b. IF Bus 16 is available, THEN go to Step 6.

IF neither bus is available, THEN go to Step 8.

Number:

1ECA-O.O STEP

Title:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE 1------t RESPONSE NOT OBTAINED 1---------

NOTE

Diesel generator lock out annunciators (47024 0703 and 47024-0706) AND power from offsite sources (lR and CTll) can be used when determining if a Unit 1 source is available.

IF both safeguards buses are available, THEN it is preferable to attempt energizing Bus 16 first.

6 Attempt To Restore Power To Any Available Safeguards Bus From Unit 1 Source:

a. Energize available bus with diesel generator:

1) Start diesel generator

2) Verify safe9uards bus automatically energized

b. Check safeguards buses AT LEAST ONE ENERGIZED

c. Start one charging pump

d. Notify Engineering to inspect RCP thermal barrier CC piping

e. Return to procedure and step in effect and implement FR procedures as necessary Page 6 of 26

1) Check any Unit 1 Source available.

IF no Unit 1 source is available, THEN go to Step 7.

2) Manually energize bus from any available Unit 1 source:

a) Place desired source breaker MAN/AUTO switch to "MANUAL".

b) Place syncroscope select switch to desired source position.

c) Close desired source breaker.

b. Go to Step 7.

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Title:

Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE---- RESPONSE NOT OBTAINED i---------

7 Attempt To Restore Power To Any Available Safeguards Bus From Unit 2:

a. Check bus tie breakers for either bus -

AVAILABLE:

1) SI actuated annunciator 47014-0604 -

OFF

2) Unit 2 SI Pumps -

BOTH OFF

3) Corresponding Unit 2 bus -

ENERGIZED

b. Place source breakers for available bus to "PULLOUT":

1) lR source

2) CTll source

3) DG source

c. Check Unit 1 SI pump breaker(s) -

OPEN

d. Close 4KV bus tie breakers for available bus:

1) Unit 2 bus tie breaker

2) Unit 1 bus tie breaker

e. Check safeguards buses AT LEAST ONE ENERGIZED

f. Start one charging pump

g. Notify Engineering to inspect RCP thermal barrier CC piping

h. Return to procedure and step in effect and implement FR procedures as necessary Page 7 of 26

1) Reset Unit 1 SI.

2) Go to Step 8.

3) Go to Step 8.

c. Manually open breaker(s).

d. Go to Step 8.

e. Go to Step 8.

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Title:

Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE---- RESPONSE NOT OBTAINED t--------""I 8

Dispatch Personnel To Locally Close Valves To Isolate RCP Seals:

a. RCP seal return isolation valve MV-32166

b. RCP seal injection throttle valves:

VC-14-1

VC-14-2

c. RCP CC return isolation valves:

CC-16-3

CC-16-2 Page 8 of 26

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Title:

Revision Number:

1ECA-0.0 LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE---- RESPONSE NOT OBTAINED 1---------.

Caution

WHEN power is restored to any safeguards bus. THEN recovery actions should continue starting with Step 29.

IF an SI signal exists or actuates during this procedure. THEN it should be reset to permit manual loading of equipment on a safeguards bus.

9 Place Following Equipment Switches In Pullout Position:

MD AFW pump

Groups A & B PRZR heaters (Off position)

RHR pumps

SI pumps

CS pumps

Containment FCUs (Off position)

CC pumps

Control room chillers and fans supplied from Unit 1 power

121 and 122 air compressors Page 9 of 26

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RC'vision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE 1------1 RESPONSE NOT OBTAINED--------

1 O Attempt To Restore Power To Any Sat eguards Bus From Unit 1 Source:

a. Place safeguards bus voltage restoration switches to "MANUAL":

Bus 15

Bus 16

b. Verify normal source breakers -

OPEN:

Bus 15 source breakers

Bus 16 source breakers

c. Check safeguards bus lock out annunciators:

Bus 15 47024-0101 -

OFF

-OR-

Bus 16 47024-0104 -

OFF

d. Check DG for available bus:

Dl lock out annunciator 47024-0703 -

OFF

-OR-

D2 lock out annunciator 47024-0706 -

OFF This Step continued on the next page.

Page 10 of 26

b. Manually open breaker(s).

c. WHEN any bus becomes available, THEN do remaining Step 10 actions.

Go to Step 13.

d. IF lR or CTll is available, THEN go to Step lOf.

IF neither lR or CTll is available, THEN go to Step 11.

Numher:

Ii tie:

R1.,vision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE t--~~~~ RESPONSE NOT OBTAINED t--~~~~~~--.

(Step 10 continued from previous page)

e. DG for available bus -

RUNNING

f. Attempt to restore power to available bus:

1) Check source to bus AVAILABLE:

lR source

-OR-

CTll source

-OR-

DG source

2) Place desired source breaker MAN/AUTO switch to "MANUAL"

3) Place syncroscope select switch to desired source position

4) Close desired source breaker

5) Check safeguards buses -

AT LEAST ONE ENERGIZED

6) Go to Step 29 Page 11 of 26

e. Manually start DG.

1) IF no source available, THEN go to Step 11.

1,).LE no source breaker can be closed, JHE~

go to Step 11.

5) Go to Step 11.

Number:

r; tie:

Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION /EXPECTED RESPONSE -----t RESPONSE NOT OBTAINED --------

11 Attempt To Restore Power To Any Safeguards Bus From Unit 2:

a. Check bus tie breakers for either bus -

AVAILABLE:

1) Unit 1 SI actuated annunciator 47014-0604 -

OFF

2) Unit 2 SI actuated annunciator 47514-0604 OFF

3) Corresponding Unit 2 bus -

ENERGIZED

b. Place safeguards bus source breakers for available bus to "PULLOUT":

lR source

CTll source

DG source

c. Close 4KV bus tie breakers for available bus:

1) Unit 2 bus tie breaker

2) Unit 1 bus tie breaker

d. Check safeguards buses AT LEAST ONE ENERGIZED

e. Go to Step 29 Page 12 of 26

1) Reset Unit 1 SI.

2) IX Unit 2 SI can NOT be reset, ]'HEN go to Step 12.

3) Go to Step 12.

c. Go to Step 12.

d. Go to Step 12.

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Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE i--~~~-1 RESPONSE NOT OBTAINED i--~~~~~~--.

12 Attempt To Start Diesel Generator(s) Using Local Actions:

a. Place diesel generator START/STOP switch(s) in "PULLOUT"

b. Locally reset DG lockout relays:

1) Shutdown relay(s):

CS-55021, Dl DSL GEN ALM AND SHTDN RESET PB (located on Dl DIESEL GEN GAUGE PANEL)

CS-55521, D2 DSL GEN ALM AND SHTDN RESET PB (located on D2 DSL GEN GAUGE PNL)

2) WHEN shutdown relay(s) reset, THEN reset 86 lockout relay(s):

Dl EMERGENCY GENERATOR LOCKOUT RELAY 86 (located on South side of Dl Metering &

Relay Cabinet in Dl room)

D2 EMER GEN LOCKOUT RELAY 86 (located on D2/GEN RLY PNL in Relay Room, East side)

c. Manually actuate SI This Step continued on the next page.

Page 13 of 26

b. Continue attempts to restore power to any safeguards bus.

Go to Step 13.

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE i------1 RESPONSE NOT OBTAINED 1---------

(Step 12 continued from previous page)

d. Place diesel generator START/STOP switch(s) in "NORMAL"

e. Verify at least one diesel generator starts

f. Manually energize safeguards bus from DG:

1) Place desired source breaker MAN/AUTO switch to "MANUAL"

2) Place syncroscope select switch to desired source position

3) Close desired source breaker

4) Check safeguards buses -

AT LEAST ONE ENERGIZED

5) Go to Step 29 Page 14 of 26

e. Attempt to manually start diesel generator(s).

_IT no diesel generator starts, }'HEN:

1) Dispatch personnel to locally start a diesel generator per 1C20.7, Dl/D2 DIESEL GENERATORS.

2) Continue attempts to restore power to any safeguards bus.

3) Go to Step 13.

4) Continue attempts to restore power to any safeguards bus.

Go to Step 13.

Number:

1ECA-O.O STEP 13 l it1f*:

Revision NumbPr:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE ____.,. RESPONSE NOT OBTAINED---------.

Check If CST Is Isolated From Hotwell:

Condenser makeup isolation valve CV-31121 -

CLOSED

Condenser makeup flow indication -

ZERO Manually or locally close valve.

14 Check SG Status:

a. MSIVs and bypass valves -

CLOSED

b. Main feed reg and bypass valves -

CLOSED

c. SG blowdown isolated:

CV-31414 -

CLOSED

CV-31415 -

CLOSED

Flow indication -

ZERO Page 15 of 26

a. Manually or locally close valves.

b. Manually close valves.

I~ NOT, JHEN locally close isolation valves (MV-32023 and MV-32024).

c. Manually close valves.

IF NOT, THEN locally close SGB containment isolation valves (MV-32044 and MV-32058).

c

(

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE 1------1 RESPONSE NOT OBTAINED i---------

Caution A faulted or ruptured SG that is isolated should remain isolated.

Steam supply to the TD AFW pump SHALL be maintained from at least one SG.

15 Check If SGs Are Not Faulted:

16

a. Check SG pressures:

NO SG PRESSURE DECREASING IN AN UNCONTROLLED MANNER

NO SG COMPLETELY DEPRESSURIZED Check If SG Tubes Are Not Ruptured:

Condenser air ejector radiation -

NORMAL

SGB radiation -

NORMAL

Main steamline radiation -

NORMAL Page 16 of 26

a. Isolate faulted SG(s):

Locally isolate AFW flow.

Locally close steam supply valve to TD AFW pump.

Verify SG PORV closed. IF NOT, THEN manually close.

IF PORV can NOT be closed, THEN locally isolate PORV.

Attempt to identify ruptured SG(s). Continue with Step 17. WHEN ruptured SG identified, THEN isolate ruptured SG ( s) :

Locally isolate AFW flow.

Locally close steam supply valve to TD AFW pump.

WHEN ruptured SG pressure less than 1050 psig, THEN verify SG PORV closed.

IF NOT closed, THEN manually close PORV.

IF PORV can NOT be closed, THEN locally isolate PORV.

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Rev"ision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE..,_~~~-1 RESPONSE NOT OBTAINED..,_~~~~~~---.

Caution JE CST level decreat;es to less than 10. 000 gallons. THEN alternate water sourcPs For AFW pumps wjll be necessary per C28.1 AOP2. LOSS OF CONDENSATE SUPPLY TO AUX FEEDWATER PUMP SUCTION.

"'17 Check Intact SG Levels:

a. Narrow range level -

GREATER THAN 596

[ATTACHMENT E]

b. Control AFW flow to maintain narrow range level between 5% and 50% [ATTACHMENT E]

Page 17 of 26

a. Maintain maximum AFW flow until narrow range level greater than 5% [ATTACHMENT E]

in at least one SG.

b. IF any SG level continues to increase in an uncontrolled manner, THEN isolate ruptured SG:

Isolate AFW flow.

Close steam supply valve to TD AFW pump.

WljEN ruptured SG pressure less than 1050 psig, THEN verify SG PORV closed.

JE NOT closed, ]'HEN manually close PORV.

IF PORV can NOT be closed, THEN locally isolate PORV.

M Number:

1ECA-O.O STEP 18 r*1t1 e:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE -----1 RESPONSE NOT OBTAINED i----------.

Check DC And AC Bus Loads:

a. Dispatch operator to perform ATTACHMENT H, DC Bus Load Shed

b. Initiate monitoring of DC supply using ERCS c..IE Control Room cooling is lost, ~HEN dispatch personnel to open Foxboro instrument rack doors

d. Power transferrable MCCs and Buses from Unit 2:

1) Check either Unit 2 safeguards 4kv bus -

ENERGIZED

2) Power transferrable 480V MCCs from available Unit 2 bus per 1C20.6

3) Power transferrable 480V Buses from available Unit 2 bus per 1C20.6

b. Dispatch personnel to locally monitor DC supply using local panels.

1) Go to Step 19.

19 Establish Battery Room Cooling:

Contact security to provide access control and block open all Battery Room doors, including fire doors between Battery Rooms

Arrange for f irewatches at the Battery Rooms Page 18 of 26

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1ECA-0.0 STEP 20

Title:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE---- RESPONSE NOT OBTAINED 1---------

Check CST Level - GREATER THAN 10,000 GALLONS Page 19 of 26 Switch to alternate AFW supply per C28.1 AOP2, LOSS OF CONDENSATE SUPPLY TO AUX FEEDWATER PUMP SUCTION.

Numb Pr:

Tit I :

Revision Number:

1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE -----1 RESPONSE NOT OBTAINED t----------.

Caulion

SG pressures should not be decreased to less than 200 psig to prevent injection of accumulator nitrogen in to the RCS.

NOTE

SG narrow range level should be maintained greater than 5% {ATTACHMENT E] in at least one intact SG.

JE level cannot be mainta.ined. J'_ffEJ'[ SG depre:=:surization should be stopped unti.I lewd is restored in at least one SG.

The SGs should be depressurized at a rate sun*.icient to maintain a cooldown rate in the RCS cold legs near 100°F/hr.

Th.is wil 1 minimize RCS.inventor-y loss while cooling the RCP Beals in a controlled manner.

PRZR.level may be lost and reactor veB:=:el upper head voiding may occur due to depressurization of SG{:=:).

Depressurization should not be stopped to prevent these occurrences.

21 Depressurize Intact SG(s) To 300 PSIG:

a. Check SG narrow range levels -

GREATER THAN 5% [ATTACHMENT E] IN AT LEAST ONE SG This Step continued on the next page.

Page 20 of 26

a. Perform the following:

1) Maintain maximum AFW flow until narrow range level greater than 5%

[ATTACHMENT E] in at least one SG.

2) WHEN narrow range level greater than 5% [ATTACHMENT E] in at least one SG, THEN do Steps 21b, 21c, 21d, 21e and 21f.

3) Continue with Step 22.

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE.,_~~~-t RESPONSE NOT OBTAINED ~~~~~~~--.

(Step 21 continued from previous page) 22

b. Manually dump steam using SG PORVs to maintain cooldown rate in RCS cold legs -

LESS THAN 100°F/HR

c. Place TD AFW pump mode selector switch to "MANUAL"

d. Check RCS cold leg temperatures -

GREATER THAN 280°F

e. Check SG pressures LESS THAN 300 PSIG

f. Manually control SG PORVs to maintain SG pressures at 300 psig Check Reactor Subcritical:

Intermediate range channels [NFM power range channels] -

ZERO OR NEGATIVE STARTUP RATE

Source range channels

[NFM source range channels] -

ZERO OR NEGATIVE STARTUP RATE Page 21 of 26

b. Locally dump steam using SG PORVs.

d. Perform the following:

1) Control SG PORVs to stop SG depressurization.

2) Go to Step 22.

e. WHEN SG pressures decreased to less than 300 psig, THEN do Step 21f.

Continue with Step 22.

f. Locally control SG PORVs to maintain SG pressures at 300 psig.

Control SG PORVs to stop depressurization and allow RCS to heatup.

Number:

1ECA-0.0 STEP

Title:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE i------t RESPONSE NOT OBTAINED 1---------

NOTE Depressurization of SGs will result in SI actuation.

SI should be reset to permit manual loading of equipment on safeguards bus.

23 Check SI Signal Status:

a. SI -

HAS BEEN ACTUATED

b. Reset SI if necessary 24 Verify Containment Isolation:

25 26

a. Containment isolation ACTUATED

b. Containment isolation valves -

CLOSED Check Containment Pressure - HAS REMAINED LESS THAN 23 PSIG Check Core Exit T/Cs - LESS THAN 1200° F Page 22 of 26

a. WHEN SI actuated, THEN do Steps 23b and 24.

Go to Step 25.

a. Manually actuate containment isolation.

b. Manually close valves.

IF valves can NOT be manually closed, THEN locally close valves.

Ref er to ATTACHMENT G for locations of valves outside containment.

Reset containment spray signal.

IF core exit temperatures greater than 1200°F and increasing, THEN go to lSACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDELINE INITIAL RESPONSE, Step 1.

Numb~r:

1ECA-O.O STEP M

27

Title:

Rev-1 $ion Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE t--~~~-i RESPONSE NOT OBTAINED....,_~~~~~~--.

Check If Alternate AFW Pump Room Cooling Should Be Established:

a. Check Unit 2 safeguards 4kv buses -

BOTH DE-ENERGIZED

b. Establish alternate AFW pump room cooling:

Notify security to provide access control for the AFW pump rooms

Notify security to block open doors 42 and 43 to AFW pump rooms

Arrange for a firewatch for the AFW pump rooms Page 23 of 26

a. IF AFW pump room cooling provided by unit cooler powered from Unit 2, THE~ go to Step 28.

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION/EXPECTED RESPONSE i------t RESPONSE NOT OBTAINED 1---------.

28 Check If AC Emergency Power Is Restored:

a. Check safeguards buses AT LEAST ONE ENERGIZED Page 24 of 26

a. Continue to control RCS conditions and monitor plant status:

1) Verify status of manual and local actions:

AC power restoration.

RCP seal isolation.

DC power supply.

Containment isolation, if required.

2) Check BAST temperature greater than 145°F.

iE NOT, TH~N consult plant engineering staff for method to reduce BAST boron concentration.

3) Periodically check status of spent fuel cooling:

SFP level greater than 752.5'

SFP temperature IF level less than

""'i52.5', JHEN dispatch personnel to initiate makeup to the spent fuel pit per C16 AOPl, LOSS OF SFP INVENTORY.

4) Return to Step 15.

Number:

r*1t-le:

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1ECA-O.O LOSS OF ALL SAFEGUARDS AC POWER REV. 23 STEP ACTION /EXPECTED RESPONSE i-------t RESPONSE NOT OBTAINED i---------.

29 Stabilize SG Pressures:

a. SG depressurization in progress

b. Manually control SG PORVs to maintain existing SG pressure

a. Go to Step 30.

b. Locally control SG PORVs.

Caution The loads placed on the energ_ized saieguards bus shou_ld not exceed the capacity ot' the power source.

30 Verify Following Equipment Loaded On Safeguards Bus:

a. 480 volt buses:

Bus 111 and 112

-OR-

Bus 121 and 122

b. Battery charger

c. Instrument buses:

Panel 111 and 113

-OR-

Panel 112 and 114

d. Start one air compressor

e. Consult engineering staff to consider repowering the following from Unit 2:

Transferrable 480V MCCs

Opposite train 480V Buses Page 25 of 26 Manually or locally load equipment as necessary.

c

(

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1ECA-O.O STEP

Title:

Revision Number:

LOSS OF ALL SAFEGUARDS AC POWER REV. 23 ACTION/EXPECTED RESPONSE t------1 RESPONSE NOT OBTAINED i---------

NOTE IF RCP seal cooling was previously isolated. THEN further cooling of the RCP seals will be established by natural circulation cooldown as directed in subsequent procedures.

31 Select Recovery Procedure:

a. Check RCS subcooling based on core exit T/Cs -

GREATER THAN 20°F [35°F]

b. Check PRZR level -

GREATER THAN 7% [27%)

c. Go to lECA-0.1, LOSS OF ALL SAFEGUARDS AC POWER RECOVERY WITHOUT SI REQUIRED, Step 1

-END-Page 26 of 26

a. Go to lECA-0.2, LOSS OF ALL SAFEGUARDS AC POWER RECOVERY WITH SI REQUIRED, Step 1.

b. Go to lECA-0.2, LOSS OF ALL SAFEGUARDS AC POWER RECOVERY WITH SI REQUIRED, Step 1.

SG Wide Range Level (%)

90 I

87.5 1 ~

I I

85

' r-..

\\.

' 1' 82.5

' l"I I"'

80 I"'

l"I Acceptabl e Reg ion

77. 5

"" 1'

~

75 1'

~

72.5 I"'"'

,~""

f"'""

70 I

Tube Uncovery Region 67.5 I

I I I 65 Thi s Figure To Be Us ed For Adverse Conta inment Conditions 62.5 I

I 60 I"'

I lECA-0.0 REV. 23 Page 1 of 1 I ATTACHM ENT Ove rf i 11 Region I'"'

I"'

I

"" I,...

I I

E 1 ~

-~

ro,..

1"'

0 100 200 300 400 500 600 700 800 900 1000 1100 SG Press ure (psi g)

ATTACHMENT E Wide Range SG Level For Control l ing Inventory

(

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Operational LEAKAGE RCS Operational LEAKAGE 3.4.14 LCO 3.4.14 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION A. RCS unidentified LEAKAGE not within limit.

B. Required Action and associated Completion Time of Condition A not met.

Prairie Island Units 1and2 REQUIRED ACTION COMPLETION TIME A.I Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limits.

B.1 BeinMODE3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND B.2.1 Identify LEAKAGE.

54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months OR B.2.2 Be in MODE 5.

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months 3.4.14-1 Unit 1 - Amendment No. 158 Unit 2-Amendment No. 149

(

ACTIONS (continued)

CONDITION

c. RCS identified LEAKAGE not within limit for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

D. Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

Prairie Island Units 1and2 C.l AND C.2.1 OR RCS Operational LEAKAGE 3.4.14 REQUIRED ACTION COMPLETION TIME BeinMODE3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Reduce LEAKAGE to 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months within limits.

C.2.2 Be in MODE 5.

44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months D.1 AND D.2 BeinMODE3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months BeinMODE5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 3.4.14-2 Unit 1 - Amendment No. H& 1771 Unit 2 - Amendment No. -149 167

RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.14.1

N 0 TES--------------------------

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

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE within limits by performance of RCS water inventory balance.

SR 3.4.14.2

NOTE----------------------------

N ot required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is

5: 150 gallons per day through any one SG.

FREQUENCY 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 72 hours Prairie Island Units l and 2 3.4.14-3 Unit 1 - Amendment No. +§.& 1771 Unit 2 -- Amendment No. +4-9 167 !

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE l

Page 93of1541

- --*-------******-**-*---*-*'-----------~--~

Table 2.2 Radioactive Liquid Effluent Monitoring Instrumentation With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels Operable, take the actions directed in Table 2.2. Restore the inoperable instrumentation to Operable status within 30 days. If instrumentation is not restored within 30 days, explain in the next Annual Radioactive Effluent Release Report, why this inoperability was not corrected in a timely manner.

INSTRUMENT

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a.

Liquid Radwaste Effluent Line

b.

Steam Generator Slowdown Effluent Line

2. Flow Rate Measurement Devices

a.

Liquid Radwaste Effluent Line

b.

Steam Generator Slowdown Flow

3. Continuous Composite Samplers

a.

Each Turbine Building Sump Effluent Line

4. Discharge Canal Monitor

5. Tank Level Monitor

a.

Condensate Storage Tanks

b.

Temporary Outdoor Tanks Holding Radioactive Liquid

6. Discharge Canal Flow System (Daily determination and following changes in flow)

MINIMUM CHANNELS OPERABLE 1

1/Unit 1

1/Gen 1/Unit 1

1/Unit 1/Tank NA APPLICABILITY ACTION During releases 1

During releases 2

During releases requiring throttling of flow During releases During releases At all times When containing radioactive material When tanks are in use At all times 4

4 3

6 5

5

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE I

H OFFSITE DOSE CALCULATION MANUAL (ODCM)

I NUMBER:

I H41 REV:

28 Page 94 of 154 !

-*---*-*---*-*-*-*n----'~----------------------~--------~

Table 2.2 Radioactive Liquid Effluent Monitoring Instrumentation Table Notations ACTION 1 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue, provided that prior to initiating a release:

a.

At least two independent samples are analyzed in accordance with Specification 2.2.1, and

b.

At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of not more than that specified in Table 2.1 for Principal Gamma Emitters.

1. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is

_::0.01 µCi/gram DOSE EQUIVALENT 1-131, or

2.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is

<0.01 µCi/gram DOSE EQUIVALENT 1-131.

ACTION 3 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and saved for weekly composition and analysis in accordance with Table 2.1.

ACTION 4 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that the flow rate is estimated at least once per four (4) hours during actual releases.

Pump curves may be used to estimate flow.

ACTION 5 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that tank liquid level is estimated during all liquid additions.

ACTION 6 With the number of channels Operable less than required by the Minimum Channels Operable requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gamma emitters.

Number:

~,it I e:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 CONTINUOUS USE Continuous use of procedure required.

Read each step prior to performing.

Mark off steps as they are completed.

Procedure SHALL be at the work location.

PORC REVIEW DATE:

OWNER:

EFFECTIVE DATE:

5/6/10 D. Smith 5/7/10 Page 1 of 13

Numb"r:

litle:

Rcvi s ion Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 A.

PURPOSE This procedure provides actions to respond to voids in the reactor vessel head.

B.

ENTRY CONDITIONS

1. This procedure is entered from lF-0.6, INVENTORY STATUS TREE, on a YELLOW condition.

C.

ATTACHMENTS:

ATTACHMENT F:

Reactor Vessel Vent Time Calculation Page 2of13

Number:

r; tl e:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED RESPONSE -----t RESPONSE NOT OBTAINED ---------.

J_f a controlled natural circulCJtion cooldown is in progress and a void in the reactor vessel upper head is expected, l'ffJ;;lf this procedure should not be performed.

1 Check If SI Has Been Terminated:

a. SI pumps -

BOTH STOPPED 2

Check If Charging Flow Has Been Established:

a. Charging pumps -

AT LEAST ONE RUNNING

b. Instrument air to containment -

ESTABLISHED

c. Charging flow -

ESTABLISHED Page 3of13

a. Return to procedure and step in effect.

a. Perform the following:

1) _IF CC flow to RCP (s) thermal barrier is lost, 1HEN isolate seal injection to affected RCP(s) before starting charging pump.

2) Start one charging pump.

b. Establish instrument air to containment.

c. Establish 30 gpm charging flow.

lE 30 gpm charging flow can NO'l:'_ be established, ~HEN return to procedure and step in effect.

Numbc-r:

lit le:

Rev-is.ion Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED RESPONSE i------t RESPONSE NOT OBTAINED 1--------~

3 Check Letdown - IN SERVICE 4

Establish Stable RCS Conditions:

5

a. PRZR level -

GREATER THAN 90%

b. RCS pressure -

STABLE

c. RCS hot leg temperatures -

STABLE Check RCPs - BOTH STOPPED 6

Check If RCS Pressure Should Be Increased:

a. Pressure -

AT LEAST 100 PSI LESS THAN LIMIT ON FIGURE FRI3-l

b. Pressure -

LESS THAN 1550 PSIG [1875 PSIG]

c. Turn on PRZR heaters to increase RCS pressure by 50 psi Page 4of13 Establish letdown per 1C12.1, LETDOWN, CHARGING AND SEAL WATER INJECTION.

re letdown can NOT be established, ~HEN establish excess letdown per 1Cl2.l, LETDOWN, CHARGING AND SEAL WATER INJECTION.

a. Control charging and letdown as necessary.

b. Operate PRZR heaters and normal PRZR spray as necessary.

IE normal spray NOT available and letdown in service, THEN use auxiliary spray.

c. Dump steam as necessary.

Go to Step 12.

a. Go to Step 9.

b. Go to Step 9.

Number:

lit le:

Revis-ion Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV. 10 STEP ACTION/EXPECTED RESPONSE ____ _, RESPONSE NOT OBTAINED 1----------.

7 Control Charging And Letdown As Necessary To Maintain PRZR Level Greater Than 21% [41%)

8 Check RVLIS Full Range Indication:

a. Indication -

INCREASING

b. Indication -

93% OR GREATER

c. Turn off PRZR heaters as necessary to stabilize RCS pressure

d. Return to procedure and step in effect Page 5of13

a. Go to Step 9.

b. Return to Step 6.

Number:

1FR-l.3 STEP f iii e:

Revision Number:

RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 ACTION/EXPECTED RESPONSE i------1 RESPONSE NOT OBTAINED i---------

IE RCP serLl cooling had previous] y been lost. ]_JJliN the a[fected RCP(s) should noL be started prior to a status evaluation.

9 Attempt To Start One RCP:

10

a. Establish the following conditions prior to RCP start:

PRZR level -

GREATER THAN 90%

RCS subcooling based on core exit T/Cs -

GREATER THAN 55°F

[70°F]

Use PRZR heaters as necessary to saturate the PRZR water

b. Start 12 RCP per 1C3 AOPl, POST ACCIDENT START OF A RCP Check RVLIS Indication:

Dynamic range -

52% OR GREATER WITH ONE RCP RUNNING

-OR

Full range -

93% OR GREATER WITH NO RCP RUNNING 11 Go To Step 20 Page 6of13

a. LE conditions NQT established, THEN go to Step 12.

b. 1£ 12 RCP can NOT be started, THEN attempt to start 11 RCP per 1C3 AOPl, POST ACCIDENT START OF A RCP.

Go to Step 12.

Numb0r:

r i tie:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED HESPONSE 1------1 RESPONSE NOT OBTAINED ---------

12 Check If SI Signal Can Be Blocked:

a. PRZR pressure -

LESS THAN 2000 PSIG

b. Block SI 13 Record RCS Pressure 14 Establish Following RCS Conditions:

a. PRZR level -

GREATER THAN 90%

b. RCS pressure -

STABLE

c. RCS subcooling based on core T/Cs -

GREATER THAN 70° F [ 85° F]

d. RCS hot leg temperatures -

STABLE Page 7of13

a. Decrease PRZR pressure to less than 2000 psig using normal PRZR spray.

I~ normal spray NOT available and letdown in service,

.'.!'_~EN use auxiliary spray.

_IF NOT, TH_f.N use one PRZR PORV.

a. Control charging and letdown as necessary.

b. Control PRZR heaters and normal spray as necessary.

X£ normal spray NOT available and letdown in service, TJIEN use auxiliary spray.

c. Dump steam as necessary.

d. Dump steam as necessary.

Number:

!Hie:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED RESPONSE -----1 RESPONSE NOT OBTAINED ---------.

15 Prepare Containment For Reactor Vessel Venting:

a. Start containment air circulation equipment:

All CFCUs

All containment dome recirc fans 16 Determine Maximum Allowable Venting Time:

a. Containment hydrogen concentration -

LESS THAN 3.0%

b. Determine maximum venting time per ATTACHMENT F Page 8of13

a. Reduce hydrogen concentration with recombiners per C19.8, POST LOCA H2 ELECTRIC RECOMBINER CONTROL SYSTEM.

Number:

Title:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED RESPONSE t------1 RESPONSE NOT OBTAINED..,.... ______ _

17 Review Reactor Vessel Venting Termination Criteria:

RCS subcooling based on core exit T/Cs -

LESS THAN 20° F [35° F]

-OR-

PRZR level -

LESS THAN 21% [41%]

-OR-

RCS pressure -

DECREASES BY 200 PSI

-OR-

Venting time -

GREATER THAN MAXIMUM TIME CALCULATED IN STEP 16

-OR-

RVLIS indication:

Dynamic range -

100% OR GREATER WITH BOTH RCPS RUNNING

-OR-

Dynamic range -

52% OR GREATER WITH ONE RCP RUNNING

-OR-

Full range -

93% OR GREATER WITH NO RCP RUNNING Page 9of13

Number:

l l tie:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV. 10 STEP ACTION/EXPECTED RESPONSE 1------1 RESPONSE NOT OBTAINED i----------.

eau.tion IF any vt.>nt termination criterion in Step 17 is exceeded,

_'J}IEl! venUng should be stopped.

18 Vent Reactor Vessel:

19

a. Open head vent path to PRT:

1) Reactor vessel head vent valve:

SV-37037

-OR

SV-37038

2) Head vent to PRT valve SV-37039

b. Any venting termination criterion in Step 17 -

EXCEEDED

c. Close all vent valves Check RVLIS Indication:

Dynamic range -

100% OR GREATER WITH BOTH RCPS RUNNING

-OR-

Dynamic range -

52% OR GREATER WITH ONE RCP RUNNING

-OR -

Full range -

93% OR GREATER WITH NO RCP RUNNING Page 10of13

a. IF' head vent to PRT valve fails to open, 1fiEN open head vent to containment valve SV-37040.

b. Continue venting. WHEN any venting criterion is exceeded, THEN do Steps 18c, 19, 20 and 21.

Increase RCS pressure to value recorded in Step 13. Return to Step 12.

Number:

Title:

Revision Number:

1FR-l.3 RESPONSE TO VOIDS IN REACTOR VESSEL REV.10 STEP ACTION/EXPECTED RESPONSE.....,. __

--1 RESPONSE NOT OBTAINED t---------.

20 Check PRZR Level - STABLE 21 Return To Procedure And Step In Effect

-END-Page 11of13 Control charging and letdown as necessary.

~

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX Abno"'"

Rod Rtiu10 Rod Effl""'

Off.!te RIMS Conditlono Onslte Rad c--

PINGP 1576 Rev. 8 Doc. Type/Sub Type: EPIEVT Retention: Lifetime

tfdow HSl!Slinent ""Its ~ 111a-*81~ 11Iii.1r.. ol1Mcitir-1ion.

lhdci.11Hlflc11!MJnlhook:I b.ibiK<tdonRGI 21nlludolR01 I Vlhh n"'1:"uary d<<:lara:iofl5 *hOOld not be dollaoy~ *....,.,..g lffull.I 1r111do1WllJUSlif"'!tllllll>Olll1beini1111t*d la>r1pl;ttl'dirlor\\$et'IO d11l<<n1,,_ II 11'111 dasslfil:al!Ofl $hOUU ~

'ublWtQuto"l:tl 6~1i111d V"LID r-fmg on ooo or "10I'* nionitors lisltid In Table R* I lh.11 Heeads or expoc:lfld lo exceed cokinin *GE" fOf" 15 mlnU:oa °' longer RG1 2 I - -, I 2

I 3

I - 4 J" 5* I ---,f I DEF I OoGe 11sses&ment ur.ing actu:il mei.arok>gy iodlcnles do&es GREATER Tl W~ 1000 mRem TEDE er 5000 mRem lhyfoid COE 01 CT beyond the lltllbol.rdary Fi., ~It)' re!MAts indicallt dos.ad window do'ie ratea e1.ceeding 1000 n1RllJ' expected to contlooe for mc-'e th.Jo one~ at Of beyond lllte bo!nlary OR Al'l."\\fyl>ea of OolJ Bl.fVey&afl1pleo trdicale lh)toid COE or 5000 01Rem fc-'

onel'!Oll"oflnh."'!atlon. ;]lorbeyofld sitebo1.1~iwy ido511 :t!!Hnneru r*~ ""**~*bbl*.,tllv lio"io9 of~tiofl l!l<f dlull'atoon 5hould t. based on RS121111:..ad d RS1 1 vn. n.c.KUry d<<Wr..o,. tl>Ot.lld nOI be d!H;l/ed,...,.,~ 1esu1t1 1"4t~MWRn<<lla.ouldbelrlr:t.alotJ/~edillOfdWIO d.if'lrllMo ii Ille d;lu1'\\'.::11jg.n thO'.lld be *UOwqu.n:lfHCN1.J VALID reat.Hng on one or mcll"e monll.or* ltslod in Table R-1 lhnl &11eeed1 or1s eir.pecled 1011.<.oeed cf!ll,JITln "SAE" for 15 minutes or lorogor Fiekl str111yreultslndtcalflcio!le(lv..inOciwdOS4tr.1htseirceeding 100mR/lY expoctOO to oonlirua loro11J"a th.-lf1 one hou', at or beyon.:I lhemleb0t¥ld:iry*

OR AoalysM of lk!IJ W°VllY aampks lr.Jicote thyroid COE of 500 mRem fOf ON! l!Oll of iM31."\\l.ion. a1 or beyond the size b!Mrd;wy Table R*1 Entuont Monitor Cllissfflcation Tiwnhokb Monitor GE SAE Alo<t UE

!l!ll2l!l

£eM

£eM 1(2) R*SO H19h RaJ'9(t Stack Q3i; Montor

-13000mRltv-4300mRlhr NIA NIA IR-22' Sl'iald Bui~ Vent R:1d Monitor NIA NIA 100 000*1 16 E5 t 600"1 t 6E3 2R-22".$Mid Bulding Ven! Rad Monitor NIA NIA 100 000*1 1 es 1 000*1 1 EJ 1R-30° & 1R*37" Uri\\ 1 Aux 8ui<Jing Vent RAd Mol*IOl's NIA NIA 100.000'1 1 E5 1 000"/ 1 El 2R30' U11! 2 Aux Btiiding Ven! Rad Monltn NIA NIA 100.000*1 1 E5 1,000"/ I EJ 2R-37" Urwt 2 AIA. &Jldlng Venl Rad Monil0f1 NIA NIA 120,000"1 1 2 E5 I 200"1 12E3 R 35* Radwaste Blalding Venl Rad Monllnr NIA t4/A 100 000*1 1 es 1 000"/ 1 E3 R*25" & R*31

SPf!rll Fuel Pool Vent Rad Mondus NI>

WA 800,000"/ 8 E5 3000*13£3 l.l!l!!lll R* 18' Wttt1te Etnuent Llqi*! Monitor

'"A NIA 900 000*19 es 30 000"/ 3 E4 IR* 19" SG Slowdown Radltttion Monitor NIA NIA 100 000*1 1 E5 1.000*1 1 E3 2R-19' SG Blo\\Yckr.vn ROO'iattonMonilOr NIA NIA 00.000'16 e.i 600"t6E2 R-21 CircW:itorDlsclllYneMon.!or NIA NIA 800000/8E5 8000'8El Ha1u

1) (AC!I CAL Al....,.l lno;I~**

n CAL thNtllold M91 ll-too.11 U C4dd f...tllu""klation ol tt.Bdi..,_......,llo<rHdl"lil *Maffll<ltodst9<"m!M lf lhe CAL lhr.. lloldlln.~ffd*cl.

Z ) "ApjlliHwll111CM-6iolc~110tlloi.t.<I.

RA1.1 r

-1 l 2" l

-3 I

4 I

5 I e I DEF I VALID r63ding on anyelfluari monitor 1hat exceed& 200 Timfl Iha at.vm isetpoint established by a c11reo1 ratioactivlty dscharge pemiilfor15rninutHotlonger OR VALID reaciog on eftluefll moritor R-18 U1at e1.ceeda 900.000cpm fcr15mlnutesorloriger RAU I 1 I 2 I 3 I 4 I s I e I oEF J VALID readi119 Of! one Of more of till! fol~ng ndation mo,..1ors (Table R* 1) Iha\\ e1.ceods Iha reOOlng stlQ'Wrl IOf 15011.Ues or longer eonr,med s1V11ple anal)"J* rOf gaseous or iqukl releoae inc.llCale&

concentrations or relea!>e mlet1 with 3 release d\\lralion of 15 m11iute1 or longer io exc6" of 200 rimes OOCM 1peciACMk>n A VALJCLilirnl cifiont-Or ITIOle of the k>lloY.+ng radi3tion mooltors

R*25 or R.JI SFPAir Mo~o1 (HI Alarm)

R-5 FuelHardlrigAru Monitor tndllll \\HIAMrm)

R-28 New Fuel Pool Cr*ic*l*y Arn Monrtor (HI Ali.rm)

1 {2) R-11 CtmliSBV Air ParticUare Monrtor (HI Altrm)

1(2) R-12 CtmtlSBVRadiO Ga1 Moritor {HI All1m)

1{2) R-2 Conlalnment ~ne l Arn Mori101 (HI Alltm)

Water level LESS THAN 10 reel :ibove 11n hai;il..11ad fuel :i1Mtmhly fa-lhe reactCf refueling cavity, 5J!OOl loo! pool 11rwt fuel lmn&hir cannl Uul IYill re&uU mITTa..iated rue1 uncovering Release of~

Mal#iol or 111C1UH9 In RadAtion t..Y'tlll Wi_., the Fadily ThAI lml)llde* OperllUon d S~llMOS R9qt11red IO P.Wnll9n SAJ.t Opera6one or IO El&Ulhkh Of Maltsain Cold RA31 I 1

I 2

I 3

I 4

I s

I 6

I DEF I VALID rafiallon mori1or1e.ring& GREATER THAN 15 n1Rl'h" in a1113s roqUring conliruoOs occ14><-"W'IC)' lo m11irta1n plMC safety fi.n:tionl Cornrol Room (ROO moritor R-1)

OR Central Alarm Stnbon (hy portable radi.ilion n1onitOt1ng lns!Nmenl3l.ion)

RA3.2 !

1 I 2 I 3 I 4 I 5 I 6 I DEF I Any VALID radalion rtlotlllc-' reading GREATER THAN 1 R/hr in meas reqtAong 1"req1,11i; &ec.ess to ma1oloin plaril sarety fi.n:llon5 (TllbleH-1)

- Shield/CoritaifllTl6nt 61.ilding

Aux~iaryBuilding

05"06 Diesel Genen1tor Building

- Plant Suoonhouse

Control Room

-Relay Room

- Tixbir.e Bu~ding

Condensate Storage Tanks VALID re."leing Ofl one or rnore ol 11* Folk'ng *aclotiooi ITIOflllorJ fTol>le R-1) lh.lt axce<<!1 Iha re11ding 1hown for 60 minule& or longer Ru13 !

1 1 2 1 3 1 **-i--s-r,- 1 oei"l Conf.rrned sample 01'1."ll)'Sll lor gaseous or iqtjd r&l&aMI Indicates coric::anll"illions or relooso rotas, with a release d1J11;1tlon of BO mini.It&&

or long6f In eitCel5 d lwo ~niea OOCM apeclflc.aUon RU2-l&ip<<:ted i0cre3N RU2 1 [TT2 HT--l"*r---.---T *5 -,

e I DE i" I Vit.Ll0 irdt3~onof~wat81 le11tt1oecreas.e in U19teitCtOf ref'-*ing cavity ;pef'll tuotpool or fueltr11t1Clarean."llW1thall ITikfiat&d fuel ass..,1blies remai~ coverltd by water as Aie<l!ltd by level LESS THAN SFP 1oW W:'ller level oJarn1. Rllfuekng CAMI Level or Y1&ual ob5ervlllal f752 5 feet eleva~Ofl)

ANO Arrf UNPLANNED VALID Hen Radiabon Morilor reading increases asil'llkstedby:

R-5 Fuel Herdt119 Area Monrlor reading

R-28 Newfuet Pool C1n1c.~1V ArH Mon1to1

1(2) R-2CorQ!nn1entVHselAIH Mol'\\llor

Olher Port11ble Area RadiatlOn Mon*orl'lg lnstrumenlalion Ru2. 2 1 1 1 l

1 3 1 '

1 s r--e--T oEFJ Any UNPLANNED VALID Area Radiahon Morltor reading Increase;; by a loct0t or 1000 fNet normal" levels

' Noml:i! le'.'els can be conslde1fld as !he highest rending In !Ill! p..ist twenty*rou-to.n e*eWng the cinn pe.:1' value Tablo H-1 Plant Areas EIAU<<lt HU1.a* HU2 1' HA T:~>t:iAf. 3 *H-AT-1 HPJ5-- HA2.1 KA.3.1" HA3.2*

RA3.2

Also consider aret1S contiqoous to Hlese.

Page 1 of 8

~..,.,

Prairie Island Nuclear Generating Plant Hanrd1 Natunl &

OostrucUve Flro or Eiplo&lon Toxlc al\\d flenwnabl*

o..

PIN GP 1576. Rev. 8 Doc. Type/Sub Type: EPtEVT Retention: Lifetime.,.

Nono No~

Nono r

EMERGENCY ACTION LEVEL MATRIX

Shieki'Containrnent BL.ilding

-Auxikt1ryBti!ding

D51D6 Diesel Generator B ~ ldi ng Plant Scrnenhouse

. Control Room

Ra i a~ Room

Tixbine Building

. ConOOnsate Storage Tanks HA1.1 I 1 I 2 I 3 I 4 I 5 I i

I DEF l Seismic Evert GREATER THAN Opaoa~ng Basia Eorthqual;e (OBEI as irdic.oled by "OBE Elceedance" alarm oo SeiMnlc Monitorifv:I Panel HAU I 1 I 2 I J I 4 I 5 I ti I DEF I Tornado or l19h >Mods GREATER THAN% mphwitiVI PROTECTED AREA bourdary Mid resu!Wig in VlSIBlE DAMAGE toanyd the fole'Mng plart struella'86 / equipment or Conlrol Room indication of degraded perlorm:uu d tho&. systen11 (Tabkl H-1)

Vehicle crash will in PROTECTED AREA bovndary ond resul'ling In VISIBLE DAMAGE lo:myd tl'8 folow!ng plantstruettHaleqtJpment therein or Control Room lndlcabori ol degroded periofmance of tll058 syslems(TubleH-1)

HA1.4 I 1 I 2 I 3 I 4 I 5 I II I DEF I Turllioe fai111e*gll0tlfn!ed missile1 res!Jt in any VISIBLE DAMAGE to or penetrn!iorlorarryofthe rollooN!r:g pltlnl nreaa (Table H*1)

HA1.5 I 1 I 2 I 3 I

I 5 I 0 I DEF I Uncomol4!Jlloodng ln anyTnble M* l weaor theplant ltlol rH~tsin dogroded &.1fety system performorv:e aa lrd'ieated in 11.e COl"W'ol Room or Thal etMtes irlduStrlot sarecy ha::11rds (e g electric *hoclo.) that pr9CluJes acceu Mees&11ryto operate<< monilor urety ~I'll HA1" I 1 I 2 I 3 I.. I 5 I 0 I DEF I High or loWriVGf WOlf!f le'1el OCCl.IT..-.CU affectlr-g lhe PROTECTED AREA as indicated by Rrter intt*.ele.,e!GREATER THAN 693ft MSL OR Rmw intM..e level LESS THAN 666 S fi MSL

alWI of ro~w: or Flommabli 0 3SH Within or Contlguout to a VITAL AREA Which JeopArcfies Opemlion of System1 Req1.1red to Malntoln Sare Operalion& or E1tablh1h Of Ml!intain Safe Sh.rldown.

HA3.1 I 1 I 2 I 3 I 4 I 5 I II I DEF I Report Of dot~ion or tOJ1k: g;is.5 wlt!vn or COflligl.IOu!I to T nbl& H-1 areas lo eoocen* 3tioos lN1 "'llY re~~ l in QI\\ a!mosphe10 IMMEDtATEL Y DANGEROUS TO LIFE IV~O HEALTH (IOLH)

~~ L

~~L!

gal~

~n L;,raLAR£fE~ ftA~E:.'

LO'Y'IER FLAMMABILITY LIMIT wilHn or contl{Juous to Tril>le ~

1 :."&as Table H* 1 Plant Anma HU1-6' HU2.1" HA1.2 HA1.3 HA1.4 HAl.5 HA2.1 HA3.1" HA3.2" RA3.

" Alooconsiderareascontiquous !o these EarthQuak* lelt in plane aa lndlcaled by VALID "Evanl"' alarm on Sel1111icMor1toril'lgP:tnel ttu1 2 I 1 I 2 I J I

I s I e I DEF I Rep:!f1 byplonl perwm&I of lom.'ldo or high winds GREATER THAN 95 mph &trilling y,(Jlin PROTECTED AREA bot.ndary Hu1.3 I 1 I 2 I J I " I 5 I e I DEF I Vehicle cr3&h Into plant ilttlttutU or &ygtema wldlin PROTECTED AREAt>otn!ary HU1.4 I 1 I 2 I l

I " I s I e I DEF I Repc:rt by pla!'IC persom&I or an l6'13nticipi8tecl EXPLOSION wilfwl PROTECTED AREA bouridlry r.. tilrng In VISIBLE DAMA.GE to permanent slruciure or equipruef'll:

HUU I 1 I 2 I 3 I ' I 5 I 11 I DEF I Report of llf'Ulne ri.ilt.re resUllog In casing per.natlm or dan1age to tl'btne or i:ieneraia sea!*

Hu1.& !.. 1 1.. 2._1 l._

I_

-~

- J. ~._ L _~. _ (~eF HU1.7 I 1 I 2 I J

I '

I 5 I e r oe*,: *1 High or low 1iv.twater level ocanences alfecting the PROTECTED AREA as if'di<;oted by River incak.e level GREATER TltAN 692 n. MSL HUJ.2 I -*, I 2 I J I

I s I e I DEF I Report by Local Cot.\\ty a State otrdltls for eYncu.1tion or sheltemg of sil* p<<soM81 ba&ed on an olfsile event 0

HOT & COLD Natural&

Dfftructlve Phonomonon Flroor E-ploslon Tod c~nd F~ m~blo a..

H.n:atck Page2 of8

~

Prairie Island Nuclear Generating Plant Hau rds Continued S.ci.rlty Comrol Room Evacuation

&Mreency Dlroctor Judgmont PINGP 1576 Rev. 8 Doc. Type/Sub Type: EP/EVT Retention: Lifetime +

ttG1.1 I 1 I 2 I J

I 4 I 5 I e I DEF I A llOSrtLE ACTION hos OCCllred ISUCh th.al plar( p<<somel 8"

""'ible to OJ*:tle eqiJpment requirftl.l to maintain saletyf~.

HG12 !

1 I 2 I 3 I

I 5 I II I DEF I A I IOSTILE ACTION 1'\\11!1 Cl:IU&ed falk.re oC Spent Fuel Cooing System&

ond IMMINENT fuel dan1394' is ~kely for u f1eshly otT-loodod reBCl.or core in pool

,z

-Oth&rCOiiiillO.. Exlling vmch in-lhe Judgment ol 1he Emtwgency Di'eclcr Warrant Ooci8ration ol Generel Emsgency.

HG2.1 1*-,--1 f ]- 3---r-, - I -s-1-*a-roei= I Other condlioN eiusl which In U111 futlgmert or lhe EmeigMC)' Director lndic.~te INI 8'."llnt' are In proceg& or hove OCCUITed wlich ll'!'IO!Vil uctutll or immlnenl llbslanllal core degradation or m11ltlog wHh polenllal for lot.S of COfUlrJnent ntegrily Relea&&s can be reaf;()f\\,'lbly expeclod lo exceed EPA Protectlv* Action GWJOine exposue levels offsile for mere lhnn the immedt1le sit* aren I'-,

EMERGENCY ACTION LEVEL MATRIX HS2 COnrol Room Evi>Cootlon Ha&

Carrot BeE&tablished.

lnilia1edDndPianlCOt'WfOI HS21 (

1 I

2 I

3

' *-~*

~

5 I e I DEF I Controlroom9Yacualion1,..,......,..,.,...,,....,

AND Cortrol d U'8 pla1\\I cannot be establrshetl per 1(2)C1 3 AOP-1, SIUdoWn from Outsida the Cortrol Room or F.5 Appentlilt B Cortrol Room EvaclJ3riOt'l (F"ire) "Mthin 15 ml...nas HSJ Ottw Condtlloos EXiaUng Whcii In lhe JUdgment of the Emergency Direclor WDIT'&rt Dedaralion of Sile Ar*a Emergency HSl.1 I 1

I 2

I 3

I 4

I 5

I 6

I DEF I OtlW!I' conditlol'l6 eicisl wllich Jn tho jlldgmenf of tho Em<<ventY Director indicate !hilt *"anls are in proce11& a hllYe occall'Ted which lriYolve nctu31 or hkety nmlor lailt *.-es d pl3l1t lundlona ne8<1e.J fur pro1eclion of the public Artt rele.1Ms are not e:io.pected 10 reault In &xpo&l#'e ltw11h~ wt9ci1 e11ceod EPA Protective Actlon GOOellne e1tposll'e levels beyond 1he site txxnl3ry Am

- Shield/Cootainn1en1 Blilding

- Auxifiary Blildng

- 05106 Diesel Geoorntor Building

- Plant Scraenhouse

. Control Room

- Relay Room

- Tt1bina Building

- Condensate Storage Tanks HA4.1 I 1 I 2 I s I I

5 I II I DITT A HOSTILE ACTION II OCWTing or h.1!1 OCCUTed l'<ithln the OWNER CONTROLLED AREA as reported bySeolfty Slirt

~~jion.

_1~j~2~j~,~,~.~,~,~,~

. ~,~DE~F~j A validated no(if"!Calion rrom NRC of on itirilner i.U.ock UYeal tMthln JOminutesoflhesile HAS Control ~oom Evacuo:itlon ~

BMn" l"*lalid HAS.1 CJ:"C 2 T_T _J--4---J T" l *e-I oei' I En¥)' Into 1(2)C 1 3 AOP-1 Sh.Adown from OutsiJe 11'111 Comal Room or F-5 ~

B CoNrol Room f 'lllCU!lliotl (Frre) for control room evlk:O(!tlon HA6 Othe,. Cordrlbis Existing Which In U'lll Judgmet'i Of lhe Emergency Director W1uran1 Dedemtion of en Alert HA8.1 1 1 1 2 1 3 1

,-1~ --n-- r oeF 1 Ottier conditions elllst which In the judgnltlnt d lh* Em01gency [)lreclof" Indicate 111-11 evonls nre in pmces.1 or hMe occurr*d wtlich involve t'lctual or lilt.ely potentll'll subslllntlal degrl'ldrllion of 1/111 level of u fely ol 1he plan!

Any relea608 are expecied to be im~etl to 1n1al rraclioN olthe EPA Pro1ect1Ve Aclion Guideline Hpo&lre level1 Table H-1 Plant Areas HU1.6" HU2.1* HA 12 HAl.3 HA1.4 HA1.5--HA2.1--HA3.F -HA3.:2'-RA3.

AJso considef' are... 1s conLi~uous to those.

A va&ooted no~ficntlori from NRC provkf1119 Wormalion ol ari11ircr'lllllhr11N1l HU5 Other Condotlona EXWlflii w~

In u.. J\\M:lgmen1 oru*

En1er119ncy Director WRrnnl Oec:llnlion d

UE HU5.1[1 -rT J *3-T--,- I 5 I s I DEF I Other COl'dltlona e!llal v.t1id1 in tl'lll JutJgrnent of the EmW981"CY Oll'&dOI' lrdcnte 1h.1\\ even1s !V11 In proce1111 or have OCWTIKI wNch indeato a polenUQ! degr.Jdt'lllon of Ille lovet of ta I Illy of \\he plant No tel.asea of ra1:f1011Ctwe n1ater1ttl reqLJring onal10 re&ponse or monllotlng rtr* e*fl9cted unl&H r1.11her degradltllon or aafety syctema ocetTs

/'"\\

HOT & COLD Socur1ty Control Room Ev11cuatlon Ern..-gonc:y Dlroctor Judgment Ka:zard5 Continued Page 3 of B

Prairie Island Nuclear Generating Plant Sys tom Malfunct.

Lot.I~

RPS

,allure liubNltyto Rh<hM Malnt*tn Shutdown Condl!ICH'ls Inst.I Comm.

PH~GP 1576. Rov (J Doc. TypE!.ISub Type: EP/EVT Retention: Lifetime.-

L.Dss.ofpowertoorfromTmn5foonerSC T::-w-cr. 12, 1RY and2RY ltult resulls ln a ton of oil off site pow9' to bolh S.1regu., n1s Buses 15 aod 16 (25 and iii)

AND Failure ol OiMel Gerwalcn 0111NI 02 (05and 06) to aup~ypowe1 1o Safeguards Buses 15 ond 16125 and 26)

AND E1theror1119follov.1n1r a

Ra&1cra~on of Saleguiwds Bus 15 Of 16 {25ot 26) wl!hln 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sls!!Ql.ikely OR b Contirting degrocl:tliol't of core cooling bnsed on F1&&lon Prodtlct Barrier mof"itOl'lng i)S lrn'CDhtd by COie Conlmg-HEO or ORANGE patlt SG2 Failla"& of tt* RHCior Protec:lion System 10 COmp1tt111 an Automatic Tnp Mid M3rtiol Trtp was NOT SucclUflA 1md There ii tndicUUon ol *n Exnme ChOhnge 10th& Ablllly 10 Cool lt1' Cor*.

SG2 1 !

I I 2 I I

I I

lnr.k fttion(&) exisl lhol :t1~onK1lic :ird manual trip wete NOT succes.;flA in reducing powot to LESS THAN 5%

AND Eiln&rofthe fol!O\\'Afig a Co-e cooll'lQ Ill u.treniely chslleng&d M lndicutt>t1 l1y Core Cooling - RED p..'\\lh OR b Heat removlll Is e11:tremely ch.illenged a.s indic.itfld by I-teat Siok - RED pntt1 N~

No"'

.'~

EMERGENCY ACTION LEVEL MATRIX L0&s Of power loorfrnm Tranafor~ CT* \\ I CT*12. tRY, ontJ 2RY tl1.."tt resuha In a loss of ()II oll'sile power to bOth Sof119L1.1rds Buses 15

tnd 16 (25And26)

AND Foilu"eofbothOl&sel Generator& 0 1 and02 (05ond DG)tos~

power lo Srlfagu:wds 8u&e3 15 11nd 16 {25and 26)

AND Fa~11e 10 restore power tti Safegintds Bus 15or 16 (25 or 26) 'Mlhn 15 min.CH tom the bme ofloH of bolh otr&lle andONite AC powtW SS3 lOM d M Vital DC POMW.

I SSJ.1 !

1 I 2 I 3 I 4

I I

Lou ol all Safo9uird1 DC POl'tU ba.!WKI Ofl LESS THAN 112 VDC on 125VOC Panel1 11 and 12 (21 and 22) for GREATER THAN 15mlnl,ite11 SS2 F.!likH of Rfmtl<< Pl'lltecilOO Sptem lnllt\\lrl'letto1ion to Compiet* or l~llt* an Aucom1t1ie Reaaor Trip Once o Reactor Protecion System s.tpoint HM Bffn ~

afld M.11"1Ua1 TtlJI Was t-IOT SUccassluf I ss2.1 !

1 I 2 I I

I I

lflic...,lior(s} e.-Jst lhat al.(On'l:lltic and nlarw.ml trip were NOT tuccef>SflA In radut;ing power to LESS THAN 5%

1abilty Losa or oou1 cooling oro heat \\\\Ink u lrrlcnted by:

~ Core Cooling - RED P.."!h, AND b Heat Smk

RED path.

1nabilo1yio Mcritor8

,N$1fNf ln Pmgreu ssa 1 !

1 I 2 I J

I

.t I I

I I

Loos of mo&I (appro!Cimaiely >75%) or 1'18 3nntJnCi..i1or1 o&&OC~.,tod wilh~fety11~ten11*

MaWi Conlrot Bo;udsA, 5-1(2) C*1(2l 0*1(2} E*l(2) Mt2)

G-1(2) NISRaclllll, ll Ill IV arodERCSAJorois AND A SIGNJFICANTTRANSlENT in progre&a, AND Cornpens31Dfy non-31.,rming lndicollore are unavoioble AND lrdcallons needed to monitor tt* ability lo shi.A ~It. reactor ma.l'llaln the 0Clf9 cooled makibln the reaclor ooolant 1ylll;:;m inlaci, and ni:ilnl.'.lin corLilnment ili:M:t are lNVailoble SA5.1 I 1 I 2 I 3 I 4 I I

I I

AC power Ct1jl3bility10 53fegoa1d8 Buses 15 and 16 (25 nnd 26) tl!Juced to orly one of lhe follovo(ng &OllCes for GREATER THAN 15mlra4B&'

Tra115forme1 CT-ti

T111'16IOrfllt!tCT-t 2

T1anslonner l RY

T11rw;lo1me1 2RV

Diesel Generator DI (05),

!Msel Gene1ato1 02 t06).

AND A"'f :tdcibONll 1lngle l:ill.ft ~D resull In stalion bbcll.out Fahl'* of Reactor ProtactlOn S)'3(em IMll'l.fllenlo\\ion to Complete<< Initiate on AutonmtlcReaaor Tnp Once a Reactor ProtecUon Syslern Seq>oint Hat Befln Excecdtld

f'd~Trl>Waa~llflJ.

I sA.2.11 1 I 2 I 3 I I

I I

I NOTE: A failed ITIOf\\tJlll lrip fc:tONed by 3 IUCC&lillf\\i mal'Ulll trip reducing reac10r po'Mll' 10 IHI lli.""1 5% meets this EAL lndica!IOO(.s)exlst th:~I o Reactor Prot&c.tlonSyatern setpointw.u A ND RPS :wlOnl('lllC trip dd Jl2l reWce poWGf' ta LESS THAN 5%*

AND Any of tho following operotOf acllOM nr* succesdul in reducing po'/41 to LESS THAN 5%, M3nu.al ~'1ol 6oarU

Rnctor Tnp

A~SACIOSS A.c!Uilticm

Tlrll!neTnp UNPLANt~ED LO&I Of Motil or All Safety S)'Slem AfYUIOOllon

<< lrdallon In Coraol Room Wllh Eitt* (1) :1 SGNIFICANT TRANSIENT 'In Progresa, or (2) Compensatory Non-Al3fming l~la~

11 UMv:tilabla.

I SA4.1 I 1 I 2 I 3 I 4 I I

I I

UNPLANNED los!I d most (appro:drnn\\ftty >75%) or oll amunci.Jtor&

or indicator& nsso<:iated with s.ifety 1yaton1r; for GREATER THAN 15minutes:

Main Conlrol Bo11rds.A, ll-1(2). C-1(:?/ D-1(2). E-1(2), F-1(2)

G*1(2) Nl$Rackst U, 111. IV anclERCSAlarms AND E~l18rtltlW)following :

a A SIGNIFICANT TRANSIENT in progress*

OR

b. Coniperulory rion-313rmlng ~liol'llll Me Uf'l3'1ailable Tab4o C*1 On.lta CommunlcaUona Syctoms PCIW9fedPhones

- Plnr(PagingSystem

- PlUnt Telephone Network

- Plan! RadioS)"&teni I SU1.1 I 1 I 2 I 3 I 4

I I

LoJa of POWOr lo or rrom Tronsfcrm<<1 CT-11, CT-12 1RY, &r.-J 2RY lhalresutts irt a lols of a*oflde po.vet lo bolh Safeguard& Bu1M 15 and 16 (25 3nd lel for GREATER THAN 15 mlnules AND Two Oleoel Genetalora (01 02 05 06) *e ~pPlylng ~r lo sareguards BU&fls 15 and 16 (25 ond 2e1 Nono lri<al Plant 11 not tniughl to r~ed oPiiffilin!j-mo.:M Mt!-tnTOChrtcal Speoticcuion& LCO Action Stntenient Time SU3 UNPLANNED LoUOf Mt>>1 or All Safety Syil9n ~

or lnclcalon n lh* eor.ol ROOIT\\ kw ar..it* ThGn 15 mhiwa.

I SUJ.1 !

1 I 2

I 3

I 4

I I

UNPLANNED loll of moot (npprolCirnacely ~75%) or al 31YlJnclators or Jndcatcn a11odotfkl with t.al'aty 1)'5lems for GREATER THAN 15nlioutes*

Main Con1rol Board A. 8-1(2), C-1(2). 0-1(2) E-1(2), F-1(2) G-1(21 NIS R3Cll.& 1 II IU. IV. ard ERCS Aliwm1 sue UNPLANNED LON o( Al Onlil* Of orr.i. Commun.eatior.

SU&.1 I 1 I 2 I 3

I 4 I I

I I

LossolaA TableC*1 <N11ecommtricaliorscap.."lbittyan.cung the abilitytoporformrO\\.tlneoperall(llW.

sue.21 1 I 2 I l

I.t I I

I I

Losis of oil Talle C*2 olf!Mte commlricatlnns caoubl~..

Tablo C-2 Offalta CommunlC111iona Syst11m

- Pinnt Te!ephono NatWOrll.

- F't."Jnt Radio Systen1 (dedk:ated offsite channels)

ENSNutwor~

HOT Lou of Pow or RPS F:lllure lnoblU1y10 Roi.ch Of Malolaln Shutdown Condllk>ns Intl.I Comm.

System MaHuncl Page 4 of 8

Prairie Island Nuclear Generating Plant Sys tom Mam.met.

IEYC1nta FuolCllld Dogt*datlon RCS Lukago lnad'fllN'tOnt Crhle.Hty c...

Confine

?1NGP 1576, Rev 6 Doc. Type/Sub Type: EP/EVT Reten!ion: Lifetime.,.

Nono None None Nono None Nooe

"~

\\

EMERGENCY ACTION LEVEL MATRIX N,..

None None

-i~1o c-1 O~itrc0mmunica~s'Syatcms ';J

- Sound Powered Phones

- Plant Paging System Pl.::lr~ T<!lr*phooo Nct\\*J[rl\\

Nono I SU4.1 I 1 I 2 I 3 I

I I

I I

Radi~tlon Mooltor 1(2lR*9 GREATER THAN 1 2 RJll' nlic11~ng fuel cl3c:l dolgradnlkln I SU4.2 f 1 I 2 I 3 I

I I

I I

Coolool aample nciivily GREATER THAN Tedric.a! SpeciflCOUon 3 4.17 Condition C nilowtible Um!tr; ntbllng f\\Jel dad degradation.

~itge An UNPLANNED IUltlJinltd pollitrV* 5tartup rme ob&efved on nuc:learlnBlrU'llflnl31ion Tab!* C-2 OffsJt* Communlgtfons Sy&ttm

- Plant f eleptioneNtitv.or' - ----

Pl:inl Rado Systern (dedicated ofl'slte channelg}

ENS Nal'M:lr' Nal\\lal ptvinmiero evel'(S al'l'ICling a loaded eaak CONFINEMENT aour~OARY 31 irdicaled by VISIBLE DAMAGE lo the tall'

earthql.i!'"

tornado (* rod tom1do mlulle)

flood

llgltninp lil'OW f lCI!

Accident cord11lorl1 Al'fecllng a ID3dod c:a!lk CONFINEMENT BOUNDARY OS indlc.111ed by VISIBLE DAMAGE r.o the caak*

dropped ~ak

tippedov*rCHk

~Wial

Uplo~ri Any concitlon in lhe opinion of the En1erg;mcy Dt-ector lt\\al lfl<ilcn!H loss ofloadocl fuol r;.torage cask CONFINEMENT BOUNDARY

/\\

HOT Fuotctad Oograd41don RCS Leak.go lnadYcftent Critk.Allty Sy*i.om MaHunct.

MODE-NA Caa*

ConliM

"°""""'

15'Sl Page 5 of 8

Prairie Island Nuclear Generating Plant EMERGENCY ACTION LEVEL MATRIX Losa or Polenlinl loss or ANY two Bwrier1 (Tilble F* 1).

le r -

D ER ET NOTE Del or mine wllth combinntlon or the 1hree b:wriora nre losl or h:wo a po1ential lo8s an<J 1.168 Ute fol~ng key to classify the evont Aloo an event for mulllple event11 could occur which result m the cooclusion that oir.ceeding U118 Lon OI' Polential Losa ltnsholds ls lrnminent (1.e wilhin 1 lo 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) In this immlneot loss 1ituat!oo US& judgmerl and classify as ii the thresholds are e11ceeded Fuel Cladding Barrier RCS Barrier Containment Barrier Fission Product Barrkif1

" !*!GP 1:J"rf. ~<r*';. 8 Der-. T y~IS uh Type: E?;E:VT Retention: Lihi time +

0 Loss I Cri!Jca!SnfetyfunctlonStntUS Core-Cooling Red 2 primary Coohut Attiyjty Lev!!

Cool;'!nl Acllvity GREATER THAN 300 µC~'gm 1-131 eq1M11alent 3 Core ExM Tharn10('!0!Jp!e Raai!na<1 GREATER THAN 1200 Uegroe F R11m;torV6S<111f W1!1.orlttvOI NorAppllcable 5 ConL'linm*!!l! R.odpljon

~

Cont11lnmon1 rod monitor 1(2)R.A8 or.&9 readng GREATER TliAN 200 Rll*

6~

Not Appicable 7 Emergency [)nclOr JyclgrMn!

My cordlion in the opirion ol U1e Eniorgency Olrecior 11131 in<kates Losa or Che Fuitl

°""""""'

0 Potential Loss 1

CrilicaJS<!fe lyfuncljonS1mu~

Core Cooling-Orange, OR Hent S1t*-Red 2 P!'lmarvCoolant.4.ctiv1tyleV!I No1Applk:..'1b!G J, Core Eli! Thermncourfo RM<fnqg OREA TER Tl-IAN 700 dogree F.

0 4

Rfl.~tnr V!!:S!iP.l W"!8f l ftval Leval LESS THAN*

-.io-.t. RVUS F~ Ra1"1941 (no RCPs),

- 32% RVUS DyMmic M83tl Range 11 RCP):

- 62% RVUS Oynarnit MeM Rango (2RCP&)

5. Cq!!airvnen! Radi:ition Morj!Of!!"Q Nol.Applicable IDe~

NolAppfic."lbfe 0

7. Emorooncy Dirnctor Judgmo!!I My conc.lilion in lho opiroJon ol the Emorg4tncy0..ect0flhtllindk:atas Potential Locs oflhe Fuel Clad Barner 0 Lo.ss

1. Critical SMety Func!lon Sta!IJ!l NolApplicabkl 2~

GREATER TliAN av<Mable makeup eapacMy asinlica1edby111oooof RCS &ib:ooling LESS THAN OR EOU.4.L T020 f35\\' dogrooF.

'M*91'HCOtll*"'rwo!c:ondlOOl'l*"*d*rinW<l H*QOOlllCln!tnlP<*M.. *QRa:<<lhlM't5;151')

01c:ol'll*l'W"Mtnl,....,:.orik,..,.gt1!a:*r11Wr1 1E<IR/Hr 3~

SGTR U\\Dl ref.vii& Jn an ECCS !SI)

Actuation 4

C9ntnim1A01 R/Mfi:ilio!'! Monitorlrn Cortaimienl rad morilOf' 1(2)R-18 or 49 rMding GREATER THAN 1 R.lhr 5~

Not Applicable 6

Emeraency Prector Juqgn1enl Any c~itlon In the opinion o( lhe Emergency OireclO' that lndicrites Loss d the RCS 8.-mief 0 Potential Loss I

Cril!f'.alSufetvEuncUOl'IS!!!lli'J RCS tnlegrity*R1t1;I OR Heal Srnk.-Red 2~

u,.:so1able leo ~ exceeding 60gpm 3 ~

NotAppl"tcnble 4

CorlalMlent R"'1!!atlon Montoriro NotApplk:able 5~

r~or Applicable

6. Emmgency Qjr&c!q Judgmen!

Arrt oondltk>nintllooplnlonofthe Emergency Director that indca!es Polenlial lo&&d 11'8 RCS D Loss 1

Crit@I Snfetv function St11h1&

No1Applit.'1ble 2

Contair.116!'1! Pfe&111e R:tj'lld unexpl.11ned decroose folkrwing inluo!

inCl'e.asa.

OR Containment preH u*e or s1*np 18'.lel response notoonsistent withLOCAc:ondibol'W>

3 Core E~it Thermocouple Ra..'\\l'fim~

NotApplicoble

4. SG SecroJart 5'!:1!! R!!:kl&.<:e Yo!Ch P-!Q-S Le~!'I!!!

RUPTURED SIG ls alto FAULTED O!Aside of tortaimlent OR Ptimat)'-tl).Secor.Ulfy leak rate GREATER THAN 10 gpm with norisoloble st1.nm reloase from affected SIG to the onvironnlent.

5 CNMT 1'>0lation y;ilyes St:itlfi Alt!?!' Cl-MT t'jC!latlon Containment isolatlonVal\\19(s) r'Ol c:loled.

ANO Drecl p:ttl~y lo the enWorY"OAM! exists

iftt!f Conlalrvnont lsobilon 1i9nnl 6

SigrWfqnt Raclio.1ctl'1t lnven1ory In Con1ninmefll NotApplicabl1 7 ~

Not Applicable 8

Emi:rqonr:y Plrtctor Judgm!f'I\\

Any oordtion In ti* opinion of the Em*gency DitectOf' !hat JrwJ~tes Loss of the Conttirme111 BMW 0

Potential Loss I

Crlbc:4 Safoty Func1!orl StnM Containn1&nt*Red 2

Ccntajnmeni Pres1nn 46 PSIG am 1nr.re.-.lling OR Con\\uirment hyJrogen oonc;entmtlot\\ MEAT ER TIIAN OR EQUAL TO 6%

OR Conta+nmert prs&5jll GREATER THAN 23 psig wnh LESS THAN one llAI train cl dep'essaautkln eq\\jpmeni opemtmg 3

Cor&Eli!Themiocou[!lnRel\\tfnao; Core ell.it lhefmoc0\\4)1M lnexceHof 1200 degr86& F and restoration procedures rot effecli"* wiltlin ISminU:e1 OR Core exit 1h11rmoc~H in excess of 700 degrees F wilh re!IClor Yit11t1el l1Yal below40% RVLIS Fl..41 Range o!ldrestorationprocetk#Hnotonectivewithln 15minl(e.s

-1 SG Se90!!i'1tyS!de Relll:i1t1 with P-to. S L'"all!19e Nol-5 CNMT IF.C'l!a!IC'lt'I V@t'I SJ1tlU'l AflM CNMT l:fS!btion NolAppic:lble 6

Sig?flcant R:x1i99dry1 !n;'f!ntory to Conl!jrvntf1!

ConlaiMJOnt rnd "'onitor 1(2)R48 or 49 r&adllig GREATERWAN800RIY 7~

Nol Applicoble 8

Emugnncv Pints;!Qf !Yt!gmM!

Nry condlion in uie opinion d the Em<<"V*ncy Oree.tor tl\\al incftealH Potential loss d the Containment Barner HOT Fission Product Page 6 of B

Prairie Island Nuclear Generating Plant Cold SDI Rolu.I Syst11m Malfunct.

p :~j<..;JP 15711 "ffi. 8 Loss of Pow ctr RH ct or Vessol Lovef D ;:..~; TypWS:ib 'i' Y',>e.: F.:PiEVT Retention: Lifetime+

CG l LOH Cl( RP\\/ lrtYenttry A.lfecting Fuel Clid lntogiiy with COnlalmlenl ~\\Qfleog6d v.ith In~

Fuel In lhe RPV.

cG11 I I

I I

I s I s I I

1 Loss of RPV inven&ory as lndica1ed by une*plllined level increase in Cof'll:tlnmert Suo1pgA or C. a-Westo Hoklup Tt1nk as indic.."tod by s1.wnp pump tun 11n1as, levels, or alarms AND 2 RPV level*

11. LESS THAN 63% RVUS Fill! RBnge for GREATER THAN 30 n1inutes; DR b cal'V'IOI be momored, wilt1 inckation or co.a uncovery for GREAT ER THAN 30mimAea na evJdenced byono or more of the following

Co111ailmer4 Vni;elAreai Mondor A-2 reading GAEA TEA THAN 1000 mR/hr

E111lic Soll'ce Range Monito1 Indication AND 3 Indication of CONTAINMENT ch.llletlged os indicated by one or m<Fe oflhelollowing*

Corit<1irvnent hyclrogen eoncenlrt1Tion GREATER THAN OR EQUAL T06%

TS B.3.9.4 CONTAINMENT CLOSURE !!!1 *1111t!Jlshecl

Conla1nrnent pleriillle GREATER THAN 1 0 Ptig 'liith TS B 39..i CONTAINMENT CLOSURE estnlA1tied EMERGENCY ACTION LEVEL MATRIX CS 1 LOii ol RPV lnv.nlory Affecbng Core Dec.'ty Heat Removal Capability CS1.1 I I

I I

I """5--r:=r::-1 With CONT AJNMENT CLOSURE QS!1 eslabllshed:

o RPV invenwy us Jn6caled by RPV level LESS THAN 13% RVUS Full Range OR

b. RP\\/ level c:."lmat be monitored f1,7 GREATEFt THAN 30 nl!WtM l'Alh a lnls of RPV irwenlory as irdtwled tiy unexplarnod Imel Increase 11'1 ConluinrnentSump1 Aor C or WasteHnldq> Took as lndiealed by !IUlllp PU!llP run lin1es levels. or alarm s.

cs1.2 (

I I

I I s I

"(

I With CONTAINMENT CLOSURE estabUsh&d" 0 RPV lrr1ernory OS ifdcated by RP\\/ level LESS THAN 63".4 RVUS F~ Range OR b RP\\/ level C:.'UVlOI be morutored fer GREATER THAN 30 n1lntJIM >Miil a lo&a of RPV inventory.u irdcatctd by either:

l/nel(plalned level Increase in Conlainmerc Sumps A or C. or Waate Holdup Tank as indicated by !1Ufl1p pump ru"\\ t1mea levels,ort'!!(Yms

E1ra1lc SoUlce Range Monl\\or lndicatton CS2 Los. d RPV Inventory AIJ<<q Cof* 09c:ay He.at R~al Cap:lbillty Wllh lrra<.bted Fuel inlhl RPV NOTE* CS2 1 and CS2 2 shoi>ld !!II.I bit uWd fO' d11uifttation u*lli* RPV i.vel Is ~low Ille bOtlnn Ins id* d111rwt<< (lOJ of Ill* RCS l'lol i.g pe°":nmon lfi.wtt ii 11tor 1MIOV1t1t1"19onon C. CUlorCA2 stiould bot uJed for FJfn! t laS11iftc:Micm In th* R*lue~ng nod*

CS21 r-----. - r--T-1 I *. r I

Wllh CONTAINMENT CLOSURE !l2l e;tabliahed, nnd RPV leval camot be nionilored. with inlie:ition of eore urw:overy as ellidel'109d by 01"18 or more of tho following*

Containmtnl V.ssel Area Morito1 R-2 111d1ng GREATER THAN 1000 mR/tY

Eriatk; So...-u Range Monitor lndcahon cs2.2 I I

I I

& I I

Wilh 001-ITAJNMENT CLOSURE e&tabllshed, and RPV level cannoC be moAlored with indication cf core 1.1'\\COYllfY as evkklriced by ooe or 01ore or the follOWing*

Cot'Uiomenl Vessel Area W.Onltor R-2 reading GREATERTltAN 1000mRhit

Erratic So\\J"ce Rillngl' Monitor lndit*tton LOH or power IOOf From Tr1molorn1ers CT*11 CT-12 \\RY, and 2RY U1.i! r05ults in 3 loss of 31 off site po\\YOr to bolh Safeguards Buses 15 and 16(25and26):

AND Faihn or Di0sol Gonern1er1 DI and 02 (05 Ord 06) lo supply pow8f to Snreguards 9t.£es 15 ond 16 (25 and 26)

AND F!lilLH to ros11,7e power to Sa!eguards Bus 15 or 16 (25or 261 Yottt~n 1Smlrutos frcrn the time oflofis of bolh offsit* and onslte AC poww CA1 losaofRCS\\nventory CAI I I 1 -1 I

l 5 I I

l LOH of RCS lrwOOlory ;t;i lodicnled by RP\\/ Jovel nl 0 inches Refueling Cal\\Ut I RCS t~lJITOW Range I Ullrasooic (OI Of LESS THAN 75*Ai RVLIS Fua Range).

CA1.2 I I

I I

I 5 I I

I Loss of RCS lnveniory ns lrdetlted by uneiplained level ini:rease inCont.ainmenl Si.mpsAorC, 01Waste Holdup T;arV: as indic.i:r!ed by ; ump purnp nn limH levels, or al."Wm5, AND RCS level e.1nnot be monitored for GREATER TMAN 15 minutes CA2 Lon of RPV lnvenlay wiih Irradiated FUii in 1118 RPV.

CA2 I I J -1 I

J._... !

I I

Loss of RPV ttwentory 115 indicated by RPV lev&l al 0 inches RehJel ng Canal f RCS Nooow Rnnge f Ullrnsonk:

cA2.2 I I

I I

s I I

loss of RCS inventory as indcaled by unexplained level incraase In Contoinmenl Sumps A er C or Wasle Moldup TaM 11s indicated by sump pimp ri...i t!mes leveta er !Jl~ms AND RPVtevef~ be rrioritored rer GREATER THAt~ 15 minules Los.s ofpoMii--10 C11 ffom Transrern1er1 CT.fl CT-12 1RY, ari(f2RY that r4tsUla in a k>M ot' al ottslte power to both Safeguards Buses 15 and 16 (25rmtl26)for GREATER THAN 15n1itlules AND Al leas! one DleHI Generator (01 02 D5 06) Is &UpPlying power to ooe of the affecled safagu.Tdi> buses CU7 UNPLANNED Lon of Recped DC Power fOf GREATER THAN 15""1ulee f11lkse to re&lor* PO'IWf to at kliMt OM reqlM"ed DC p:inel W!hn 15 niin.it1111 lrom th9 lme of loss CU2 UNPLANNEOLossofRCSffweilioy~hl~ FLM!in

"'RPV CU>1 c::r-*-Tu::::r-. l I

r I

UNPLANNED RCS level decrease below the RPVflnnge for OR EATER THAN OR EOUAL TO 15 minut.u; CU2.2 I I

I I

6 I I

Loss of RPV ln...ntory as lrdcated by lr!OXplained level 1nc:re.3H in Cor(alM1ent St.rn1>5 A or C, orw1111e llol00p Tank as indcoled bysumppumpf1.Wlllmes tevM oralarme.

ANO APV level carnal be rnMtor&t.l COLD LOH of Pow*

Reactor veuol LCYet CoklSDI Refuot Sy*lom M*lfunct.

Page 7 of 8

Prairie Island Nuclear Generating Plant Cold SDI Ratuol Sy11em Malfunct.

RCS Tomp.

c~.

FuolCl3d Oogradatlon RCS L.e*Us*

lnadvortont Ctltle11llty PlflGf' 15TI\\_ R~.v _ 3 Ox. Typf;/$ub i°W": EP/D:r Retention: Lifetime +

NoM NOM Not,.

None None EMERGENCY ACTION LEVEL MATRIX NOTES 1f an RCS h"1 rtnOv*I ~!er1 is In OPl!l'llllOn 11>11nt1 lfll.;; tr..

ftarw *nd RCS 1-npowt'lllUr* ~ bemg._dU<:~l lhH trus EAL It t\\Ol.iopllilc9bl*

'f1tw1ProHS..u.r*$Dldl~nonlylhot RCSw"'9!'11turw 1tu~1ncild is *ol!6atM 1o CM 3 cM2 I l

-T I

I s I 6 I I

With CONT A!NMENT CLOSURE estoblishOO mf RCS il'llegrily ~

esl~blishetJ.!!: RCS lnv1111lotY reduced an UNPLANNED e~ent res ills In RCS !efllpernt1.1e excootJlng 2oo*F for GREATER THAN 20mlnu!es',

CA4l i~~,~~,~~,~~,~

, ~,-.~,~~j An UNPLANNED eveol r115Ulls. in RCS temperature v.ceaJing 200'F for GREATER THAN 60 minutes ' or rll!lilts in an RCS preasu-e

~

NOM

""~

T~blo c.1 -'onstto C~unl~~* Sysk~?:ii:'.'

- Sour1d f-'0werorl PiVJMS

- Pla:-il P::igi1'19Systcm

- Plant T alephone Network Pl~nl Radi0Sysle1*n GREAlER UNPLANNED LOl9s ol All OO&ilo or Off11ile C(iiT!mtn00lion&

Ctapabili!IH cu&. 1

.-------.---~----I s I 11 I

I Loss of all Tajje C-1 onslte c0tnml.ricallo"6 capabllltyaffecting the al.illy ro perfOfrrt roulne Clpe'lllions cuu !

I I

I I s I 6 I I

Loas of all Tnblt! C-2 oftslle communication& capabiily 1rad.."liloft.

RCS letdown Rad MOrii(i i(.2)R:9 or port."!Ae radalion moritorlrg lnalru'tleoblion GREA TEA THAN 1 2 R.1Y iro:icating fl.el cl.ad degraclahon.

I CUS.2 r-1 --,,.---.,---.,--,-1-::-, --.-1 -,:--,....--.,

Coolant sample octl'lityGREATER THAN Technical Specification 3 4 17 ConJitloo C ollCM1able Umlts lnrllcat!ng fl.Ill! clatJ degmc.totion.

Ut RCS Leakage.

AA UNPLANNED 1~:m.d po5itive SW~ rmte DbteNed on l'luciearlns t~ienml!Ol'I

-<!'at>>o c -2::,>.0ffsl1e COf:T!munlcation~ Sys!~

Pia"1!f;)ki:i)hOf-.il-f:fP.1\\iiif1<.

Pl~:-il ROOio System (decHcntod of!site cl ;anrltllS)

ENSNetworl<.

COLD RCS r-~

Comm.

Flllrictad Degradation RCS lnadvortent CrtUe11Hty ColdSOI Rofu!M System Motfuncl.

Page 8 of 8

(

3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation RTS Instrumentation 3.3.1 LCO 3.3.1

The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

APPUCABILITY:

According to Table 3.3.1-1.

ACTIONS

NOTE--------------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION A.

One or more Functions with one or more required channels or trains inoperable.

B.

One Manual Reactor Trip channel inoperable.

Prairie Island Units 1and2 A.I B.1 OR B.2 REQUIRED ACTION COMPLETION TIME Enter the Condition Immediately referenced in Table 3.3.1-1 for the channel(s) or train(s).

Restore channel to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months OPERABLE status.

BeinMODE3.

54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months 3.3.1-1 Unit 1 -Amendment No. 158 Unit 2-Amendment No. 149

ACTIONS (continued)

CONDITION C.

One channel or train inoperable.

D.

One Power Range Neutron Flux channel inoperable.

Prairie Island 1and2 REQUIRED ACTION C.l Restore channel or train to OPERABLE status.

OR C.2.1 Initiate action to fully insert all rods.

AND C.2.2 Place the Rod Control System in a condition incapable of rod withdrawal.

NO'f:E---------------

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing and setpoint adjustment of other channels.

D.1.1 Place channel in trip.

AND R TS Instrumentation 3.3.1 COMPLETION TIME 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 48 hours 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months 6 hours 3.3J 1 --- Amendment No. 158 nit 2 - Amendment No. l

ACTIONS CONDITION D. (continued)

E. One channel inoperable.

Prairie Island Units 1and2 REQUIRED ACTION D.1.2 ------------NOTE------------

Only required to be performed when THERi\\1AL POWER is

> 85% RTP and the Power Range Neutron Flux input to QPTR is inoperable.

Perform SR 3.2.4.2.

OR D.2 Be in MODE 3.

N ()rfE---------------

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

E.1 Place channel in trip.

OR E.2 Be in MODE 3.

R TS Instrumentation 3.3.l COMPLETfON TIME Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 12 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours 3.3.l Unit 1 Amendment No. 158 Unit 2-Amendment No. 149

ACTIONS (continued)

CONDITION F.

One Intermediate Range Neutron Flux channel inoperable.

G.

Two Intermediate Range Neutron Flux channels inoperable.

Prairie Island Units 1and2 F.1 OR F.2 G.1 AND G.2 REQUIRED ACTION Reduce THERMAL POWER to < P-6.

Increase THERMAL POWER to> P-10.

NOTE------------

Limited plant cooldown or boron dilution is allowed provided the change is accounted for in the calculated SDM.

Suspend operations involving positive reactivity additions.

Reduce THERMAL POWER to < P-6.

R TS Instrumentation 3.3.1 COMPLETION TIME 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hours Immediately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 3.3.1-4 Unit 1 -Amendment No. 158 Unit 2-Amendment No. 149

ACTIONS (continued)

CONDITION H.

One Source Range Neutron Flux channel inoperable.

I.

Two Source Range Neutron Flux channels inoperable.

J.

One Source Range Neutron Flux channel inoperable.

Prairie Island Units l and 2 REQUIRED ACTION H.l

N 01' E------------

Limited plant cooldown or boron dilution is allowed provided the change is accounted for in the calculated SDM.

Suspend operations involving positive reactivity additions.

I.1 Open Reactor Trip Breakers (RTBs).

J.l Restore channel to OPERABLE status.

OR J.2.l Initiate action to fully insert all rods.

RTS Instrumentation 3.3.1 COMPLETION TIME Immediately Immediately 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 48 hours AND l2.2 Place the Rod Control 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months System in a condition incapable of rod withdrawal.

3,3.1 Unit l **

Amendment No. 158 Unit 2 - Amendment No, 149

ACTIONS (continued)

CONDITION K.

One channel inoperable.

L.

One or both channel(s) inoperable on one bus.

Prairie Island Units 1 and 2 REQUIRED ACTION

NOTE----------------

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

K.1 Place channel in trip.

OR K.2 Reduce THERMAL POWER to < P-7 and P-8.

N.OTE----------------

One inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels.

L. l Place channel(s) in trip.

OR L.2 Reduce THERMAL POWER to < P-7 and P-8.

RTS Instrumentation 3.3.1 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours

. l-6

- - Amendment No. 1 Unit 2 - Amendment No. 149

ACTIONS (continued)

CONDITION M. One Reactor Coolant Pump Breaker Open chaimel inoperable.

N.

One Turbine Trip channel inoperable Prairie Island Units 1 2

REQUIRED ACTION M.l Restore channel to OPERA.BLE status.

OR M.2 Reduce THERMAL PO\\VER to < P-7 and P-8.

1\\JOTE----------------

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channel(s ).

N.1 Place channel in trip.

OR N.2 Reduce THEfilv1AL POWER to< P-9.

R TS Instrumentation 3.3.1 COMPLETION TIME 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 54 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours

'1-7 Unit l -- Amendment No. 158 2 - Amendment No. 1

ACTIONS (continued)

CONDITION

0.

One train inoperable.

P.

One RTB train inoperable.

Prairie Island Units 1and2 REQUIRED ACTION

NOTE----------------

Onc train may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing provided the other train is OPERABLE.

0.1 Restore train to OPERABLE status.

0.2 Be in MODE 3.

NOTES---------------

1.

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing, provided the other train is OPERABLE.

2.

One RTB may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE.

P. l Restore train to OPERABLE status.

RTS Instrumentation 3.3.l COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

' 1 Unit 1 - Amendment No. 158 Unit 2 - Amendment No. 149

ACTIONS (continued)

CONDITION Q.

One or more channels inoperable.

R.

One or more channels inoperable.

S.

One trip mechanism inoperable for one RTB.

Prairie Island Units 1 and 2 Q.l OR Q.2 R.1 OR R.2 S. l OR S.2 REQUIRED ACTION Verify interlock is in required state for existing unit conditions.

Be in MODE 3.

Verify interlock is in required state for existing unit conditions.

Be in MODE 2.

Restore inoperable trip mechanism to OPERABLE status.

Be in MODE 3.

RTS Instrumentation 3.3.1 COMPLETION TIME 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 7 hours 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 7 hours 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 54 hours Unit 1 /\\mendment No. 158 tJnit 2 -Amendment No. 149

(

SURVEILLANCE REQUIREMENTS R TS Instrumentation 3.3.1

N()TE--------------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 ---------------------------N ()TES----------------------------

Prairie Island Units 1and2

1.

Adjust NIS channel if absolute difference is

>2%.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL P()WER is:::: 15% RTP.

Compare results of calorimetric heat balance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months calculation to Nuclear Instrumentation System (NIS) channel output.

3.3.1-10 Unit I -Amendment No. 158 Unit 2 -Amendment No. 149

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.3.1.3

NOTES---------------------------

1.

Adjust NIS channel if absolute difference is

2%.

2.

Not required to be perfonned until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after THERMAL POWER is 2.: 15%) RTP.

Compare results of the core power distribution measurements to NIS AFD.

SR 3.3.1.4

NOTE---------------------------

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform T ADOT.

SR 3.3.1.5 Perform ACTUATION LOGIC TEST.

R TS Instrumentation 3.3. l FREQUENCY 31 effective full power days (EFPD) 31 days on a STAGGERED TEST BASIS 31 days on a STAGGERED TEST BASIS Prairie Island Units 1and2

.3.1-11 Unit 1 Amendment No.~* 201 I Unit 2 - Amendment No. +49 188 I

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.3.1.6 ----------------------------N()TE----------------------------

N ot required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL P()WER is 2: 75% RTP.

Calibrate excore channels to agree with core power distribution measurements.

SR 3. 3.1. 7

N ()TE----------------------------

N ot required to be performed for source range instrumentation prior to entering M()DE 3 from M()DE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into M()DE 3.

Perform C()T.

RTS Instrumentation 3.3.1 FREQUENCY 92 EFPD 92 days Prairie Island Units 1and2 3.3.1-12 Unit 1 - Amendment No.~ 201 j Unit 2 - Amendment No. -149 188

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3. 3.1. 8

NOTES--------------------------

NOTE------

Only required when not performed within previous 92 days Prairie Island Units 1and2

1. This Surveillance shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

2. Not required to be performed for intermediate and source range instrumentation prior to reactor startup following shutdown _:::: 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Perform COT.

3.3.1-13 Prior to reactor startup Twelve hours after reducing power below P-10 for power and intermediate range instrumentation Four hours after reducing power below P-6 for source range instrumentation Every 92 days thereafter Unit I -Amendment No. 158 Unit 2-Amendment No. 149

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.3.1. 9

NOTE----------------------------

V crification of sctpoint is not required.

Perfonn TADOT.

SR 3.3. 1.10 ---------------------------NOTE-----------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.1.11 --------------------------NOTE------------------------------

N eutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALTBRA TION.

R TS Instrumentation 3.3.l FREQUENCY 92 days 24 months 24 months Prairie Island Units 1and2 3.3.l Unit 1 - Amendment No. 1 Unit 2 - Amendment No. 149

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.3.1.12 ----------------------------NOTE----------------------------

This Surveillance shall include verification of Reactor Coolant System resistance temperature detector bypass loop flow rate.

Perform CHANNEL CALIBRATION.

SR 3.3.1.13 Perform COT.

SR 3.3.1.14 ----------------------------NOTE----------------------------

V erification of setpoint is not required.

Perform TADOT.

SR 3.3.1.15 ----------------------------NOTE----------------------------

Verification of setpoint is not required.

Perform TADOT.

R TS Instrumentation 3.3.1 FREQUENCY 24 months 24 months 24 months Prior to exceeding the P-9 interlock whenever the unit has been in MODE 3, if not performed within the previous 31 days Prairie Island Units 1and2 3.3.1-15 Unit 1-AmendmentNo. 158 Unit 2 Amendment No. 149

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE SR 3.3.1.16 ----------------------------NOTE----------------------------

N eutron detectors are excluded from response time testing.

Verify R TS RESPONSE TIME is within limits.

RTS Instrumentation 3.3.1 FREQUENCY 24 months Prairie Island Units 1and2 3.3.1-16 Unit 1 -Amendment No. 158 Unit 2-Amendment No. 149

1.

2.

3.

4.

APPLICABLE MODES OR OTHER SPECIFIED FUNCTION CONDITIONS Manual Reactor I, 2 Trip 3(a), 4(a), 5(a)

Power Range Neutron Flux

a.

High 1, 2

b.

Low 1(b),2 Power Range Neutron Flux Rate

a. High Positive 1, 2 Rate b.High 1, 2 Negative Rate Intermediate l(b), 2(c)

Range Neutron Flux Table 3.3.1-1 (page 1 of 8)

Reactor Trip System Instrumentation REQUIRED CHANNELS CONDITIONS 2

B 2

c 4

D 4

D 2

F,G R TS Instrumentation 3.3.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE SR 3.3.1.14 NA SR 3.3.1.14 NA SR 3.3.1.1 S 110%RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.l.1 S40% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 S6%RTPwith SR 3.3.1.11 time constant SR 3.3.1.16

2 sec SR 3.3.1.7 S8%RTPwith SR 3.3.1.11 time constant SR 3.3.1.16

2 sec SR 3.3.1.1 S40%RTP SR 3.3.1.8 SR 3.3.1.11 (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(b)

Below the P-10 (Power Range Neutron Flux) interlocks.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

Prairie Island Units I and 2 3.3.1-17 Unit I -Amendment No. 158 Unit 2-Amendment No. 149

FUNCTION

5.

Source Range Neutron Flux

6. Overtemperature LiT

7. Overpower Li T

8.

Pressurizer Pressure

a.

Low

b.

High Table 3.3.1-1 (page 2 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED CONDITIONS CHANNELS CONDITIONS 2(d) 2 H,I 3(a), 4(a), 5(a) 2 I, J 1, 2 4

E I, 2 4

E 1 (e) 4 K

1, 2 3

E R TS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11 SR3.3.l.16 SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 ALLOWABLE VALUE

~ l.OE6 cps

~ l.OE6 cps Refer to Note 1 (Page 3.3.1-23)

Refer to Note 2 (Page 3.3.1-24)

1845 psig

~2400 psig (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

( e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

Prairie Island Units 1and2 3.3.1-18 Unit 1 -Amendment No. 162 Unit 2 -Amendment No. 153

Table 3.3. l-l (page 3 of8)

Reactor Trip System lnstrnmentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED FUNCTION CONDITIONS CHANNELS CONDITIONS

9.

Pressurizer Water l (c) 3 K

Level-High

10. Reactor Coolant I (t) 3 per loop K

Flow-Low

11. Loss of Reactor Coolant Pump (RCP)

a.

RCP Breaker I (f) l per RCP M

Open

b. Under-l (t) 2 per bus L

frequency 4 kV Buses 11 and 12 (21 and 22)

12. Undervoltage on l ( e) 2 per bus L

4 kV Buses 11 and 12 (21 and

22)

L2 3 per SG E

13. Steam Generator (SG)

Water Level -

Low Low (e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

RTS Instrumentation 3.3.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE SR 3.3.1.l

S90%

SR 3.3.1.7 SR 3.3. LI 0 SR 3.3.1.J

91%

SR 3.3. l.7 SR 3.3.1.10 SR 3.. i.1.14 NA SR 3.3.1.9

'.': 58.2 Hz SR 3.3.1.10 SR 3.3.1.9

76% rated SR 3.3.1.10 bus voltage SR 3.3.1.1 SR 3.3. J.7

'.': l 1.3'Yo SR3.3.l.IO (f)

Above the P-8 (Power Range Neutron Flux) or P-7 (Low Power Reactor Trips Block) interlocks.

Prairie Island Units land 2 3.3.1-19 Unit l -- Amendment No. 158 2 ****Amendment No. 149

Tabk3.3.i-l (pagc4of8)

Reactor Trip System Instrumentation R TS Instrumentation 3.3.1 SURVEILLANCE FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

14. Turbine Trip

a.

Low Autostop Oil Pressure

b.

Turbine Stop Valve Closure

15. Safety Injection (SI)

Input from Engineered Safety Feature Actuation System (ESFAS)

I (g) 3 1(g) 2 L2 2 trains (g)

Above the P-9 (Power Range Neutron Flux) interlock.

Prairie Island Units I and 2 3.3. l N

N 0

SR3.3.l.IO SR 3.3. l.!5 SR 3.3.1.10 SR 3.3.1.15 SR3.3.l.14 Closed NA Unit 1 Amendment No. 158 Unit 2 Amendment No. 149

FUNCTION

16. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6

b. Low Power Reactor Trips Block, P-7 I. Power Range Neutron Flux

2. Turbine Impulse Pressure

c. Power Range Neutron Flux, P-8

d. Power Range Neutron Flux, P-9

e. Power Range Neutron Flux, P-10 I 7. Reactor Trip Breakers(h) (RTBs)

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE 2(d) 2 Q

SR 3.3.1.11 2: I.OE-IO amp SR 3.3.1.13 4

R SR 3.3.1.11

12%RTP SR3.3.l.13 2

R SR 3.3.1.7

12% Full SR 3.3.1.10 Load 4

R SR 3.3.1.11

11%RTP SR 3.3.1.13 4

R SR 3.3.1.11

12%RTP SR 3.3.1.13 1,2 4

Q SR 3.3.1.1 I 2: 9% RTP SR3.3.l.13 I, 2 2 trains p

SR 3.3.1.4 NA 2 trains c

SR 3.3.1.4 NA (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(h)

Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

Prairie Island Units 1and2 3.3.1-21 Unit 1 -Amendment No. 158 Unit 2-Amendment No. 149

FUNCTION

18. Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms

19. Automatic Trip Logic Table 3.3.1-1(page6 of 8)

Reactor Trip System Instrumentation R TS Instrumentation 3.3.1 SURVEILLANCE APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE 1, 2 l, 2 1 each per RTB 1 each per RTB 2 trains 2 trains S

SR 3.3.1.4 NA c

SR 3.3.1.4 NA 0

SR 3.3.1.5 NA c

SR 3.3.1.5 NA (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Prairie Island Units I and 2 3.3.1-22 Unit 1 - Amendment No. 158 Unit 2 -Amendment No. 149

Table 3.3.1-1(page7of8)

Reactor Trip System Instrumentation Note 1: Overtemperature ~ T R TS Instrumentation 3.3.1 The Overtemperature ~T Function Allowable Value is defined by the following Trip Setpoint.

Where: ~Tis measured Reactor Coolant System (RCS) ~T, °F.

~To is the indicated ~Tat RTP, °F.

sis the Laplace transform operator, sec-1.

T is the measured RCS average temperature, °F.

T' is the nominal Tavg at RTP, = *°F.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, =

psig K1 S

K2 = *1°F K3 = */psig

't' 1 =*sec

't' 2 =*sec when qt - qb S *% RTP f1(~I) =*{*+(qt-qb)}

% ofRTP

{ (qt - qb) - *}

when*% RTP <qt - qb S *% RTP when qt - qb > *% RTP Where qt and qb are percent R TP in the upper and lower halves of the core, respectively, and qt+ qb is the total THERMAL POWER in percent RTP.

As specified in the COLR.

Prairie Island Units 1and2 3.3.1-23 Unit I -Amendment No. 162 Unit 2 -Amendment No. 153

R TS Instrumentation 3.3.1 Table 3.3.1-1(page8of8)

Reactor Trip System Instrumentation Note 2: Overpower ~T The Overpower ~ T Function Allowable Value is defined by the following Trip Setpoint.

Where:

~Tis measured RCS ~T, °F.

~To is the indicated ~Tat RTP, °F.

sis the Laplace transform operator, sec-1*

Tis the measured RCS average temperature, °F.

T' is the nominal Tavg at RTP, = *°F.

K5 = */°F for increasing Tavg

= * /°F for decreasing T avg K6 = */°F when T > T'

= */°F when T :S T'

't3 =*sec

As specified in the COLR.

Prairie Island Units 1and2 3.3.1-24 Unit 1 Amendment No. 162 Unit 2-Amendment No. 153

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 1 of20 WO: --------

RESULT$1COMMENTS: I Work Request Card or Work Order Initiated: YES NO No. ______ _

Test Performance:

Performed By: ______________ _

Date: _______ _

(Signature or Initials)

Additional Requirements:

NONE SP Completion:

l&C Supervisor:

Date: ------

SP Surveillance Schedule Satisfied. YES/NO Surv. Admin: ---------

Other Actions for Consideration:

System Engineer Review:------------

Date: --------

O.C. REVIEW DATE:

OWNER:

EFFECTIVE DATE

(

2/11/05 T. Wadley 2/11/05

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 2 of20 1.0 PURPOSE AND GENERAL DISCUSSION CONTINUOUS USE Continuous use of procedure required.

Read each step prior to performing.

Mark off steps as they are completed.

Procedure SHALL be at the work location.

1.1 Purpose 1.1.1 The purpose of this test is to demonstrate the operability of the NIS Power Range PB, P9, and P10 permissive functions, and the 25% High Flux Low Setpoint Reactor Trip function prior to reactor startup OR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10..!.E power is below 10% but above P-6, THEN perform SP 1198.2[2198.2] in place of this procedure.

1.1.2 This test fulfills Technical Specification surveillance requirements T.S. SR 3.1.8.1 and T.S. SR 3.3.1.8 for Table 3.3.1-1, Function 2b, and T.S. SR 3.3.1.13 for Table 3.3.1-1, Functions 16b1, 16c, 16d and 16e.

1.2 Acceptance Criteria General In the event Acceptance Criteria (designated with an*) cannot be met, refer to Ops Manual Section G "Surveillance and Periodic Test Program" for additional guidance.

2.0 REFERENCES

Tech Spec Table 3.3.1-1, Functions 2b, 16c, 16d and 16e.

3.0 PRECAUTIONS AND LIMITATIONS 3.1 This test may be performed on only one of the Nuclear Instrument System channels at a time.

3.2 All steps in this procedure are to be performed in sequence. The procedure must be completed for the individual channel prior to testing another channel.

3.3 Notify the Shift Supervisor whenever unresolvable problems are encountered OR Acceptance Values are not met. Criteria designated "ACCEPTABLE VALUES" are directly related to Technical Specifications. Refer to Appendix A for guidance when these are not met. "Desired Value" criteria are not directly related to Tech Specs unless the channel is definitely inoperable.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 (2198)

REV:

18 Page 3 of20 4.0 PERSONNEL AND SPECIAL EQUIPMENT REQUIREMENTS Manpower:

4.1 One ( 1 ) l&C Specialist 4.2 One (1) Control Room Operator 5.0 SPECIAL CONSIDERATIONS 5.1 Technical Specification SR 3.3.1.8 for T.S. Table 3.3.1-1, Function 2b, requires this test to be performed within 92 days prior to reactor startup OR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10.

5.2 This procedure affects Technical Specification related equipment. If "As Found" data is discovered out of tolerance, the following actions are required:

Notify an l&C Supervisor.

Notify the System Engineer.

Obtain Concurrent Verification prior to and during adjustments.

Document that Concurrent Verification was performed on the affected data sheet(s) in Appendix A.

Initiate a CAP to document the "As Found" out of tolerance condition(s).

If the instrument cannot be adjusted to within tolerance, notify the Shift Supervisor immediately.

6.0 PREREQUISITES AND INITIAL CONDITIONS 6.1 Test SHALL be performed in Mode 3, Hot Standby; Mode 4, Hot Shutdown; Mode 5, Cold Shutdown OR Mode 6, Refueling.

6.2 NIS drawers energized for at least 30 minutes.

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP 7.0 PROCEDURE NOTE:

NIS POWER RANGE STARTUP TEST Within the text of this procedure Ch refers to Channel No.

under test and ERCS refers to ERCS Address No.

7.1 Preparation 7.1.1 Obtain Shift Supervisors permission to perform test.

7.1.2 Verify the following annunciators are as specified.

Box 47014 [475141 A.

P-9 REACTOR TRIP BLOCKED 47014-0104 [47514-0104] "ON" B.

P-8 LO POWER LO FLOW TRIP BLOCKED 47014-0204 [47514-0204] "ON" C.

P-7 LO POWER TRIP BLOCKED 47014-0304 [47514-0304] "ON" D.

POWER RANGE LO SETTING TRIP BLOCKED 47014-0402 [47514-0402] "OFF" E.

P-10 NUCLEAR AT POWER PERMISSIVE 47014-0501 [47514-0501] "OFF" Box 47013 [475131 F.

NIS POWER RANGE LO SETPOINT CHANNEL ALERT 47013-0202 [47513-0202] "OFF" SP 1198 [2198)

REV:

18 Page 4 of 20 SS OPS OPS OPS OPS OPS OPS

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 5 of20 7.1.3 On NIS COMPARATOR AND RATE drawer, INITIALS/DATA place COMPARATOR CHANNEL DEFEAT N41 N42 N43 N44 switch to "DEFEAT" channel in test.

7.1.4 On PR A drawer, position METER RATE switch to "FAST" position.

7.1.5 On PR B drawer, in sequence, position:

A.

DETECTOR A TEST LEVEL and DETECTOR B TEST LEVEL potentiometers fully CCW.

B.

OPERATION SELECTOR switch to "DET A & B" position.

C.

METER RANGE/RATE SWITCHES to "400 µA FAST" or "4000 µA FAST", as necessary, to read current test signals.

7.1.6 Verify

A.

PR B drawer CHANNEL ON TEST lamp "ON".

OPS OPS OPS OPS B.

On the COMPARATOR AND RATE drawer COMPARATOR DEFEAT lamp "ON."

OPS OPS OPS OPS C.

Annunciator NUCLEAR INSTR SYSTEM CHANNEL TEST 47013-0601 [47513-0601]

"ON".

OPS OPS OPS OPS

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 6 of20 INITIALS/DATA N41 N42 N43 N44 7.2 Power Range Permissives 7.2.1 Raise Detector A and B test signals until POWER ABOVE PERMISSIVE PB lamp lights.

A.

Verify PB lamp LIT.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter.

C.

Check value from Step 7.2.1.B

~ 11.0%.

D.

lE value from Step 7.2.1.B is outside calibration tolerance band (B.9% to 9.9% ), THEN complete Appendix A, using Appendix B as reference, otherwise NA.

E.

Verify either the value from Step 7.2.1.B or "As Left" value using Appendix A is ACCEPTED B.9% to 9.9%.

F.

Trip status light 44178 [44510]

POWER RNG PB NC Ch N "ON."

-0102

-0202

-0302

-0402 G.

ERGS PWR RNG CHAN Ch N PB PERM. "CH SET."

F0495D F0496D F0497D F0498D

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 7.2.2 Lower detector A and B test signals until P8 lamp de-energizes.

A.

Verify P8 lamp de-energized.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter DESIRED 7.8% to 8.go/o C.

Trip status light 44178 [44510]

PWR RNG P8 NC Ch N "OFF."

D.

ERGS PWR RNG CHAN Ch N P8 PERM. "CH RESET."

7.2.3 Raise Detector A and B test signals approximately equal until POWER ABOVE PERMISSIVE pg lamp lights.

A.

Verify pg lamp LIT.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter.

C.

Check value from Step 7.2.3.B

~ 12.0%.

(Step 7.2.3 continues on next page... )

Page 7 of20 INITIALS/DATA N41 N42 N43 N44

-0102

-0202

-0302

-0402 F0495D F0496D F0497D F0498D

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 8 of20 (Step 7.2.3 continued from previous page... )

INITIALS/DATA N41 N42 N43 N44 D.

!E value from Step 7.2.3.B is outside calibration tolerance band (g.5% to 10.5%) THEN complete Appendix A, using Appendix B as reference, otherwise NA.

E.

Verify either the value from Step 7.2.3.B or "As Left" value using Appendix A is within ACCEPTED g_5% to 10.5%.

F.

Trip status light 44205 [44512]

PWR RNG >Pg NC Ch S "ON."

-0112

-0212

-0312

-0412 G.

ERGS Ch POWER BELOW pg "CH TRIP."

Y9185D Y9186D Y9187D Y9188D 7.2.4 Lower Detector A and B test signals until pg lamp de-energizes.

A.

Verify pg 1amp de-energized.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter DESIRED 8.8% to g_g%.

c.

Trip status light 44205 [44512]

PWR RNG >Pg NC Ch S "OFF."

-0112

-0212

-0312

-0412 D.

ERGS Ch POWER BELOW pg "NT CH TRIP."

Y9185D Y9186D Y9187D Y9188D

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE c

NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 9 of20 INITIALS/DATA N41 N42 N43 N44 7.2.5 Raise Detector A and B test signals approximately equal until POWER ABOVE PERMISSIVE P10 lamp lights.

A.

Verify P10 lamp LIT.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter.

C.

Check value from Step 7.2.5.B

12.0%.

D.

IF value from Step 7.2.5.B is outside calibration tolerance band (10.4% to 11.4%) THEN complete Appendix A, using Appendix B as reference, otherwise NA.

E.

Verify either the value from Step 7.2.5.B or "As Left" value using Appendix A is within ACCEPTED 10.4% to 11.4%.

F.

Trip Status light 44178 [44510]

PWR RNG P10 NC Ch M "ON."

-0103

-0203

-0303

-0403 G.

ERGS PWR RNG CHAN Ch M P10 PERM "NOT BLOCK."

N0011D N0012D N0013D N0014D

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 10 of20 INITIALS/DATA N41.

N42 N43 N44 7.2.6 Lower Detector A and B test signals until P10 lamp de-energizes.

A.

Verify P10 lamp de-energized.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter.

C.

Check value from Step 7.2.6.B

~9.0%.

D.

IF value from Step 7.2.6.B is outside calibration tolerance band (9.5% to 10.5%) THEN complete Appendix A using Appendix B as reference, otherwise NA.

E.

Verify either the value from Step 7.2.6.B or "As Left" value using Appendix A using is within ACCEPTED 9.5% o 10.5%.

F.

Trip Status light 44178 [44510]

PWR RNG P10 NC Ch M "OFF."

-0103

-0203

-0303

-0403 G.

ERGS PWR RNG CHAN Ch M P10 PERM "BLOCK."

N0011D N0012D N0013D N0014D

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE c

NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 (2198]

REV:

18 Page 11of20 INITIALS/DATA N41 N42 N43 N44 7.3 Power Range Trip 7.3.1 Raise Detector A and B test signals approximately equal until OVERPOWER TRIP LOW RANGE lamp lights.

A.

Verify OVERPOWER TRIP LOW RANGE lamp LIT.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter.

c.

Check value from Step 7.3.1.B s 40.0%.

D.

IF value from Step 7.3.1.B is outside calibration tolerance band (23.9% to 24.9%) THEN complete Appendix A, using Appendix B as reference, otherwise NA.

E.

Verify either the value from Step 7.3.1.B or "As Left" value using Appendix A is within ACCEPTED 23.9% to 24.9%.

F.

Trip Status light 44178 [44510]

PWR RNG LO Q - HI F NC Ch P "ON."

-0106

-0206

-0306

-0406 G.

ERGS PWR RNG CHAN Ch PHI Q LO SP "CH TRIP."

N0006D N0007D N0008D N0009D

(

H.

Annunciator NIS POWER RANGE LO SETPOINT CHANNEL ALERT 47013-0202

[47513-0202] "ON."

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE

(

NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 12 of 20 INITIALS/DATA N41 N42 N43 N44 7.3.2 Lower Detector A and B test signals until OVERPOWER TRIP LOW RANGE lamp de-energizes.

A.

Verify OVERPOWER TRIP LOW RANGE lamp de-energized.

B.

Record REACTOR POWER PERCENT from PR A digital panel meter DESIRED 21.8% to 22.9%. --

c.

Trip status light 44178 [44510]

PWR RNG LO Q - HI F NC Ch P "OFF."

-0106

-0206

-0306

-0406 D.

ERGS PWR RNG CHAN Ch P HI Q LO SP "NT CH TRIP."

N0006D N0007D N0008D N0009D

E.

Annunciator NIS POWER RANGE LO SETPOINT CHANNEL ALERT 47013-0202 [47513-0202] "OFF."

(

c PRAIRIE ISLAND NUCLEAR GENERATING PLANT SP NIS POWER RANGE STARTUP TEST 7~3.3 On PR 8 drawer, position in sequence:

A.

DETECTOR A TEST LEVEL and DETECTOR 8 TEST LEVEL potentiometers fully CCW.

8.

OPERATION SELECTOR to "NORMAL".

7.3.4 On PR A drawer, after 30 seconds, position RATE MODE switch to "RESET" AND verify switch returns to "NORMAL".

7.3.5 On PR A drawer, position METER RATE switch to "SLOW".

7.3.6 On PR 8 drawer, position METER RANGE/RATE switches to "400 µA SLOW".

7.3.7 On COMPARATOR AND RATE Drawer place COMPARATOR CHANNEL DEFEAT switch to "NORMAL".

7.3.8 Verify the following:

On the PR 8 drawer A.

CHANNEL ON TEST lamp "OFF".

On the PR A drawer

8.

POSITIVE RATE TRIP lamp "OFF".

C.

NEGATIVE RATE TRIP lamp "OFF".

D.

On the COMPARATOR AND RATE DRAWER, verify COMPARATOR DEFEAT lamp is OFF.

SURVEILLANCE PROCEDURE NUMBER:

SP 1198 [2198]

REV:

18 Page 13 of 20 INITIALS/DATA N41 N42 N43 N44 OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS

PRAIRIE ISLAND NUCLEAR GENERATING PLANT 8.0 SP NIS POWER RANGE STARTUP TEST On the Annunciator Panels E.

Trip Status Light 44205 [44512] PWR RNG HI F RT NC Ch U/K "OFF".

F.

Annunciator NIS POWER RANGE POSITIVE FLUX RATE CHANNEL ALERT 47013-0101 [47513-0101]

"OFF".

G.

NIS POWER RANGE NEGATIVE FLUX RATE CHANNEL ALERT 47013-0201

[47513-0201] "OFF".

H.

NUCLEAR INSTR SYSTEM CHANNEL TEST 47013-0601 [47513-0601] "OFF".

7.4 Repeat Steps 7.1.3 through 7.3.8 for remaining channels.

ATTACHMENTS 8.1 Appendix A-Guidelines for Resolving Out-Of-Tolerance Conditions 8.2 Appendix B - Guidelines for Bistable Calibration SURVEILLANCE PROCEDURE NUMBER:

SP 1198 [2198]

REV:

18 Page 14 of 20 INITIALS/DATA N41 N42 N43 N44

-0104 -0204 -0304 -0404 OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS OPS

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 15 of 20 Appendix A Guidelines for Resolving Out-Of-Tolerance Conditions This Appendix provide the guidelines for resolving "As Found" data out-of-tolerance conditions for T.S. related setpoints. Listed below are the Required Actions to be taken when "As Found" data is outside the Acceptable Range.

Power Range Permissive P-8 Bistable No.

Value(%)

Concurrent Required As Found As Left Verification Action D

[]

D D

D G

~*

[2J I

0

, ",[]

D G l] ',i I

'1~!'1fft

i~; '

,, Cf 4........

~*

~~

N/A 8.87 8.90 9.90 9.93 11.00 Required Actions:

1 =No Action 2 = Readjust Setpoint 3 = Initiate AR to "Trend" data 4 = Initiate AR to "Evaluate" data 5 = Declare inoperable, contact Shift Supervisor

r

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 16 of 20 Appendix A Guidelines for Resolving Out-Of-Tolerance Conditions Power Range Permissive P-9 N/A Required Actions:

1 =No Action Bistable No.

9.47 2 = Readjust Setpoint 3 = Initiate AR to "Trend" data 4 = Initiate AR to "Evaluate" data Value(%)

As Found As Left 9.50 10.50 5 = Declare inoperable, contact Shift Supervisor Concurrent Verification 10.53 Required Action 12.00

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE

~-*

NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 17 of 20 Appendix A Guidelines for Resolving Out-Of-Tolerance Conditions Power Range Permissive P-10 TRIP (P-7 Input) i Bistable No.

Value(%)

Concurrent Required I

As Found As Left Verification Action

(

[]

'[]

1,

- o*

[]

[ f '[J '

I

~

I.

Ill !1

~ '

I 8 []

8

'[J I ll'i l

~

'z*'** **[]

i'..' '

)ff I 111

i

" *.(,

~ If*! I

- ~

,*JI;~*.,:

~,iiii','iil1i I,, I ' *1;iili ' I 4~

~

.......... ~

N/A 10.37 10.40 11.40 11.43 12.00 Required Actions:

1 = No Action 2 = Readjust Setpoint 3 = Initiate AR to "Trend" data 4 = Initiate AR to "Evaluate" data 5 = Declare inoperable, contact Shift Supervisor

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE

(

NUMBER:

5-p SP 1198 [2198]

NIS POWER RANGE STARTUP TEST REV:

18 Page 18 of 20 Appendix A Guidelines for Resolving Out-Of-Tolerance Conditions Power Range Permissive P-10 RESET Bistable No.

Value(%)

Concurrent Required As Found As Left Verification Action

(

.i:!il *r*

~""r *

')

~. '

,I.If..

'J f/t I

.(

[]

~ D

D 0

er

'.o 0

,, [J l'"

ii,,

E

.! I I!

f~

r

~

1,;

If I '

1

'~!!1!~

1 "

!l11.

I' ii;!1"1' 1"': :I,.

0

,/

('

it j ~

~

~ ~*

~*

9.00 9.47 9.50 10.50 10.53 N/A Required Actions:

1 =No Action 2 = Readjust Setpoint 3 = Initiate AR to "Trend" data 4 = Initiate AR to "Evaluate" data 5 = Declare inoperable, contact Shift Supervisor

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE

~--**

I NUMBER:

SP 1198 [2198]

SP NIS POWER RANGE STARTUP TEST REV:

18 Page 19 of 20 Appendix A Guidelines for Resolving Out-Of-Tolerance Conditions Power Range Trip-Low Setpoint N/A Required Actions:

1 =No Action Bistable No.

23.87 2 = Readjust Setpoint 3 = Initiate AR to "Trend" data Value(%)

As Found As Left 23.90 24.90 4 = Initiate AR to "Evaluate" data 5 = Declare inoperable, contact Shift Supervisor Concurrent Required Verification Action 24.93 40.00

PRAIRIE ISLAND NUCLEAR GENERATING PLANT SURVEILLANCE PROCEDURE NUMBER:

SP NIS POWER RANGE STARTUP TEST SP 1198 [2198]

REV:

18 Page 20 of20 Appendix B Guidelines for Bistable Calibration Calibration of the trip of each bistable is done with the TRIP control on the bistable and the reset calibrated with the LOOP control. Adjust to setpoint +/- 0.01VDC Bistable Function DESIRED DESIRED TRIP RESET

%/VDC

%/VDC NC304 PERMISSIVE PB 9.4/0.783 8.4/0.700 N41 1 [2]

F0495D CH SET CH RESET N42 F0496D N43 F0497D N44 F0498D NC305 OVERPOWER TRIP LOW 24.4/2.033 22.4/1.867 RANGE N41 1 [2]

N0006D CH TRIP NT CH TRIP N42 N0007D N43 NOOOBD N44 N0009D NC307 PERMISSIVE P9 10.0/0.833 9.4/0.783 N41 1 [2]

Y9185D CH SET CH RESET N42 Y9186D N43 Y9187D N44 Y9188D NC 308 PERMISSIVE P10 10.9/0.908 10.0/0.833 N41 1 [2]

N0011D NT BLOCK BLOCK N42 N0012D N43 N0013D N44 N0014D

(

3.3 INSTRUMENTATION 3.3.3 Event Monitoring (EM) Instrumentation EM Instrumentation 3.3.3 LCO 3.3.3 The EM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.

APPUCABILTIY:

MODES 1 and 2.

ACTIONS

NOTE---------------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION A.

NOTE-----------

A.I Not applicable to core exit temperature Function.

One or more Functions with one required channel inoperable.

Prairie Island Units 1 and2 REQUIRED ACTION Restore required channel to OPERABLE status.

COMPLETION TIME 30 days 3.3.3-1 Unit 1 -Amendment No.+.§:.&, 1671 Unit 2 -Amendment No. +/-49, 157

ACTIONS (continued)

CONDITION B. One or more required Core Exit Thermocouple (CET) channel(s) inoperable.

AND At least 4 CET channels OPERABLE in the center region of the core.

AND At least one CET channel OPERABLE in each quadrant of the outside core region.

C. Required Action and associated Completion Time of Condition A or B not met.

D.

NOTE------------

Not applicable to CET channels.

One or more Functions with two required channels inoperable.

Prairie Island Units 1and2 B.l C.1 D.1 REQUIRED ACTION Restore required CET channel(s) to OPERABLE status.

Initiate action in accordance with Specification 5.6.8.

Restore one channel to OPERABLE status.

EM Instrumentation 3.3.3 COMPLETION TIME 30 days Immediately 7 days 3.3.3-2 Unit 1 - Amendment No..+§.& 1631 Unit 2 - Amendment No..f.49 154

ACTIONS (continued)

CONDITION E.

Three or more required CET channels inoperable m one or more quadrants.

AND Less than four CETs OPERABLE in the center region of the core.

F.

Three or more required CET channels inoperable m one or more quadrants.

AND Less than one CET OPERABLE in each quadrant of the outside core region.

G.

Required Action and associated Completion Time of Condition D, E, or F not met.

Prairie Island Units 1and2 E.1 F.l G.1 REQUIRED ACTION Restore required channels to OPERABLE status.

Restore required channels to OPERABLE status.

Enter the Condition referenced in Table 3.3.3-1 for the channel.

EM Instrumentation 3.3.3 COMPLETION TIME 7 days 7 days Immediately 3.3.3-3 Unit 1 - Amendment No. H& 1631 Unit 2-Amendment No..f.49 154

(

ACTIONS (continued)

CONDITION H.

As required by Required ActionG.1 and referenced in Table 3.3.3-1.

I.

As required by Required Action G.1 and referenced in Table 3.3.3-1.

Prairie Island Units 1and2 H.1 1.1 REQUIRED ACTION Be in MODE 3.

Initiate action in accordance with Specification 5.6.8.

EM Instrumentation 3.3.3 COMPLETION TIME 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Immediately 3.3.3-4 Unit 1 - Amendment No. H-8 1631 Unit 2 - Amendment No. +/-49 154

SURVEILLANCE REQUIREMENTS EM Instrumentation 3.3.3

N()TE--------------------------------------------------

SR 3.3.3. l and SR 3.3.3.2 apply to each EM instrumentation Function in Table 3.3.3-1.

SURVEILLANCE SR 3.3.3.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

SR 3.3.3.2

N ()'fE----------------------------

N eutron detectors are excluded from CHANNEL CALIBRA 'fl()N.

Perform CHANNEL CALIBRATION.

FREQUENCY 31 days 24 months Prairie Island Units 1 and 2 3.3.3-5 Unit l - Amendment No. +§.& 163

1 Unit 2 -Amendment No. -!49 154

EM Instmmentation 3.3.3 Table 3.3.3-l (page l of I)

Event Monitoring Instrumentation

l.

2.

3.

4.

5.

6.

7.

8.

9.

10.

FU0iCTION Power Range Neutron Flux (Logarithmic Scale)

Source Range Neutron Flux (Logarithmic Scale)

Reactor Coolant System (RCS) Hot Leg Temperature RCS Cold Leg Temperature RCS Pressure (Wide Range)

Reactor Vessel \\Vatcr Level Containment Sump Water Level (Wide Range)

Containment Pressure (Wide Range)

Penetration Flow Path Automatic Containment Isolation Valve Position Containment Area Radiation (High Range)

11.

Not used

12.

13.

14.

15.

16.

Pressurizer Level Steam Generator Water Level (Wide Range)

Condensate Storage Tank Level Core Exit Temperature Rdueling Water Storage Tank Level REQUIRED CHANNELS 2

2 2

2 2 per p<:nelration flow path(a)(b) 2 2

2 per steam generator 2

(c) 4 per quadrant 2

CONDITION REFERENCED FROM REQUIRED ACTION G.l H

H H

(a)

Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(b)

Only one position indication channel is required for penetration flow paths with only one installed control room indication chann<::!.

(c)

A channel consists of one core exit thermocouple (CET).

Prairie Island Units 1 and 2 3.3.3-6 Unit I - Amendment No.~ 163 I Unit 2--Amendment No. +49 154 I

3.2 PGWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO 3.2.4 The QPTR shall be S: 1.02.

QPTR

3.2.4 APPLICABILffY

MODE 1 with THERMAL POWER> 50% RTP.

ACTIONS CONDITION A. QPTR not within limit.

Island Units l and.2 REQUIRED ACTION A. l Reduce THERMAL POvVER 2: 3% from RTP for each 1 % of QPTR

> 1.00.

A.2 Perform SR 3.2.4. l.

COMPLETION TIME 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after each QPTR determination Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.2.4-J Urnt l -* Amendment No. 158 2-- Amendment No. 149

ACTIONS CONDITION A. (continued)

Prairie Island Units 1 and 2 REQUIRED ACTION A.3 Perfonn SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1.

AND A.4 Re-evaluate safety analyses and confirm results remain valid for duration of operation under this condition.

QPTR 3.2.4 COMPLETION TIME 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.I Once per 7 days thereafter.

Prior to mcreasmg THERMAL POWER above the limit of Required Action A.l Amendment No. I 58 2 --Amendment t49

ACTIONS CONDITION A. (continued)

Prairie Island l and2 REQUIRED ACTION A. 5 -------------NOTES------------

1. Perform Required Action A.5 only afl::er Required Action A..4 is completed.

2. Required Action A.6 shall be completed when Required Action A.5 is perfonned.

Normalize excore detectors to restore QPTR to within limits.

QPTR 3.2.4 COMPLETION TIME Prior to mcreasmg THERMAL POWER above the limit of Required Action A.I 32.4-3 1 -- Amendment No. I Unit 2 Amendment No. 149

ACTIONS CONDITION A. (continued)

B. Required Action and associated Completion Time not met.

Prairie Island Units 1and2 REQUIRED ACTION QPTR 3.2.4 COMPLETION TIME A.6 -------------NOTE-------------

B.1 Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1, SR Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.2.1.2 and SR 3.2.2.1.

after achieving equilibrium conditions at R TP not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after mcreasmg THERMAL POWER above the limit of Required Action A.I Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months POWER to S 50% RTP.

3.2.4-4 Unit 1-AmendmentNo. 158 Unit2-AmendmentNo.149

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2. 4.1 -------------------------NOTES------------------------------

l.

With input fl-om one Power Range Neutron Flux channel inoperable and THERMAL POWER

S 85% RTP, the remaining three power range channels can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

V crify QPTR is within limit by calculation.

SR 3. 2. 4.2 --------------------------NOTE------------------------------

N ot required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER

> 85% RTP.

Verify QPTR is within limit using core power distribution measurement information.

QPTR 3.2.4 FREQUENCY 7 days 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Prairie Island Units 1 2

Unit I **-Amendment No. -l-§.g. 201 nit 2 No. +49 188

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM C21.3-10.1 121 LOW LEVEL GAS DECAY TANK REV:

21 Page 1of17 Continuous use of procedure required.

Read each step prior to performing.

Mark off steps as they are completed.

Procedure SHALL be at the work location.

PORC REVIEW DATE:

OWNER:

EFFECTIVE DATE NR D. Smith 6/15/12

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM C21.3-10.1 121 LOW LEVEL GAS DECAY TANK REV:

21 Page 2of17 Section TABLE OF CONTENTS Title Page 1.0 PURPOSE AND GENERAL DISCUSSION.......................................................... 3

2.0 REFERENCES

..................................................................................................... 3 3.0 PRECAUTIONS AND LIMITATIONS.................................................................... 3 4.0 PERSONNEL AND SPECIAL EQUIPMENT REQUIREMENTS........................... 4 5.0 SPECIAL CONSIDERATIONS............................................................................. 4 6.0 PREREQUISITES AND INITIAL CONDITIONS................................................... 5 7.0 PROCEDURE....................................................................................................... 5 8.0 ATTACHMENTS................................................................................................. 16 LIST OF ATTACHMENTS Attachment A 121 Gas Decay Tank Release Data Sheet.............................,........... 17

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM C21.3-10.1 121 LOW LEVEL GAS DECAY TANK REV:

21 Page 3of17 1.0 PURPOSE AND GENERAL DISCUSSION Emptying a low level waste gas decay tank may become necessary due to a build-up of nitrogen. This procedure is used by an operator to release gas from the low level loop to the atmosphere through the sample sink exhaust ducts.

2.0 REFERENCES

2.1 Developmental References 2.1.1 H4, ODCM Section 3.0, Table 3.2 2.1.2 Flow Diagram XH-1-124 2.2 Implementing References 2.2.1 C21.3.2, Low level Waste Gas Loop 2.2.2 C37.1, Auxiliary Building Normal and Special Ventilation Systems 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Gas decay releases should not be made any time precipitation is occurring.

H4, ODCM, defines the arc in Step 3.2 as 5° west of north to 45° east of north at 1 O MPH or less.

3.2 IF ALL the following conditions are met, THEN Gas Decay Tank releases SHALL NOT be made:

ANY Cooling Tower is in operation.

Wind direction is within arc defined as 330° to 360° and 0° to 60°.

Wind speed is~ 10 mph.

c PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM C21.3-10.1 121 LOW LEVEL GAS DECAY TANK REV:

21 Page 4of 17 4.0 PERSONNEL AND SPECIAL EQUIPMENT REQUIREMENTS 4.1 Personnel 4.1.1 One Control Room operator is required to monitor meteorological conditions.

4.1.2 One outplant operator is required for various valve manipulations.

4.1.3 One Radiation Protection Specialist is required for source testing 2R-30 and 2R-37 radiation monitors.

4.2 Equipment Requirements 4.2.1 11 mCi Cs-137 Bug Source 4.2.2 Key# 25 for WGDTS to U1 and U2 Sample Sink Exhaust Valve Lock.

5.0 SPECIAL CONSIDERATIONS 5.1 Gas decay tanks may only be released during certain weather conditions (see Section 3.0, Precautions and Limitations). Due to the unpredictability of the weather, Waste Gas releases may need to be suspended or terminated before the tank has emptied entirely.

5.2 Suspending a waste gas release until desirable conditions exist is acceptable, however once a release procedure is begun, it should be terminated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.3 If unfavorable weather occurs prior to releasing any gas to the environment (up to Section 7.13) it is acceptable to terminate the release, detach the Effluent Release Permit from the old release procedure marking applicable steps N/A, and reattach the Effluent Release Permit to a new release procedure. If a release is terminated after Section 7.13 has begun, a new gas decay tank sample and Effluent Release Permit should be generated before another release is attempted.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 6.0 PREREQUISITES AND INITIAL CONDITIONS 6.1 Hydrogen concentration of the tank should be less than 4 % as indicated on 121 Gas Analyzer.

6.2 121 Low Level Gas Decay Tank has been isolated per C21.3.2.

6.3 121 and 122 Hot Lab Sample Sink Exhaust Fans should be running prior to testing CV-31271. At least one of these fans must be running to satisfy the interlock to OPEN CV-31271.

7.0 PROCEDURE 7.1 Log the following:

Date and time Tank Pressure 7.2 Route this procedure to the Duty Chemist.

7.3 Sample 121 Low Level Gas Decay Tank to determine activity of tank.

NOTE:

If gas activity is too high, determine decay time needed based on isotopes in tank. Hold paperwork until tank can be released.

7.4 WHEN 121 Low Level Gas Decay Tank activity is low enough for release, THEN complete a Gas Decay Tank Gaseous Effluent Release Permit.

7.5 Attach the Effluent Release Permit to this procedure.

7.6 Route paperwork to Radiation/Chemistry Manager or designee for approval.

7.7 Route paperwork to Unit 1 SS for release.

C21.3-10.1 REV:

21 Page 5of17 Date/Time psig CHEM CHEM CHEM CHEM CHEM

c

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 7.8 Pre-release Set-up:

7.8.1 Conduct a pre-job brief in accordance with PINGP 1112 with Shift Supervisor or designee.

7.8.2 Check wind conditions AND enable ERCS wind condition alarm by performing the following:

NOTE:

7.8.3 A.

Select Operator Utilities/Operator Entry/Miscellaneous.

B.

Enter new value 1, for point 1K4100 WASTE GAS RELEASE ALARM.

C.

Select apply.

D.

Monitor ERCS server group OPWIND_U1, to verify wind conditions are appropriate per Section 3.0.

121OR122 SMPL RM EXHT FAN must be running to satisfy the interlock to OPEN CV-31271.

Verify the following Control Room Misc. Vent.

Annunciators are NOT LIT:

44071-0106, 121 SMPL RM EXHT FAN STOPPED.

44071-0107, 122 SMPL RM EXHT FAN STOPPED.

7.9 Verify proper operation of 2R-30 by performing the following steps:

7.9.1 Using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, OPEN CV-31271 until 5333402 red OPEN light is LIT.

7.9.2 Record 2R-30 background count rate:

7.9.3 Notify the Radiation Protection Specialist to place the bug source in the monthly bug position (position 2) for 2R-30 radiation monitor AND record the count rate:

C21.3-10.1 REV:

21 Page 6of17 cpm cpm

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE r~*-*--*--------*-----*----*------*

I NUMBER:

I i

C I RELEASING RADIOACTIVE GAS FROM !

C21.3-10.1 L----*----~--J ___ 121 LOW LEVEL GAS DECAY_T_A_N_K

__c ___

R:_:_~e-7_o_f2_1~~

7.10 IF 2R-30 does not fall in the required range or does not trip CV-31271 shut, THEN Gas Decay Tank Releases SHALL NOT be made until the requirements of H4, ODCM Table 3.2 are satisfied.

7.9.4 Check that 2R-30 bug source reading minus background is within 3.5 X 103 to 1.8 X 104 cpm.

___ cpm -

cpm =

cpm bug source reading background reading 3.5 X103 to 1.8 X 104 from Step 7.9.3 from 7.9.2 7.9.5 Using 53334, GAS DCY TNKS TO PLNT VNT CV-31271 POSN IND, check that CV-31271, CLOSES on the high radiation signal.

7.9.6 Notify the Radiation Protection Specialist to remove the Bug Source from the test position.

7.9.7 Depress the ESF EQUIP/RESET Pushbutton for 2R-30 to reset the Hi Rad Alarm.

7.9.8 Independently verify Step 7.9.7.

7.9.9 Stop Aux Bldg Special Ventilation per C37.1 returning Aux Bldg Ventilation to normal.

7.9.10 Using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, CLOSE CV-31271 until 5333403 white light is NOT LIT.

Verify proper operation of 2R-37 by performing the following:

7.10.1 Using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, OPEN CV-31271 until 5333402 red OPEN light is LIT.

7.10.2 Record 2R-37 background count rate:

7.10.3 Notify the Radiation Protection Specialist to place the bug source in the monthly bug position (position 2) for 2R~37 radiation monitor and record the count rate:

IV cpm cpm

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK NOTE:

IF 2R-37 does not fall in the required range or does not trip CV-31271 shut, THEN Gas Decay Tank Releases SHALL NOT be made until the requirements of H4, ODCM Table 3.2 are satisfied.

7.10.4 Check that 2R-37 bug source reading minus background is within 4 X 103 to 2.5 X 104 cpm.

____ cpm bug source reading from Step 7.10.3

_ cpm =

cpm background reading 4 X1 o3 to 2.5 X 104 from 7.10.2 7.10.5 Check that CV-31271 (RCV-014) GAS DCY TNKS TO PLNT VNT CV, CLOSES on the high radiation signal.

7.10.6 Notify the Radiation Protection Specialist to remove the Bug Source from the test position.

7.10. 7 Depress the ESF EQUIP/RESET pushbutton for 2R-37 to reset the Hi Rad alarm.

7.10.8 Independently verify Step 7.10.7.

7.10.9 Stop Aux Bldg Special Ventilation per C37.1 returning Aux Bldg Ventilation to normal.

7.10.10 Using HC-53332, GAS DCY TNKS PLT VNT ISOL CV MAN LOADER, CLOSE CV-31271 until 5333403 white light is NOT LIT.

C21.3-10.1 REV:

21 Page 8of17 IV

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 7.11 Shift Supervisor release review and approval:

7.11.1 Verify Chemistry Manager or designee has approved the release on the Gas Decay Tank Gaseous Effluent Release Permit.

7.11.2 Check wind conditions as specified in the Limitations section are satisfied (from ERGS server group "OPWIND"):

10-meter average wind speed ------- mph 10-meter average wind direction ------ 0 7.11.3 Release Approval:--------

Shift Supervisor Date:

Time:

7.11.4 Assign the key for valves WG-15-1 and WG-15-2, WGDTS TO SMPL SINK EXH (Key Hook #25).

7.12 Complete the following line-up of the Waste Gas System:

7.12.1 Verify that 21 Auxiliary Building General Exhaust Fan is running.

7.12.2 Verify the following valves are CLOSED:

A.

CV-31271, GAS DCY TNKS TO PLNT VNT CV, (Control located at the Waste Gas Panel)

B.

CV-31265, 125 LO LVL GAS DCY TNK INLT ISOL PCV.

C.

CV-31455, 121/125 WGDT TO eves HUT CV 7.12.3 Set CV-31270, LO LVL GAS DCY TNKS TO PLNT VNT PCV, at a setpoint of 7 psig (approximately 13 on the dial). This ensures a constant release rate throughout the tank's discharge. (Located in Waste Gas Compressor area.)

C21.3-10.1 REV:

21 Page 9of17

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE I NUMBER:

I I

I l C I RELEASING RADIOACTIVE GAS FROM

___________ L 121 LOW LEVEL GAS DECAY TANK I

C21.3-10.1 I

REV:

21 I

Page 10of17 7.12.4 Verify CLOSED the following valves:

A.

WG-3-21, AT 125 GAS DECAY TANK-CV31265 INLET.

B.

WG-3-22, 125 GAS DECAY TANK-ISOL VALVE.

C.

WG-3-3, 127 GAS DECAY TANK-OUTL TO EXHAUST DUCTS.

D.

WG-3-8, 128 GAS DECAY TANK-OUTL TO EXHAUST DUCTS.

E.

WG-3-13, 129 GAS DECAY TANK-OUTL TO EXHAUST DUCTS.

F.

WG-3-18, 123 GAS DECAY TANK.

G.

WG-3-28, 126 GAS DECAY TANK-OUTL TO EXHAUST DUCTS.

7.12.5 Verify OPEN the following valves:

A.

WG-21-4, FROM 121 GAS DECAY TANK TO SMPL SINK EXHST DUCT.

B.

WG-3-23, 125 GAS DECAY TANK-OUTL OUTL TO EXHAUST DUCTS.

7.12.6 Verify proper operation of Sample Room Exhaust Hoods by CLOSING the exhaust hood glass until there is noticeable suction from the hot sample room. Maintain the glass window at this point during release.

7.12.7 Hang "Caution: Waste Gas Decay Tank Release in progress" tags on the Unit 1 and Unit 2 sample hoods.

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 7.13 Release Procedure If unfavorable conditions arise, suspend the release per Section 7.14 of this procedure. If extended unfavorable conditions exist, terminate the release per Section 7.15 of this procedure.

7.13.1 Inform the Duty Chemist that a Gas Decay Tank release is about to commence.

7.13.2 Unlock and OPEN the following valves:

CAUTION:

A.

WG-15-1, CV31271 OUTL TO U1 SMPL SINK - EXHAUST DUCT.

B.

WG-15-2, WASTE GAS TO SAMPLE SINK-EXHAUST DUCTS.

THE RELEASE SHALL BE SUSPENDED OR TERMINATED PER SECTION 7.14 OR 7.15 IF THE CONDITIONS IN SECTION 3.0 ARE NOT SATISFIED.

7.13.3 Notify the Control Room that the release is beginning and that wind conditions are to be periodically monitored (approx. every 15 minutes). Evaluate using one of the following methods for monitoring atmospheric conditions:

A.

ERCS server group "OPWIND".

B.

ERCS alarm point 1U4101D, Waste Gas Release Wind Condition.

7.13.4 Instruct the Control Room Operator to place 2R-30 and 2R-37 on an ERCS trend plot using Point ID's 2R0030A and 2R0037 A. Monitor until the release has been completed.

A new Gaseous Effluent Release Permit. should be generated if the release is terminated prior to depressurizing the tank to less than 5 psig after this point in the procedure.

7.13.5 Record the time and pressure at the start of the release on the data sheet on Attachment A of this procedure.

C21.3-10.1 REV:

21 Page 11 of 17

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK NOTE:

Opening CV-31271 finishes the valve line-up for the release and will begin sending waste gas to the atmosphere. Use caution to avoid actuation of Aux. Building Special Ventilation.

7.13.6 7.13.7 Using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, crack OPEN CV-31271, to a demand signal of about 8.

Check that 2R-30 and 2R-37 are below the expected reading indicated on the Gas Decay Tank Release Authorization Form by observing the 2R-30 and 2R-37 meters at the Aux. Bldg. operator's shack.

7.13.8 Contact the Control Room Operator to verify that 2R-30 and 2R-37 are below the expected reading as indicated on the Gas Decay Tank Release Authorization Form, by observing the Control Room 2R-30 AND 2R-37 meters OR by monitoring the digital ERCS display Point ID's 2R0030A AND 2R0037A.

NOTE:

7.13.9 7.13.10 7.13.11 The release should be terminated if count rate cannot be maintained below the expected count rate.

Using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, slowly OPEN CV-31271 until the expected count rate is achieved OR until the valve is fully OPEN.

Verify CV-31270 has a process pressure of 7 psig.

Periodically monitor and check that no unexpected lowering of waste gas tank pressures is occurring.

C21.3-10.1 REV:

21 Page 12of17

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK A release should be terminated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after it was first initiated.

7.13.12 WHEN the pressure in 121 Gas Decay Tank has decreased to less than 5 psig, OR extended unfavorable weather exists, THEN using HC-53332, GAS DCY TNKS PL T VNT ISOL CV MAN LOADER, CLOSE CV-31271.

7.13.13 Complete one of the following:

A.

Suspend the release per Section 7.14.

B.

Terminate the release per Section 7.15.

C21.3-10.1 REV:

21 Page 13of17

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 7.14 Suspending a Waste Gas Release While releasing a waste gas decay tank to the atmosphere, weather conditions may arise that violate the specifications outlined in the ODCM. This section is used to suspend a waste gas release if desired by the Shift Supervisor due to plant or weather conditions when the release is expected to resume in the near future.

A Waste Gas Release should be active for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . IF weather conditions are unlikely to become desirable for an extended period, THEN terminate the release per Section 7.15.

7.14.1 Verify CLOSED CV-31271, GAS DCY TNKS TO PLNT VNTCV.

Since the release is expected to resume, the locks for WG-15-1 and WG-15-2 may be left OFF at this time. These

xy: ::'""':';>r~ valves should be locked when the release is terminated in Section 7.15 of this procedure.

7.14.2 CLOSE WG-15-1, CV31271 OUTL TO U1 SMPL SINK - EXHAUST DUCT.

7.14.3 CLOSE WG-15-2, WASTE GAS TO SAMPLE SINK-EXHAUST DUCTS.

7.14.4 Complete one (1) of the following:

A.

WHEN desirable conditions arise, THEN resume the release per Section 7.13.

B.

Terminate the release per Section 7.15.

C21.3-10.1 REV:

21 Page 14of17

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE

-*******------*----***--*-*,-------------------~--------,

I NUMBER:

c RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK 7.15 Terminating a Waste Gas Release 7.15.1 Record the remaining entries on Attachment A and notify the Control Room and the Duty Chemist that the release has ended.

7.15.2 CLOSE and lock the following valves:

A.

WG-15-1, CV31271 OUTL TO U1 SMPL SINK - EXHAUST DUCT B.

WG-15-2, WASTE GAS TO SAMPLE SINK-EXHAUST DUCTS 7.15.3 Remove "Caution: Waste Gas Decay Tank Release in progress" tags on the Unit 1 and Unit 2 sample hoods.

7.15.4 CLOSE WG-3-23, 125 GAS DECAY TANK-OUTL OUTL TO EXHAUST DUCTS.

7.15.5 OPEN the following valves:

A.

WG-3-22, 125 GAS DECAY TANK-ISOL VALVE B.

WG-3-21, AT 125 GAS DECAY TANK-CV31265 INLET 7.15.6 CLOSE CV-31270, LO LVL GAS DCY TNKS TO PLNT VNT PCV.

7.15.7 Log the required entries in the Aux Building Operator Log.

7.15.8 Disable ERGS waste gas release alarm as follows:

A.

Select Operator Utilities/Operator Entry/Miscellaneous.

B.

Enter new value 0, for point 1K4100 WASTE GAS RELEASE ALARM.

C.

Select apply.

7.15.9 Review the completed release procedure and return the procedure and the key for WG-15-1 and WG-15-2 to the Shift Supervisor (Key Hook# 25).

7.15.10 Shift Supervisor review release procedure and ensure that the Gas Decay Tank Gaseous Effluent Release Permit and Gas Decay Tank Release Data Sheet are complete.

C21.3-10.1 REV:

21 Page 15 of 11 \\

SS

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE I NUMBER:

c RELEASING RADIOACTIVE GAS FROM I

C21.3-10.1 121 LOW LEVEL GAS DECAY TANK REV:

21 Page 16of17 7.15.11 Return the Gas Decay Tank Gaseous Effluent Release Permit with this procedure to the Duty Chemist.

8.0 ATTACHMENTS Attachment A - 121 Gas Decay Tank Release Data Sheet

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OPERATING PROCEDURE NUMBER:

RELEASING RADIOACTIVE GAS FROM 121 LOW LEVEL GAS DECAY TANK Attachment A 121 Gas Decay Tank Release Data Sheet A.

Date Tank Placed on HOLD: ------------~

B Start of Release Data Tank Pressure at Start of Release: _________ psig Date at Start of Release: -------------~

Time at Start of Release: -------------~

Operator's Signature:---------------

C.

End of Release Data Final Tank Pressure at End of Release: ________ psig Date at End of Release: ---------------

Time at End of Release: --------------

Operator's Signature:---------------

Comments:

C21.3-10.1 REV:

21 Page 17of17

SRO APPLICANT EXAMINATION

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

76. P8170L-002 136/015/017 AA2.10/3.7/3.7/3H/YES/P8100/1C3 AOP2/T.S. 3.4.5/2014 ILT NRC S76 Given the following conditions:

- Unit 1 is in Mode 3, HOT STANDBY.

- 11 and 12 RCPs are running.

- The RCP indications on Panel B (CVCS Letdown) are as follows:

Question continued on next page.

Page 1

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

76. P8170L-002 136/015/017 AA2.10/3.7/3.7/3H/YES/P8100/1C3 AOP2/T.S. 3.4.5/2014 ILT NRC S76 Question continued on next page.

Question continued from previous page.

- The CC indications of Panel A (Component Cooling) are as follows:

After completing the actions of the appropriate AOP, the Shift Supervisor will declare _____________________ INOPERABLE.

ONLY the "A" RCS loop ONLY the "B" RCS loop BOTH RCS Loops NEITHER RCS Loops A.

B.

C.

D.

Page 2

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for T.S. 3.4.5.

Justifications:

a. Incorrect. Plausible as seal injection is loss to the 11 RCP; however, 11 RCP will NOT be secured because seal cooling is not lost to 11 RCP as indicated by bearing temperatures and CC flows to 11 RCP.

b. Correct. 12 RCP has lost seal cooling (Seal injection and CC to the bearings); therefore, 12 RCP will be secured per 1C3 AOP2. Once 12 RCP is secured, the "B" RCS Loop is INOPERABLE per T.S. 3.4.5.

c. Incorrect. Plausible if examinee incorrectly believes both RCPs will be stopped based on loss of seal injection flow alone. This is incorrect per 1C12.1 AOP1.

d. Incorrect. Plausible if examinee is not familiar or does not recognize 12 RCP has exceeded the bearing water temperature limit of 200F and determines NO RCPs need to be tripped at this time.

K/A Number:

015/017 Reactor Coolant Pump (RCP) Malfunctions AA2.10:

Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

When to secure RCPs on loss of cooling or seal injection Technical Reference(s): 1C3 AOP2 page 4, 1C12 AOP1 page 4, TS LCO 3.4.5 Proposed references to be provided to applicants during examination: None Learning Objective: P8170L-002 Obj. 3H Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 __ ___

55.43 __2___

Comments:

Page 3

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

77. P8172L-001A 133/022 2.4.8/3.8/4.5/7B/YES/P8100/1E-0/1C12.1 AOP1/SWI O-10/2014 ILT NRC S77 Given the following conditions:

- The crew is performing 1C12.1 AOP1, Loss of RCP Seal Injection.

- Unit 1 Reactor is tripped.

- Both Unit 1 RCPs are tripped.

- Immediate operator actions of 1E-0, Reactor Trip or Safety Injection, are complete.

- CC flow to each RCP is 210 gpm.

- The crew can NOT restore any Unit 1 Charging Pumps.

The SS will...

direct the Lead RO to perform Attachment L and enter 1C3 AOP2, Loss of RCP Seal Cooling.

direct the Lead RO to perform Attachment L and enter 1C18 AOP1, Makeup or Boration of the RCS Using a Safety Injection Pump.

transition to 1ES-0.1, Reactor Trip Recovery and enter 1C3 AOP2, Loss of RCP Seal Cooling.

transition to 1ES-0.1, Reactor Trip Recovery and enter 1C18 AOP1, Makeup or Boration of the RCS Using a Safety Injection Pump.

A.

B.

C.

D.

3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes SI will be actuated during E-0 in order to provide makeup to the RCS and incorrectly believes RCP seal cooling is lost.

b. Incorrect. Plausible as the SS will enter 1C18 AOP1; however, the SS will NOT direct the Lead RO to perform Attachment L.

c. Incorrect. Plausible as the SS will transition to 1ES-0.1; however, the SS will NOT enter 1C3 AOP2.

d. Correct.

Page 4

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

022 Loss of Reactor Coolant Makeup 2.4.8:

Knowledge of how abnormal operating procedures are used in conjunction with EOPs Technical Reference(s): 1C12.1 AOP1 page 4, 1E-0 pages 4 -5, SWI O-10 pgs 5, 10, 14.

Proposed references to be provided to applicants during examination: None Learning Objective: P8172L-001A Obj. 7B Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X_

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 5

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

78. P8180L-003 058/025 AA2.07/3.4/3.7/7B/YES/P8100/1C15 AOP1 / AOP2/1C15 AOP3 / D2 AOP1/2014 ILT NRC S78 Given the following conditions:

- Unit 1 is in Mode 5.

- ERCS DP is 151" and stable.

- RVLIS is 100% and stable.

- An out-plant operator is in the field performing a valve lineup in the Auxiliary Building.

- 11 RHR Pump is in standby.

- 12 RHR Pump is running with the following indications:

- Discharge pressure is oscillating between 0 and 100 psig.

- Flow to the RCS is oscillating between 0 and 400 gpm.

The Shift Supervisor will enter...

1C15 AOP1, RHR Flow Restoration, and stop 12 RHR pump.

1C15 AOP2, Loss of Coolant Inventory with RHR in Operation, and isolate the leak.

D2 AOP1, Loss of Coolant While In A Reduced Inventory Condition, and make up to the RCS.

1C15 AOP3, RHR Operation without CR Instrumentation or Flow Control, and manually throttle CLOSE the 11/12 RHR HX Bypass Flow Valve.

A.

B.

C.

D.

3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires the knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Justifications:

a. Correct. Indications given are showing the 12 RHR Pump cavitating due to loss of suction.

b. Incorrect. Plausible if examinee incorrectly believes a loss of level is what caused the loss of RHR flow; however, RVLIS and ERCS DP are stable.

c. Incorrect. Plausible if examinee incorrectly believes the RCS is at reduced inventory; however, the RCS is NOT considered at reduced inventory until ERCS DP level is below 52.25 inches (corresponds to 3 feet below the reactor vessel flange) per 1D2.

d. Incorrect. Plausible as the 12 RHR pump is cavitating; however, NOT due to the flow control valve failing open.

Page 6

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

025 Loss of Residual Heat Removal System (RHRS)

AA2.07:

Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System:

Pump cavitation Technical Reference(s): 1C15 AOP1 pages 3 & 4, 1C15 AOP2 pages 3 & 4, 1C15 AOP3 pages 3 & 4, D2 AOP1 pages 3 & 4, 1D2 page 18.

Proposed references to be provided to applicants during examination: None Learning Objective: P8180L-003 058 Obj. 7B Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 7

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

79. P8197L-012 219/E12 2.2.44/4.2/4.4/38/YES/P8100/ECA-2.1//2014 ILT NRC S79 Given the following conditions:

- Unit 1 has experienced a major secondary system break.

- Both MSIVs are OPEN and can NOT be closed remotely.

- The crew has entered 1ECA-2.1, Uncontrolled Depressurization of Both Steam Generators.

- An Out-plant Operator closes 11 MSIV locally.

- 11 SG pressure is 550 psig and RISING.

- 12 SG pressure is 575 psig and LOWERING.

- 11 SG WR level is 45% and stable.

- 12 SG WR level is 48% and slowly lowering.

- Secondary radiation is normal.

- RWST level is 43% and slowly lowering.

- RCS pressure is 1600 psig and slowly lowering.

The Shift Supervisor will transition to...

1ES-1.2, Transfer To Recirculation.

1E-3, Steam Generator Tube Rupture.

1E-2, Faulted Steam Generator Isolation.

1E-1, Loss of Reactor or Secondary Coolant.

A.

B.

C.

D.

Page 8

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. This is NOT a direct entry into a major EOP because the SS will have entered E-2, then transitioned to 1ECA-2.1, and then transition back to E-2.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes the switchover criteria is 43% RWST level; however, switchover criteria is 33% RWST level.

b. Incorrect. Plausible as 11 SG pressure rising is an indication of a SG Tube Rupture; however, during a SG Tube Rupture, level would also rise.

c. Correct. ECA-2.1 directs the operator to go to E-2 if one of the SGs pressures start to rise.

d. Incorrect. Plausible as RCS pressure is lower than normal; however, transition from 1ECA-2.1 only occurs if RCS is pressure is below 250 psig (no adverse containment).

K/A Number:

E12 Uncontrolled Depressurization of all Steam Generators 2.2.44:

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Technical Reference(s): ECA-2.1 pages 5, 6, and information page.

Proposed references to be provided to applicants during examination: None Learning Objective: P8197L-012 Obj. 38 Question Source:

Bank # _______

Modified Bank # _P8197L-012 046__

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 9

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

80. P8197L-011 105/055 EA2.03/3.9/4.7/7/YES/P8100/1ECA-0.0//2014 ILT NRC S80 Given the following conditions:

- The crew has entered 1ECA-0.0, Loss of All Safeguards AC Power.

- Offsite power is NOT available.

- Bus 15 is locked out.

- D2 Diesel Generator is OOS.

- The Bus 15 green load rejection lights are LIT.

- The Bus 16 green load rejection lights are NOT LIT.

- The Unit 2 Safeguard busses are energized from their respective diesels.

- 1ECA-0.0 is provided.

On which step of 1ECA-0.0 will the Shift Supervisor direct the Lead Reactor Operator to restore power to a Unit 1 Safeguards Bus?

step 6 step 7 step 10 step 11 A.

B.

C.

D.

2-RI EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes a Unit 1 source is available; however, since Bus 15 is locked out, D2 OOS, and NO offsite power, there is NO unit 1 source available.

b. Incorrect. Plausible as it is a common misconception to state Bus 16 is available; however, it is NOT available for sequencer loading. Therefore, ECA-2.1 directs operator to take isolate RCP seals and take loads to pullout prior to energizing Bus 16 from Unit 2 to prevent block loading the Unit 2 Diesel.

c. Incorrect. Plausible if examinee incorrectly believes a Unit 1 source is available.

d. Correct. Since the load sequencers are NOT available, Unit 1 sources are NOT available, and Unit 2 safeguards buses are energized, the crew will restore power on step 11.

Page 10

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

055 Loss of Offsite and Onsite Power (Station Blackout)

EA2.03:

Ability to determine or interpret the following as they apply to a Station Blackout:

Actions necessary to restore power Technical Reference(s): 1ECA-0.0 pages 5 - 12 Proposed references to be provided to applicants during examination: All steps of 1ECA-0.0, but no background information.

Learning Objective: P8140L-247 Obj. 7 Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 11

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

81. P8197L-011 017/E04 2.4.18/3.3/4.0/A3/YES/P8100/E-0/ECA-1.2/2014 ILT NRC S81 Given the following conditions:

- A LOCA has occurred on Unit 1.

- Containment pressure is 0.1 psig and stable.

- RCS subcooling is 92°F and stable.

- Total feed flow to SGs is 250 gpm and stable.

- RCS pressure is 1840 psig and stable.

- Pressurizer level is 0% and stable.

- Auxiliary building radiation alarms are the ONLY radiation alarms occurring.

- Additional equipment failures result in the break NOT being isolable from the RCS.

The Shift Supervisor will transition from 1E-0 to ________________________ and the Unit will be cooled down to Cold Shutdown using _____________________________.

1ECA-1.2, LOCA Outside Containment 1ECA-1.1, Loss of Emergency Coolant Recirculation 1ECA-1.2, LOCA Outside Containment 1ES-1.1, Post-LOCA Cooldown and Depressurization 1ES-0.2, SI Termination 1ES-1.1, Post-LOCA Cooldown and Depressurization 1ES-0.2, SI Termination 1C1.3, Unit 1 Shutdown A.

B.

C.

D.

Page 12

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Correct. Since the LOCA is in the Auxiliary Building, transition is made directly to 1ECA-1.2 based on adverse radiation levels in the Auxiliary Building. The transition to 1ECA-1.1 is made because all RCS water is going to the Auxiliary Building instead of sump B in containment; therefore, loss of recirc capability.

b. Incorrect. Plausible as the SS will transition from 1E-0 to 1ECA-1.2; however, the unit will not be cooled down using ES-1.1.

c. Incorrect. Plausible if the examinee incorrectly believes SI termination criteria is met and incorrectly believes the unit will be cooled down using 1ES-1.1. SI termination criteria is not met because pressurizer level is below 7% and RCS pressure is below 2000 psig. During a normal small break LOCA (i.e. inside containment), the unit would be cooled down using 1ES-1.1.

d. Incorrect. Plausible as the crew would transition to 1ES-0.2 and the unit would be cooled down using 1C1.3; however, this procedure flow path would only be used in this situation if SI termination criteria was met. SI termination criteria is not met because pressurizer level is below 7% and RCS pressure is below 2000 psig.

K/A Number:

E04 LOCA Outside Containment 2.4.18:

Knowledge of the specific bases for EOPs Technical Reference(s): 1E-0 page 11, 1ECA-1.2 pages 1 -4.

Proposed references to be provided to applicants during examination: None Learning Objective: P8197L-011 Obj. A3 Question Source:

Bank # _P8197L-011 017_

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 13

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

82. P8182L-003 030/036 2.4.4/4.5/4.7/3C/YES/P8100/C1.6 AOP1 /D5.2 AOP1/B17 / C17/2014 ILT NRC S82 Given the following conditions:

- Core reload refueling activities are in progress.

- An irradiated fuel assembly is being lowered into the core with the HOIST JOG SWITCH.

- The ENTERING CORE SLOW ZONE light has just extinguished.

- The manipulator crane operator continues lowering the irradiated fuel assembly into the core, now using the HOIST CONTROL LEVER.

- The hoist abruptly stops and the mast support tube is shaking noticeably.

- The manipulator crane camera shows the fuel assembly is FULLY inserted.

- The following indications are present:

- ENTERING CORE SLOW ZONE light is OFF.

- INTERMEDIATE CORE ZONE light is ON.

- BOTTOM CORE SLOW ZONE light is OFF.

- UNDERLOAD light is ON.

- SLACK CABLE light is ON.

- Gas bubbles are visible rising from the vicinity of the fuel assembly.

The _________ _______ control signal has failed. The Containment SRO will implement _________ __________ AND _________ _________.

UNDERLOAD D5.2 AOP1, Damaged Fuel Assembly C1.6 AOP1, Containment Evacuation UNDERLOAD D5.2 AOP1, Damaged Fuel Assembly D5.2 AOP4, Spent Fuel Pool Area Evacuation-Refueling BOTTOM CORE SLOW ZONE D5.2 AOP1, Damaged Fuel Assembly C1.6 AOP1, Containment Evacuation BOTTOM CORE SLOW ZONE C1.6 AOP1, Containment Evacuation D5.2 AOP4, Spent Fuel Pool Area Evacuation-Refueling A.

B.

C.

D.

Page 14

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(7) Fuel handling facilities and procedures. This question requires knowledge of the Refuel Floor SRO responsibilities.

Justification:

a.

Incorrect. This is the wrong control signal failure, since the underload light is illuminated, even though the procedures are correct.

b.

Incorrect. This has both the wrong control signal failure and incorrect procedures to implement.

c.

Correct. This is the correct control signal failure and the correct procedures to be implemented.

d.

Incorrect. Although this is the correct control signal failure, the implemented procedures are not correct.

K/A Number:

036 Fuel Handling Accident 2.4.4:

Ability to recognize abnormal indications for system operating parameters that are entry level conditions for emergency and abnormal operating procedures.

Technical Reference(s): C1.6 AOP1 page 3, D5.2 AOP1 page 3, B17 pages 10 - 14, C17 page 27.

Proposed references to be provided to applicants during examination: None Learning Objective: P8182L-003 Obj. 3C Question Source:

Bank #: P8182L-003 030 Modified Bank #:

New:

Question History: Last NRC Exam: 2012 ILT NRC EXAM Question Cognitive Level:

Memory or Fundamental Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content:

55.41:

55.43: 7 Comments:

Page 15

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

83. P8171L-007 071/037 AA2.10/3.2/4.1/6/YES/P8100/TS 3.4.14//2014 ILT NRC S83 Given the following conditions:

- Unit 1 is at 100% power.

- The following RCS leakage indications were determined at 1100 on 8/4/14:

- IDENTIFIED leakage is 9.1 gpm.

- UNIDENTIFIED leakage is 0.8 gpm.

- Primary to Secondary leakage is 432 gallons per day.

- T.S. LCO 3.4.14 is provided.

Technical Specification LCO 3.4.14 requires Unit 1 to be in Mode 5 by ________ on 2300 8/5/14 0700 8/6/14 2300 8/7/14 0300 8/8/14 A.

B.

C.

D.

3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for T.S. 3.4.14.

Justifications:

a. Correct. Since primary to secondary leakage is 432 gallons per day, LCO 3.4.14 condition D is entered requiring the unit to be in Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. Incorrect. Plausible if examinee incorrectly believes ALL leakage combined meets TS 3.4.14 for >10 gpm IDENTIFIED leakage; however, in this case IDENTIFIED leakage is 9.4 gpm total.

c. Incorrect. Plausible if examinee incorrectly believes UNIDENTIFIED leakage is "unidentified" plus "pri to sec" at 1.1 gpm AND incorrectly applies Cond B only; however, in this case "unidentified" leakage is limited to 0.8 gpm and Cond A & B would be entered if >1 gpm.

d. Incorrect. Plausible if examinee incorrectly believes UNIDENTIFIED leakage is "unidentified" plus "pri to sec" at 1.1 gpm; however, in this case "unidentified" leakage is limited to 0.8 gpm.

Page 16

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Statement:

037 Steam Generator Tube Leak AA2.10:

Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:

Tech-Spec limits for RCS leakage Technical Reference(s): TS 3.4.14 Proposed references to be provided to applicants during examination: TS 3.4.14 Learning Objective: P8171L-007 Obj. 6 Question Source:

Bank # ____

Modified Bank # _____

New ___X____

Question History: Last NRC Exam ____N/A___

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 ___ __

55.43 __2___

Comments:

Page 17

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

84. P8182L-002 001/2.2.38/3.6/4.5/10C/YES/P8100/H4//2014 ILT NRC S84 Given the following conditions:

- Steam Generator blowdown is aligned to the river.

- Secondary coolant specific activity is <0.01uCi/gram DOSE EQUIVALENT I-131.

- R-21, CIRC WATER DISCH MONITOR, fails low and is declared inoperable.

- Table 2.2 of H4, Offsite Dose Calculation Manual, is provided.

The Shift Supervisor will ensure...

flow rate is estimated every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

grab samples are collected and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

grab samples are collected and analyzed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

grab samples are collected and saved for weekly composition and analysis every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

A.

B.

C.

D.

2-RI EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for the Offsite Dose Calculation Manual. ODCM required actions is a SRO only function.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes R-21 is a flow monitor for Steam Generator Blowdown; however, R-21 is a radiation monitor.

b. Correct. Per Table 2.2 of H4, if the discharge canal monitor is inoperable than action 6 is required.

c. Incorrect. Plausible if examinee incorrectly believes R-21 is used to measure Steam Generator Blowdown effluent; however, R-19 is used to measure blowdown effluent.

d. Incorrect. Plausible if examinee incorrectly believes R-21 is the rad monitor used to measure Turbine Building Sump effluent; however, there are seperate rad monitors for this.

Page 18

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

APE 059 Accidental Liquid Radwaste Release 2.2.38:

Knowledge of conditions and limitations in the facility license.

Technical Reference(s): H4 pages 93 - 94.

Proposed references to be provided to applicants during examination: Table 2.2 of H4 Learning Objective: P8182L-002 Obj. 10C Question Source:

Bank # _ X _

Modified Bank # _

New _______

Question History: Last NRC Exam ___N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __2__

Comments:

Page 19

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

85. P8197L-012 225/E03 EA2.1/3.4/4.2/17/YES/P8100/1ES-1.1//2014 ILT NRC S85 Given the following conditions:

- A small break LOCA has occurred on Unit 1.

- The crew is in 1E-1, Loss of Reactor or Secondary Coolant.

Based on the following information:

The Shift Supervisor will transition to ___________________________.

1ES-0.2, SI Termination 1ES-0.1, Reactor Trip Recovery 1ECA-1.2, LOCA Outside Containment 1ES-1.1, Post LOCA Cooldown and Depressurization A.

B.

C.

D.

Page 20

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires the knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes subcooling is sufficient and CTMT is NOT adverse; however, in this case CTMT is adverse, also RCS pressure and PRZR level are not sufficient to terminate SI.

b. Incorrect. Plausible if examinee incorrectly believes a SI did NOT occur and a transition to 1ES-0.1 is warranted.

c. Incorrect. Plausible if examinee incorrectly believes the LOCA is outside containment; however, the LOCA is inside containment as indicated from Containment Pressure and water level.

d. Correct. For the given conditions, per step 20 of 1E-1, a transition to 1ES-1.1 is appropriate.

K/A Number:

E03 LOCA Cooldown and Depressurization EA2.1:

Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization):

Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Technical Reference(s): 1E-1 Info Page & page 11, 1ES-1.1 info page & pages 2 & 3.

Proposed references to be provided to applicants during examination: None Learning Objective: P8197L-012 Obj. 17 Question Source:

Bank # P8197L-012 225 Modified Bank #

New Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content:

55.41 55.43 5 Comments:

Page 21

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

86. P8197L-014 116/003 2.1.7/4.4/4.7/50/YES/P8100/1FR-I.3//2014 ILT NRC S86 Given the following conditions:

- Both RCPs are secured.

- Prior to both RCPs being secured, 12 RCP seal cooling was lost.

- A status evaluation of 12 RCP has NOT been completed.

- The crew is on step 8 of 1FR-I.3, Response to Voids in Reactor Vessel.

- RCS pressure is 1185 psig.

- RCS cold leg temperatures are 500°F.

- RCS subcooling is 65°F.

- Containment Pressure is 3.8 psig.

- RVLIS Full Range is 70% and lowering.

- Pressurizer level is 92%.

- The water in the pressurizer is saturated.

- 1FR-I.3 is provided.

What is the NEXT action the Shift Supervisor will direct?

Start 11 RCP.

Start 12 RCP.

Block Safety Injection.

Dump steam as necessary.

A.

B.

C.

D.

3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Correct.

b. Incorrect. Plausible if examinee disregards Caution on top of page 6 and also because 12 RCP is the preferred RCP.

c. Incorrect. Plausible if examinee incorrectly believes containment is adverse and goes to step 12 per RNO on step 9a. However, containment is not adverse because it is below 5 psig.

d. Incorrect. Plausible as step 14 will require dumping steam if subcooling is not greater than 70F; however, this would NOT be the next step performed.

Page 22

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

003 Reactor Coolant Pump System (RCPS) 2.1.7:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation Technical Reference(s): 1FR-I.3 pages 5 - 7.

Proposed references to be provided to applicants during examination: 1FR-I.3 (no bases)

Learning Objective: P8197L-014 Obj. 50 Question Source:

Bank #

Modified Bank #

New X Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content:

55.41 55.43 5 Comments:

Page 23

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

87. P8180L-004 024/006 A2.13/3.9/4.2/4B/YES/P8100/1C18 AOP2//2014 ILT NRC S87 Given the following conditions:

- Unit 1 is in Mode 3, Hot Standby.

- RCS Tavg is 380°F and stable.

- RCS pressure is 400 psig and stable.

- RCS subcooling is 70°F and stable.

- PT-945, 1 CNTMT PRESS 1 NARROW RANGE, has failed HIGH.

- When going to trip the bistable for PT-945, I&C inadvertently tripped the SI high pressure bistable for PT-947, 1 CNTMT PRESS 3 NARROW RANGE.

- Containment pressure is 0 psig on all Control Room indications.

The Shift Supervisor will enter _____________________________________ and direct a Control Room Operator to ______________________________.

1ES-0.2, SI Termination stop SI pumps ONLY 1ES-0.2, SI Termination stop SI and RHR pumps 1E-0, Reactor Trip or Safety Injection perform Attachment L 1C18 AOP2, Inadvertent Safety Injection When Shutdown place running SI pumps in PULLOUT A.

B.

C.

D.

3-PEO EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Incorrect. Plausible as this procedure would be used for inadvertent SI actuation after transition from 1E-0; however, in this case 1E-0 entry conditions are NOT met and 1ES-0.2 would not be entered.

b. Incorrect. Plausible if examinee incorrectly believes the RHR pumps should be secured and incorrectly believes 1ES-0.2 should be used.

c. Incorrect. Plausible as this procedure would be used for inadvertent SI actuation at power; however, in this case 1E-0 entry conditions are NOT met.

d. Correct.

Page 24

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Statement:

006 Emergency Core Cooling System (ECCS)

A2.13:

Ability to Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Inadvertent SIS actuation Technical Reference(s): 1C18 AOP2 pages 2 & 3, 1E-0 page 2, 1ES-0.2 page 2.

Proposed references to be provided to applicants during examination: None Learning Objective: P8180L-004 Obj. 4B Question Source:

Bank # P8180L-004 024 Modified Bank #

New Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content:

55.41 55.43 5 Comments:

Page 25

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

88. P8180L-009H 048/022 2.2.22/4.0/4.7/9B/YES/P8100/T.S. 3.6.5 BASES//2014 ILT NRC S88 Given the following conditions:

- Unit 2 is at 100% power.

- 21 & 23 CFCUs are running in SLOW and aligned to the DOME.

- 22 & 24 CFCUs are running in FAST and aligned to the GAP/SUP CLG.

- Containment Fan Coil Units (CFCUs) are being shifted per 2C19.2, Containment System Ventilation Unit 2.

- 22 CFCU fails to start in SLOW speed.

- 22 CFCU is re-started and running in FAST speed.

- 23 CFCU fails to start in FAST speed.

- 23 CFCU is re-started and running in SLOW speed.

The 22 CFCU is _______________ AND the 23 CFCU is _______________.

OPERABLE OPERABLE OPERABLE INOPERABLE INOPERABLE OPERABLE INOPERABLE INOPERABLE A.

B.

C.

D.

1-B EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the knowledge of TS bases that is required to analyze TS required actions and terminology.

Justifications:

a. Incorrect. Plausible as 23 FCU is operable based on slow speed capability; however, in this case 22 FCU is NOT operable.

b. Incorrect. Plausible if examinee incorrectly believes that fast speed operation is required by safety analysis instead of slow speed.

c. Correct.

d. Incorrect. Plausible as 22 FCU is inoperable; however, in this case 23 FCU is still operable because it can fulfill its required slow speed function.

Page 26

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

022 Containment Cooling 2.2.22:

Knowledge of limiting conditions for operations and safety limits.

Technical Reference(s): T.S. 3.6.5 Bases Proposed references to be provided to applicants during examination: None Learning Objective: P8180L-009H Obj. 9B Question Source:

Bank #:

Modified Bank #: ____

New: X__

Question History: Last NRC Exam: N/A Question Cognitive Level:

Memory or Fundamental Knowledge: X Comprehension or Analysis:

10 CFR Part 55 Content:

55.41:

55.43: 2 Comments:

Page 27

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

89. P8182L-002 136/039 A2.03/3.4/3.7/3J/YES/P8100/PINGP 1576//2014 ILT NRC S89 Given the following conditions:

Question continued on next page.

Page 28

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

89. P8182L-002 136/039 A2.03/3.4/3.7/3J/YES/P8100/PINGP 1576//2014 ILT NRC S89 Question continued from previous page.

- Radiation levels are expected to remain as shown for at least 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

- PINGP 1576, Emergency Classification Tables, is provided.

Based ONLY on the ERCS information given, which of the following EAL classifications will the Shift Manager declare?

RU 1.2 RU 2.2 RA 1.2 RS 1.1 A.

B.

C.

D.

3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(6) Procedures and limitations involved in alterations in core configuration. This question requires evaluating emergency classifications based on core conditions.

Justifications:

a. Incorrect. Plausible as conditions are met for this NUE; however, in this case the indications given exceed the ALERT threshold.

b. Incorrect. Plausible as conditions are met for this NUE; however, in this case the indications given exceed the ALERT threshold.

c. Correct.

d. Incorrect. Plausible if examinee incorrectly believes indications given meet the SAE threshold.

Page 29

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

039 Main and Reheat Steam System (MRSS)

A2.03:

Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Indications and alarms for main steam and area radiation monitors (during SGTR)

Technical Reference(s): PINGP 1576 Proposed references to be provided to applicants during examination: PINGP 1576 Learning Objective: P8182L-002 Obj. 3J Question Source:

Bank #

Modified Bank #

New X Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis X

10 CFR Part 55 Content:

55.41 55.43 6 Comments:

Page 30

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

90. P8182L-002 137/073 A2.02/2.7/3.2/5C/YES/P8100/D5.1 AOP1//2014 ILT NRC S90 Given the following conditions:

- Both Units are at 100% power.

- Fuel handling is in progress in the Spent Fuel Pool (SFP) area.

- A fuel assembly is dropped.

- R-25, Spent Fuel Pool Air Monitor A, fails low.

- R-31, Spent Fuel Pool Air Monitor B, is in alarm.

The Shift Supervisor will enter __________________________________ and direct _____________________________________________________.

D5.1 AOP1, SFP Area Evacuation - Non-Refueling raising the R-25 test current signal at the radiation monitor racks D5.1 AOP1, SFP Area Evacuation - Non-Refueling placing 122 Spent Fuel Special & 21 In-Service Purge Exhaust Fan in START D5.2 AOP4, SFP Area Evacuation - Refueling raising the R-25 test current signal at the radiation monitor racks C47047 R-25, Spent Fuel Pool Air Monitor A placing 122 Spent Fuel Special & 21 In-Service Purge Exhaust Fan in START A.

B.

C.

D.

3-SPK EXPLANATION:

This question is linked to 10 CFR 55.43(b)(7) Fuel handling facilities and procedures. This question requires knowledge of the Refuel Floor SRO responsibilities Justifications:

a. Correct.

b. Incorrect. Plausible as the SS will enter D5.1 AOP; however, the SS will not direct starting 122 SFP Special Fan.

c. Incorrect. Plausible as the SS will direct raising R-25 test current signal; however, the SS will not enter D5.2 AOP4.

d. Incorrect. Plausible as entering C47047 R-25 would occur; however, that procedure will not direct starting 122 SFP Special.

Page 31

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

073 Process Radiation Monitoring (PRM) System A2.02:

Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Detector failure Technical Reference(s): D5.1 AOP1 pages 3 - 5, D5.2 AOP1 pages 1 -5, D5.2 AOP4 page 3, C47047 (R25) pages 1 & 2.

Proposed references to be provided to applicants during examination: None Learning Objective: P8182L-002 Obj. 5C Question Source:

Bank #:

Modified Bank #:

New: X Question History: Last NRC Exam: N/A Question Cognitive Level:

Memory or Fundamental Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content:

55.41:

55.43: 7 Comments:

Page 32

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

91. P8170L-006 058/011 A2.03/3.8/3.9/10C/YES/P8100/T.S. 3.3.1/C51/2014 ILT NRC S91 Given the following conditions:

- Unit 1 is at 100% power.

- 1LT-428, Blue Channel Pressurizer LEVEL, fails LOW.

- Actions per 1C51.3, Instrument Failure Guide, are in progress.

- Bistables cannot be tripped within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

- T.S. LCO 3.3.1 is provided.

Technical Specification LCO 3.3.1 requires Unit 1 thermal power to be reduced to...

MODE 3 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

less than 10% in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

less than 10% in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

A.

B.

C.

D.

3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for TS 3.3.1.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes T.S. LCO 3.3.1 does not apply and therefore, T.S.

LCO 3.0.3 must be entered, which requires to be in Mode 3 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

b. Incorrect. Plausible if examinee incorrectly misapplies T.S. and believes Condition E should be entered.

c. Correct.

d. Incorrect. Plausible if examinee incorreclty applies the NOTE for Condition K of LCO 3.3.1 and believes that 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months may be added to the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months requirement.

Page 33

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

011 Pressurizer Level Control System (PZR LCS)

A2.03:

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of PZR level Technical Reference(s): T.S. LCO 3.0.3, T.S. LCO 3.3.1 Proposed references to be provided to applicants during examination: T.S. LCO 3.3.1 Learning Objective: P8170L-006 Obj. 10C Question Source:

Bank # _______

Modified Bank # _____

New ___X___

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 __ ___

55.43 __2___

Comments:

Page 34

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

92. P8184L-002 081/2.2.12/3.7/4.1/10D/YES/P8100/T.S. 3.3.1/SP-1198/2014 ILT NRC S92 Given the following conditions:

- SP-1198, NIS Power Range Startup Test, is being performed.

- N42, PR Nuclear Instrument, trip function setpoint was found set at 37%.

- The as left setpoint was recorded as 25.2%.

- T.S. LCO 3.3.1 and SP 1198 are provided.

What is the status of N42 and SP-1198?

N42 SP-1198 AS FOUND ACCEPTANCE OPERABILITY CRITERIA OPERABLE MET OPERABLE NOT MET INOPERABLE MET INOPERABLE NOT MET A.

B.

C.

D.

3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of surveillance requirements.

Justifications:

a. Incorrect. Plausible as the as found status of N42 is operable; however, the acceptance criteria for N42 as left is not met.

b. Correct.

c. Incorrect. Plausible if examinee incorrectly believes as found status must be below 25% and therefore is inoperable. Also, if examinee incorrectly believes that as left must be below 40% and the acceptance criteria is therefore met.

d. Incorrect. Plausible if the examinee incorrectly believes the as found and as left status must be below 25%.

Page 35

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

015 Nuclear Instrumentation 2.2.12 Knowledge of surveillance procedures.

Technical Reference(s): T.S. 3.3.1, T.S. 3.3.1 Bases pgs 1-3, SP-1198.

Proposed references to be provided to applicants during examination: T.S. 3.3.1, SP-1198.

Learning Objective: P8184L-002 Obj. 10D.

Question Source:

Bank #: P8184L-002 081 __________

Modified Bank #:

New:

Question History: Last NRC Exam: N/A Question Cognitive Level:

Memory or Fundamental Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content:

55.41:

55.43: 2 Comments:

Page 36

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

93. P8182L-002 089/072 A2.01/2.7/2.9/10C/YES/P8100/TS LCO 3.3.3//2014 ILT NRC S93 Given the following conditions:

- Both units are at 100% power.

- 1R-48, U1 CNTMT HI RNG AREA MON B, is OOS for the past 10 days.

- T.S. LCO 3.3.3 Condition A was entered 10 days ago.

- 1R-49, U1 CNTMT HI RNG AREA MON A, fails LOW due to a loss of power.

- T.S. LCO 3.3.3 is provided.

The Shift Supervisor will enter Technical Specification LCO 3.3.3 Condition ____ and perform required actions.

C D

H I

A.

B.

C.

D.

3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for TS 3.3.3.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes that a second channel inoperable causes Condition A required action not to be met.

b. Correct. Per Table 3.3.3-1, two Containment High Range Area Monitors are required to be operable.

With two required channels inoperable, Condition D is entered with required action of restoring one channel to operable status within 7 days.

c. Incorrect. Plausible if examinee implements T.S. 3.3.3 similiar to TS 3.3.1 and 3.3.1 by going to the table first and entering the condition listed in the table (in this case "I") AND uses containment pressure instead of containment area radiation.

d. Incorrect. Plausible if examinee implements T.S. 3.3.3 similiar to TS 3.3.1 and 3.3.1 by going to the table first and entering the condition listed in the table (in this case "I").

Page 37

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

072 Area Radiation Monitoring (ARM) System A2.01:

Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system-and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Erratic or failed power supply Technical Reference(s): T.S. LCO 3.3.3 Proposed references to be provided to applicants during examination: T.S. LCO 3.3.3 Learning Objective: P8182L-002 Obj. 10C Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __2___

Comments:

Page 38

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

94. P8184L-002 103/2.1.7/4.4/4.7/10C/YES/P8100/T.S. 3.2.4//2014 ILT NRC S94 Given the following conditions:

- Unit 1 is at 96% power.

- Quadrant Power Tilt Ratio (QPTR) is 1.05.

- T.S. LCO 3.2.4 is provided.

Technical Specification LCO 3.2.4 requires thermal power to be reduced to less than _____ within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of QPTR determination.

91%

87%

85%

81%

A.

B.

C.

D.

3 - SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(2) Tech Specs. The question can NOT be answered by solely knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS actions OR by solely knowing the LCO "above the line" information OR by solely knowing TS Safety Limits. The question requires the application of required actions for TS 3.2.4.

Justifications:

a. Incorrect. Plausible if examinee incorrectly subtracts 1.02 from 1.05 instead of 1.00 from 1.05.

100 - [3(3)]% = 91%

b. Incorrect. Plausible if examinee incorrectly reduces power by the correct amount but from the current power level instead of RTP and incorrectly subtracts 1.02 from 1.05 instead of 1.00 from 1.05.

96% - [3(3)]% = 87%

c. Correct.

100% - [5(3)]% = 85%

d. Incorrect. Plausible if examinee incorrectly reduces power by the correct amount but from the current power level instead of RTP.

96% - [5(3)]% = 81%

Page 39

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

Conduct of Operations 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Technical Reference(s): T.S. 3.2.4 Proposed references to be provided to applicants during examination: T.S. 3.2.4 Learning Objective: P8184L-002 Obj. 10C Question Source:

Bank #:

Modified Bank #: P8184L-002 099 New:

Question History: Last NRC Exam: N/A Question Cognitive Level:

Memory or Fundamental Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content:

55.41:

55.43: 2 Comments:

Page 40

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

95. P9150L-024 038/2.2.5/2.2/3.2/2/YES/P8100/FP-E-SE-03/FP-E-MOD-03/2014 ILT NRC S95 Given the following conditions:

- Unit 2 is at 100% power.

- 2HD-4-3, SCAV STM FROM 1B MSR TO 25B FW HEATER, has a body to bonnet leak.

- The leak cannot be stopped by torquing the body to bonnet studs.

- Furmanite will perform an INJECTION LEAK SEAL to stop the body to bonnet leak.

- 2HD-4-3 will be replaced during the Unit 2 scheduled outage.

- The Unit 2 scheduled outage is in 135 days.

Which of the following is REQUIRED to perform the Furmanite repair to 2HD-4-3?

50.59 Temporary Screening Modification NO NO NO YES YES NO YES YES A.

B.

C.

D.

1-F EXPLANATION:

This question is linked to 10 CFR 55.43(b)(3) Facility licensee procedures required to obtain authority for design and operating changes to the facility. This question requires knowledge of 10CFR50.59 screening process and the administrative process for temporary modifications.

Justifications:

a. Incorrect. Plausible if examinee is not familiar with the requirements for 50.59 screening and t-mods and incorrectly believes neither is required.

b. Incorrect. Plausible as a T-Mod is required; however, a 50.59 screening is also required.

c. Incorrect. Plausible as a 50.59 screening is required; however, a T-Mod is also required.

d. Correct.

Page 41

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

Equipment Control 2.2.5 Knowledge of the process for making design or operating changes to the facility.

Technical Reference(s): FP-E-SE-03 page 19-23, FP-E-MOD-03 page 3.

Proposed references to be provided to applicants during examination: None Learning Objective: P9150L-024 Obj. 2 Question Source:

Bank #

Modified Bank # __________

New X Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehension or Analysis 10 CFR Part 55 Content:

55.41 55.43 3 Comments:

Page 42

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

96. P7410L-002 048/2.2.44/4.2/4.4/3/YES/P8100/PINGP 1576/F3-2/2014 ILT NRC S96 Given the following conditions:

- PINGP 1576, Emergency Action Level Matrix, is provided.

Based ONLY on the ERCS STAT screen above, what is the status of the following fission product barriers per the Emergency Action Level Matrix?

Fuel Cladding RCS Containment POTENTIAL LOSS LOSS INTACT POTENTIAL LOSS INTACT INTACT INTACT LOSS INTACT INTACT INTACT POTENTIAL LOSS A.

B.

C.

D.

Page 43

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(6) Procedures and limitations involved in alterations in core configuration. This question requires evaluating emergency classifications based on core conditions.

Justifications:

a. Correct. A potential loss of fuel cladding barrier is indicated by the Core Cooling CSF being orange and by RVLIS full range less than 40% with both RCPs stopped. A loss of RCS is indicated by subcooling being less than 35F (containment is adverse because pressure is greater than 5 psig). Containment is intact because pressure is less than 46 psig. Also both trains of depressurization equipment (CFCU &

CS) are operating.

b. Incorrect. Plausible as there is a potential loss of fuel cladding; however, the RCS is NOT intact.

c. Incorrect. Plausible as there is a loss of RCS; however, fuel cladding is NOT intact.

d. Incorrect. Plausible if examinee incorrectly believes RVLIS level is ok because it is green instead of red on the STAT screen and misses that Core Cooling is orange. Also, if examinee does not notice the RCS is at saturation indicating a LB LOCA condition. Also, if examinee incorrectly believes containment is potentially loss because containment pressure is above 23 psig; however, this will only cause containment to be potentially loss if there is less than one full train of depressurization equipment running.

According to the STAT screen in this case, both trains are operating.

K/A Number:

Equipment Control 2.2.44:

Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Technical Reference(s): PINGP 1576, F3-2 Proposed references to be provided to applicants during examination: PINGP 1576 Learning Objective: P7410L-002 Obj 3 Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 __ ___

55.43 __6___

Comments:

Page 44

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

97. P8182L-001C 137/2.3.6/2.0/3.1/6/YES/P8100/H4 ODCM/C21.3-10.1/2014 ILT NRC S97 Given the following conditions:

- Preparations for a gaseous radioactive waste release from 121 Low Level Gas Day Tank are in progress.

- C21.3-10.1, Releasing Radioactive Gas from 121 Low Level Gas Decay Tank, is provided.

The Shift Supervisor will _________________ the release because _______________________________________.

NOT approve precipitation is occurring NOT approve wind speed is greater than 10 mph approve wind direction is from 110° approve cooling towers are not in operation A.

B.

C.

D.

Page 45

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. This question requires knowledge of the process for gaseous release approvals.

Justifications:

a. Incorrect. Plausible if examinee incorectly believes that precipitation is occurring.

b. Incorrect. Plausible as wind speed is greater than 10 mph; however, the permit would be approved.

c. Correct. Four limits apply in C21.3-10.1: Permit SHALL NOT be approved if ALL 3 of the following are met: 1) ANY CT in operation 2) Wind direction between 330-360 or 0-60 AND 3) wind speed < 10 mph; also 4) permits should NOT be approved if precipitation is occurring.

With the given information the examinee must identify that CTs are in operation based on U1 in Mode 1 during the April - Oct summer months, wind direction is from 110, wind speed is 15 mph, AND it is NOT raining.

d. Incorrect. Plausible as the permit will be approved based on wind speed and direction; however, cooling towers are in operation but would not be a factor.

K/A Statement:

Radiation Control 2.3.6 Ability to approve release permits.

Technical Reference(s): H4 ODCM pages 32 & 33, C21.3-10.1 page 3.

Proposed references to be provided to applicants during examination: C21.3-10.1 Learning Objective: P8182L-001C Obj. 6 Question Source:

Bank # __P8182L-001C 005__

Modified Bank # _______

New _______

Question History: Last NRC Exam ___N/A_____

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 _4___

Comments:

Page 46

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

98. P8197L-013 114/2.3.14/3.3/4.0/7/YES/P8100/1E-3 STEP 7 BASES//2014 ILT NRC S98 Given the following conditions:

- The crew is on step 7, Initiate RCS Cooldown, of 1E-3, Steam Generator Tube Rupture.

The Shift Supervisor will direct the crew to initiate RCS cooldown by releasing steam via the ________________________ from the _____________ steam generator because this path will _____________________________________________.

Condenser Steam Dump ruptured prevent pressurizing the steam generator to RCS pressure Condenser Steam Dump intact minimize radiological release PORV ruptured prevent radiological contamination of the Main Steam system PORV intact ensure adequate capacity to cooldown the RCS A.

B.

C.

D.

1-B This question is linked to 10 CFR 55.43(b)(4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. This question requires knowledge interpretation of radiation and activity readings as they pertain to selection of emergency procedures.

Justifications:

a. Incorrect. Plausible as steam will be released from the ruptured steam generator as needed; however, in this case it is to prevent overpressurization of the SG. The intact SG would be used for RCS cooldown.

b. Correct.

c. Incorrect. Plausible as the ruptured steam generatorwill maintain pressure using PORVs as needed AND radiological concerns are appropriate;however, in this case PORVs are to prevent overpressurization of the SG and cooldown would be from the intact SG to condenser.

d. Incorrect. Plausible as the intact steam generator is the appropriate choice for the cooldown; however, in this case the PORVs would not be used; also if examinee incorrectly believes capacity of the steam dumps is not adequate.

Page 47

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Statement:

Radiation Control 2.3.14:

Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Technical Reference(s): 1E-3 Step 7 Bases pages 5 & 6.

Proposed references to be provided to applicants during examination: None Learning Objective: P8140L-242 Obj. 7 Question Source:

Bank #: P8197L-013 114 Modified Bank #:

New:

Question History: Last NRC Exam 2010 ILT NRC Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehension or Analysis 10 CFR Part 55 Content:

55.41 55.43 4 Comments:

Page 48

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR

99. P8197L-011 116/2.4.19/3.4/4.1/7/YES/P8100/1ECA-0.0/SWI O-10/2014 ILT NRC S99 Given the following conditions:

- A loss of all AC power has occurred on Unit 1.

- The crew is performing step 18 of 1ECA-0.0, Loss of All Safeguards AC Power.

- Bus 15 is locked out.

- Bus 16 is locked out.

- Containment pressure is 0.6 psig and stable.

- Containment radiation is 1.6 x 100 R/H.

- 11 SG indications are as follows:

- NR level is 6% and stable.

- WR level is 58% and stable.

- AFW flow is 50 gpm.

- 12 SG indications are as follows:

- NR level is 55% and rising rapidly.

- WR level is 62% and rising.

- AFW flow is 50 gpm.

- 1ECA-0.0 is provided.

The NEXT action the Shift Supervisor will direct is...

establish Battery Room Cooling per step 19.

isolate AFW flow to 12 SG per step 17.b RNO.

raise total AFW flow to 200 gpm per step 17.a RNO.

raise ONLY 11 SG AFW flow to 150 gpm per step 17.b.

A.

B.

C.

D.

Page 49

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 3-SPR EXPLANATION:

This question is linked to 10 CFR 55.43(b)(5) Assessment and selection of procedures. The question can NOT be answered by solely knowing "systems knowledge" OR by solely knowing immediate operator actions OR by solely knowing entry conditions for AOPs or plant parameters that require direct entry to MAJOR EOPs OR by solely knowing the purpose overall sequence of events or overall mitigative strategy of a procedure. The question requires assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

Justifications:

a. Incorrect. Plausible if examinee does not understand that step 17 is a continuous action step.

b. Correct. Step 17 is annotated as a continous action step with a triangle. Since 12 SG level is rising uncontrollable, Step 17.b. RNO directs the operator to isolate AFW flow to the ruptured SG.

c. Incorrect. Plausible if the examinee incorrectly believes that containment is adverse and uses Attachment E values for required SG levels in step 17.a.

d. Incorrect. Plausible if examinee incorrectly believes 11 SG level is too low and level needs to be raised in 11 SG.

K/A Number:

Emergency Procedures / Plan 2.4.19 Knowledge of EOP layout, symbols, and icons.

Technical Reference(s): 1ECA-0.0 pages 17 & 18, SWI O-10 page 10.

Proposed references to be provided to applicants during examination: All steps of 1ECA-0.0, but no background information.

Learning Objective: P8140L-247 Obj. 7 Question Source:

Bank # _______

Modified Bank # _______

New __X_____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge Comprehension or Analysis

__X__

10 CFR Part 55 Content:

55.41 _____

55.43 __5___

Comments:

Page 50

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR 100. P9150L-011 003/2.4.40/2.7/4.5/5/YES/P8100/F3-2//2014 ILT NRC S100 Given the following conditions:

- The Control Room observes indications of a LOCA at 0200.

- The Shift Manager declares an Alert at 0205.

- The PINGP-577 is completed at 0210.

What are the LATEST ALLOWABLE notification times?

States and Counties NRC 0215 0300 0220 0305 0220 0310 0225 0310 A.

B.

C.

D.

1-F EXPLANATION:

This question is linked to 10 CFR 55.43(b)(1) Conditions and limitations in the facility license. This question requires knowledge of government notification requirements per 10CFR50.72.

Justifications:

a. Incorrect. Plausible if examinee incorrectly believes the clock starts for notifications at time of accident instead of time of classification.

b. Correct. 10CFR50, App. E requires state and local government notification to be made within 15 minutes from time of emergency declaration. Also, 10CFR50.72 (a)(3) requires NRC notification to be made immediately after the notification of the state and local governments and not later than one hour after the emergency declaration.

c. Incorrect. Plausible if examinee incorrectly believes the NRC notifications clock starts when the PINGP-577 is completed.

d. Incorrect. Plausible if examinee incorrectly believes the offsite notification and NRC notification clock starts when the PINGP-577 is completed.

Page 51

2014 NRC INITIAL LICENSE WRITTEN EXAM SENIOR REACTOR OPERATOR K/A Number:

Emergency Procedures / Plan 2.4.40 Knowledge of SRO responsibilities in emergency plan implementation.

Technical Reference(s): F3-2 pages 9 -11.

Proposed references to be provided to applicants during examination: None Learning Objective: P9150L-011 Obj. 5 Question Source:

Bank # P9150L-011 003 Modified Bank #

New Question History: Last NRC Exam N/A Question Cognitive Level:

Memory or Fundamental Knowledge X

Comprehension or Analysis 10 CFR Part 55 Content:

55.41 55.43 1 Comments:

You have completed the test!

Page 52

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