ML033510582

4.4.23 RCS Operational LEAKAGE

LCO 3.4.43 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE.;

c. 10 gpm identified LEAKAGE;

d. 576 gallons per day total primary to secondary LEAKAGE through all

steam generators (SGs); and

e. 150 gallons per day primary to secondary LEAKAGE through any

one SG.

APPLICABILITY: MODES 2 , 2,3, and 4.

CONDITION COMPLETION TIME

A. RCS LEAKAGE not A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

within limits for reasons within limits.

other than pressure

boundary LEAKAGE.

B. Required Action and B.l Be In MODE 3.

associated Completion

Time of Condition A not AND

met.

8.2 Be in MODE 5.

-

Pressure boundary

LEAKAGE exists.

~

Catawba Units I' and 2 3.4.13-1 Amendment Nos. 1731165

RCS Operational LEAKAGE

3.4.13

SURVEILLANCE REQUIREMENTS

Verify RCS Operational LEAKAGE within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

performance of RCS water inventory balance.

SR 3.4.13.2 Verify steam generator tube integrity is in accordance In accordance with

with the Steam Generator Tube Surveillance Program.

Generator Tube

Catawba Units f and 2 3.4.$3-2 Amendment Nos. 173/%65

Duke Power Company 'rocedurc No.

Catawba Nuclear Station PT/1/Af460O/OU9

levision No.

067

Loss of Operator Aid Computer

~

Electronic Reference No.

* * * * * * * * * UNCONTRQLLEDFORPRINT * * * * * * * * * *

(ISSUED) - PDF Format

PT/~/A/~GCIO/OW

Page 2 of 8

Loss of Operator Aid Computer

1. Purpose

To document Technical Specifications requirements normally perfonned by the Operator Aid

Computer in the event that the unit Operator Aid Computer is out of service.

2. Reference

2.1 OP/1/N6700/003 (Operation with the Operator Aid Computer Out of Service)

2.2 Catawba TS and SLC Requirements:

2.2.1 TS3.1.4

2.2.2 TS 3.1.6

2.2.3 TS 3.2.3

2.2.4 TS 3.2.4

2.2.5 TS 3.3.1

2.2.6 TS 3.4.2

2.2.7 TS 3.4.5

2.2.8 TS 3.4.13

2.2.9 TS 3.4.15

2.2.10 SR3.1.4.1

2.2.11 SR 3.1.6.2

2.2.12 SR 3.2.3.1

2.2.13 SR 3.2.4.1

2.2.14 SR3.4.2.1

2.2.15 SLC 16.5-7

2.2.16 TS 3.6.3

2.2.17 TS 3.1.3

PT/!'1/A/4600/009

Page 3 of 8

3. Time Required

3.1 Manpower - One Operator

3.2 Time - Until the Operator Aid Computer is restored to service.

3.3 Frequency When the Operator Aid Computer is out of service.

4. Prerequisite Tests

None

'5. Test Equipment

None

6. Limits and Precautions

6.1 If an acceptance criteria is NOT met, the Operations Shift Manager and the Operator at

the Controls should be notified immediately.

6.2 If the unit status or system condition prevents the performance of a surveillance item, the

item should be noted on the affected data sheet with an explanation and the Operations

Shift Manager and the Operator at the Controls should be notified immediately.

7. Required Unit Status

None

8. Prerequisite System Condition

__- Verify the Operator Aid Computer is out of service.

9. Test Method

A visual inspection of various system instrumentation wiil be made until the computer is

returned to service.

10. Data Required

Complete Enclosures as required.

11. Acceptance Criteria

No data taken shall exceed limits listed on the Enclosures.

PT/l/A/4600/009

Page 4 of 8

12. Procedure

12.1 IF in Modes 5

_. 6, EVERY 15 MINUTES document the critical core parameters listed

on Enclosure 13.1 (Critical Core Parameters Sheet) (Reference OEP).

12.2 -IF Start Up OfND System During Plant Cooldown (OP/I/A/6200/004) is in progress

~AND KCHX Maximized Cooling Temperature Monitoring is being performed, within

15 minutes and every 15 minutes thereafter record parameters on Enclosure 13.2 (KCHX

Maximized Cooling Temperature Monitoring).

12.3 EVERY 15 MINUTES record on Enclosure 13.3 (Auxiliary Building Ventilation Supply

Unit Status) the status of the Auxiliary Building Ventilation System supply units.

12.4 IF in Modes 1-4, within 30 minutes of Loss of OAC and once per hour thereafter, verify

and record on Enclosure 13.4 (Ventilation Unit Condensate Drain Tank Input Rate

Determination) that the rate of increase in VUCDT level is 1% per hour. (TS 3.4.13

and 3.4.15)

12.5 -

IF in Modes 1-4, within 30 minutes of Loss ofOAC, begin performing Enclosure 13.5

(Conbinment Floor and Equipment Sumps Input Rate Determination) to verify input to

the Containment Floor and Equipment Sump is less than 1 gpm. (TS 3.4.13 and 3.4.15)

12.6 -

IF in Modes 1-4, within 30 minutes of Loss of OAC and once per hour thereafter, verify

and record on Enclosure 13.6 (IEMF-38 Delta Count Rate Determination) that the

change in count rate on 1EMF-38 is < 750 cpm in one hour. (TS 3.4.13 and 3.4.15)

12.7 -

IF in Modes 1-4, within 30 minutes of Loss of QAC and once per hour thereafter, verify

and record on Enclosure 13.4 (1EMF-39 Delta Count Rate Determination) that the

change in count rate on IEMF-39 is < 6700 cpm in one hour. (TS 3.4.13 and 3.4.15)

12.8 -

IF ALL the following conditions exist (Reference SK 3.4.2.1):

Reactor Critical

T~~(i<561'F

T ~ - TAU^

F H i h Alarm Present, Annunciator 1AD2 A/4

EVERY 30 MbWTES verify Reactor Coolant loops TAVG ? 55 1"F by completing

Enclosure 13.8 (TAVC; Data Sheet).

NOTE: The Y C Operable But Degraded Condition is normally active during the winter months

based on Lake Wylie and SNSWP temperatures.

12.9 the YC Operable But Degraded Condition is active, perform Enclosure 13.9 (YC

Operable But Degraded Temperature Monitoring).

PT/1iAi4600/009

Page 5 of 8

12.10 E both trains of the plasma display monitor are inoperable in Modes 1-6, EVERY 60

MINUTES or after 10% change in power, complete Enclosure 13.10 (Subcooling Data

Sheet) to monitor subcooling margin.

12.1I Unit 1 net generation CANNOT be obtained from the Unit 1 operator aid computer,

perform the following:

12.1 1.1 At the top of the first hour during loss of OAC, notify SOC that they will not

be getting station or unit MWH hourly values from both CNS units.

Person notified

12.11.2 EVERY HOUR on the HOUR complete Enclosure 13.1 1 (Eiectrical Data

Sheet).

NOTE: I. If pressure (primary and secondary) are verified < 200 psig, then temperatures are

-

NOT required to be taken nor recorded.

2. Use a calibrated pyrometer to obtain SiG shell temperatures.

12.12 NC Tc is > 80°F a NC pump is operating, then the secondary side temperature

is > 80°F and documentation of shell temps is NOT necessary. E in Modes 5 , 6 No

Mode, EVERY 60 MINUTES complete Enclosure 13.12 (Steam Generator Data Sheet)

(Reference SLC 16.5-7).

12.I3 in Mode I less than 50% rated power, prior to exceeding 50% rated power and

every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter, with the AFD monitor alarm inoperable, monitor and log the

indicated Axial Flux Difference for each operable excore channel on Enclosure 13.13

(Axial Flux Difference (Yd Flux) Following Loss of AFD Monitor A l m ) . (Reference

SR 3.2.3.1 and TS 3.2.3).

12.14 E in Mode 1 AND 2 50% rated power, once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter

with the AFD monitor alarm inoperable, monitor and log the indicated Axial Flux

Difference for cach operable excore channel on Enclosure 14.13 (Axial Flux Difference

(%A Flux) Following Loss of AFD Monitor Alarm). (Reference SR 3.2.3.1 and TS

3.2.3).

12.15 E in Modes I 2, EVERY 4 HOURS verify by signing off on Enclosure 13.14 (Rod

Verification Checklist) that the Digital Rod Position indication for all rods are within

f 12 steps of their group step counter denland position and operable (Reference SK

3.1.4.1).

12.16 in Mode 1 2 E D KEFF2 1.0, EVERY 4 HOURS verify and record on Enclosure

13.15 (Rod Insertion Limit Checksheet) that each coetrol bank of rods is above the rod

insertion limit (Reference SR 3.1.6.2).

PT/1/A/4600/009

Page 6 of 8

12.17 E in Modes 1,2,3, Mode 4, when steam generators are being used for heat removal,

EVERY 4 HOURS record CA suction source temperatures measured locally using a

calibrated Keithley 872 digital thermometer, Type J or its equivalent, as required, per

Enclosure 13.16 (CA Suction Source Temperature Monitoring Data)

12.18 E in Modes 1-4, within 4 HOURS and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter, monitor the CF

containment isolation valves N2 accumulator pressures on Enclosure 13.17 (CF

Containment Isolation Valve N2 Accumulator Pressure Monitoring).

12.19 in Modes 1-4, EVERY 6 HOURS, document data needed for primary to secondary

leakage calculation on Enclosure 13.18 (Primary to Secondary Leakage Calculation

Data) and provide data to Chemistry. Notify Secondary Chemistry to perform

PT/l/B/4600/028 (Determination Of Steam Generator Tube Leak Rate For Unit 1).

12.20 E Auxiliary Spray is being used for pressurizer pressure control, EVERY 12 HOURS

complete Enclosure 13.19 (Pressurizer Spray AT Data Sheet).

12.21 E in Mode 1 AND above 50% ratedpower, once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

thereafter, document Quadrant Power Tilt Ratio, as calculated by PT/Q/A/4600/08B

(Man. Cal. of Quad. Tilt), in Enclosure 13.1 of PT/l/A/4600/002A (Mode 1 Periodic

Surveillance Items). (Reference SR 3.2.4.1)

12.22 in Modes 1-3, within 12 HOURS of the Loss of OAC and every 12 ~ O U F Sthereafter,

monitor the CA piping surface temperatures. Perform OP!l/A/6250/002, Enclosure 4.12

(Checking Pipe Surface Temperatures).

12.23 E in Modes 1-2, within 12 HOURS of the Loss of OAC and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thercafter,

monitor the Overteniperature Delta T parameters and record on Enclosure 13.20

(Overtemperature Delta T Setpoint Channel Check). (Reference SK 3.3.1.1)

12.24 in Modes 1-4, EVERY 24 HOURS perform a manual leakage calculation of the NC

System in accordance with PT/I/A/4150/0011 (NC Manual Leakage Calculation).

(Reference TS 3.4.15, Required Action A. 1).

12.25 Update Enclosure 13.21 (Chemistry Data Sheet) as information becomes available from

Chemistry.

12.26 WHEN the OAC is returned tu service, notify Shift Work Manager to coordinate with

Lccal IT and Reactor Group Duty Engineer to ensure OAC is updating properly.

12.26.1 Notify SOC that MWM data should be valid at the top of the next hour.

Person notified

12.26.2 Give a copy of Enclosure 13.1 1 to the SSA to assist them in editing the switch

board logs.

PT/1/A/4600/009

Page 7 Of 8

12.27 Evaluate the acceptance criteria by performing one of the following:

_ I 12.27.1 Verify the acceptance criteria specified in Section 11 is met.

~ 12.27.2 @the acceptance criteria is met, perform the following:

CS Notify the Unit/WCC SRO that the acceptance criteria is NOT met.

--J.--.

Unit/WCC SRO Contacted Date Time

61 Initiate a PIP to document the test failure.

0 Document all issues on a procedure discrepancy sheet.

12.28 E any discrepancy is noted during the performance of this test that does NOT keep the

test from meeting the acceptance criteria, it shall be given to the Unit/WCC SRO for

evaluation via a discrepancy sheet.

- 12.29 Submit PT/I/A/4600/009 (Loss of Operator Aid Computer) to the Unit'WCC SRO.

13. Enclosures

13.1 Critical Core Parameters Sheet

13.2 KCHX Maximized Cooling Temperature Monitoring

13.3 Auxiliary Building Ventilation Supply Unit Status

13.4 Ventilation Unit Condensate Drain Tank Input Rate Determination

13.5 Containment Floor and Equipment Sumps Input Rate Determination

13.6 1EMF-38 Delta Count Rate Determination

13.7 IEMP-39 Delta Count Rate Determination

13.8 TAFrG Data Sheet

13.4 YC Operable But Degraded Temperature Monitorhg

PT/1/A/4600:009

Page 8 of 8

13.10 Subcooling Data Sheet

13.1 1 Electrical Data Sheet

13.12 Steam Generator Data Sheet

13.13 Axial Flux Difference (%A Flux) Following Loss of AFD Monitor Alarm

13.14 Rod Verification Checklist

13.15 Rod Insertion Limit Checksheet

13.16 CA Suction Source Temperature Monitoring Data

13.14 CF Containment Isolation Valve N2 Accumulator Pressure Monitoring

13.18 Primaq to Secondary Leakage Calculation Data

13. I9 Pressurizer Spray AT Data Sheet

13.20 Overtemperature Deita T Setpoint Channel Check

13.21 Chemistry Data Sheet

Enclosure 13.1 PT/llA/4600/009

Critical Core Parameters Sheet Page 1 of 1

NOTE: Make additional copies of this sheet as necessary.

Enclosure 13.2 PT/1/A/4hOO/009

KCHX Maximized Cooling Temperature Page 1 of 1

Monitoring

to be > 4YF. KCHX outlet piping temperature measurements are required only when RN

Essential header temperature is less than 4 S T . A pyrometer is required to obtain these

readings.

2. Ifany ofthe following occur, terminate maximized cooling to the KCHXs per

(OP/ 1/A/6200/004)

VCT Outlet temperature decreases to 65°F.

Enclosure 13.3 PTll/Ai4600!009

Auxiliary Building Ventilation Supply Unit Page 1 of 1

Status

for Control Room D/P verification and OP/O/A/6450lOO3

Enclosure 13.4

PT!1 /A/4600/009

Ventilation Unit Condensate Drain Tank Page 1 of I

Input Rate Determination

Acceptance Criteria - Rate of increase must be l%ihour.

NOTE: 1. Either of the following instruments may be used to perform this surveillance, however,

the same instrument should be used for the duration of time the procedure is in effect:

"UNIT 1 VUCDT LEVEL" on Auxiliary Waste Processing Control Panel

(1ELCC0013) (AB-543, MM, 53-54) .

1WLP5771 (AB-543, BB-CC, 50) CA Pump Room (Next to WCDT)

2. If the rate of increase is L l%/hour, the W C D T input rate is 1 gpm. Refer to TS 3.4.13

and TS 3.4.15 and determine if NC System leakage is > 1 gpm.

3. Coordinate with Radwaste Chemistry as required when the VUCDT needs to be pumped

I I I

Enclosure 13.5 PTil/A;4600/009

Containment Floor and Equipment Sumps Page 1 of 4

Input Rate Determination

1. Procedure

I I

NOTE: If any containment floor and equipment sump pump starts during the 10 minute test period

the test results will NOT be valid. The test should be repeated until valid results are

- 1.1 Stop the following sump pumps and place in "Manual":

"Pump 1Al Cont Floor & Equip Sump"

"Pump 1A2 Cont Floor & Equip Sump"

"Pump 1Bl Cont Floor & Equip Sump"

"Pump 1B2 Cont Floor & Equip Sump"

"Manual" position.

1.2 at any time during the performance ofthis test the sump level reaches 2 15", perform the

following:

1.2.1 Place the following sump pumps in "AUTO":

"Pump 1Al Cont Floor & Equip Sump"

"Pump 1A2 Cont Floor & Equip Sump"

"Pump 1B1 Cont Floor & Equip Sunip"

"Pump 1B2 Cont Floor & Equip Sump"

I

NOTE: A level less than 4" is below the calibration range of the Containment Floor and Equipment

Sump level instrumentation, therefore the Leakage Detection Systems must be declared

1.2.2 Verify the affected sump level is lowered to 10" as indicated on 1WLP5740 (Cont

Floor and Equipment Sump A Level) or lWLP5750 (Cont Floor and Equipment

Sump B Level).

1.2.3 Return the following sump pumps to "Manual" and stopped

"Pump IAl Cont Floor & Equip Sump"

"Pump 1A2 Cont Floor & Equip Sump"

"Pump 1B1 Cont Floor & Equip Sump"

"Pump 1B2 Cont Floor I% Equip Sump"

Enclosure 13.5 PT/1/M4600/009

Containment Floor and Equipment Sumps Page 2 of 4

Input Rate Determination

- 1.3 Record initial sump readings on the "Containment Floor and Equipment Sump Inleakage

Rate Log Sheet".

1.4 Once per hour, record sump level readings on the "Containment Floor and Equipment Sump

Inleakage Rate Log Sheet".

1.5 Calculate the leakage rate using the "Sump Volume vs. Level Indication Table".

1.6 Verify leakage is < 1 gpm.

1.7 the input to the Containment Floor and Equipment Sumps is > 1 gpm, perform the

following:

RefertoTS3.4.13 andTS3.4.15.

e Determine if NC System leakage is > 1 gpm.

~ 1.8 WHEN the OAC is returned to service, place the following sump pumps in "AUTO":

"Pump I A1 Cant Floor L?L Equip Sump"

"Pump 1A2 Cont Fioor & Equip Sump"

"Pump 1B1 Cont Floor & Equip Sump"

"Pump 1B2 Cont Floor & Equip Sump"

Enclosure 13.5 PT/1/A/4600/009

Containment Floor and Equipment Simps Page 3 of 4

Input Rate Determination

A Containment Floor and Equipment Sump Inleakage Rate Log Sheet

Enclosure 13.5 PT/1/A/4600/009

Containment Floor and Equipment Sumps Page 4 of 4

Input Rate Determination

NOTE: 1. TI is the data from the previous reading.

Enclosure 13.6

PT/1/A/4600/009

1EMP-38 Delta Count Rate Determination Page 1 of 1

I I I

Acceptance Criteria - Change in count rate < 750 cpm per hour.

NOTE: 1. If the change in count rate per hour is 2 750 cpm, refer to TS 3.4.13 and TS 3.4.15 and

determine if NC System leakage is > 1 gpm.

Enclosure 13.7

PT/1/A/46OO/O09

1EMF-39 Delta Count Rate Determination Page 1 of 1

Ledcage Acceptable

Counts Rate/Hour

Acceptance Criteria - Change in count rate < 6700 cpm per hour.

Enclosure 13.8

PT/hM4600/009

TAW;Data Sheet Page 1 of2

ACCEPTANCE CRITERIA: With the Reactor Critical and TA\,G< 561°F and TWF- TAU^^

W i h a l m present, verify NC System loops TAVG> 551°F. (SR 3.4.2.1)

Loop TAVG

Enclosure 13.8 PT/I /A/4600/009

TAVG Data Sheet Page 2 of 2

Enclosure 13.9 PT/1 /Ai4600/009

YC Operable But Degraded Temperature Page 1 of 2

Monitoring

1. Procedure

- 1.i Record the temperature limits as documented on the NSD 203 Operability Notification

Form for the YC Operable But Degraded Condition.

SNSWP RN Essential Header

Hi Temperature Limit

%-Hi Temperature Limit

1.2 Record the required temperatures on the the "YC Operabie But Degraded Temperature

Monitoring" data sheet.

1.a Determine the monitoring frequency as follows:

1.3.1 any temperature is within 5 O F of either Hi Limit, the temperatures are to be

recorded every hour.

1.32 all of the temperatures are > 5 OF below both Hi Limits, the temperatures are

to be recorded every four hours.

1.4 -

IF any ofthe temperatures exceed either Hi Limit determined in step 1.1, contact the YC

System Engineer for reanalysis of the YC Operable But Degraded Condition.

1.5 -

IF any of the temperatures exceed either Hi-Mi Limit determined in step I. I , the YC

chillers should be considered inoperable unless the YC Operable But Degraded

Conditions are reanalyzed.

1.6 Repeat steps 1.2 through 1.5 at the required frequency determined in step 1.3

Enclosure 13.9 PT/ 1/A/46OO/OO9

YC Operable But Degraded Temperature Page 2 o f 2

Mnnitnring

(1) IfORNP8130 (SNSWP Temp) gauge is inoperable, SNSWP temperature reading may be obtained

per PT/O/A/4400/024 (SNSWP Temperature Monitoring).

Enclosure 13.10

PT/1/A/4600/009

Subcooling Data Sheet Page I of 1

CALCULATION SHEET FOR NC SYSTEM DEGREES SUBCOOLED

ACCEPTANCE CRITERIA:

Subcool limit is IO'F while at power.

30'F while shutdown.

INFORMATION:

NC Pressure - Record lowest indicated system pressure.

-

T-SAT Using NC pressure, determine saturation temperature from the Unit One Revised Data

Book Figure 57 or Figure 58.

Nighest NC Temp Determine the highest NC Temp:

In Modes I and 2, use Loop Tmr.

kn Modes 4-6:

Compare the average of the 5 highest reading operable core exit

T/CSto Loop Trior.

Use the operating train(s) of ND inlet temperature, Loop T H ~ T

and/or the operable core exit T/Cs.

"F Subcooled - Calculate by subtracting "HIGHEST NC TEMP" from "F-SAT".

Enclosure 13.12

Steam Generator Data Sheet Page 1 o f 1

ACCEPTANCE CRITERLA: If S/G shell temperature is < 80"F, SIG pfirnary or secondary pressure must NOT exceed 200 PSIG.

I NOTE: If pressure (primary and secondary) are verified < 200 pig, then temperatures are NOT required to be taken nor recorded I

Enclosure 13.13 PTll/A/4600/009

Axial Flux Difference (YOA Flux) Following Page 1 of 1

Loss of AFD Monitor Alarm

( I ) Record the current Axial Flux Difference limit.

ACCEPTANCE CRITERIA Each channel within the listed limit. Reference the COLR for each

channels limit

Enclosure 13.14 PT/1/k/46OO/OO9

Rod Verification Checklist Page I of 1

ACCEPTANCE CRITERIA - Ali rods are positioned k 12 steps uftheir step demand counters and

operable.

E ~ c ~ Q 13.15

su~~

Rod Insertion Limit Checksheet Page 1 of 1

Enclosure 13.16 PT/ 1/A/4600/009

CA Suction Source Teniperature Monitoring Page 1 of 4

Data

I. Procedure

1.1 Every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , record the following temperatures on the table of this enclosure:

UST IA at 1CSTT5990 (TB-640, ID-30)

UST 1B at 1CSTT6000 (TB-640,1D-29)

Hotwell Femperture at 1CMTT7260 (TB-577, 15-26)

1.2 Every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , measure CST temperature at 1CSTX5020 (TB-573, IC-33) by using a

Caiibrated Kiethly 872 Digital Thermometer, Model "J" or equivalent, and record on the

table of this enclosure.

1.3 -

IF UST temperature is 2 134 OF,

1.3.1 Notify CRSRO

1.3.2 Verify 1CM-127 (CM-CF Cleanup Flow Ctrl) is:

A. Isolated

B. E in Modes 3 OR 4 1CM-I27 is open, immediately close

1CM- 127.

1.3.3 Ensure Aux Steam is isolated to the UST by verifying the following valves

closed

1AS- 15 (AS to UST BTR Isol) (TB 609, 1 C-29)

m IAS-22 (AS to 1A USF HTR Ctri Inlet) (TB-640, 1C-30)

IAS-25 (AS to 1A UST HTR Ctrl BYP) (TB-640,1C-30)

IAS-28 (AS to 1B UST HTR Ctrl Inlet) (TB-640, 1C-29)

IAS-31 (AS to 1B UST HTR Chl BYP) (TB-640, 1'2-29)

1.3.4 Verify 1CM-33 (Hotwell High Levei Control) (TB-581, 1 6 2 5 ) is isolated and

-

NOT leaking past seat by checking downstream pipe temperature 5 120 OF.

1.3.5 Notify MSE.

1.3.6 temperature of 1A OR 1B UST is 2 I36'F, declare CA System inoperable

per TS 3.7.5.

Enclosure 13.16

PT/1/A/4600/009

CA Suction Source Temperature Monitoring Page 2 Of4

Data

1.4 E CST is 2 134 "F

1.4.1 Notify CRSRO

1.4.2 -

IF UST is overflowing, as determined by UST level 2 130%, measure

overflow pipe temperature (TB-594, IC/ID-29) using Calibrated Kiethly 872

Digital Thermometer, Model "J" or equivalent and record on table of this

enclosure.

1.4.3 -

IF overflow pipe temperature 2 134 OF, refer to step 1.3.

1.4.4 Verify Manual Loader for Condensate Storage Tank (IASML0170, CST

Temp Ctrl) (TB 573, 1E-32) is set per OP/O/Bi6250/007A (Auxiliary Steam

System Alignment).

1.4.5 -

IF Aux Steam Manual Loader for CST is malfunctioning, then:

Ensure IAS-I6 (AS to CST HTR Ctrl Inlet) (TB-582, 1E-33) is closed.

NotifyMSE.

1.4.6 -

IF temperature of CST is 5 136"F, secure both CST pumps and

A. UST is overflowing to the CST as indicated by UST level 2 130%,

measure temp on UST overflow line (TB-594, IWD-29).

B. E temp is 5 136'F, declare CA System inoperable.

C. E the UST is NOT overflowing to the CST as indicated by UST level

S 130%, measure UST temperature at 1CSTT5990 (TB-640, 1I)-30)

and lCSTT6000 (TB-640, ID-29). E temp is 2 136 "F, declare CA

system inoperable.

NOTE: Step 1.5 is only applicable if ICM-33 is open in Modes 2 and 3 and Mode 4 when SiGs

1.5 E Hotwell pump discharge temperahire reaches > 136°F Main Condenser Vacuum

decreases to <27" Hg vac, immediately close 1CM-33 (Hotwell High Level Control).

Enclosure 13.16 PT/l/A/4600/009

CA Suction Source Temperature Monitoring Page 3 of4

Data

NOTE: ICM-33 is required to be isolated when CA is aligned for Standby Readiness. Refer to

OP/l/A/6250/002 (Auxiliary Feedwater System).

1.6 Hotwell teniperature reaches 2 134"F, then:

1.6.1 E ICM-33 is isolated, then no further action is required.

1.6.2 E Unit 1 S/Gs are NOT relied upon for heat removal, then no hrther action is

required.

1.6.3 E Unit 1 is in modes 1,2,3, 4 with CA required operable ICM-33 is

-

NOT isolated, then isolate ICM-33.

1.I Rotweli temperature reaches 2 136'F, then:

1.7.1 1CM-33 is isolated, then no further action is required.

1.7.2 E Unit 1 S/Gs are relied upon for heat removal, then no further action is

required.

1.7.3 E Unit 1 is in modes 1,2,3, 4 with CA required operable 1CM-33 is

-

NOT isolated, then declare auxiliary feedwater system inoperable and apply

Tech Spec 3.7.5.

I---

Enclosure 13.17 PT/1/A/4600/009

CF Containment Isolation Valve N2 Page 1 o f 4

Accumulator Pressure Monitoring

1. Procedure

1.1 Every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , monitor the CF isolation valves N2 accumulator pressures on the table of this

enclosure for the following:

1CP-33 (1A S/G Feedwater Containment Isol), 1CFPT5710, (DH, 597, EE-43)

ICF-42 (1B S/G Feedwater Containment Isol), ICFPT5420, (DH, 597, EE-52)

1CF-51 (1C S/G Feedwater Containment Isol), 1CFPT5730, (DH, 596, EE-53)

1CF-60 (ID S/G Feedwater Containment Isol), 1CFPT5740, (SH, 597, EE-44)

1.2 N2 pressure is c: I285 psig for any ofthe CF containment isolation valves, then perform

the following:

1.2.1 Notify the Control Room SRO that the applicable CF containment isolation valve

is inoperable per Technical Specifications 3.6.3 and 3.7.3.

1.2.2 Notify the Shift Work Manager

1.2.3 Issue the following Model W/O for the applicable CF containment isolation valve

for WE to recharge the N2 accumulator:

1CF-33, Model W/O #91003804

ICF-42, Model W/O #91003805

1CF-51, Model W/O #91003806

1CF-60, Model W/O #91003807

Enclosure 13.17 PT/ 1/A/46OO/OO9

CF Containment Isolation Valve N2 Page 2 of 4

Accumulator Pressure Monitoring

1.3 @ N2 pressure is 2 1285 psig < 1625 psig for any of the CF containment isolation

valve, then perfom the following:

1.3.1 Notify the Cont~olRoom SRO.

1.3.2 Notify the Shift Work Manager

1.3.3 Issue the following Model WiO for the applicable CF containment isolation valve

for IAE to recharge the N2 accumulator:

1CF-33, Model W/Q #91003804

ICF-42, Model W/O #91003805

1CF-51; Model W/O #91003806

ICF-60, Model W/O #91003807

1.4 @ N2 pressure is > 2425 psig AND 5 2415 psig, then perfom the following:

e Notify the Control Room SRO.

failure due to overpressurization.

Dispatch operator to the applicable doghouse to investigate if a local steam leak in the

area could be the cause of the high pressure.

Notify the Shift Work Manager that the applicable CF N2 accumulator needs to be

vented.

Enclosure 13.17 $Ti I lN46OOM)OY

CF Containment Isolation Valve N2 Fage 3 of 4

Accumulator Pressure Monitoring

1.5 N2 pressure is > 2445 psig, then perform the following:

I NOTE: The CF containment isolation valve is inoperable when N2 pressure is > 2475 psig. I

Notify the Control Room SRO that the applicable CF containment isolation valve is

inoperable per Technical Specification 3.6.3 and 3.7.3.

NOTE: Due to system design limits, there is no immediate concern with catastrophic equipment

failure due to overpressurization. I

Dispatch operator to the applicable doghouse to investigate i f a local steam leak in the

area o f the applicable CF

Notify the Shift Work Manager that the applicable CF isolation valve N2 accumulator

needs to be vented.

Enclosure 13.17 PT/1!N4600/009

CF Containment Isolation Valve 112 Page 4 o f 4

Accumulator Pressure Monitoring

NOTE: The normal operating range for the CF containment isolation valves N2 accumulator pressure

is approximately 2200-2300 psig.

_______

(1) With N2 pressure < 1285 p i g , the CF containment isolation valve is inoperable.

(2) With N2 presume > 2475 psig, the CF containment isolation valve is inoperable.

Enclosure 13.18

Primary to Secondary Leakage Calculation Page 1 of 1

Data

Enclosure 13.19 PT/l/N4600/009

Pressurizer Spray AT Data Sheet Page 1 of 1

ACCEPTANCE CRITERIA - Differential temperature between the pressurizer and auxiliary spray

water must be < 260'F.

NOTE: When using Residual Heat Removal Pump 1A (ND Pump 1A), use Residual Heat

Removal Heat Exchanger "A" Inlet Temperature on 1MC7. When using Residual Heat

Removal Pump 1B (ND Pump lB), use Residual Heat Removal Heat Exchanger "B" Inlet

temperature on IMC7. When using Chemical Volume and Control System, use

Regenerative Heat Exchanger charging temperature on 1MC5.

Enclosure 13.20

Overtemperature Delta T Setpoint Channel Page 1 of 1

Check

1. Procedure

1.1 Gauges to be used:

NCkoopA 1NCP5422

NCLoopB lNCP5462

NCLoopC 1NCP5502

0 N C h o p D 1NCP5542

NOTE: At lower power levels (<;. 85%), the instrumentation will be overranged (> 150%). Ifthe

instrumentation is overranged, use a value of 150% when performing the Channel Check.

1.2 Verify that the difference between the highest and lowest reading loop is less than or equal

to 10% and record on table below.

1.3 the difference is > IOOh, notify Reactor Group Duty Engineer to perform a qualitative

assessment of channels to determine operability.

Enclosure 13.21 PTlliA/4600/009

Chemistry Data Sheet Page 1 o f 4

Enclosure 13.21 PT/1lA/4600/009

Chemistry Data Sheet Page 2 of 4

!

Analysis CONC DATE/TIME j

I I PPB I

NC Chloride PPB

NC Fluoride

INC Dose Equiv

1-131

Enclosure 13.21

PT/1/A/4600/009

Chemistry Data Sheet Page 3 of 4

CF Cation

BE Sodium

BB Cation

(Conductivity)

Enclosure 13.21

Chemistry Data Sheet

Analysis DateITinie

i i

Seal Inj. Filter 1A

Seal Inj.Filter 1B

NC Filter IA

MC Filter 1I3

NV Mixed Bed Demin. 1A

NV Mixed Bed Demin. 1B

NV Cation Bed Demin.

Comments:

DUKE POWER CATAWBA OPERATIONS TMINlNC

c) Aux Feedwater (CA) System Response to a Loss of Instrument Air

With a Loss of Instrument Air, the flow control valves for all SIG's

will no longer fail open. A Mod has been completed on both

units which added air receiver tanks to the CA flow control valves

with enough air that will allow closure of these valves for 6Q

minutes after a loss of VI. This will also preclude SIG overfill on a

SGTR with a loss of AC Power and a subsequent loss of VI.

d) Transient

1) Air will be lost to:

Main feedwater control

Steam dump control

e SIG PORVs in AUTO

e NC PQRVs and PZR sprays

2) Reactor trip will occur on 'SIG LO LO bevel due to CF control

valves failed closed.

3) Steam dumps will not operate. SIG POWVs will not open in

'AUTO. Heat removal will be via SIG safeties.

4) NC system pressure may increase beyond PZR P O W setpoint.

N2 may have to be aligned to PZR PORVs.

2.2 Station Air System (VS)

A. System Purpose (Obj. #13, 14)

I. Supplies air to miscellaneous tools and equipment throughout the plant.

2. Backup air supply for the Instrument Air system.

3. Interconnections with other systems.

a) Recirculated cooling water (KR)cools the compressor

1) Effluent coding water temperature controls the KR flow rate

2) Inlet isolation valves closes automatically when the associated

compressor is shutdown.

b) Low Pressure Sewice Water (RL) cools the after coolers

c) instrument Air (Obj. #8, 14)

1) Backup supply for VI

2) IVS-78 will automatically open if VI pressure drops to 76 p i g .

3) Via two oil removal filters

d) Logic from the Engineered Safety Features Actuation System closes

the containment isolation valves upon receiving a Phase A (ST)

containment isolation signal. (VS-54B) (Ob]. #ti')

OP-CN-SS-VI FOR TRAiNING PURPOSES ONLY REV. 26

Page 20 of 36

Bank Question: 988 Answer: B

1 Pi(s) Unit 1 was operating at 100% powcr when the pressurizer spray line

developed a leak. Which one of the following conditions would cause the

safety injection system to be inoperable during this accident?

A. 1NI-118A and 1NI-150B are open with power applied

B. 1NI-162A is closed with powcr applied

C. 1NI-121A and lNI-152B are closed with power removed

D. INI-100B is open with power removed

Distracter Analysis:

A. Incorrect: This is a nonnal alignment for the NI cold leg injection

vaives.

Plausible: If the candidate thinks that these valves open on a safety

injection signal. These valves tic the two trains of safety injection

together so the candidate may think that cross-tying the two coid leg

injection paths constitutes a loss of train separation

B. Correct: INI-162 is the common isolation line for NI. The proper

position would be for the valve to be open with power removed.

C. Incorrect: INI-121A and 1NI-152B are the NI hot leg isolation lines

-the proper position is closed with power removed as listed.

Plausible: If the candidate thinks that power should not be

disconnected or the hot leg injection path should be open

D. Incorrect: 1NI-100B is the FWST supply iSOkdtiOII line. Normal

position is open with power removed as listed.

Plausible: Ifthe candidate conkses 1NI-100B with 1NI-103A --

which is a different FWST suppiy isolation valve. INI-103A is

normally open with power removed.

Levcl: SRO Only IWFR55.43@)5

KA: APE 008 AA2.10(3.6/3.6)

Lesson Plan Objective: NI Obj: 8,13

Source: New

Level of knowledge: comprehension

References:

1. OP-CN-ECCS-NI page 8

2. Tech Spec 3.5.2

3. Tech Spec Bases 3.5.2

DUK PO WEB CATAWBA OPERATiQNS TRAINING ~

OBJECTIVES

I

Objective S

S

State the pu~poseand explain the operation of the NI Svstem. / X

State the ournose of the safeh, iniection D U ~ D Sand the FWST. Ix

Draw a block diagram of the ECCS system from the FWST to the NC X

Svstem per the lesson plan simplified drawina.

3escribe the indications to enter and actions taken for Reactor Trip or

SI Actuation durina plant shutdown.

State from memory all B.S.actions for the applicable systems,

subsystems and components, which require remedial action to be

!aken in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

OP-CN-EGCS-NI 6 0 R TRAINING PURPOSES ONLY REV. 35

Page 3 of 20

__-_

DUKE.. POWER ...

CATAWBA OPERATIONS - -- .

TRAINING

3) 'This recirculation flow path must be isolated during CoId/Hot leg

recirculation to prevent putting sump water back into the FWST.

The possibility exists for removing water from the sump and

contaminating the FWST, which would result in an uncontroliec!

radioactive release via FWS'T vent.

6 . The following is the power arrangement for the NI and other ECCS

pumps; all are powered from ETAIB. The sequencer loads are:

a) LOCA SEQUENCER: ND, NV, and NI Pumps

b) Blackout SEQUENCER: NV Pumps

B. The purpose of Refueling Water Storage Tank (FWS'I) is to provide the

initial source of borated water to the NI pumps, other ECCS components,

and the containment spray pumps. Limited in volume, this tank will

eventually be isolated. The continued suction source of the ECCS pumps

will be the containment sump

C. NI system motor operated valves (Unit l&2-the same)

1. NI-1006 FWST S ~ ~ p pIsolation,

ly open with power rernoved.

2. NI-I03/Ni..I 35 A&B pump's FWST suction isolation valves, normaliy

open with power available.

3. Discharge Isolations

a) Cold Leg injection isolation vaives, NI-I 18A and 150B are norrnally

open, tying the two trains of pimp discharge togethei.

b) NI-l62A, common isolation to the cold legs, is required open with

cmtrol board switches disconnected.

c) Globe throttle valves are locked in a position, determined by the

ECCS FLOW BALANCE TEST, to prevent puntp runout and to

equalize the flow to the loops.

d) M i Leg injection valves, NI-121A arid 1528, are riomially closed

with control board switches disconnected until going into Hot leg

Recirculation

2.2 Standby Alignment and procedure Limits a!id Precautions, refer to latest copy of

OP~l(2)/A!6200/006 (Obj. #5)

A. The system is placed in standby when al! NC cold legs teniperatures w e

greater than 3OC"F Tavg is less than 350°F.

6 . Norrnnl Standby Condition

1. An injection path exists from the FWST (NI-100B) to the cold legs (NI-

128A, 2508, 162A). Valves NI-1005 and NI-162A have a"5isconnect"

switch used to meet operability per T.S. surveillance. In the "disconnect"

position, the breaker control circuit is shut off and the control board

switch will not funccioc.

OP-CN-ECCS-NI FOR TRAINING PURPOSES ONLY REV. 35

Page 8 of 20

ECCS-Operating

3.5.2

3.5 EMERGENCY CORE CQOLBNG SYSTEMS (ECCS)

3.5.2 ECCS-Operating

K O 3.5.2 Two ECCS trains shall be OPERABLE*.

APPLICABILITY: MODES I;2,and 3.

ACTIONS

CONDlTlON I REQUIRED ACTION COMPLETION TIME

A. One or mere trains A.1 Restore train@)to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s*

inoperable. OPERABLE status

&NJ

At le& 100% of the

ECCS flow equivalent to

a single OPERABLE

ECCS train available.

8. Required Action and B.l Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

associated Completion

Time not met.

8.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

'For each ECCS train on Unit 2. the Completion Time that one ECCS train can be inoperable as specified by

Required Action A . l may be extended beyond th6 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to 2R8 hours as part of the NSWS system upgrades.

System upgrades include maintenance and modification activities associated with cleaning of NSWS piping. vaI\es

ard branch lines, necessary repairs andior replacement, valve repair andlor replacement, and replacement of

portions of the NSWS piping to the AFW system. Upon completion of the cleaning. upgrades. and system restoratton

in refueling outage Z EOC12, this footnote is no longer applicable.

Catawba Units 1 and 2 3.5.2-1 Amendment Nos. 1891182

ECCS - Oaeratino

. 3.5.2

SURVEILLANCE REQUIREMENTS e _ _ _ _

SURVEILLANCE FREQUENCY

SR 3.5.2.1 Verify the following valves are in the listed position with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

power to the valve operator removed.

Number Position Function

N1162A Open SI Cold beg

Injection

N1121A Closed SI Hot Leg

Injection

N1152B Closed SI Hot beg

Injection

Nil838 ClQSed RHR Hot Leg

Injection

N1173A Open RHR Cold Leg

lnjection

Nil788 Open RHR Cold Leg

Injection

N1100B Open SI Pump Suction

from RWST

N1147B Open SI Pump

Mini-Flow

SR 3.5.2.2 Verify each ECCS manual, power operated, and 31 days

automatic valve in the flow path, that is not locked,

sealed, or otherwise secured in position, is in the correct

PQskbl.

SI? 3.5.2.3 Verify ECCS piping is full of water. 32 days

SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with

flow point Is greater than or equal to the required the Inservice

developed head. Testing Program

(continued)

Catawba Units 1 and 2 3.5.2-2 Amendment Nos. 1931165

ECCS -Operating

3.5.2

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is 18 months

not locked, sealed, or otheiwise secured in position,

actuates to the correct position on an actual or simulated

actuation signal.

SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual 18 months

or simulated actuation signal.

SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each 18 months

position stop is in the correct position.

Centrifugal Charging Safety Injection

Pump Injection Throttle Pump Throttle

Valve Number Valve Number

N114 NI164

N116 N1166

N118 MI168

N120 N1170

SR 3.5.2.8 Verify, by visual inspection, each ECCS train containmeni 18 months

sump suction inlet is not restricted by debris and the

suction inlet trash racks and screens show no evidence

of structural distress or abnormal corrosion.

Catawba Units Iand 2 3.5.2-3 Amendment Nos. 173/165

ECCS-Operating

B 3.5.2

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS-Operating

BASES

BACKGROUND The function of the ECCS is to provide core cooling and negative

reactivity to ensure that the reactor core is protected after ~ R ofY the

following accidents:

a. boss of coolant accident (LOCA), coolant leakage greater than the

capability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam

or feedwater release; and

d. Steam generator tube rupture (SGTR).

The addition of negative reaetivlty is designed primarily for the loss of

secondary coolant accident where primary cooldown could add enough

positive reactbity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold leg

recirculation, and hot leg recirculation. In the injection phase, water is

taken from the refueling water storage tank (RWST) and injected into the

Reactor Coolant System (RCS) through the cold legs. When sufficient

water is removed from the RWST to ensure that enough boron has been

added to maintain the reactor subcritical and the containment sumps

have enough water to supply the required net positive suction head to the

ECCS pumps, suction is switched to the containment sump for cold leg

recirculation. When the core decay heat has decreased to a level low

enough to be successfully removed without direct RHR pump injection

flow, the RHR cold leg injection path is realigned to discharge to the

auxiliary containment spray header. After approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . part of

the ECCS flow is shifted to the hot leg recirculation phase to provide a

backflush which. for a cold leg break, would reduce the boiling in the top

of the core and prevent excessive boron concentration.

The ECCS consists of three separate subsystems: centrifugal charging

(high head), safety injection (Si) (intermediate head), and residual heat

removal (RHR) (low head). Each subsystem consists of two redundant,

100% capacity trains. The ECCS accumulators and the RWST are also

part of the ECCS, but are not considered part of an ECCS flow path as

described by this K O .

Catawba Units -I

and 2 B 4.5.2-1 Revision No. 0

ECCS - Operating

B 3.5.2

BASES

BACKGROUND (continued)

The ECCS flow paths consist of piping, valves, heat exchangers, and

pumps such that water from the RWST can be injected into the RCS

following the accidents described in this LCO. The major components of

each subsystem are the centrifugal charging pumps, the RHR pumps,

heat exchangers, and the SI pumps. Each of the three subsystems

consists of two 100% capacity trains that are interconnected and

redundant such that either train is capable of supplying 100% of the flow

required to mitigate the accident consequences. This interconnecting and

redundant subsystem design provides the operators with the ability to

utilize components from opposite trains to achieve the required 100% flow

to the core.

During the injection phase of LOCA recovery, a suction header supplies

water from the RWST to the ECCS pumps. Mostly separate piping

supplies each subsystem and each train within the subsystem. The

discharge from the centrifugal charging pumps combines, then divides

again into four supply lines, each of which feeds the injection line to one

RCS cold leg. The discharge from the SI and RHR pumps divides and

feeds an injection fine to each of the RCS cold legs. Throttle valves in the

SI lines are set to balance the flow to the RCS. This balance ensures

sufficient flow to the core to meet the analysis assumptions following a

LOCA in one of the RCS cdd legs. The flow split from the RHR lines

cannot be adjusted. Although much of the two ECCS trains are

composed of completely separate piping, certain areas are shared

between trains. The most important of these are 1) where both trains flow

through a single physical pipe, and 2) at the injection connections to the

RCS cold legs. Since each train must suppiy sufficient flow to the RCS to

be considered 100% capacity, credit is taken in the safety analyses for

flow to three intact cold legs. Any configuration which, when combined

with a single active failure, prevents the flow from either ECCS pump in a

given train from reaching all four cold legs injection points on that train is

unanalyzed and might render both trains of that ECCS subsystem

For LOCAs that are too small to depressurize the RCS below the shutoff

head of the SI pumps, the centrifugal charging pumps supply water until

the RCS pressure decreases below the SI pump shutoff head. During

this period, the steam generators are used to provide part of the core

cooling function.

During the recirculation phase of LOCA recovery, RHR pump suction is

transferred to the containment sump. The RHR pumps then supply the

other ECCS pumps. Initially, recirculation is through the same paths as

the injection phase. Subsequently, for large LOCAs, the recirculation

phase includes injection into both the hot and cold legs.

Catawba Units 1 and 2 B 3.5.2-2 Revision No. 0

ECCS - Operating

B 3.5.2

BASES

BACKGROUND (continued)

The high and intermediate head subsystems of the ECCS also functions

to supply borated water to the reactor core following increased heat

removal events, such as a main steam line break (MSLB). The limiting

design conditions occur when the moderator temperature coefficient is

highly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed on

the maximum number of ECCS pumps that may be OPERABLE. Refer to

the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection

(LTBP) System," for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an St signal. The

actuation of safeguard loads is accompbished in a programmed time

sequence. If offsite power is available, the safeguard loads start

immediately in the programmed sequence. If offsite power is not

availabie, the Engineered Safety Feature (ESFJbuses shed normal

operating loads and are connected to the emergency diesel generators

(EDGs). Safeguard loads are then actuated in the programmed time

sequence. The time delay associated with diesel starting, sequenced

loading, and pump starting determines the time required before pumped

flow is available to the core following a safety injection actuation.

The active ECCS components, along with the passive accumulators and

the RWST covered in LCO 3.5.2, "Accumulators," and LCO 3.5.4,

"Refueling Water Storage Tank (RWST)," provide the cooling water

necessary to meet GDC 35 (Ref. I).

APPLICABLE The LCO helps to ensure that the following acceptance criteria for the

SAFETY ANALYSES ECCS, established by 10 CFR 50.46 (Ref. 21,will be met following a

small break LOCA and there is a high level of probability that the criteria

are met following a large break LOCA:

a. Maximum fuel element cladding temperature is 5 2200°F;

b. Maximum cladding oxidation is 5 0.17 times the total cladding

thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is

m< 0.01 times the hypothetical amount generated if all of the metal in

the cladding cylinders surrounding the fuel, excluding the cladding

surrounding the plenum volume, were to react;

Catawba Units 1 and 2 B 3.5.2-3 Revision No. 9

ECCS - Operating

B 3.5.2

APPLICABLE SAFETY ANALYSES (continued)

d. Core is maintained in a coolable geometry; and

e. Adequate long term core cooling capability is maintained.

The LCQ also limits the potential for a post trip return to power foilowing

an MSLB event and ensures that containment pressure and temperature

limits are met.

Each ECCS subsystem is taken credit for in a large break LQCA event at

full power (Refs. 3 and 4). This event has the greatest potential to

challenge the limits on runout flow set by the manufacturer of the ECCS

pumps. It also sets the maximum response time for their actuation. Direct

flow from the centrifugal charging pumps and SI pumps is credited in a

small break LQCA event. The RHR pumps are also credited, for larger

small break LQCAs, as the means of supplying suction to these higher

head ECCS pumps after the switch to sump recirculation. This event

establishes the flow and discharge head &.the design point for the

centrifugal charging pumps. The MSbB analysis also credits the SI and

centrifugal charging pumps. Although some ECCS flow is necessary to

mitigate a SGTR event, a single failure disabling one ECCS train is not

the limiting single failure for this transient. The SGTR analysis primary to

secondary break flow is increased by the avaiiabiiity of both centrifugal

charging and SI trains. Therefore, the SGTR analysis is penalized by

assuming both ECCS trains are operable as required by the LCQ. The

QPEBABILITY requirements for the ECCS are based on the following

LQCA analysis assumptions:

a. A large break LQCA event, with loss of offsite power and a single

failure disabling One ECCS train; and

b. A small break LBCA event, with a loss of offsite power and a single

failure disabling one ECCS train.

During the blowdown stage of a LQCA. the RCS depressurizes as

primaly coolant is ejected through the break into the containment. The

nuclear reaction is terminated either by moderator voiding during large

breaks or control rod insertion for small breaks. Following

depressurization, emergency cooling water is injected into the cold legs,

flows into the downcomer, fills the lower plenum, and refloods the core.

The effects on containment mass and energy releases are accounted for

in appropriate analyses (Ref. 3). The LCO ensures that an ECCS train

will deliver sufficient water to match boiloff rates soon enough to minimize

the consequences of the core being uncovered following a large LQCA.

~

Catawba Units 1 and 2 B 3.5.2-4 Revision No. 0

ECCS - Operating

B 3.5.2

BASES

APPLICABLE SAFETY ANALYSES (continued)

It also ensures that the centrifugal charging and SI pumps will deliver

sufficient water and boron during a small LOCA to maintain core

subuiticality. For smaller LOCAs, the centrifugal charging pump delivers

sufficient fluid to maintain RCS inventory. For a small break LOCA, the

steam generators continue to serve as the heat sink, providing part of the

required core cooling.

The ECCS trains satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).

LCO In MODES 1,2, and 3, two independent (and redundant) ECCS trains are

required to ensure that sufficient ECCS flow is available, assuming a

single failure affecting either train. Additionally, individual components

within the ECCS trains may be called upon to mitigate the consequences

of other transients and accidents.

In MODES I, 2, and 3, an ECCS train consists of a centrifugal charging

subsystem, an SI subsystem, and an RHR subsystem. Each train

includes the piping, instruments, and controls to ensure an OPERABLE

flow path capable of taking suction from the RWST upon an SI signal and

automatically transferring suction to the containment sump.

During an event requlring ECCS actuation, a flow path is required to

provide an abundant supply of water from the RWST to the RCS via the

ECCS pumps and their respective supply headers to each of the four cold

leg injection nozzles. In the long term, this flow path may be switched to

take its supply from the containment sump and to supply its flow to the

RCS hot and cold iegs. The flow path for each train must maintain its

designed independence to ensure that no single failure can disable both

ECCS trains.

APPLICABILITY In MODES 1,2, and 3, the ECCS OPERABILITY requirements for the

limiting Design Basis Accident, a large break LOCA, are based on dulI

power operation. Although reduced power would not require the same

level of performance, the accident analysis does not provide for reduced

cooling requirements in the lower MODES. The centrifugal charging

pump performance is based on a small break LOCA, which establishes

the pump performance curve and has less dependence on power. The SI

pump performance requirements are based on a small break LOCA. For

both of these types of pumps, the large break LOCA analysis depends

only on the flow value at containment pressure, not on the shape of the

flow versus pressure curve at higher pressures. MODE 2 and MODE 3

requirements are bounded by the MODE 1 analysis.

Catawba Units 1 and 2 B 3.5.2-5 Revision No. 0

ECCS - Operating

B 3.5.2

BASES

APPLICABILITY (continued)

This LCO is only applicable in MQDE 3 and above. Below MODE 3, the

SI signal setpoint is manually bypassed by operator control, and system

functional requirements are relaxed as described in LCO 3.5.3,"ECCS-

Shutdown."

As indicated in the Note, the flow path may be isolated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in

MODE 3, under controlled conditions, to perform pressure isolation valve

tesfhg per SR 3.4.14.1. The flow path is readily restorable from the

control room.

In MODES 5 and 6, plant conditions are such that the probability of an

event requiring ECCS injection is extremely low. Core cooling

requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-

MODE 5, Loops Filled," and LCQ 3.4.8, "RCS Loops-MQDE 5, hoops

Not Filled." MODE 6 core cooling requirements are addressed by

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-

High Water Level," and LCO 3.9.5, "Residual Heat Removal (RHR) and

Coolant Circulation-Low Water Level."

ACTIONS

With one or more trains inoperable and at least 100% of the ECCS flow

equivalent to a single OPERABLE ECCS train available, the inoperable

components must be returned to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The

72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on an NRC reliability evaluation

(Ref. 6) and is a reasonable time for repair of many ECCS components.

An ECCS train is inoperable if it is not capable of delivering design flow to

the RCS. individual components are inoperable if they are not capable of

performing their design function or supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent

subsystems Due to the redundancy of trains and the diversity of

subsystems, the inoperability of one component in a train does not render

the ECCS incapable of performing its function. Neither does the

inoperability of two different components, each in a different train,

necessarily result in a loss of function for the ECCS. The intent of this

Condition is to maintain a combination of equipment such that 100% of

the ECCS flow equivalent to a single OPERABLE ECCS train remains

available. This allows increased flexibility in plant operations under

circumstances when components in opposite trains are inoperable.

Catawba Units 1 and 2 0 3.5.2-6 Revision No. 0

ECCS - Operating

3 3.5.2

BASES

ACTIQNS (continued)

An event accompanied by a loss of offsite power and the failure of an

EBG can disable one ECCS train until power is restored. A reliability

analysis (Ref. 6) has shown that the impact of having one full ECCS train

inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Reference 7 describes situations in which one component, such as an

RHR crossover valve, can disable both ECCS trains. With one or more

compenent(s) inoperable such that 100% of the flow equivalent to a

single OQEfWBLE ECCS train is not available, the facility is in a condition

outside the accident analysis. Therefore, LCO 3.0.3 must be immediately

entered.

B.1 and 8.2

If the inoperable trains cannot be returned to OPERABLE status within

the associated Completion Time, the plant must be brought to a MODE in

which the LCQ does not apply. To achieve this status, the plant must be

brought to MQDE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The

allowed Compietion Times are reasonable, based on operating

experience, to reach the required plant conditions from full power

conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.2.1

REQUIREMENTS

Verification of proper valve position ensures that the flow path from the

ECCS pumps to the RCS is maintained. Misalignment of these valves

could render both ECCS trains inoperable. Securing these valves using

the power disconnect switches in the correct position ensures that they

cannot change position as a result of an zctive failure or be inadvertently

misaligned. These valves are of the type, described in Reference 7,that

can disable the function of both ECCS trains and invalidate the accident

analyses. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered reasonable in view of other

administrative controls that will ensure a mispositioned valve is unlikely.

SR 3.5.2.2

Verifying the correct alignment for manual, power operated, and

automatic vaives in the ECCS flow paths provides assurance that the

proper flow paths will exist for ECCS operation. This SR does not apply

to valves that are locked, sealed, or otherwise secured in position, since

these were verified to be in the correct position prior to locking, sealing,

~

Catawba Units 1 and 2 B 3.5.2-7 Revision No. 0

ECCS - Operating

0 3.5.2

BASES

SURVEILLANCE REQUIREMENTS (continued)

or securing. A valve that receives an actuation signal is allowed to be in a

nonaccident position provided the valve will automatically reposition

within the proper stroke time. This Surveilrance does not require any

testing or valve manipulation. Rather, it involves verification that those

valves capable of being mispositioned are in the correct position. The

32 day Frequency is appropriate because the valves are operated under

administrative control.

This Frequency has been shown to be acceptable through operating

experience.

SR 3.5.2.3

With the exception of the operating centrifugal charging pump, the ECCS

pumps are normally in a standby, nonoperating mode. As such, flow path

piping has the potential to develop voids and pockets of entrained gases.

Maintaining the piping from the ECCS pumps to the RCS full of water by

venting the ECCS pump casings and accessible discharge piping high

points ensures that the system will perform properly, injecting its full

capacity into the RCS upon demand. This will also prevent water

hammer, pump cavitation, and pumping of noncondensible gas (e.g.. air,

nitrogen, or hydrogen) into the reactor vessel foliowing an SI signal or

during shutdown cooling. The 31 day Frequency takes Into consideration

the gradual nature of gas accumulation in the ECCS piping and the

procedural controls governing system operation.

- -

SR 3.5.2.4

Periodic surveillance testing of ECCS pumps to detect gross degradation

caused by impeller structural damage or other hydraulic component

problems is required by Section XI of the ASME Code. This type of

testing may be accomplished by measuring the pump developed head at

only one point of the pump characteristic curve. This verifies both that the

measured performance is within an acceptable tolerance of the original

pump baseline performance and that the performance at the test flow is

greater than or equal to the performance assumed in the plant safety

analysis. SRs are specified in the Inservice Testing Program, which

encompassesSection XI of the ASME Code.Section XI of the ASME

Code provides the activities and Frequencies necessary to satisfy the

requirements.

Catawba Units 1 and 2 B 3.5.2-8 Revision No. 0

ECCS - Operating

B 3.5.2

BASES

SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.25 and S R 3.5.2.6

These Surveillances demonstrate that each automatic ECCS valve

actuates to the required position on an actual or simulated SI and

Containment Sump Recirculation signal and that each ECCS pump starts

on receipt of an actual or simulated SI signal. This Surveillance is not

required for valves that are locked, sealed, or otherwise secured in the

required position under administrative controls. The 18 month Frequency

is based on the need to perform these Surveillances under the Conditions

that apply during a plant outage and the potential for unplanned plant

transients if the Surveillances were performed with the reactor at power.

The 48 month Frequency is also acceptable based on consideration of

the design reliability (and confirming operating experience) of the

equipment. The actuation logic is tested as part of ESF Actuation System

testing, and equipment performance is monitored as part of the Inservice

Testing Program.

SR 3.5.2.7

The position of throttle valves in the flow path on an SI signal is

necessary for proper ECCS performance. These valves have mechanical

iocks to ensure proper positioning for restricted flow to a ruptured cold

leg, ensuring that the other cold legs receive at least the required

mlnimum flow. The 18 month Frequency is based on the same reasons

as those stated in SR 3.5.2.5 and SR 3.5.2.6.

SR 3.5.2.8

Periodic inspections of the containment sump suction inlet ensure that it

is unrestricted and stays in proper operating cundition. The 18 month

Frequency is based on the need to perform this Surveillance under the

conditions that apply during a plant outage and on the need to have

access to the location. This Frequency has been found to be sufficient to

detect abnormal degradation and is confirmed by operating experience.

~~ ~

Catawba Units 1 and 2 B 3.5.2-9 Revision No. 1

ECCS - Operating

B 3.5.2

BASES

REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.

2. 10 CFR 50.46.

3. UFSAR, Section 6.2.4.

4. UFSAR, Chapter 15.

5. 10 CFR 50.36, Technical Specifications, (cj(2)(ii).

6. NRC Memorandum to V. Stello, Jr., from R.L. Baer,

"Recommended Interim Revisions to LCOs for ECCS

Components," December 1, 1975.

7. IE Information Notice No. 87-01.

Catawba Units 1 and 2 B 3.5.2-10 Revision No. 0

Rank Question: 969 Answer: C

1 Pt(s) Unit 1 is operating at 65% power following a turbine runback caused by a feed

pump failure. Given the following events and conditions:

e The RO reports that rod control bank D has inserted past the rod insertion

limit (FUL)

b Annunciator 1AD-2, FV9, Control Rod BankLo-Lo Limit, has not alarmed.

0 IAE has reported that a failed anrimciator card must be replaced but the part

will not be available until next week.

Which one of the following acti0nr is required in response to this failed alarm?

A. The shift work manager must initiate a temporary modification to

change the ControfItodBank Lo Limit (1AD-2, A/9) annunciator

setpoint to the ContPulRodBankLo-Lo rod insertion limit.

B. The unit supenisor must initiate a control panel information tag for

the 1AD-2, Bl9.

C. The operations s h i t manager shall ensure that aiternate indications are

monitored to duplicate the function of the failed annunciator.

D. The reactor operator must enter the requirement to verify RIL limits

manually during transients in the shift turnover log.

____s__________________s__s_____s__s____---------------------------------------------------

Distracter Analysis:

A. Incorrect: An increased surveillance sheet must be initiated.

Plausible: If the candidate does not know the requirement, this is a

logical alternative.

J3. Incorrect: An increased surveillance sheet must be initiated.

Plausible: Ifthe candidate does not know the requirement, this is a

logical alternative.

C. Correct:

D. Incorrect: An increased surveillance sheet must be initiated.

Plausible: If the candidate does not h o w the requirement, this is a

logical alternative.

Level: SRO Only 10CFW55.43@)5

K A G 2.4.33 (2.4D.8)

Lesson Plan Objective: ADM-NSOS Obj: 29

Source: New

Level of howledge: memory

References:

I . OQ-CN-ADM-NSOSpage 11

2. OMP 2-3 1 page 2

DUKE BOWER CATAWBA QPERATlONS TRAlNlNG

I N L L P

Objective S b P P T

S O R S R

-

0 0 0

12 rocess for approval and extension of Technical

I

13

14 Describe the Work List Priority scheme and how priority conflicts x x

- are resolved (OMP 2-5)

15 Describe how the Work List is completed (OMP 2-5)

- 1 jx x

16 Explain peBonnel responsibilities for key control (OMP2-9)

- 1x1 I x l x l x

27 Describe restrictionslrequirementsfor the use of file keys.

- (OMP 2-9)

18 Describe restrictions/requirementsfor long term key removal.

l x I lxlxlx

19

-

20

-

21

-

22

I

23

control (OMP 2-16, SD 3.1.20,).

24 Describe the expectations for conduct of personnel while in the

Control Room (OMP 2-16)

25 Explain the process of maintaining the Unified Logbook in

accordance with OMP 2-17 (as appropriate to job level).

-

26

before, during and after the formal transfer of shift responsibilities.

(OMP 2-22)

27 Explain the process of maintaining Control Room Instrumentation

Status Logs and actions required for failed annunciators.

- (QMP 2-32)

! 1 lXIXlX

28 Describe how to initiate a corrective work request. (WFM 401)

- IX I 1X I I X

OP-CN-ADM-NSO5 FQR TRAINING PURPOSES ONLY Rev11

Page 5 of 12

I-

DUKE POWER

_

CATAWBA OPERATIONS TRAlNlNG

5. Alarm response during execution of EFs/AFs

6. Alarm response during Crew Briefs.

7. Exceptions to normal alarm response protocol.

2.6 Unit Unified Logbooks

A. Using OMF 2-17, (Unit Unified Logbook Maintenance), review responsibilities

and processes associated with the Unit Supervisor and Control Room

subiogs. (Obj. #25):

1. Personnel responsible for maintaining Control Room and Unit Supewisor

and FT sublogs.

2. General instructions and entry procedures.

3. Definition of QFEN ITEM and how OPEN ITEMS are documented.

4. Logbook required entries.

5. Test Logbook (PT Sublogs)

El. Using OMP 2-34, (Control Room Instrumentation Status), review the process

involved in maintaining the Control Room Annunciator Status Cog, audits and

increased surveillance. (Qbj. #27)

2.7 Key Control

A. Using OMP 2-9, (Administration and Control of Keys), review personnel

responsibilities for control of keys. (Obj. #16)

6. Using OMF 2-3,(Administration and Control of Keys), describe

requirements/restrictiorssfor:

1. Use of File Keys (Obj. # I T )

2. Long Term Key Removal (Qbj. #18)

3. Interlock Keys (Obj. #19)

2.8 Work Requests (Obj. #28)

A. Refer to WFM 401 (Problem Communication Guidelines)

B. Problem Communication

1. Normally problems with plant equipment are communicated to the Work

Control Single Point Of Contact (SFOC) via the WMS Problem identifier

Screen (RQQ5 or Rat 0).

2. For emergency work, SPOC should be notified of problems by telephone.

SPOC will complete work request documentation as required.

3. Problem complexity may require that the problem be communicated in

person in the Work Control Center.

QP-CN-ADM-NSOS FOR TRAINING PURPOSES ONLY Rev71

Page l f of 12

Operations Management Procedure 2-3 1 Page 2 of 3

4. Responsibilities

4.1. The Reactor Operators (ROs) are responsible for review and knowledge of

Control Room annunciator status.

4.2. The Unit Lead is responsible for Increased Surveillance Items.

A. The Unit Lead should work with Engineering to ensure there are no

problems with using the Increased Surveillance.

H. The IJnit Lead should work with Work Control to ensure timely

completion of any work order creating the need for an Increased

Surveillance.

4.3. 'The Operations Shift Manager (OSM) is responsible for actions on failed

instrumentshannunciators.

5. Annunciator Status Log Instructions

5.1. Control Room annunciator status shall k reviewed at turnover per

CbMP 2-22 (Shift Turnover').

5.2. "Strapped" Bistable Inputs

0 Bistable with "strapped" inputs shall be reviewed at turnover per

OMP 2-22 (Shift 'turnover).

0 As directed by IAE procedures, distinctive red flags shall be placed on

"strapped bistable windows.

5.3. Annunciators with inputs disabled by Temporan1 Modifications ( t M ) shaall be

recorded on Attachment 8.1 (TMs Affecting Control Room Annunciators).

e Attac.hnient shall be filed in Ops Shift Routine Logbook.

6. Action on Failed ~nstrumentation/Annunciators

6.1. The OSM shall ensurc that alternate indications are monitored to duplicate the

function of the failed instruinentation/annuncia~or.

0 Alternate indications to be monitored shall be recorded on an Increased

Surveillance Sheet (Attachment 8.2) per Section 7.

6.2. The Shift Work Manager shall be notified of any instrumentation determined

to be failed.

PANEL: 1AD-2

CONTROL ROD BANK LO-LO LIRIIl B/9

SETPOPNT: Variable, ealcuiated based on auctioneered high D:T.

ORIGIN: Kod insertion limit computer.

PROBABLE 1. Boron concentration too low for power andjor Xenon conditions.

CAUSE: 2. Reactor Shutdown.

3. Rod insertion limit computer or instnlment malfunction.

4. Major load rejection or transient.

AUTOMATIC hone

ACTIONS:

IMMEDIATE 1. Stop any dilution in progress.

ACTIONS: 2. Refer to AI/1/.45500/013 (Boron Dilution).

3. Detennine affected control bank(s) by referring to lRFM17 in

Cable Spading Room (AUX 5 4 7 ) and acknowledge the a l m .

SUPPLEMENTARY 1. Kefer to TS 3.1.6 (Contra1 Bank Insertion Limits).

ACTIONS: 2. Refer to TS 3.1, I (Shutdown Margin (SDM)).

REFERENCES: I. TS 3.1.I (Shutdown Margin (SDM)) and

TS 3.1.6 (Control Bank Insertion Limits).

2. FSAR, Figure 7-2 (Part 9 of 16)

OPiI iRi6100iOlOC

PANEL: 1AD-2 Page 12 of 68

CONTROL ROD BANK LO LIMIT

SETPOINT: Vanable calculated based on auctioneered high DiT

ORIGIN: Rod insertion limit computer.

PROBABLE 1. Boron concentration too low for power andior Xenon conditions.

CAUSE: 2. Rod insertion limit computer or instrument malfunction.

3. Major load rejection or transient.

AUTOMATIC None

ACTIONS:

IMMEDIATE i. Stop any dilution in progress.

ACTIONS: 2. Determine affected control bank(s) by referring to 1RFM16 in

Cable Spreading Room (AtJX 577) and acknowledge the alarm.

SUPPLEMENTARY 1. Borate coolant system as necessary until alami clears per

ACTIONS: OPil/A%150i009 (Boron Concentration Control).

2. -

IF due to temperature instrument malfunction, defeat affected

instrument using "T-AVG DEITAT" and "ATEMP DEFEAT"

switch.

3. Refer to TS 3.1.6 (Control Bank Insertion Limits).

4. -

IF due to instniment failure, have IAE trip bistable on affected

channel per Tech Spec Action.

REFERENCES: 1. TS 3.1.1 (Shutdown Margin (SDM) and

Ths 3.1.6 (Control Rank Insertion Limits)

2. FSAR, Figure 1-2 (Part 9 of 16)

Bank Question: 962 Answer: D

1 Pt(s) Unit 1 is conducting a containment ais release at 85% power. Given the

following events and conditions:

The last containment atmosphere sample at 0600 was consistent with

the current gaseous waste release (GWR) pennit.

The latest air release was started at 1400.

The release flow indicator on l h K - 5 reads 200 CFM.

1EMF-36 (UnitVent Gas Monitor) is at the trip 1 setpoint.

The VQ flow totalizer is out of service.

Which of the following conditions reported at 1700 requires isolation of &e

air release until a new GWR permit is approved?

A. The VQ air release flow indicator on 1MC-5 fails high.

B. IEMF-39 (Containment Gas Monitor) increases to Trip 2 alarm

due to a power supply fluctuation.

C. The 1800 containment atmosphere sample must be delayed for 4

hours due to sample valve repairs.

D. IEm-37 (UnitVent Iodine Monitor) increases to Trip 2 aIarm

due to a fuel pin failure.

--

Distracter Analysis:

A. Incorrect: The flow indicator is not required to conducVcontinue an

air release.

Plausible: If the candidate believes that release flow is critical with

the totalizer out of service.

B. Incorrect The release can be reset and continued on the current

GWR.

Plausible: If the candidate does not know that spurious nips can be

reset twice.

C. Incorrect: The release can continue on the current GWR as long as

the sample is analyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the previous sample.

Plausible: Ifthe candidate thinks the sample frequency (12 hrs) is the

GWR validation requirement.

D. Correct:

Level: SRO Only lOCFR55.43@)4

KA: SYS 029 G 2.3.8 (233.2)

Lesson Plan Objective: CNT-VQ Obj: 4

Source: New

Level of knowledge: comprehension

References:

1. OP-CN-CNT-VQ pages 9

2.OP/I/A/64.50/17

3. OP/I/A/6450/17 EncI 4.2

4.OP/UA/6450/17 Enci 4.3

I

Objective S

s

State the P L I ~ C JofS the

~ VQ system Ix

Describe the sources of pressure fluctuations in containment during X

normal operations

Describe the simal that will auto stop a VQ fan I

Given appropriate plant conditions, apply Limits and Precautions

assodated with related station procedures.

Exalain how an ais addition from containment is ammDlished I

Describe the effect a S, signal has on VQ

Given a set of specific plant conditions and access to reference materials, I

determine the actions necessary

...

to comply

....

with._

Tech Spets/SI.Cs

...... ..

..... ...

'li

TIME: 1.O Hour

OP-CN-CNT-VQ FOR TRAINING PURPOSES ONLY R N . 20

Page 3 of If

DUKE BO WEW CATAWEA OPERATIONS TRAINING

2. Air Release Mode is performed if containment pressure is greater than

0.09 psig and a release is desired. Refer to OP/1/A16450/017 Enclosure

4.2 (OBJ. 6)

UNIT

VENl

I

UPPER CONTAINMENT

3. Initiating and terminating a GWR form refer to OP/l/AI6450iO17

Enclosure 4.3

4. Signals or conditions that terminate a release or addition (OBJ. 7)

a) At any time by manual operator action from the control room.

b) Automatically when containment pressure returns to 0 psig by closing

VQ-to or VQ-23.

c) The Containment Ais Release Fans automatically shut off when low

flow is detected at fan discharge.

d) EMF-35,36,37 trig 2 will auto dose VQ-10.

OP-CM-CMT-VQ KIR TRAINING PURPOSES ONLY REV. 20

Page 9 of 1f

OP/1/N6450/0 17

Page 2 of 3

Containment Air Release and Addition System

1. Purpose

To outline the proper operation of the Containment Air Release and Addition System.

2. Limits and Precautions

2.1 Do NOT exceed Containment Pressure Limits of -0.08 psig and +0.25 pig. Tech Spec

Containment Pressure Limits are -0.1 p i g to +0.3 p i g .

2.2 When manually operating any motor operated valve, minimize the torque applied to the

handwheels.

2.3 ARer manual operation, maintenance or packing adjustment of any motor operated

Safety Related valve, it shali be cycled electrically to ensure reliable automatic operation.

2.4 Pressure switches for valve operation should m T be manuaily overridden since ice

condenser doors are very sensitive to over or under pressure conditions.

2.5 When Containment Air Release Filter unit pre-filter or absolute filter differential

pressure reaches 2.5 inches H20,the standby fan should be placed in service and action

initiated to replace the dirty filter(s).

2.6 A new Gaseous Waste Release (GWR) sample is required if:

E 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has elapsed since the last sample.

e VQ release is automatically terminated due to a valid controlling EMF actuation. If

actuation is due to an EMF spike, the release may be re-attempted twice before a new

sample is required.

2.1 A VP; VQ or Unit Vent Sample is required if:

e Rx Trip or Startup occurs.

E Rated Themial Power change of 2 15% in one hour occurs followed by a Thermal

Power Stabilization bower level constant at desired power level).

2.8 If 1EMF-37 or 1EMF-40 has reached the Trip 1 setpoint, RP should be notified to

change the c'artridge before a release is attempted.

3. Procedures

Refer to Section 4 (Enclosures).

OP/llA/645O/O17

Page 3 of 3

4. Enclosures

4.1 Air Addition Mode

4.2 Air Release Mode

4.3 lnitiation and Termination of a GWR Permit Report

4.4 Auxiliary Building Valve Checklist

4.5 Reactor Building Valve Checklist

4.6 Auxiliary Building Separate Verification Valvc Checklist

4.7 Reactor Building Separate Verification Valve Checklist

4.8 Manual Air Addition Mode

Enclosure 4.2 OP/1/N6450/0 17

Air Release Mode Page 1 of 6

1. Initial Conditions

1.1 Review the Limits and Precautions.

1.2 Verify Conkainment Pressure > 0.09 psig.

__ 1.3 Verify CR SRO has signed and dated the appropriate sheet of the Gaseous Waste

Release (GWR) Record authorizing releases:

c3 "VQ release monitored by EMF 39(L)"

0 "VQ release monitored by EMF 36(L)"

__ 1.4 Verify Containment pressure increase is NOT due to a LOCA or steam line break.

1.5 Review the "SPECIAL INSTRUCTIONS FOR RELEASE" section on the GWR Permit

Report.

2. Procedure

__ 2.1 Ensure the following enclosures are complete:

Enclosure 4.4 (Auxiliary Building VaIve Checklist)

0 Enclosure 4.5 (Reactor Building Valve Checklist)

0 Enclosure 4.6 (Auxiliary Building Separate Verification Valve Checklist)

0 Enclosure 4.7 (Reactor Building Separate Verification Valve Checklist)

2.2 Perform the following to sign off the "EMF Operable and Source Checked" blank on the

appropriate Gaseous Waste Release (GWR) Record

2.2.1 1EMF-39 (low range) is operable, perform the following:

2.2.1 .I Verify EMF-39 is specified for use on the GWR Permit Report.

2.2.1.2 Verify 1EMF-39 is operable per SLC 16.11-7 using

OP/O/N6500/080 (EMF W86A Output Modules).

2.2.1.3 Set 1EMF-39 (low range) setpoints to the value specified on the

GWR Permit Report using OP/O/N6500/080 (EMF RP86A

Output Modules).

2.2.1.4 Verify trip setpoints are set to the values as specified on the GWR

Permit Report using OPI0IN65001080 (EMF RP86A Output

Modules).

Enclosure 4.2 OP/hA/6450/0 17

Air Release Mode Page 2 of 6

__ 2.2.2 IEMF-39 (low range) is inoperable &Q 1EMF-36 (low rangc) is to be

used to monitor this release, perform the following:

CI 2.2.2.1 Verify 1EMF-36 is specified for use on the GWR Permit Report.

0 2.2.2.2 Verify 1EMF-36 is operable using OP/O/N6500~080(EMF

RP86A Output Modules).

NOTE: 1EMF-36 (low range) trip setpoints are pre-established for offsite dose.

E! 2.2.2.3 Verify trip setpoints are set to the values as specified on the GWR

Permit Report using OP/O/N6500/080 (EMF RE'86A Output

Modules).

2.2.2.4 N/A the "Wblank on the "VQ release monitored by EMF 36(L)"

sheet of the Gaseous Waste Release (GWR) Record.

_ I 2.2.3 -

IF 1EMF-39 1EMF-36 are both inoperable, perform the following:

Kl 2.2.3.1 Verify EMF-39 and EMF-36 are both N/Aed on the GWR Permit

Report.

0 2.2.3.2 Notify RP to take grab samples per HP/O/B/1004/005

(Containment Air Release and Addition (VQ) And Containment

Purge Ventilation (VP) System Release).

Person notified

El 2.2.3.3 N/A the "EMF Operable and Source Checked" blanks on the

Gaseous Waste Release (GWR) Record.

l

l 2.2.3.4 N/A Step 2.3.

-2.3 Set u- EMF Chart re Jrder as follows:

2.3.1 Ensure the paper drive is on for the applicable EMF chart recorder:

1MICR6640if IEMF-39 (L) is used

1MICR6650 if 1EMF-36 (L) is used

2.3.2 Stamp and record the following on the chart paper:

Date

Time

GWW#

Initials

Enclosure 4.2

OPi~lN64501017

Air Release Mode Page 3 of 6

~

sv - - .. 2.4 Adjust "IVQ-10 VQ Fans Disch To Unit Vent" manual loader (IMC5) to I the

"Recommended Release Rate (cfm)" on the GWR Permit Report.

2.5 Record the following on the appropriate Gaseous Waste Release (GWR) Record.

2.5.1 IF the tobilker is operable, reset it and enter "0" in the "Initial Integrator

Reading" blank

2.5.2 the totalizer is inoperable, N/A the "Initial Integrator Reading" blank.

2.6 Open the following valves (1MC5):

Record date/time f k t valve is opened 1

0 IVQ-2A (VQ Fan Suct From Cont Isol)

0 IVQ-3B (VQ Fan Suct From Cont Isol)

2.7 Place one VQ train in service as foilows (1MC5):

-2.7.1 To place A train in service, perform the following:

0 2.7.1.1 Place "VQ Filt Htr A" in the "AUTO" position.

E l 2.7.1.2 Start "Cont Air Re1 Fan 1A".

-2.7.2 To place B train in service, perfonn the following:

63 2.7.2.1 Place "VQ Filt Htr B" in the "AUTO" position.

0 2.7.2.2 Start "Cont Air Rei Fan IB".

2.8 Notify Rp that the VQ release has been started and give start time as recorded in

Step 2.6.

Person notified

m _ 2.9 Record the VQ start dateltime on the following:

3 Appropriate Gaseous Waste Release (GWR) Record

0 Control Room Autolog

Enclosure 4.2 OP/I/A/6450/OI7

Air Release Mode Page 4 of 6

-2.10 the OAC OR Computer Point C I P l l l 2 (Average Containment Pressure, best) is out

of service, record containment pressure as read on 1VQP5040 (Containment Pressure) on

BMCS every 30 minutes in the Control Room Log for the duration of the VQ Release.

{PIP 93-0074)

__ 2. I 1 the VQ fan does NOT automatically shutdown at approximately 0 psig, perfom the

following:

N/A Step 2.12.

Perform Step 2.13.

__ 2.12 WHEN Containment pressure decreases to approximately 0 psig, verify that

"lVQ-IO VQ Fans Disch To Unit Vent" closes, by no flow indicated on the manual

loader (black needle).

2.13 Revet "IVQ-10 VQ Fans Disch To Unit Vent" by adjusting the manual loader fully

SV

counterclockwise.

2.14 Secure the VQ train placed in service in Step 2.7 as follows:

__ 2.14.1 To secure A train, perform the following:

5 2.14.1.1 Ensure "Cont Air Re1 Fan 1A" has stopped.

0 2.14.1.2 Place "VQ Filt Htr A" in the "OFF" position

2.14.2 To secure B h i n , perform the following:

c3 2.i4.2.1 Ensure "Cont Air Re1 Fan 1B" has stopped.

2.14.2.2 Place "VQ Filt Btr B" in the ""OFF" position

2.15 Close the following valves:

Record ddte/time both valves are closed i

IVQ-2A (VQ Fan Suct From Cont Isol)

[3 IVQ-3B (VQ Fan Suct From Cont Isol)

Enclosure 4.2 OP/1/A/6450/017

Air Release Mode Page 5 of 6

-2.16 Stamp and record the following on the chart paper:

Date

Time

GWRX

Initials

-2.17 Notify RP that the VQ release has been terminated and give termination time as recorded

inStep2.15.

Person notified

2.18 Record the VQ terminate datehime on the following:

I3 Appropriate Gaseous Waste Release (GWR) Record

0 Control Room Autolog

2.19 Record the following on the appropriate Gaseous Waste Retease (GWR) Record

2.19.1 -

IF the totalizer is operable, perform the following:

2.19.1.1 Record totalizer reading

2.19.1.2 Record totalizer value in "Final Integrator Reading" blank of the

appropriate Gaseous Waste Release (GWR) Record sheet.

2.19.1.3 Enter the volume releawd in the "Volume" blank.

Volume = final integrator reading X 10.

2.19.2 -

IF the totalizer is inoperable, perform the following:

2.19.2.1 NiA the "Finai Integrator Reading" blank.

2.19.2.2 Enter the volume released in the "Volume" blank.

Volume = 350 CPM X Release Time (in minutes).

e Release Time = DateiTime initiated DateRime terminated.

2.19.3 Enter the "Highest EMF Reading" during the release as read on the chart

recorder.

2.19.4 Sign the "Control Room Operator" blank.

Enclosure 4.2 o P i l i ~ 6 4 5 o i 17

o

Air Release Mode Page 6 of 6

e nearest 100 prior to

2.20 IEMF-39 was used for this release, reset IEMF-39 (low range) trip setpoints using

OP/O/A/6500/080 (EMF RP86A Output Modules):

__ 2.20.1 in Mode 5 or 6, the trip setpoints shall be as follows:

Trip 2 = 17,400 cpm + Existing reading, Rounded down to the

nearest 100 cpm

Trip 1 = Trip2X.70

___ 2.20.2 in Mode 1 , 2 , 3 or 4, the trip setpoints shall be set as follows:

Trip 2 = 3 X Containment Atmosphere Activity (sampled at all three

locations) as indicated by EMF allowing about I S minutes for

indication to stabilize.

Trip 1 = Trip 2 X .70.

2.20.3 Signoff "EMF Setpoints Reset" blank on the "VQ release monitored by EMF

39(L)" sheet ofthe Gaseous Waste Release (GWR) Record.

2.20.4 Verify trip setpoints are reset as described in Step 2.20.1 or 2.20.2 using

OP/O/N6500/080(EMF W86A Output Modules).

2.2 1 1EMF-36 was used for this release, NIA the "EMF Setpoints Reset" blank on the

"VQ release monitored by EMF 36(L)" sheet of the Gaseous Waste Release (GWR)

Record.

2.22 Do NOT file this enclosure in the Control Copy folder of this procedure.

Enclosure 4.3 OP/1/N645O/O 14

Initiation and Termination of a GWR Page 1 of 1

Permit Report

Information Use

1. Initial Conditions

1.1 Review the Limits and Precautions.

1.2 Verify Radiation Protection has taken a sample and has issued a new GWR Permit

Report.

2. Procedure

2.1 Request the CR SRO sign, date and time the appropriate sheet o f the Gaseous Waste

Release (GWR) Record authorizing releases.

be started and stopped as often as desired as long as the GWR Permit Report remains

valid. Subsequent steps in this section will terminate the GWR Permit Report when

2.2 The following steps are to aid the operator in temlinating the GWR Permit Repoa:

2.2.1 Add all values in the "VQ Volume" column of the appropriate sheet o f the

Gaseous Waste Release (GWR) Record. Record total in "Total Volume

Released" blank.

2.2.2 Ensure the CR SRO performs the following:

2.2.2.1 Sign, date and time the "Termination o f GWR Release

Acknowledged by Shift Supervisor or Designee" blank of the

appropriate sheet of the Gaseous Waste Release (GWR) Record.

2.2.2.2 Place the completed VQ GWR Permit Report in the completed

release box.

Bank Question: $36.1 Answer: A

1 Pt(s) Unit 1recently shut down. Given the following events and conditions:

NCS temperature is 230°F.

Both trains of ND are operable with A train in senrice

KC loops A and B are operating.

CAPT is tagged for maintsnance.

CA pump 1B and CA pump 1A are isolated for merentia1 pressure

testing.

Engineering has reported that a test on the 1B KC heat exchanger showed

that fouling had reduced its heat transfer capability beiow the minimum

design value.

The plant is expected to remain in mode 4 for another 7 days fer special

testing.

What is the EARLIEST action required for the operators to take?

REFPRENCESSPROWDED: Tech Spec's 3.4.6,3.7.5,3.7.7 and Bases

A. Return at least one motor driven CA pump t~ operable status or

be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Return at least one motor driven CA pump to operable status or

be in mode 5 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

C. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in mode 5

within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months o r be in mode 5

W h i n the newt 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months .

__

Distracter Analysis: One loop of KC is operable therefore one loop of ND

is operable. With both MDCA pumps isolated, the S/Gs do not have

a source of feedwater. Tech Spec 3.4.6 requires CA to be operable

for the NC loops to be operable. Action statement B of Tech S p c

3.4.6 applies.

A. Correct answer: Without m operable CA pump to supply S/Gs, the

NC loops are not operable in mode 4 (TS 3.4.6 bases). Action

statement B applies. Tech Spec 3.7.5 action D (do not change

modes) does not apply because the plant is not in modes 1-3.

B. Incorrect: This does not restore CA and meet Tech Spec 3.4.6

Plausible: Meets the Tech Spec 3.7.5 action B for one train of CA

Ques-936.1 .doc

C. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Plausible: This is the correct answer for TS 3.7.7 action R.

D. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Plausible: if the candidate adds the time allowed to reach mode 3 (6

hrs) to the mode 5 action time (36 hrs) in TS 3.7.7 action B.

Level: SRO Only 10CFR55.43@)3

KA: G2.1.33 (3.4'4.0)

Lesson Plan Objective: PSS-KC Obj: 13

Source: Mod Ques-936 NRC McGuire 2002

Level of knowledge: comprehension

References:

1. Tech Spec 3.4.6 and bases - PROVIDED

2.Tech Spec3.7.5 andBases-PROVIDED

3. Tech Spec 3.7.9 and bases - PROVIDED

- ... DUKE POWER .. .... - ....-...... - CATAWBA

._ -_. .....

OPERATIONS TRAlNlNG

...

I

Objective s

1

1 State the Durgose of the KC Svstem. l x

2 I Describe how the KC Svstem is cooled. I x

1

3 Describe the normal flowpath of the KC System, including

each header and the t w e of loads serviced by each.

X

- Safety Injection (Ss)

- Phase A Containment Isolation (St)

Phase B Containment Isolation (Sp)

I

6 State the tvpical values of the KC pump discharge I x

I pressure, KC Hx outlet temperature and KC pump flow.

of the KC system per the KC

concernincs the KC svstem.

System and what is indicated by a high level radiation

OB-Cff-PSS-KC FOR TRAINING PURPOSES ONLY REV. 42

Page 3 of 26

RCS LOOPS- MODES 4

3.4.6

3.4.6 RCS LOOPS -MODE 4

LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat

removal (RHR) loops shall be OPERABLE, and one loop shall be in

operation.

NOTES-------------------------~~-----------~----------

1. All reactor coolant pumps (RCPs) and RHR pumps may be

de-energized for 5 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would came reduction of the

RCS boron concentration; and

b. Core outlet temperature is maintained at least 10°F below

saturation temperature.

2. No RCP shall be started with any RCS cold leg temperature 5 285°F

unless the secondary side water temperature of each steam

generator (SG) is 5 50°F above each of the RCS cold leg

temperatures.

_s_______________s____________ss__s_____--~----------------------~~--~--------------------~-

APPLICABILITY: MODE 4

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. One RCS loop A.1 Initiate action to restore a Immediately

OPERABLE. second loop to

OPERABLE status.

ANB

Two RHR loops

. 1

(continued)

Catawba Units 4 and 2 3.4.6-1 Amendment Nos. 173165

RCS LOOPS - MODES 4

3.4.6

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

B. One RHR loop 3.1 Be in MODE 5 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

&NJ

ALL RCS loops

I

C. Both required RCS or C.l Suspend all operations lmmediateiy

RHR loops inoperable. involving a reduction of

RCS boron concentration.

-

QR

No RCS or RHR loop in

operation. C.2 Initiate action to restore immediately

one loop to OPERABLE

status and operation.

_e

SR 3 4 6 2 Verify SG secondary side water levels are 2 12% narrow 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

range for required RCS loops

SR 3.4.6.3 Verify correct breaker alignment and indicated power are 7 days

available to the required pump that is not in operation.

Catawba Units 1 and 2 3.4.6-2 Amendment Nos. 173/165

RCS LooPs-MQDE 4

B 3.4.6

B 3.4 REACTOR COOLANT SYSTEM (RCS)

E3 3.4.6 RCS LooPs-MODE 4

BASES

BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of

decay heat and the transfer of this heat to either the steam generator

(SG)secondary side coolant or the component cooling water via the

residual heal removal (RHR] heat exchangers. The secondary function of

the reactor coolant is to act as a carrier for soluble neutron poison, boric

acid.

The reactor coolant is circulated through four RCS loops connected in

parallel to the reactor vessel, each loop containing an SG,a reactor

coolant pump (RCP), and appropriate flow, pressure, level, and

temperature instrumentation for control, protection, and indication. The

RCPs circulate the coolant through the reactor vessel and SGs at a

sufficient rate to ensure proper heat transfer and to prevent boric acid

stratification,

In MODE 4, either RCPs or RHR loops can be used to provide forced

clrculation. The intent of this LCO is to provide forced flow from at least

one RCP or one RHR loop for decay heat removal and transport. The

flow provided by one RCP loop or RHR loop is adequate for decay heat

removal. The other intent of this LCO is to require that two paths be

available to provide redundancy for decay heat removal.

APPLICABLE In MODE 4, RCS circulation is considered in the determination of the

SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The

RCS and RHR loops provide this circuiation.

RCS Loops-MODE 4 satisfy Criterion 4 of 10 CFR 50.36 (Ref. 1)

LCO The purpose of this LCO is to require that at least two loops be

OPERABLE in MODE 4 and that one of these loops be in operation. The

LCO allows the two loops that are required to be OPERABLE to consist of

any combination of RCS loops and RHR loops. Any one loop in operation

provides enough flow to remove the decay heat from the core with forced

circulation. An additional loop is required to be OPERABLE to provide

redundancy for heat removal.

Note 1 permits all RCPs or RHR pumps to be deenergized for 5 'Ihour

per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests that are

designed to validate various accident analyses values. One of the tests

Catawba Units 1 and 2 3 3.4.6-1 Revision No. 0

RCS Loops - MODE 4

B 4.4.6

BASES

LCO (continued)

performed during the startup testing program is the validation of rod drop

times during cold conditions, both with and without flow. The no flow test

may be performed in MODE 3 , 4 , or 5 and requires that the pumps be

stopped for a short period of time. The Note permits the de-energizing of

the pumps in order to perform this test and validate the assumed analysis

values. If changes are made to the RCS that wouM cause a change to

the flow characteristics of the RCS, the input values must be revalidated

by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to

perform the test, and operating experience has shown that boron

stratification is not a problem during this short period with no forced flow.

Utilization of Note i is permitted provided the following conditions are met

along with any other conditions imposed by initial startup test procedures:

a. No operations are permitted that would dilute the RCS boron

concentration, therefore maintaining the margin to criticality. Boron

reduction is prohibited because a uniform concentration distribution

throughout the RCS cannot be ensured when in natural circulation;

and

b. Core outlet temperature is maintained at least I O T below

saturation temperature, so that no vapor bubble may form and

possibly cause a natural circulation flow obstruction.

Note 2 requires that the secondary side water temperature of each SG be

s 50°F above each of the RCS cold leg temperatures before the start of

an RGP with any RCS cold leg temperature s 285°F. This restraint is to

prevent a low temperature overpressure event due to a thermal transient

when an RCP is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an

OPERABLE SG in accordance with the Steam Generator Tube

Surveillance Program, which has the minimum water level specified in

SR 3.4.6.2.The water levei is maintained by an OPERABLE AFW train in

accordance with LCO 3.7.5,"Auxiliary Feedwater System."

Similarly for the RHR System, an OPERABLE RHR loop comprises an

OPERABLE RHR pump capable of providing forced flow to an

OPERABLE RHR heat exchanger. RCPs and RHR pumps are

OPERABLE if they are capable of being powered and are able to provide

forced flow if required.

Catawba Units 1 and 2 B 3.4.6-2 Revision No. 0

RCS LOOPS - MODE 4

B 3.4.6

BASES

APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to

remove decay heat from the core and to provide proper boron mixing.

One loop of either RCS or RHR provides sufficient circulation for these

purposes. However, two loops consisting of any combination of RCS and

RHR loops are required to be OPERABLE to meet single failure

considerations.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES Z and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.7, "RCS Loops-MQDE 5,hoops Filled";

LCO 3.4.8, "RCS Loops-MODE 5,Loops Not Filled";

LCO 3.4.1 7, "RCS Loops-Test Exceptions";

LCO 3.9.4, "Residual Heat Removal (RHR)and Coolant

Circulation-High Water Level" (MODE 6); and

LCO 3.9.5. "Residual Heat Removal (RHR) and Coolant

Circulation-Low Water Level" (MODE6).

ACTIONS A.1

If only one RCS loop is OPEMBLE and two RHR loops are inoperable,

redundancy for heat removal is lost. Action must be initiated to restofe a

second RCS OF RHR loop to OPERABLE status. The immediate

Completion Time reflects the importance of maintaining the availability of

two paths for heat removal.

If only one RHR loop is QPERABLE and in operation and there are no

RCS loops OPERABLE, an inoperable RCS or RHR loop must be

restored to OPERABLE status to provide a redundant means for decay

heat removai.

If the parameters that are outside the limits cannot be restored, the unit

must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to

MODE 5 is a conservative action with regard to decay heat removal. With

only one RHR loop OPERABLE, redundancy for decay heat removal is

lost and, in the event of a loss of the remaining RHR loop, it would be

safer to initiate that loss from MODE 5 (< 200°F) rather than MODE 4

(200 to < 350OF). The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time,

based on operating experience, to reach MODE 5 from MODE 4 in an

orderly manner and without challenging plant systems.

Catawba Units 1 and 2 B 3.4.6-3 Revision No. 0

RCS LOOPS- MOB 4

E 3.4.6

BASES

ACTIONS (continued)

C.l and C.2

If no loop is OPERABLE or in operation, except during conditions

permitted by Note 1 in the LCO section, all operations invoiving a

reduction of RCS boron concentration must be suspended and action to

restore one WCS or RHR loop to OPERABLE status and operation must

be initiated. RCP seal injection flow is not considered to be an operation

involving a reduction in RCS boron concentration. Boron dilution requires

forced circulation for proper mixing, and the margin to criticality must not

be reduced in this type of operation. The immediate Completion Times

reflect the importance of maintaining operation for decay heat removal.

The action to restore must be continued untll one loop is restored to

OPERABLE status and operation.

SURVEILLANCE SR 3.4.61

REQUIREMENTS

This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one RCS or RHR loop is

in operation. Verification includes flow rate, temperature, or pump status

monitoring, which help ensure that forced flow 1s providing heat removal.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and

alarms available to the operator in the control room to monitor RCS and

RHR loop performance.

SW 3.4.6.2

SR 3.4.6.2requires verification of SG OPEBAEILITY. SG OPERABILITY

is verified by ensuring that the secondary side narrow range water level is

2 12%. If the SG secondary side narrow range water level is c 12%, the

tubes may become uncovered and the associated loop may not be

capable of providing the heat sink necessary for removal of decay heat.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other

indications available in the control room to alert the operator to the loss of

SG level.

SR 3.4.6.3

Verification that the required pump is OPERABLE ensures that an

additional RCS or RHR pump can be placed in operation, if needed, to

maintain decay heat removal and reactor coolant circulation. Verification

is performed by verifying proper breaker alignment and power available to

the required pump. The Frequency of 7 days is considered reasonable in

view of other administrative controls available and has been shown to be

Catawba Units I and 2 B 3.4.6-4 Revision No. 0

RCS LOOPS - MODE 4

B 3.4.6

BASES

acceptable by operating experience.

Catawba Units 1 and 2 B 3.4.6-5 Revision No. 0

RCS LOOPS - MODE. 4

B 3.4.6

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.4.6-6 Revision No. 0

AFW System

3.7.5

3.7 PLANT SYSTEMS

3.7.5 Auxiliary Fedwater (AFW) System

APPLICABILITY: MQDES I,2, and 3,

MODE 4 when steam generator is relied upon for heat removal.

ACTIONS

CONDI%[QN REQUIRED ACTION COMPLETIQM TIME

A. Qne steam supply to A.l Restore steam supply to 7 days

turbine driven AFW QPERA%LEstatus.

pump inoperable.

10 days from

discovery of

failure to

meet the LCO

B. One AFW train B.l Restore AFW train to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s: I

inoperable in MODE 1 , 2 OPERABLE status.

or 3 for reasons other

than Condition A.

10 days: from I

discovery of

failure to

meet the LCO

(conthued)

-each A F \ a a i n on Unit 2. the Com&tion Time that one A F \ n r t r a i n m bempetahleas3PeGifiedpy &Lluir.&'

'For ~ ~

~

Action Aa I.y be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 10 davs froAdis!z!!Leryofailureto m~eAthe&O:mo 288

hours a a a t d h e N4!S m t e n u p g r a d e s . S y s t e m . u ~ ~ d e s ~ c l u d e m . a ~ n t e n a n . c emdificati0.n~

and advitles

associatedwithae NSWS e i ~ ~ n g , v a l v e ~ a n d ~ n ~ ~ i n e ~ n . e c e ~ a ~ ~ ~ a i r I&acement ~~~d!o~r@~ia~e~

ofeonions of the NSWS Oiping to&e~AFW s v s t e m , ~ o n x m p M o n o f h e c&nine_upgra&s,and SYSW

~- nolongemx?k&b!e

r e s t o r a m in refuel~imoutw I~ . C l . 2 h i ~ s f o o t n n ~ i.s

Catawba Units 1 and 2 3.7.5-1 Amendment Nos. 1891182

AFW System

3.7.5

ACTIONS (continued)

CONDiTlON REQUIRED ACTION COMPLETION TIME

6. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

associated Completion

Time for Condition A y

d

J

or E! not met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

-

Two AFW trains

inoperable in MODE 1,

2, or 3.

D. Three AFW trains D. ------------NOTE-------------

inoperable in MODE 1, LCO 3.0.3 and all other

2, or 3. LCO Required Actions

requiring MODE changes

are suspended until

one AFW train is restored

to OPERABLE status.

..................................

Initiate action to restore Immediately

one AFW train to

OPERABLE status.

E. Required AFW train =.IInitiate action to restore Immediately

inoperable in MODE 4. AFW train to OPERABLE

status.

Catawba Units 1 and 2 3.7.5-2 Amendment Nos. 1731165

AFW System

3.9.5

SURVEILLANCE REQUIREMENTS

SL6RVElbLANCE FREQUENCY

SR 3.7.5.1 ____________~ __s______l__ _______

s__-_ NOTE____----__ ---__

_s__s_ s__s~--

Not applicable to automatic valves when THERMAL

POWER is 5 10% RTP.

_____________________________s__s_______~--~~----------------------------

yr

Verify each AFW manual, power operated, and automatic 31 days

valve in each water flow path, and in both steam supply

flow paths tu the steam turbine driven pump, that is not

locked, sealed, or otherwise secured in position, is in the

correct position.

SR 3.7.5.2 ------------------------------------NOTE--~~---------~--------------

Not required to be performed for the turbine driven AFW

pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 2 600 p i g in the steam

generator.

Verify the developed head of each AFW pump at the flow In accordance

test point is greater than or equal to the required with the Inservice

deveioped head. Besting Program

98 months

(continued)

Catawba Units 1 and 2 3.9.5-3 Amendment Nos. 1731165

AFW System

3.7.5

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

SB 3.7.5.4 ____- s__~___s____ NOTES--_____--____---

_-_-_

s_---

3. Not required to be performed for the turbine

driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after Z 600 psig

in the steam generator.

2. Not applicable in MODE 4 when steam generator

is relied upon for heat removal.

Verify each AFW pump starts automatically on an actual

or simulated actuation signal. 18 months

SR 3.7.5.5 Verify proper aiignment of the required AFW flow paths Prior to entering

by verifying flow from the condensate storage system to MODE 2,

each steam generator. whenever unit

has been in

MODE 5 or 6 for

> 30 days

Catawba Units 1 and 2 3.7.5-4 Amendment NOS. 1941165

AFW System

3 3.7.5

B 3.7 PLANT SYSTEMS

B 3.7.5 Auxiliary Feedwater (AFW) System

BASES

BACKGROUND The AFW System automatically supplies feedwater to the steam

generators to remove decay heat from the Reactor Coolant System upon

the loss of normal feedwater supply. The AFW pumps take suction

through suction lines from the condensate storage system (CSS)

(LCO 3.7.6) and pump to the steam generator secondary side. The

normal supply of water to the AFW pumps is from the condensate

system. The supply valves are open with power removed from the valve

operator. The assured source of water to the AFW System is supplied by

the Nuclear Service Water System. The turbine and motor driven pump

discharge lines to each individual steam generator join into a single line

outside containment. These individual lines penetrate the containment

and enter each steam generator through the auxiliary feedwater nozzle.

The steam generators function as a heat sink for core decay heat. The

heat load is dissipated by releasing steam to the atmosphere from the

steam generators via the main steam safety valves (MSSVs) (kCO 3.7.1)

or SG PQRVs (LCO 3.7.4). If the main condenser is available, steam

may be released via the steam dump valves and recirculated to the

hotwell.

The AFW System consists of two motor driven AFW pumps and one

steam turbine driven pump configured into three trains. Each of the motor

driven pumps supply 155% of the flow requirements to two steam

generators, although each pump has the capability fo be realigned to feed

other steam generators. The turbine driven pump provides 250% of the

flow requirements and supplies water to all four steam generators. Travel

stops are set on the steam generator flow control valves such that the

pumps can supply the minimum flow required without exceeding the

maximum flow allowed. The pumps are equipped with independent

recirculation lines to prevent pump operation against a closed system.

Each motor driven AFW pump is powered from an independent Class 1E

power supply. The steam turbine driven AFW pump receives steam from

two main steam lines upstream of the main steam isolation valves. Each

of the steam feed lines will suppiy 100% of the requirements of the

turbine driven AFW pump.

~~

Catawba Units 1 and 2 B 3.7.5-1 Revision No. 0

AFW System

B 3.7.5

BACKGROUND (continued)

The AFW System is capable of supplying feedwater to the steam

generators during normal unit startup, shutdown, and hot standby

conditions. One turbine driven pump at full flow is sufficient to remove

decay heat and cool the unit to residual heat removal (RHR) entry

conditions. During unit cooldown, SG pressures and Main Steam

pressures decrease simultaneously. Thus, the turbine driven AFW pump

with a reduced steam supply pressure remains fully capable of providing

flow to all SGs. Thus, the requirement for diversity in motive power

sources for the AFW System is met.

The AFW System is designed to supply sufficient water to the steam

generalor(s) to remove decay heat with steam generator pressure at the

iowest setpoint of the MSSVs plus 3% accumulation. Subsequently, the

AFW System supplies sufficient water to cool the unit to RHR entry

conditions, with steam released through the SG PORVs or MSSVs.

The motor driven AFW pumps actuate automatically on steam generator

water level low-low in 1 out of 4 steam generators by the ESFAS

( K O 3.3.2).The motor driven pumps also actuate on loss of offsite

power, safety injection, and trip of all MFW pumps. The turbine driven

AFW pump actuates automatically on steam generator water level iow-

low in 2 out of 4 steam generators and on loss of offsite power.

The AFW System is discussed in the UFSAR, Section 10.4.9 (Ref. I].

APPLICABLE The AFW System mitigates the consequences of any event with loss

SAFETY ANALYSES of normal feedwater.

The design basis of the AFW System is to supply water to the SkaM

generator to remove decay heat and other residual heat by delivering at

least the minlmurn required flow rate to the steam generators at

pressures corresponding to the lowest steam generator safety vaive set

pressure plus 3%.

In addition, the AFW System must supply enQUghmakeup water to

replace steam generator secondary inventory lost as the unit coois to

MODE 4 conditions. Sufficient AFW flow must also be available to

account for flow losses such as pump recirculation valve leakage and line

breaks.

The limiting Design Basis Accidents (DBAs) and transients fer the AFW

System are as follows:

Catawba Units 1 and 2 B 3.7.5-2 Revision No. 0

AFW System

B 3.7.5

APPLICABLE SAFETY ANALYSES (continued)

a. Feedwater Line Break (FWLB); and

b. Loss of MFW.

In addition, the minimum available AFW flow and system characteristics

are Considered in the analysis of a Small break loss of coolant accident

(LOCA) and events that could lead to steam generator tube bundle

uncovery for dose considerations.

A range of AFW flows is considered for the analyzed accidents, with the

Main Steam Line Break being the most limiting for the maximum AFW

flowrate.

The AFW System design is such that it can perform its function following

a FWLB between the steam generator and the downstream check valve,

combined with a loss of offsite power following turbine trip, and a single

active failure of the steam turbine driven AFW pump. in such a case, one

motor driven AFW pump would deliver to the broken MFW header at the

pump runout flew until the problem was detected, and flow terminated by

the operator. Sufficient flow would be delivered to the intact steam

generators by the redundant AFW pump.

The ESFAS automatically actuates the AFW turbine driven pump and

associated power operated valves and controls when required to ensure

an adequate feedwater supply to the steam generators during loss of

offsite power.

The AFW System satisfies the requirements of Criterion 3 of 20 CFR

50.36 (Ref. 2).

LCO This LCO provides assurance that the AFW System will perform its

design safety function to mitigate the consequences of accidents that

could result in overpressurization of the reactor coolant pressure

boundary. Three independent AFW pumps in three diverse trains are

required to be OPERABLE to ensure the availability of RHR capability far

all events accompanied by a loss of offsite power and a single failure.

This is accomplished by powering two of the pumps from independent

emergency buses. The third AFW pump is powered by a different means,

a steam driven turbine supplied with steam from a source that is not

isolated by closure of the MSBVs.

Catawba Units 1 and 2 B 3.7.5-3 Revision No. 0

AFW System

B 3.7.5

BASES

K O (continued)

The A W System is configured into three trains. The AFW System is

considered OPEFPABLE when the components and flow paths required to

provide redundant AFW flow to the steam generators are OPERABLE.

This requires that the two motor driven AFW pumps be OPERABLE in

two diverse paths, each supplying AFW to separate steam generators.

The turbine driven AFW pump is required to be OPERABLE with

redundant steam supplies from two main steam lines upstream of the

MSIVs, and shall be capable of supplying AFW to any of the steam

generators. The piping, valves, instrumentation, and controls in the

required flow paths also are required to be OPERABLE. T M S A S

a s s u r e d m d !&laktsupply~iscMfiguredwvi- etubine

driu?nApumP recei\LesM!&Sfrnm hatb_trainsnthlSmh&rK

the&ss.afmetminafassuredsame r&ersnuIylmeAEWtPain

n~aPERARLEassurd

i r S a p e r a h L T h r e m a'njng~hlslnLstcainpra\Lidesa

saurcEFhthe~atbEiMnotmdmenpump xdth&rhinedriwt~puq~

The L e 0 is modified by a Note indicating that one AFW train, which

includes a motor driven pump, is required to be OPERABLE in MODE 4.

This is because of the reduced heat removal requirements and short

period of time in MODE 4 during which the AFW is required and the

insufficient steam available in MODE 4 to power the turbine driven AFW

pump.

APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in

the event that it is called upon to function when the MFW is lost. In

addition, the AFW System is required to supply enough makeup water to

repiace the steam generator SeCQRdaiy inventory, lost as the unit cools to

MODE 4 conditions.

In MODE 4 the AFW System may be used for heat removal via the steam

generators.

In MODE 5 or 6, the steam generators are not normally used for heat

removal, and the AFW System is not required.

ACTiONS &1

If one of the two steam supplies to the turbine driven AFW train is

inoperable, action must be taken to restore OPERABLE status within

7 days. The 7 day Completion Time is reasonable, based on the

following reasons:

Catawba Units 1 and 2 B 3.7.5-4 Revision No. i

AFW System

B 3.7.5

BASES

ACTIONS (continued)

a. The redundant OPERABLE steam supply to the turbine driven AFW

pump;

b. The availability of redundant OPERABLE motor driven AFW

pumps; and

c. The low probability of an event occurring that requires the

inoperable steam supply to the turbine driven AFW pump.

The second Completion Time for Required Action A.1 establishes a limit

on the maximum time allowed for any combination of Conditions to be

inoperable during any continuous failure to meet this LCO.

The 10 day Completion Time provides a limitation time allowed in this

specified Condition after discovery of failure to meet the LCO. This limit

is considered reasonable for situations in which Conditions A and B are

entered concurrently. The Bhill connector between 7 days and 10 days

dictates that both Completion Times apply simultaneously, and the more

restrictive must be met.

&1

With one of the required AFW trains (pump or flow path) inoperable in

MODE 1, 2, or 3 for reasons ether than Condition A, action must be taken

to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This Condition includes the

loss of two steam supply lines to the turbine driven AFW pump. The

72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on redundant capabilities

afforded by the AFW System, time needed for repairs, and the low

probability of a DBA occurring during this time period.

The second Completion Time for Required Action B.l establishes a limit

on the maximum time allowed for any combination of Conditions to be

inoperable during any continuous failure to meet this LCO.

The 10 day Completion Time provides a iimitation time allowed in this

specified Condition after discovery of failure to meet the LCO. This limit

is considered reasonable for situations in which Conditions A and B are

entered concurrently. The AND connector between 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and 10 days

dictates that both Completion Times apply simultaneously, and the more

restrictive must be met.

Catawba Units 1 and 2 B 3.7.5-5 Revision No. 0

AFW System

B 3.7.5

BASES

ACTIONS (continued)

c l

a n d u

When Required Action A.1 or B.Z cannot be completed withln the

required Completion Time, or if two AFW trains are inoperable in

MODE 1, 2, or 3, the unit must be placed in a MODE in which the LCO

does not apply. To achieve this status, the unit must be placed in at least

MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating

experience, to reach the required unit conditions from full power

conditions in an orderly manner and without challenging unit systems.

In MODE 4 with two AFW trains inoperable, operation is allowed to

continue because only one motor driven pump AFW train is required in

accordance with the Note that modifies the LCO. Although not required,

the unit may continue to cool down and initiate RHR.

1Li

If all three AFW trains are inoperable in MODE 1, 2, or 3, the unit is in a

seriously degraded condition with no safety related means for conducting

a cooldown, and only limited means for conducting a coddown with

nonsafely related equipment. In such a condition, the unit should not be

perturbed by any action, including a power change, that might result in a

trip. The seriousness of this condition requires that action be started

immediateiy to restore one AFW train to OPERABLE status.

Required Action B.1 is modified by a Note indicating that all required

MODE changes or power reductions are suspended until one AFW train

is restored to OPERABLE status. In this case, LCO 3.0.3 is not

applicable because it could force the unit into a less safe condition.

&1

In MODE 4, either the reactor coolant pumps or the RHR loops can be

used to provide forced circulation. This is addressed in LCO 3.4.6, "RCS

Loops-MODE 4." With one required AFW train inoperable, action must

be taken to immediately restore the inoperable train to OPERABLE

status. The immediate Completion Time is consistent with LCO 3.4.6.

Catawba Units 1 and 2 B 3.9.5-6 Revision No. 0

AFW System

B 3.7.5

SURVEILLANCE SEZ-KEd

WEQULREMENTS

Verifying the correct alignment for manual, power operated, and

automatic valves in the AFW System water and steam supply flow paths

provides assurance that the proper flow paths will exist for AFW

operation. This SR does not apply to valves that are locked, sealed, or

otherwise secured in posltion, since they are verified to be in the correct

position prior to locking, sealing, or securing. This SR also does not

apply to valves that cannot be inadvertently misaligned, such as check

valves. This Surveillance does not require any testing or valve

manipulation; rather, it involves verification that those valves capabie of

being mispositioned are in the correct position. The St3 is also modified

by a note that excludes automatic valves when THERMAL POWER is

5 10% RTP. Some automatic valves may be In a throttled position to

support low power operation.

The 31 day Frequency is based on engineering judgment, is consistent

with the procedural controls governing valve operation, and ensures

correct valve positions.

x5RLU52

Verifying that each AFW pump's developed head at the flow test point is

greater than or equal to the required developed head ensures that AFW

pump performance has not degraded during the cycle. Flow and

differential head are normal tests of centrifugal pump performance

required by Section XI of the ASME Code (Ref. 3). Because it is

undesirable to introduce cold AFW into the steam generators while they

are operating, this testing is performed on recirculation flow. This test

confirms one point on the pump design curve and is indicative of overall

performance. Such inservice tests confirm component OPERABILITY.

trend performance, and detect incipient failures by indicating abnormal

performance. Performance of Inservice testing discussed in the ASME

Code,Section XI (Ref. 3) (only required at 3 month intervals) satisfies this

requirement.

This SR is modified by a Note indicating that the SR should be deferred

until suitable test conditions are established. This deferral is required

because there is insufficlent steam pressure to perform the test.

Catawba Units 1 and 2 B 3.7.5-7 Revision No. 0

AFW System

B 3.7.5

SURVEILLANCE REQUIREMENTS (continued)

sL3.759

Phis SR verifies that AFW can be delivered to the appropriate steam

generator in the event of any accident or transient that generates an

ESFAS, by demonstrating that each automatic valve in the flow path

actuates to its correct position on an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or

othewise secured in the required position under administrative controls.

The 18 month Frequency is based on the need to perform this

Surveillance under the conditions that apply during a unit outage and the

potential for an unplanned transient if the Surveillance were performed

with the reactor at power. The 18 month Frequency is acceptable based

on operating experience and the design reliability of the equipment.

Phis SR is modified by a Note that states the SF? is not required in MODE

4. In MODE 4, the required A f W train may already be aligned and

operating.

SaEu5.4

This SR verifies that the AFW pumps will start in the event of any

accident or transient that generates an ESFAS by demonstrating that

each AFW pump starts automatically on an actual or simulated actuation

signal in MBBES 2 , 2, and 3. In MODE 4, the required pump may

already be operating and the autostart function is not required. The

18 month Frequency is based on the need to perform this Surveillance

under the conditions that apply during a unit outage and the potential for

an unplanned transient if the Surveillance were performed with the

reactor at power.

This SR is modified by two Notes. Note 1 indicates that the SR can be

deferred until suitable test conditions are established. This deferral is

required because there is insufficient steam pressure to perform the test.

Note 2 states that the SR is not required in MODE 4. In MODE 4, the

required pump may already be operating and the autostart function is not

required. In MODE 4, the heat removal requirements would be less

providing more time for operator action to manually start the required

AFW PUMP if It were not in operation.

Catawba Units 1 and 2 B 3.7.5-8 Revision No. 0

AFW System

0 3.7.5

SURVEILLANCE REQUIREMENTS (continued)

sE32235

This SR verifies that the AFW is properly aligned by verifying the flow

paths from the CSS to each steam generator prior to entering MODE 2

after more than 30 days in MODE 5 or 6. OQEfWBlblTY of AFW flow

paths must be verified before sufficient core heat is generated that would

require the operation of the AFW System during a subsequent shutdown.

The Frequency is reasonable, based on engineering judgement and other

administrative controls that ensure that flow paths remain OPERABLE.

To further ensure AFW System alignment, flow path OPERABILITY is

verified following extended outages to determine ne misalignment of

valves has occurred. This SR ensures that the flow path from the CSS to

the steam generators is properly aligned.

~ ~

REFERENCES 1. UFSAR, Section 10.4'9.

2. 10 CFR 50.36, Technicai Specifications, (c)(2)(ii).

3. ASME, Boiler and Pressure Vessel Code,Section XI.

Catawba Units 1 and 2 B 3.7.5-9 Revision No. 0

cew System

3.7.7

3.7 PLANT SYSTEMS

3.7.7 Component Cooling Water (CCW) System

LCO 3.7.7 Two CCW trains shall be OPERBBLE.:

APPLICABILITY: MODES 1. 2, 3, and 4

ACTIONS

CONDITION REQUIRED ACTION 1 COMPLETION TIME

A. One CCW train A.1 _______

________-----NOTE------

inooerable. Enter applicable

Conditions and Requit-ed

Actions of LCO 3.4.6,

"RCS LOOPS-MODE 4,"

for residual heat removal

loops made inoperable by

ccw.

Restore CCW train to 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s:

OPERABLE status.

B. Required Action and 3.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

associated Completion

Time of Condition A not AND

met.

8.2 Be in MODE 5. i 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

Catawba Units 1 and 2 3.7.7-1 Amendment Nos. 1891282

CCW System

B 3.7.7

B 3.7 PLANT SYSTEMS

B 3.7.7 Component Cooling Water (CCW) System

BASES

- -

BACKGROUND The CCW System provides a heat sink for the removal of process and

operating heat from safety related components during a Design Basis

Accident (BBA) or transient. During normal operation, the CCW System

also provides this function for various nonessential components, as well

as the spent fuel storage pool. The CCW System serves as a barrier to

the release of radioactive byproducts between potentially radioactive

systems and the Nuclear Service Water System (NSWS), and thus to the

environment.

The CCW System is arranged as two independent, full capacity cooling

loops, and has isolatable nonsafety related components. Each safety

related train includes two 50% capacity pumps, surge tank, heat

exchanger, piping, valves: and instrumentation. Each safety related train

is powered from a separate bus. An open surge tank in the system

provides sufficient inventory to protect the pumps from a [ack of net

positive suction head available (NPSHA) due to a moderate energy line

break. The pumps have sufficient NPSHA with the surge tank empty

provided the piping up to the tank is filled. The pumps on each train are

automatically started on receipt of a safety injection signal, and all

nonessential components are isolated.

Additional information on the design and operation of the system, along

with a list of the components served, is presented in the UFSAR,

Section 9.2 (Ref. 1). The principal safety related function of the CCW

System is the removal of decay heat from the reactor via the Residual

Heat Removal (RHW) System. This may be during a normal or post

accident cooldown and shutdown.

APPLICABLE The safety related design basis function of the CCW System is to remove

SAFETY ANALYSES waste heat from various components essential in mitigating design basis

events which require Emergency Core Cooling System (ECCS)

operation. The CCW System is also used to support normal operation.

The normal temperature of the CCW is 9QF, and, during unit cooldown to

MODE 5 (TmIde 20O0F),a maximum temperature of 128°F is

Catawba Units 1 and 2 B 3.7.7-1 Revision No. 0

CCW System

B 3.7.7

BASES

APPLICABLE SAFETY ANALYSES (continued)

assumed (Ref. I). This 120°F limit is to prevent thermal degradation of

the large pump motors supplied with cooling water from the CCW

System.

The CCW System is designed to perform its function with a single failure

of any active component, assuming a loss ob offsite power.

The CCW System also functions to cool the unit from RHR entry

conditions (Teald 350°F), to MODE 5 (Icod l a 2OO0F),during normal and

post accident operations. The time required to cool from 350°F to 200.F

is a function of the number of CCW and RHR trains operating. One CCW

train is sufficient to remove decay heat during subsequent operations with

TWlaa 200°F. This assumes a maximum service water temperature of

100°F occurring simultaneousiy with the maximum heat loads on the

system.

The CCW System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2).

LCO The CCW trains are independent of each other to the degree that each

has separate controls and power supplies and the operation of one does

not depend on the other. In the event of a DBA. one CCW train is

required to provide the minimum heat removal capability assumed in the

safety anabysis for the systems to which it supplies cooling water. To

ensure this requirement is met, two trains of CCW must be OPEFWBLE.

At least one CCW train will operate assuming the worst case single active

failure occurs coincident with a loss of offsite power.

A CCW train is considered OPERABLE when:

a. Both pumps and associated surge tank are OPERABLE; and

b. The associated piping, valves, heat exchanger, and instrumentation

and controls required to perform the safety related function are

The isolation of CCW from other components or systems not required for

safety may render those components or systems inopsrable but does not

affect the OPERABILITY of the CCW System.

Catawba Units 1 and 2 B 3.7.7-2 Revision No. 0

CCW System

B 3.7.7

BASES

APPLICAB ILlTY In MODES 1,2,3, and 4,the CCW System is a normally operating

system, which must be prepared to perform its post accident safety

functions, primarily RCS heat removal, which is achieved by cooling the

RHR heat exchanger.

in MODE 5 or 6, the requirements of the CCW System are determined by

the systems it supports.

Required Action A.l is modified by a Note indicating that the applicable

Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"

be entered if an inoperable CCW train results in an inoperable RHR loop.

This is an exception to LCO 3.0.6 and ensures the proper actions are

taken for these components.

If one CCW train is inoperable, action must be taken to restore

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining

OPERABLE CCW train is adequate to perform the heat removal function.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on the redundant

capabilities afforded by the OPERABLE train, and the low probability of a

DBA occurring during this period.

BLamiR2

If the CCW train cannot be restored to OPERABLE status within the

associated Completion Time, the unit must be placed in a MODE in which

the LCO does not apply. To achieve this status, the unit must be placed

in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The

allowed Completion Times are reasonable, based an operating

experience, to reach the required unit conditions from full power

conditions in an orderBy manner and without challenging unit systems.

SURVEILLANCE SR2ZZ.l

REQUIREMENTS

This SR is modified by a Note indicating that the isolation of the CCW

flow to individual components may render those components inoperable

but does not affect the OPERABILITY of the CCW System.

Verifying the correct alignment for manual, power operated, and

automatic valves in the CCW flow path to safety related equipment

provides assurance that the proper flow paths exist for CCW operation.

Catawba Units 1 and 2 B 3.7.7-3 Revision No. 0

CCW System

r3 3.7.7

BASES

SURVEILLANCE REQUIREMENTS (continued)

This SR does not apply to valves that are locked, sealed, or otherwise

secured in position, since these valves are verified to be in the correct

position prior to iocking, sealing, or securing. This SR also does not

apply to valves that cannot be inadvertently misaligned, such as check

valves. This Surveillance does not require any testing or valve

manipulation; rather, it involves verification that those valves capabie of

being mispositioned are in the correct position.

The 31 day Frequency is based on engineering judgment, is consistent

with the procedural controls governing valve operation, and ensures

correct valve positions.

sB3LL2

This SR verifies proper automatic operation of the CCW valves on an

actual or simulated actuation safety injection, Phase A isolation, or

Phase 5Isolation signal. The CCW System is a normally operating

system that cannot be fully actuated as part of routine testing during

normal operation. This Surveillance is not required for valves that are

locked, sealed. or othewise secured in the required position under

administrative controls. The 18 month Frequency is based on the need to

perform this Surveillance under the conditions that apply during a unit

outage and the potential for an unplanned transient if the Surveillance

were performed with the reactor at power. Operating experience has

shown that these components usually pass the Surveillance when

performed at the 18 month Frequency. Therefore, the Frequency is

acceptable from a reliability standpoint.

sR?Lz23

Phis SR verifies proper automatic operation of the CCW pumps on an

actual or simulated actuation signal. The CCW System is a normally

operating system that cannot be fully actuated as part of routine testing

during normal operation. The 48 month Frequency is based on the need

to perform this Surveillance under the conditions that apply during a unit

outage and the potential for an unplanned transient if the Surveillance

were performed with the reactor at power. Operating experience has

shown that these components usually pass the Surveillance when

performed at the 18 month Frequency. Therefore, the Frequency is

acceptable from a reliability standpoint.

Catawba Units 1 and 2 B 3.7.7-4 Revision No. 0

CCW System

B 3.7.7

BASES

REFERENCES 1. UFSAR, Section 9.2.

2. 10 CFW 50.36, Technical Specifications, (c)(a)(ii).

Catawba Units 1 and 2 B 3.7.7-5 Revision No. 0

1 Pt(s) Unit 1 is shutting down, in mode 4. Engineering reported that a recent test on

the 1B KC heat exchanger showed that fouling had reduced its heat transfer

capability below the minimum design value. Given the following events and

conditions:

NCS tempertture is 260 O F .

ND loops A and B are operating.

CAPT is tagged for maintenance.

CA pump 1B and CA pump 1A are isolated for differential pressure

testing.

"he plant is expected to remain in mode 4 for another 7 days.

What is the EARLIEST action required for the operators to tnke?

REFEREArCESPROF7DED: Tech Spec's 3.4-6,3.%6

A. Be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

B. Return at least one CA pump to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

C. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months o r be in mode 5

within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

D. Clean the 1B KC heat exchanger within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in mode 5

within 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months .

Distracter Analysis:

A. Correct answer: TS 3.7.6 cascades to TS 3.4.6 (see note).

B. Incorrect: Any action on the CA system would have immediate

response required.

Plausible: Restoring a CA pump couki resolve the TS 3.4.6 action

but not the TS 3.7.6 action.

C. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Plausible: This is the correct answer for TS 3.7.6, disrcganling the

cascading note.

D. Incorrect: must be in mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Plausible: if the candidate adds the time allowed to reach mode 3 to

the mode 5 action time in TS 3.7.6.

Level: RO Only

KA: ADM G 2.2.22 (3.5 / 4.1)

Lesson Plan Objective: PSS-KC SEQ 15

Source: New

Level of knowledge: comprehension

References:

1. Tech Spec's 3.4.6 and 3.7.6 - PROVIDED

. . ,. ,. .. . .

.......

Bank Question: 718 Answer: B

I Pt(s) Unit 2 is conducting a plant shutdown fkom 100% power. Given the following

events and conditions:

Reactor power is 6%

All manual actions have been taken as required in the procedures

Intemiediate Range channel N-36 fails HIGH.

Which ofthe following statements correctly describes how this failure affects the

reactor shutdown and subsequent operation of the Nuclear Instrumentation System?

A. The reactor will trip; the source range detectors will reenergize when N-35

decreases to the proper setpoint.

B. The reactor will trip; the source range detectors will have to be manually

reenergized.

C. The reactor will not trip; the source ranges will reenerghe when N-35

decreases to the proper setpoint.

D. The reactor WiIl not trip; the source ranges will have to be manually

reenergized.

Distracter Analysis:

A. Incorrect: The source range instruments will not automatically reenergize

Plausible: If the operator believes the remaining IR energizes the Source

Ranges, but knows the reactor trips.

B. Correct: The IR trip will occur when either IR channel increases to > 25%

equivalent. However this trip is blocked manually when P-10 is satisfied.

Since reactor power is given as 6%, when N-36 fails high, the IR high flux

trip will occur.

The Source Range instmnents will automatically reenergize when:

1. P-10 is not satisfied, 3/4 NIS PW 10% and

2. P-6, both IR < 10." amps

Othenvise they will need to be MANUALLY reenergized. With a high

failure of IR N-36, they will not automatically reenergize.

C. Incorrect The reactor will trip and the source range instruments will not

automatically reenergize

Plausible: Operator believes the IR trip is blocked and only one IR is

necessary bo energize the SR

D. Incorrect: The reactor will trip and the source range instruments will not

automatically reenergize

Plausible: Operator believes the IR trip is blocked hut knows both LR are

necessary to energize the SR.

Level: SRO Only 10CFR55.43@)5

KA: APE 033 AA2.08 (3.3/3.4)

Lesson Plan Objective: ENB Qbj: 9

Source: bank

Level of knowledge: analysis

References:

1. OP-CN-IC-ENB page 10,11,12,13

Ques-718.doc

DUKE BOWER CATAWBA OPERATlONS TRAiNlhlG

OBJECTIVES

-

L

Objective P

S

I I

5 Explain the function of each portion of the individual ranges when

given a block diag%rarnof each range.

1 6 Explain the function of all indications and controls associated with

(ENB.

-.+I 1

9 Describe the plant response to a given detector or instrtirnent

failure.

-

X

OP-CN-IGENB MIR TRAINING PURPOSES ONLY REV. 29

Page 3 of 41

DUKE BOWER

= . - ..... - .-_ . ...

CA TAWBA OPERATONS TRAlNlNG

. . . . , I - _

-

.

.A .

.

- ..

15. Level Adiust Potentiometer - Adji!stable test signal into level amp.

Enables adjustment of the trip level of various bistables.

16. &ghJFlux at Shutdowrr Switch - Two position switch.

a) Normal -allows circuit to provide High Flux at Shdtdown and

Containment Evacuation alarm when setpoint is exceeded.

b) Block-used during startup - Blocks High Flux at Shutdown Alarm and

Containment Evacuation Alarm.

c) With the Coritrol Power fuses biown or rernoved the containment

evacuation alarm wil! alarm and cannot be BLOCKED. The OFF

switch on the main control board must be depressed and held until

the fkses can be replaced or reinstalied.

E. Source Range interlocks

1. During reactor startup as power increases into the Intermediate Range

and increases above amps (1 out of 2 channels) perrnissive P-6 is

energized ailtorriatically. This permissive allows the operator to block the

Sf3 High Flux Reactor Trip ( I O 5 cps increasing).

2. When reactor power is reduced to less than P-6 the Si3 detectors are

autoniaticaily energized and the SR tiigh Flux Reactor Trip is placed

back in service for both channels.

3. When reactor power is increased greater than P-10 (2/4 PR charinels

greater than 10% reactor power) the SR detector power is further

assured to be de-energized by b!ocking voltage to the source range

detectors. This blocking of the voltage to the source range detectors is

automatically removed when reactor power is reduced below the P-10

setpoint and P-10is de-energized. Should P-10not de-energize (either

due to 3 out cf 4 PH chanriel P-20 bistabies not clearing or the 2 out of 4

and gate not goirig to the NOT P-10 state) then the SR detectors can

not be re-energized either automatically or manually. iAE intervention is

required should this event occur.

F. Intermediate Range Detector (OM#2)

-

1. Detector Compensated Ion Chamber - (gamma wmpensatiop required

on!y beiow 105amps).

a) Two Volume Detector

I) Outer volume Boron lined, sensitive to neiitroris and gammas.

2) Inner volume sensitive only to gamma - not boron lined.

3) Each volume generates current output

b) Inner volume current due to gamma interactions with N? gas.

c) Outer volume current generdted from neutron interactions with Boron

10 lining and gamma interactions with N2 Qas.

QP-CN-IGENB FOR TRAINING PURPOSES ONLY REV. 29

Page I O of41

_____

DUKE POWER CA TAWBA OPERA- TlONS TRAINING

-

IR COMPENSATED ION CHAMBER ~

I

NITROGEN FlLLE

GROMO

d) A high DC voltage is applied to the detector to provide for the

collection of all charged particles for each ionizing event. A voltage in

opposition to this voltage is applied to the compensation electrode.

Compensation is necessary because after sustained full power

operation of the reactor, there is an appreciable amount of residual

gamma flux. The compensation voltage is adjusted to cancel out the

signal due to gamma flux leaving an output from the detector which is

proportional to neutron flux only. Due to the high rate of neutron

pulses detected in the Intermediate Range the output from the

compensated ion chamber detector is direct current, and is coupled

directly to log current amplifies. (Obj. #7)

2. Under Compensation (OB4 #8)

a) Higher power indication than actual'power level.

b) Can result in attaining an extremely high SUR without seeing it on

SUR meters.

c) Can possibly prevent P-6 clearing and prevent automatic re-

energizing of SR detectors following a Reactor Trip.

d) Will energize P-6 earlier than expected during Reactor startup.

3. Over compensation (OBJ #8)

a) indicated power Iowes than true power, possibly pegged low.

b) Indicates a much higher SUR than actual SUR.

6) Will clear $4at higher actual flux level than normal.

OP-CN-IGENB FOR TRAINING PURPOSES ONLY REV. 29

Pagellof41

DUKE POWER CATAWBA OPEUATlONS TRAlNlNG

G. Intermediate Range Channels (IR) N35 and N36 (OBJ #5)

INTERMEDIATE RANGE

MGH VOLTKGE

CMAPENSKTWQ

VOLTKGE

MnER

P-6

LOW POWER ROD STOP

REACTOR TRIP

1. Range of indication IO-2 to I O 9 amps. (Startup to full power.)

2. Bistable Circuits

a) P-6 (112 SI? greater than 10-a amps) - allows the operator to block

the SW High bevel Reactor Trip.

b) bow Power Rod Stop (amps equivalent to 20% power) - prevents any

outward rod motion.

e) Reactor Trip (amps equivalent to 25% power).

3. Isolation Amp Feeds:

a) SUR Circuit

b) CIB Recorder - NIR 45

c) CIB Meter

4. Local Meter - On BR Panel Drawer

H. If?Circuits Outputs (OW #4)

1. Loa Level Arndifier

a) Receives De signal from detector.

b) Provides logarithm voltage output proportional to linear input current.

c) Output voltage to:

1) CIB and local meter calibrated in amps (IO- to I O 3 Amps)

OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV. 29

Page f 2 of 4f

DUKE POWR CATAWEA OPERATIONS TRAINING

2) Isolation Amp

3) Bistable Relay Drivers

2. Bistable Relay Drivers

a ) P-6 (I/% IR greater than 10 -" amps)

b) bow Power Rod Stop

1) Current equivalent to 20% full power (1/2 channels).

2) Rod withdrawal stop in manual or automatic.

3) Blockable at P-10 (2/4 PR greater than 10% power).

c) Reactor Trip

1) Current equivalent to 25% full power (f/2 channels)

2) Blockable at P-I0

3. Isolation Amplifier

a) isolates kR channel from remote equipment.

b) Provides output for following:

1) SUR Circuit

(a) Converts rate of change of power level to SUR in DPM.

(b) Reads out on C/B.

2) C/B Indication - Meter calibrated in amps (1Q-I to IO3).

3) C/B Recorder

(a) 1NR-45 two pen recorder.

(b) Records IR level in amps when selected.

I. la Drawer Panei(0BJ #6)

1. Ampere Neutron Level Meter

a) Indicates current output of detector

b) Indicates in amps - Eight decades (IOmf1to amps)

2. Instrument Power "ON" Lam0 - 118 voits AC instrument power applied to

drawer.

3. Control Power "ON" Lame - 118 voits AC control power applied to driver

assembly controi circuits.

4. - Indicates OPERATION SELECTOR switch is in

a position other than "NORMAL".

OP-CMiGENB FOR TaAlNING PURPOSES ONLY REV. 29

Page73of41

Bank Quesstion: 502.1 Answer: A

1 Pt(s) A LOCA occurred on Unit 1 at 2:00 AM. Given the following events and

conditions:

0201 A containment air release was in progress and was immediately

terniinated.

0205 The control room operators are responding to the events in E- 1.

02 I0 The OSM has assumed the role ofthe Emergency Coordinator,

0215 A Site Area Emergency is declared

0230 Completed initial notification of the State and Local authorities for

declaration of a site area emergency

0240 The OSM notes the following plant conditions and determines the

appropriate classification for the following indications:

Containment pressure indicates 18 psig

Containment Hydrogen concentration is 1.5%

Containment valve VQ-2B failed to isolate on phase A signal.

Core exit T/Cs indicate temperatures of 1 3 W T

RVLIS lower range level indicates 90%

IEMF-53AB indicate 85 R/hr

The Operators enter FR-C. 1 (Response to Loss ofcore Cooling)

If the BSM determines the appropriate classification when the conditions are

first indicated (do not assume it takes 15 minutes for the OSM to classify the

events), which one of the following statements correctly describes the first

notification that is reauired?

REFERENCES PRO WDED: W/Q/M5000/01 (Class$cution of

EmergenqJ

A. Notify the State and Local authorities that the plant has declared

a general emergency no Later than 0255.

B. Notify the NRC Headquarters Operations Center of plant

conditions no later than 0300.

C. Notify the State and Local authorities that the plant has declared

a general emergency no later than 0310.

D. Provide the first follow-up notification for a site area emergency

to the State and Local authorities no later than 0315.

Dlstraeter Analysis: Comparison to RP/O/A/5000/001 EALS shows:

- Cladding - 5 pts - T/Cs > I200 'F = core cooling red path (note:however,

EMF-53AW do not meet criteria for clad failure -must be >114 WIT)

Ques-SOt. 1.doc

- NC integrity - 5 pts - LOCA has occurred

- Containment integrity - 1 point for potential loss

- 18 psig - criteria is I5 psig for potential loss - YES

- 1.5% If2 - criteria is 9% H2 far potential loss - NO

- In FR-C. 1Red Path for 15 minutes -Not until 0255

The OSM meets the criteria for a GE at 0240

A. Correct: Meets criteria for a GE at 0240 .The OSM has 15 minutes

to notify the State and Local Autholdtics - must be completed by

0255.

B. Incorrect: NRC is required to be notified within one hour of the

declaration of an emergency clawification (event time), no Later than

03 15 - not from when the event first occurs.

Plausible: Notification o f m C is a high priority communication. If

the candidate thinks that NRC notification must occur within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

of the time that the LOCA occurred at 0200.

B. Incorrect: The events are classified at the General Emergency level

at 0240. The OSM has 15 minutes to notify the State and Local

Authorities - must be completed by 0255.

Plausible: The core cooling red path must be in for 15 minutes if the

escalation tu S E is made on the basis of the core cooling red path -

this wodd be correct if the candidate does not recognize that the high

containment pressure is suficient to declare a GE at 0240.

D. Incorrect: Follow-up reports for lesser classifications are not made

Plausible: One hour from notification of State and Local authorities

for the S A E - this is correct if the candidate does not recognize the

classification of the general emergency.

Level: SRO Only 10CFR55.43(b)5

KA: APE WE03 64.35 (2.213.6)

Lesson Plan Objective: SEP Obj: 2,16

Source: Mod Ques-502, Ques-109

Level of knowledge: comprehension

References:

1. OP-CN-EP-SEP pages I I Bi 12

2. RP/O/A/5000/001 page 1-3

3. Rp/O/A/5000/OOI End 4.1pages 1-5

4.lZP/O/B/5000/013 Encl 4.1 page 2

DUKE POWER CATAWBA OPERATlONS TRAINING

Objective

List the four levels of emergency classification at Catawba Nuclear

Station.

When given a set of plant conditions and access to reference

materials, correctly classify an event using RP/O/N5QOO/OO1.

Explain personnel responsibilities during an emergency at Catawba

Nuclear Station.

0 List the persons (by title) who may assume the role of emergency

coordinator.

e State who is responsible for activation of the emergency warning

system (sirens).

e State who is responsible for making recommendations to offsite

agencies.

State your assembly point.

State the emergency tevels that require various actions to be taken.

OSC activation.

0 TSC activation

EQF activation

0 Site Assembly

e Mandatory site evacuation

Name the locations for emergency evacuation sites.

Explain how to determine which emergency evacuation site to use in

case of an emergency.

List the offsite agencies which should be notified during an emergency.

Explain the meaning of emergency sirens sounding.

State the size of the Emergency Planning Zone (EPZ).

Describe how to conduct site assembly/evacuation according to

RP/O/N5000/010 and NSD 114.

OP-CN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22

Page 3 of 16

DUKE POWER - CA TAWBA OPERA TIONS TRAINING

Objective

Describe the symptoms that may or will require a site assembly or

evacuation Der RPIQ/A/5000/010 and NSB 114.

Describe the procedure for securing from a site assembly or

evacuation.

Explain the purpose of the NSB for Site Assernbly/Evacuation per NSD

114.

Describe the procedure for site assembly during the following per NSD

1q4:

inability to reach assembly point

working in RCA

Summarize the procedure for site assembly per NSD 114 including:

who may enter the plant during a site assembly

the procedure lo account for personnel

State the time frames in which immediate and follow-up notifications

are to be made to various ofkite agencies.

Prepare and evaluate Emergency Notification Forms for both initial and

follow-up notification for any given accident scenario.

When given a copy of RP/QIN5500/002,apply the Immediate Actions

required for a Notification of Unusual Event.

Summarize the subsequent actions required for the Notification of

Unusual Event procedure per RP/O/N5000/002:

Describe the procedure to terminate the emergency.

Describe the procedure to give a follow-up message.

Z . For events lasting greater than one hour.

2. Significant change in the situation.

3. Escalation to a higher classification.

When given a copy of RP/0/A/5(300/003, apply the Immediate Actions

required for an Alert.

OP-CN-EP-SEP FOR TRAINING PURPOSES ONLY REV. 22

Page 4 of 16

DUKE POWER CA TAWBA OPERATIONS TRAINING

4. Immediate Actions Burinq an Emerqency

a) Compare actual plant conditions to the Emergency Action bevel(s)

listed in Enc. 4.1, RPIO/Af5000/001 then declare the appropriate

Emergency Class. (Obj. #2)

NOTE: Using the appropriate forms, have students classify an event and

fill out an Emergency Notification Form (ENF). (Qbj. # 2 ) (Obj.

I 7)

b j Make notifications to state and county agencies within 15 minutes

of declaring the event. Notify the NRC immediately after the states

and counties, not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Notify all others as soon as

possible after the states, counties, and NRC are notified. (Qbj.

16) (Obj. #7)

1) NC and SC

2) York, Gaston, Mecklenburg counties

3) NRC (ENS phone)

c) Refer to the applicable Emergency Response Procedure (RP) for

the classification found in Enc. 4.1, RP/O/N5000/001:

Notification of Unusual Event, WP/O/A/50Q0/002(Obj. 18)

Alert, RP/O/Ai5000/003 (Obj. 28)

Site Area Emergency, RP/8/A/5000/004 (Obj. 22)

General Emergency, RP/0/8/5000/005 (Obj. 24)

d) In the event of an "Urgent Condition" problem that has radiological

implications, but does not require EP/RP implementation,

Radiation Protection Shift Supervisor shall be contacted to respond

with Protective clothing and equipment to accommodate 1 RP

Technician and 3 workers in an "Urgent Response Kit".

5. Subsequent Actkms Durina an Emeruency

a) Refer to the applicable Emergency Response Procedure (KP) for

the classification:

1) Notification of Unusual Event, RP/O/A/5000/002 (Obj. 19)

2) Alert, RP/O/N5000/083 (Obj. 21)

3) Site Area Emergency, RP/O/A/5000/004 (Obj. 23)

4) General Emergency, RPIO/N5000/005 (Obj. 25)

b) Make follow-up notifications to states and counties (Obj. 16), (Obj.

26)

I)Refer to the most recent revision of KP/O/A/5500/006A and B.

OP-CN-EP-SEP FOR TRAlNl" PURPOSES ONLY REV. 22

Page 11 of 76

_. DUKE POWER.......~ ..._.-- CATAWEA

.......... -OPERATIONS TRAINING

__. ....

.....

2) Notification of Unusual Event, Alert, Site/Area Emergency and

General Emergency

(a) Hourly until emergency closed out

or

(b) If any significant change in situation

or

(c) As agreed upon with each agency (not to exceed 4

hours).

c) Assess plant conditions to determine the need to change

emergency classifications :/Qbj. 27)

1) Raise classification by declaration of Alert, Site Area or

General Emergency. Make initial notification within 15 minutes

of declaration.

2 ) Lower classification by declaration of an Alert or NOUE per

Encl. 4.3 of RP/O/A/500Q/004,Site Area Emergency or

RP/O/A/5008/003, Alert respectively. Make initial notification

within 15 minutes of declaration.

3) Termination of an emergency varies with the classification from

which the emergency is being terminated. Refer to

RF/0/,4/5000/002 through 006A.

d) Brief plant personnel as to status of the emergency via the plant

page.

e) Make Protective Action recommendations to states and counties,

as appropriate.

f) Ensure emergency worker doses do not exceed the limits of

RPfOfA/50O0/018, Emergency Worker Dose Extension, as

appropriate.

g) Provide turnover to the TSC Emergency Coordinator, as

appropriate.

C. Personnel Responsibilities (Qbj. #3)

1. Initially the OPS Shift Manager is the individual who assumes the role

of Emergency Coordinator.

2. Subsequently the Station Manager or his designee will assume the role

of Emergency Coordinator. After EQF activation, the EOF Director is

responsible for overall emergency management.

OP-CKEPSEP FOR TRAINING PURPOSES ONLY REV. 22

Page 12 of 16

Procedure No.

Duke Power Company

Catawba Nuclear S k i o n

015

Classification of Emergency

(ISSUED) .. PDF Format

Classification of Emergency

1. Symptoms

1.1 Notification of Unusual Event

1.1.1 Events are in process or have occurred which indicate a potential degradation

of the level of safety of the plant.

1.1.2 No releases of radioactive material requiring offsite response or monitoring

are expected unless further degradation of safety occurs.

1.2 Alert

1.2.1 Events are in process or have occurred which involve an actual or potential

substantial degradation of the level of safety of the plant.

1.2.2 Any releases are expected to be Iiniited to small fractions of the EPA

Protective Action Guideline exposure levels.

1.3 Site Area Emergency

1.3.1 Events are in process or have occurred which involve actual or likely major

failures of plant functions needed for protection of the public.

1.3.2 Any releases are not expected to exceed EPA Protective Action Guideline

exposure levels except near the site boundary.

1.4 General Emergency

I .4.1 Events are in process or have occurred which involve actual or imminent

substantial core degradation or melting with potential for loss of containment

integrity.

1.4.2 Releases can be reasonably expected to exceed EPA Protective Action

Guidelines exposure levels offsite for more than the immediate site area.

2. Immediate Actions

2.1 Determine operating mode that existed at the time the event occurred prior to any

protection system or operator action initiated in respome of the event.

2.2 the plant was in Mode 1-4 and a valid condition affects fission product barriers,

proceed to Enclosure 4.1.

RP/~/k~5000/001

Page 3 of 3

__ 2.3 -

IF a General Emergency is NOT declared in Step 2.2 the condition does not affect

fission product barriers, review the listing of enclosures to determine if the event is

applicable to one the categories shown.

__ 2.4 Compare achal plant conditions to the Emergency Action Levels listed, then declare the

appropriate Emergency Class as indicated.

m _ 2.5 Implement the applicable Emergency Response Procedure (RP) for that classification

and continue with subsequent steps of this procedure.

Notification of Unusual Event RP/O/N5000/002

Alert RP/O!N5000/003

Site Area Emergency RFVO~A/5000/004

General Emergency RP/O/N5000/005

3. Subsequent Actions

3.1 To escalate, de-escalate, or terminate the Emergency, compare piant conditions to the

Initiating Conditions of Enclosures 4.1 through 4.7.

- 3.2 Refer to enclosure 4.9, Emergency Declaration Guidelines, as needed.

4. Enclosures

4.1 Fission Product Barrier Matrix

4.2 System Malfunctions

4.3 Abnormal Rad Levels/Radiological Effluent

4.4 Loss of Shutdown Functions

4.5 Loss of Power

4.6 Fires/Explosions and Security Events

4.7 Natural Disasters, Hazards and Other conditions Affecting Plant Safety

4.8 DefinitiondAcronyms

4.9 Emergency Declaration Guidelines

4.10 Radiation Monitor Reading for Enclosure 4.3 EALs

Enclosure 4.1

RP/O/NSOOO/OOI

Fission Barrier Matrix Page 1 of 5

Use E.4Ls to determine Fission Product Barrier status (Intact, Potential Loss, or Loss). Add points for all 3 barriers. Classify according to the table below.

Note 1: This table is only applicable in Modes 1-4

Note 2: Also, an event (or multiple events) could QCCW which results in the conclusion that exceeding the Loss or Potential. Loss tkresholds is I&IMINENT

(ie., within 1-3 hours). In this IMMINENT LOSS situation, use judgement and classify as if the thresholds are exceeded.

Note 3: When determining Fission Product Barrier status, the Fuel. Clad Barrier should be considered to be lost or potentially lost if the conditions for the

Fuel Clad Barrier loss or potential loss EALs were met previously during the event, even if the conditions do not currently exist.

Note 4: Critical Safety Function (CSF) indications are not meant to include transient alarm conditions which may appear during the start-up of engineered

safeguards equipment. A CSF condition is satisfied when the alarmed state is valid and sustained. The STA should be consulted to affirm that a CSF has

been validated and the appropriate functional restoration procedure has been implemented prior to the CSF being used as a basis to classify an emergency.

-- - -- - _.

EAL # Unusual Event E m# Alert EAL# Site Area Emergency EAL # General Emergency

- - -

4.1.11.1 Potential Loss of 4.1 A . 1 JASS Potential Loss 4.13.1 LOSS =Potential LOSS 1.1.G. 1 Loss of All Three Barriers

Containment of 0fBoth

Nuclear Coolant System Nuclear Coolant System

Fuel Clad

-- -- -

4.1.U.2 Loss of Containment 4.1.A.2 Loss Potential Loss 4.1.S.2 Loss 4.1 .G.2 Loss of Any Two Barriers

of An,

Fuel Clad Potential Loss Potential Loss of the Third

Combinations of ~0th

Nuclear Coolant System

ANI)

Fuel Clad

__ _ _

4.1.A.3 Potential Loss of 4.1.s.3 Loss o f Containment

Containment -

AND

-

AND Ims Potential Loss

Loss 3Potential Loss o f h y Other Barrier

of Any Other Barrier

Enclosure 4.1 RP/@A/5000/001

~ F ~ SBarrier

S ~ O QMatrix Page 2 of 5

NOTE: If a barrier is affectec it has a single point value based on a potential 10s: 3r a 10s: Not Applicable is

F

Barrier

Containment

I

included in the table as a place holder only, and has no point value assigned.

Points (1-5) I potential Loss (X)

...

I

Loss (XI

3

Classification

I

Fuel Glad 5 7- 10 Site Area

Emergency

I I

1. Compare plant conditions against the Fission Barrier Matrix on pages 3 through 6 of 6.

2. Determine the potential loss or loss status for each barrier (Containment, NCS and Fuel Clad) based on the EAL symptom description.

3. For each barrier, write the highest single point value applicable for the banier in the Points column and mark the appropriate loss column.

4. Add the points in the Points column and record the sum as Total Points.

5 . Determine the classification level based on the number ofTota1 Points.

6 . In the table on page 1 of 6, under the classification column. select the event number (e& 4.1 .A.1 for Loss ofNuclear Coolant System) that best

fits the loss of barrier descriptions.

7. Using the number (e.g. 4.1 .A.I ) select the preprinted notification form and complete the required information for Emergency Coordinator

approval and transmittal.

Enclosure 4.1

Fission Barrier Matrix

4.1.C CONTAINMENT BARRIER 4.1.N NCS BARRIER 1.1.F FUEL CLAD BARRIER

I

POTENTIAL LOSS

(1 Point)

-

I LOSS -

(3 Points)

POTENTIAL LOSS

(4 Points)

- 1,OSS -

(5 Points)

POTENTIAL. LOSS

(4 Points)

I LOSS -

( 5 Points)

1. -Safee Function Status 1. Critical Safehi Function Status 1. Critical Safetv Function Stam

- Containment-RED Not applicable NCS Integrity-Red Not applicable . Core Coolig-

Orange

Core Cooling-Red

Core cooling-RED

Path is indicated

for ,I 5 minutes

Heat Sink-Red

. Heat Sink-Red

2. Containment Conditions 2. NCS Le& Rate 2. PhZmarv Coolant Aetivitv Level

Containment

Pressure > 15 PSIG

Rapid unexplained

decrease in

Unisolable leak

exceeding the

. GREATERTHAN

available makeup

rn 9 o t applicahle Coolant Activity

GREATER THAN

containment capacity o f one capacity as 300 pCwkc Dose

H2 concenmtion > pressure following charging pump in indicated by a loss Equivalent Iodine

9% initial increase the normal of NCS subcooling. (DEI) 1-131

- Containment

pressure greater than

Containment

pressure or sump

charging mode

with letdown

isolated.

3 psiig with less than level response not

one full train ofNS consistent with

and a VX-CARF LOCA conditions.

operating.

CONTINIJED CONTINUED CONTINUED

Enclosure 4.1 RP~olA/5OOO/QO1

Fission Barrier Matrix Page 4 of 5

4.1.C CONTAINMENT B U E R

POTENTIAL I,OSS - I.OSS - POTENTIAL LOSS - LOSS - POTENTIAL LOSS - LOSS -

( I Point) (3 Points) (4 Points) (5 Points) (4 Points) (5 Points)

I

3. Containment Isolation Vy!.v Status After . SG Tribe Ruoture

I I

. Containment Raai?.ti.pMunitoring

Containment Isolation Actuation

Not applicable Containment Primary-to- Indication that a Not applicable Containment

isolation is Secondary leak SG is Ruptured ani radiation monitor

incomplete and a rate exceeds the has a Non-IsoIable 53 A or53 B

release path from capacity of one sccondary line faul reading >11.7 R h

containment exists charging pump in

the normal Indication that a

charging mode SG is ruptured and

with letdow a prolonged releas,

isolated. of contaminated

Secondary coolant

is occurring from

the affected SG to

the environment

4. SG Secondarv Side Release With Primarv-to- I. Containment Radiation Monitoring .Emergencv CoordinatormOF Director

Secondarv Leakage

Not applicable Releaseof

secondary side to

Not applicable . Not applicable

Judeement

h y condition, including inability to monitor

the barrier, that intbe opinion of the

the environment Emergency CoordinatorlEOF Director

with primary to indicates LOSS or POTENTIAL LOSS of

secondary leakage the fuel. clad barrier.

GREATER THAN

Tech Spec

allowable -

END

CONTINUED CONTINUED

?

a

Enclosure 4.1 RF/O/B/5000/013

Events Requiring IMMEDIATE NRC Notification Page 2 of 2

irements for the following events as soon a5 practical after the o( urrence becomes known to the licensee

Event Description Reporting Requirement

B Declared emergency classification as specified in Notify the NRC Operations Center immediately after

Rp/0/A/5000/00 1, "Classification of Emergency". notification ofthe appropriate state or local agencies

Emergency and not later than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the time one of the

3assification B Change from one emergency classification to another emergency classes is declared.

Yotifications

w Termination of an emergency classification Activate the Emergency Response Data System

(ERDS) as soon as possible but not later than one

Any further degradation in the level of safety of the plant or hour after declaring an Alert or higher emergency

other worsening plant conditions, including those that classification.

require the declaration of any of the emergency classes, if

such a declaration has not been previously made

The results of ensuing evaluations or assessments of plant

conditions

The effectiveness of response or protective measures taken.

Information related to plant behavior that is not understood

- As a courtesy in situations deemed necessarq..

CNS CRITICAL SAFETY FUNCTION STATUS TREES PAGE NO.

EP/l/N/5OOO/F-O Core Cooling - Page 1 of 1

FK-L . I

GO TO

FR-C .I

1

' KtACTOR VESSSI.

LOWER RANOE LEVF.1.

GREATER I R A N 41%

N y

YES

GO TO

FR-C .2

. . C0KEFXITl:Ca

LESS THAN 1200°F

GO TO

FR-C .2

RFACIOR VESSEL

LOWER RANGE LEVEL

GRFATtRTtlAN 4116

L - -

"u'

AT LEAST ONE

NC PUMP ON

- -

I

NC' SUBCOO1.ING

RASEU ON CORE

EXIT TICS GREATER

THAN 0'F

-

No '

YES

-.

I

,'.-~.-

? CSFSAT

.. i

.~*

Bank Question: $09 Answer: C

1 Pt(s) Unit 1 was operating at 100% power when a large break LOCA occurred at

0200. The control room operators are responding to the events in E-I (Loss

of Reactor or Secondary Coolant). The OSM has assumed the role of the

Emergency Coordinator.

-

Given the following conditions and events at 0245:

Containment pressure indicates 16 psig

Containment Hydrogen concentration is 8.0%

Core exit T/Cs indisate temperatures of 1 100 "F

RVLIS lower range level indicates 30%

EMF 51a and b indicate 85 R/hr

Subcooling margin indicates -25T

Assuming that all required EQP actions were taken, what is the sorrecl

classification for the emergency event?

REFERENCES PROI'IDED

RP/0/A/5700/000

F-0

A. Notification of an Unusual Event

B. Alert

C. Site Area Emergency

D. General Emergency

~~~

Distracter Analysis: Classification as follows using N W C EALs:

Containment RED path = 1 EAL point (potentid failure)

NGS Barrier failure = 5 EAL points

Core cooling RED path = 4 EAL point (potential failure)

Total = 10 EAL points = SAE

A. Incorrect: the correct classification is an SAE

Plausible: if the candidate does not recognize the orange path or the

NCS barrier failw - with the red path for containment - this results

in 1 EALpoint=NOUE

B. Incorrect: the correct classification is an SAE

Plausible: based an LOCA table only 4 EAL points from orange path

in core cooling

C. Correct Answer:

Ques-109.doc

D. Incorrect: the correct classification is an SAIJ

Plausible: based on reasonable guess and misunderstanding EAL

philosophy - conditions we very bad

Bank Question: 491.1 Answer: C

I Pt(s) Unit 1 experienced a LOCA with a breach of containment at 0200. The

OSM a~sumedthe duties of the Emergency Coordinator and declared a

general emergency at 02 10. The initial recommended protective actions at

0225 were as foilows:

Shelter zones A2,A3, B2, C2, D2, E2, F2, F3

Evacuate zones AO; Ai, B1, CI, D1, El, FI

At 0235, the initial dose projection information was presented to the OSM.

Given the following conditions at 0245:

Wind direction = 450"

Windspeed-4MPH

e Projected dose at the site boundary

TEDE = 250 mrem

CDE thyroid = 400 mrem

Containment radiation levels

1EMF53A = 950 R/hr

1EMF53B = 955 R/hr

Which one of the following protective action recommendations arc correct in

accordance with W/O/A/5000/05?

REFERENCES PROPTDED - RP/0/A/5000/005

A. Change the protective action recommendation in zones A1 and

B1 from evacuate to shelter.

B. Extend the evacuation to zones A2, A3, B2,62

C. Extend the evacuation to zones D2, E2, F2

D. The initial set of protective action rccommendations remain in

effect with no changes required.

-

Distracter Analysis:

A. Incorrect: - wrong zones for evacuation - never reduce actions

Plausible: - will get this answer if the candidate enters the wrong

table - the one that has wind speed > 5mpb

i%. Incorrect: - wrong zones for evacuation

Plausible: will get this answer if the candidate enters the PAR table

with a reciprocal wind direction

C. Correct answer

D. Incorrect: - wrong zones for evacuation

Plausible: - if the candidate thinks that no changes are required to be

made until the first 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months update - or if the candidate thinks that gap

activity has not been exceeded.

Level: SRO Only lOCFR55.43@)5

EL4: ADM G 2.4.44(2.1 / 4.0)

Lesson Plan Objective: SEP Obj: 24

Source: Bank

Level of knowledge: analysis

References:

I. OF-CN-EF-SEPpages 12,15, 16

2. RP/O/A/5800/005 -pages 1-6

3. RPi)~A/SOOO/QOSEncl4.2 and 4.3

Ques-91.1 .doc

DUKE POWER CATAWBA OPERATlONS TRAINING

Objective

. ..... . .i. '.

I

.

.,,..,,; . . :..

.:... ' .

State the exposure limits for emergency workers. .,., .~'.. :.... . .., .

.>..,

When given a copy of RPIOIA/5000/005, apply the Immediate Actions

required for a General Emergency.

Expiain the EPA Protective Action Guides and Recommendations.

Explain the on-site and off-site protective actions.

per RP/OIA/5080/005:

Shew how to use Protective Action Zone Determination Tables.

Summarize the procedure for follow-up notifications per

RP/OIAl5000/006Aand 8:

OP-GN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22

5 Of f 6

DUKE POWEff CA TAWBA OPERA TlONS TRAlNlffG

2) Notification of Unusual Event, Alert, Site/Area Emergency and

General Emergency

(a) Hourly until emergency closed out

or

(b) If any significant change in situation

Of

(6) As agreed upon with each agency (not to exceed 4

hours).

c) Assess plant conditions to determine the need to change

emergency classifications :(Obj. 27)

1) Raise classification by declaration of Alert, Site Area or

General Emergency. Make initial notification within 15 minutes

of declaration.

2) Lower classification by declaration of an Alert or NOUE per

End. 4.3 of RF/O/A/5000/004, Site Area Emergency or

RP/O/N5000/003, Alert respectively. Make initial notification

within 15 minutes of declaration.

3) Termination of an emergency varies with the classification from

which the emergency is being terminated. Refer to

WF/O/Af5000/002 through 006A.

d) Brief plant personnel as to status of the emergency via the plant

page.

e) Make Protective Action recommendations to states and counties,

as appropriate.

f) Ensure emergency worker doses do not exceed the limits of

RPIOIN50001018, Emergency Worker Dose Extension, as

appropriate.

g) Provide turnover to the TSC Emergency Coordinator, as

appropriate.

C. Personnel Responsibilities (Obj. #3)

1. Initially the OPS Shift Manager is the individual who assumes the role

of Emergency Coordinator.

2. Subsequently the Station Manager or his designee will assume the role

of Emergency Coordinator. After OF activation, the OF Director is

responsible for overall emergency management.

f

OB-CNdP-SEP FOR TRAINING PURPOSES ONLY REV. 22

Page I 2 of f 6

DUKE POWER CATAWBA OPERATIONS TRAINING

-

3. Site Evacuation (Obj. #4) May occur for Site Area Emergency and

always occurs for General Emergency, must be Dreceded bv a Site

Assembly. All non-essential site personnel should proceed to one of

two Evacuation Sites based on site selection criteria.

a) Site Newport Newport Tie Station on Mt. Gallant Road near SC

~

Hi way I 6 1 in York County (5 miles SW). (Obj. #5)

b) Site Allen - Allen Steam Plant off Southpoint Road in Gaston

County (12 miles N). (Obj. #5)

4. Evacuation Site Selection Crtieria: (Obi. #6)

a) Site Allen will be sekected if the wind speed is less than 5 mph.

b) If wind speed is greater than or equal to 5 mph

1) Select Site Allen if the wind is from 0 to 144.9 degrees or

255.1 to 360 degrees.

2 ) Select Site Newport if the wind is from 145 tu 255 degrees.

5. Evacuation of Publie - Ordered by states or counties, Duke Power can

only make a recommendation. (Qbj. #3)

6. Securing from a Site Assembly andhr a Site Evacuation (Obj. #12)

a) The decision to secure from a site assembly will be made by the

QSM or the TSC Emergency Coordinator, as appropriate.

b) The decision to secure from a Site Evacuation will be made by the

TSC Emergency Coordinatur.

F. Offsite Emergency Planning

-

I. Emeraencv Planninq Zones Evacuation plans are prepared for the

public in and about a 10 mile radius from the plant called the IO mile

Exposure Pathway Zone (EPZ), or Plume Exposure Zone. Other

planning efforts are taken in the 50 mile radius from the plant called the

50 mile Ingestion Pathway Zone (IPZ). (Obj. #9)

-

2. PromDt Alertins Svstem In order to alert the public to a problem at the

station, an outdoor warning system of sirens has been installed in the

10 mile EPZ These sirens can be activated only by the counties

warning point for either a radiological emergency or a civil emergency

(tornado, flood or nuclear attack). Sirens alert public to tune N o r

radio to AS (Emergency Alert System) for further information or

directions. (Obj. #8)

3. Notification Svstem - After hearing a siren, the public is to turn on a

radio or TV and listen for an Emergency Alert System (AS) message.

The message could be: (Obi. #8)

a) Information only about the emergency.

b) An instruction to evacuate homes, offices, schools, factories, etc.

OP-CN-EP-SEP FOR TRAINING PURPOSES ONLY REV. 22

DUKE POWER CATAWBA OPERATIONS TRAlNlNG

c) An instruction to remain indoors for sheker until further instructions

are available.

-

4. Sheltering The states and counties have designated shelter space

available (located outside the 10 mike EPZ in the host counties) for

everyone in the 10 mile EPZ (EPZ counties) with food, water, and a

place to sleep until they are instructed to return home.

2.2 Conducting a Site Assembly or Preparing the Site for an Evacuation

A. Referencing to RP/O/N5000/010 and NSD 114, discuss the foliowing:

1. Symptoms of entry. (Obj. #I 1)

2. Immediate and subsequent actions for Site Assembly and Evacuation.

(Obj. # I O )

3. Locations for Site Evacuation and basis for selection. (Obj. #5) (Obj.

6)

4. Actions to take if an individual cannot reach their assembly point. (Obj.

I 4)

5. Actions to take if working in the RCA/RCZ and wearing protective

clothing. (Obj. #$4)

6. Who may enter the protected area during a Site Assembly? (Obj. #15)

2.3 NRC Notification Requirements (Obj. #7)

A. Referencing to RPIO/N5000/013, discuss the actions required in

communicating to the NRC. Refer to enclosures to clarify notification

requirements.

2.4 Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release

A. The Radiation Protection Manager in conjunction with available medical

advice shall control the distribution of KI tablets.

B. KI tablets should be given to:

1. Persons suspected of having been in the affected area.

2. Persons present in the affected area.

3. Persons who will enter the area while a significant amount of

radioiodine is present.

3. SUMMARY

3.1 Review Lesson Plan Objectives and answer student questions.

OP-CN-EP-SEP FOR TRAlNlNG PURPOSES ONLY REV. 22

Page f6 of f 6

Duke Power Company Procedure No.

Catawba Nuclear Station RPA)/A/5000/005

Revision No.

041

General Emergency

~ ~

Electronic Reference No.

* * * * * * * * * UNCONTROLLEDFORPRINT * * * * * * * * * *

(ISSUED) - PDF Formiit

RP/O/A/5000/005

Page 2 of 7

General Emergency

1. Symptoms

1.1 Events are in process or have occurred which involve or imminent substantial core

degradation or melting with potential for loss of containment integrity.

2. Immediate Actions

NOTE: 1. Lines in left margin are for place keeping. Immediate actions may be perfornied

simultaneously.

2. Security events may require the suspension of access to and movement about the site.

Staffing and activation of the on-site emergency response facilities could complicate

or interfere with security operations resulting in unwarranted casualties.

__- IF a security event exists, discuss the feasibility of conducting a site assembly and activating the

TSC/OSC with the Security Shift Supervisor at 5765 or 5766.

__ site assembly and activation ofthe TSC/QSC are not feasible, refer to the following

procedure enclosures for guidance and NIA the associated steps in this procedure under

Immediate Actions concerning site assembly and ERO activation:

__ wP/0IBI5000/026, "Site Response to Security Events," Enclosure 4.3 - Step 5

that evaluates taking protective action

__ RP/OIB/5000/026,"Site Response to Security Events," Enclosure 4.4 ~

Activation of ERO during an Imminent Securiw Event

- IF- thc security event involves an insider threat, implement 2-person rule for access to all

vital areas.

- Consider delaying other actions in this procedure that could endanger site personnel until

the security threat is contained.

IF TSC, OSC and EOF have NOT been previously activated, notify the ERO to staff emergency

response facilities by performing the following steps (A and B):

____ A. Notify site personnel to activate the TSC and OSC by making the following

announcement twice over public address system:

"This is the Operations Shshifr Munager. A General Emergency has been declared.

Unit(s) __ is (are) affected. Activate the TSC OSC, and EOF. )'

__ B. Activate Emergency Response Organization by completing Enclosure 4.1 of this

procedure.

RF/O/A/jOoo/oos

Page 3 of 7

~ Make an immediate PROTECTIVE ACTION RECOMMENDATION (PAR) to be entered on

of the Emergency Notification Form. Determine PAR based on current lower tower

wind speed (use upper tower wind speed if lower tower wind speed is not available) as below:

Evacuate zones: AO, Al, B1, C1, B1, E l , F1

-WIND SPEEIGGREATER THAN 5 MPH

Evacuate two mile radius AND all affected zones 5 miles

downwind AND shelter in place remaining 10 mile EPZ as

shown on Enclosure 4.2, page 2 of 2.

__ Notify off-site agencies within 15 minutes of Emergency declaration time using an Emergency

Notification Form. Refer to one of the following procedures for instructions:

RP/O/A/5000/006A, Notifications to States and Counties from the Control Room

e RP/0/A/50QQ/006B9 Notifications to States and Counties from the Technical Support

Center

0 SWO/B/2000/004, Notificationsto States and Counties from the Emergency Operations

Facility

__- IF there is an indication of a radioactive release the TSC is not activated, contact RF shifi

to perfom off-site dose assessment per HP/O/B/1009/26.

__- IF a radioactive release or hazardous material spill is occurring or has occurred the TSC is

not activated, contact Environmental Management (EM), ext. 3333, for assistance in reporting

to state, local or federal authorities. After hours, contact the Environmental Duty person by

phone or pager. no answer, page 8-777-3333 which wil1 page all Environmental

Management personnel.

~ Conduct a Site Assembly using RP/O/A/5000/010, Conducting a Site Assembly or Preparing

the Site for an Evacuation.

~ Conduct a Site Evacuation using RP/O/A/jO00/015, Conducting a Site Assembly or Preparing

the Site for an Evacuation.

RP/o/A/sooo/oos

Page 4 of 7

-Notify the NRC using RP/OIB/5000/013, NRC Notification Requircments. This notification

should be made as quickly as possible but shall be made within one hour of the emergency

declaration time.

.~ -

IF Emergency Response Data System (ERDS) transmission has not been initiated (Alert or SAE

classification), initiate ERLPS within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of initial Alert or higher declaration by perforniing

the following:

-Type ERDS or select Main, then General,then ERDS on a Control Room

OAC workstation connected to the affected units OAC.

-Initiate ERDS transmission by depressing Fd or clicking Activate.

-- IF ERDS transmission will not connect to the NRC, inform the NRC using E.NS. The

TSC Data Coordinator will troubleshoot and initiate ERDS transmission upon arrival in

the TSC.

3. Subsequent Actions

__- IF a security event has occurred, perform the following to account for site personnei:

- A. WHE.N Security notifies the OSM that the security threat has been terminated,

make the following announcement over the public address system:

Thisis the Operratians SJiift Manager. The securip event has been terminated.

The security event has been terminated. I

- B. Conduct a site assembly per RP/O/A/5000/10, Conducting a Site Assembly or

Preparing the Site for an Evacuation.

~ Ensure WP has dispatched On-Site and Off-Site Field Monitoring Teams with associated

communications equipment per HP/0IB/1009/009, Guidelines for Accident and Emergency

Response.

Evaluate specific plant conditions, off-site dose projections, field monitoring team data, and

assess need to update Protective Action Recommendations made to states and counties in

previous notification. Refer to:

Enclosure 4.3, page 1 of 3, Guidance for Subsequent Protective Actions, Subsequent

Protective Action Recommendation Flowchart

e Enclosure 4.4, Evacuation Time Estimates for Catawba Plume Exposure EPZ

RP/~/A/5000/005

Page 5 of 9

-Make follow-up notifications to state and county authorities:

Every &until the emergency is teiminated

-

If there is any significant change to the situation

-

As agreed upon with an Emergency Management official from individual agency

~ RP/O/A/5000/018, Emergency Worker Dose Extension, shall be used to authorize emergency

worker doses expected to exceed normal occupational exposure limits during a declared

emergency event or exceed blanket dose extension h i t s authorized by the Radiation

Protection Manager.

__ Augment shift resources to assess and respond to the emergency situation as needed.

~ Announce over the plant public address system the current emergency classification level and

summary of plant status.

Assess the emergency conditions and the corresponding emergency classification. See

RP/O/A/5000/00 1, Classification of Emergency, then:

Remain in a General Emergency

-

Terminate the emergency (Refer to RP/O/A/50001020 or SlUO/F3/2OOO/OO3 for

Termination Criteria).

Announce any emergency classification ievel changes over the plant public address

system including a summary of plant status.

NOTE: Turnover of command and control to the TSC or EOF relieves the OShVEmergency

Coordinator of classification notification and Protective Action Recommendation (PAR)

m _ Turnover the responsibility of command and control for the emergency as follows:

__ Provide turnover to the TSC Emergency Coordinator per Enclosure 4.5.

__ - IF the emergency situation prevents activation of the TSC within 75 minutes of

declaration, contact the EOF Director and perform a turnover. Refer to EOF Director

Turnover Form in RP/0/A/5000/020, Technical Support Center (TSC) Activation,

Enclosure 4.1.

RP/@N5000/005

Page 6 of 7

__ - IF neither facility can take M o v e r , maintain command and controi until one of the

facilities is capable of accepting turnover.

__ In the event that a workers behavior or actions contributed to an actual or potential substantial

degradation of the level of safety of the plant (incidents resulting in an Alert or higher

emergency declaration), the supervisor must consider and establish whether or not a for cause

druy/alcohol screen is required. The FFD Program Administrator is available to discusdassist

with the incident.

EOF Director will terminate the emergency and reconmend entry into Recovery by briefing the

off-site authorities at the Emergency Operations Facility or if necessary by phone. Document

the termination briefing using Enclosure 4.6.

The EOF Director shall assign an individual to provide a written report within thirty days. This

report could be an LER or a written report if an LER is not required.

Person Assigned Responsibility

Enclosure 4.2 RPi0!A/S000/005

10 Mile Emergency Planning Zone (EPZ) Map Page 2 of 2

and Protective Action Zone Determination Tables

U s this table to determine the recommended zones for evacuation within the:

281.26 -303.75

Enclosure 4.3 RP/O:A/S000/005

Guidance for Subsequent Protective Actions Page 1 of 3

Subsequent Protective Action

Recommendation Flowchart

9 START

LAROE FlSSlON PROD 7

IhWNTORY OREATER MILE RADIUS AND 10 MILES

O M ACTNITY IN W W N W M D AND RCOMMEhW N-

CONTAINMENT? PLACE SHE.l.TwI FOR ZONES NOT

(ENCI..l.J,PAGE 2 o F 3 ) EVACUATED

(ENCL 4.3 PAGE 3 OF 3)

W S E S PROJECTED I

T o BE:

RECOMMEND EVACUATION OF

AND&* 5 REM coe I N ANY IDENTIFIED ZONE~S)wrnt WSE

ZONE'SHAT tIAS NO? BEEN PROJEfTlONS L I REM TEDE

PRT\'IOUSLY EVACUATED? AND/OR t 5 W.M CDE

NOTE: CHANGES IN WIND SPEED AND/OR WIND DIREC7ION MAY REQIJIRE TIlAT ADDITIONAL ZONES BE RECOMhlENDED FOR EVACUATION. 'THTSI

ADDITIONAL. RECOMMENDATIONS ARE BASED ON THE FOLLOWM(I:

. a W N D SPEED IS OREATER THAN 5 MPH bplE LAROE FISSION PRODUCT INVENTORY IS LESS THAN O M ACTIVITY IN C O N T A N F - W , USE

ENCWSllRE 4.2 PAGE Z OF 2 TO DETERMINE IF EVACUATION OF ADDITIONAL ZONES SHOULD BE RECCMMEhWED

LAROE FISSION PRODUCT INVENTORY IS OK-ATER THAN GM ACTIVSN IN COhTArNMEW USE 7 TO

DITBRMINE IF EVACUATION OF ADDITIONAL ZONS SHOUlD BE RECLHIMEWDED

Enclosure 4 3 RP/~/A/5000/005

Guidance for Subsequent Protective Actions Page 2 of3

Guidance for Determination of Gap Activity

Fission product inventory inside Containment is greater than gap activity if the containment

radiation level exceeds the levels in the table below:

I ~ TIME AFTER 1 HIGH RANGE CONTAINMENT MONITOR READING -

SHUTDOWV

EMF 53A endlor EMF 53R

(HOURS)

100 % Gap ActiviQ Release

&nclosure 4.3 RP/o/Af5000i005

Guidance for Subsequent Protective Actions Page 3 of 3

This Table Only Used For Large Fission Product Inventory Greater Than Gap Activity In Containment.

Use this table to determine the recommended zones for evacuation within the:

5 mile radius and 10 miles downwind for any windspeed.

NOTE 1, Upper tower wind direclion ispEfwred. gnat available, use lower tower wind direction. Use wind directionfront

Nutional Weaker Setvice $site meteorological informalion is not availablr. NWS: Primary: i-800-268- 7785

Backup: 864-67Y-IO85

2. Wind direction indicator in Control Room has a scde of 0 to 540 degrees. Both 0 and 360 degretv indicate North.

3. Subtracf 360from wind direction indicalions flearer :ha" 360 degrees to arrive at wind direction for table below.

Wind Direction

(Degrees from North) 5 Mile Radius 10 miles Downwind Remainder of EPZ

Bank Question: 485.2 Answer: D

1 Pt(s) Unit 1 was shutdown in mode 6 . Engineering reported that a recent test on

the l B KC heat exchanger showed that fouling had reduced its heat transfer

capability.

0 The heat exchanger was now incapable of meeting the design A

temperature with a maximum (FSAR) lake water temperature of 90 O F .

The design A temperature couid be met with lake water with a maximum

temperature of 80 OF.

0 Current lake temperature is 68 OF.

0 Lake temperature is not expected to reach 80 "F far another 30 days.

The plant is expected to remain in mode 6 for another 30 days.

Which of the following statements correctly characterizes the operability of

the KC system?

REFERENCES PROYIDED: Tech Spec 3.7.7 & Bases

A. The B train of KC is inoperable because it is unable to maintain

KC system temperatures below 120'F during a plant shutdown.

B. The B train of KC is inoperable because it is unable to maintain

normal KC temperatures below 90 OF.

C. The B train of KC is operable but degraded as long as Lake Wylie

temperature remains below 90 OF.

D. The B train of KC is operable but degraded as long as Lake Wylie

temperature remains below 80 OF.

Distracter Analysis:

A. Incorrect: It is operable but in a degraded condition because it

cannot perform its intended safety function - within the FSAR limits.

Plausible: The B train does not meet FSAR requirements for

perfomance. 120°F is the maximum KG temperature for shutdown

B. Incorrect: The FSAR function can be performed for the existing

-

situation can temporarily satisfy the FSAR requirements

Plausible: The B train does not meet FSAR requirements for

performance. 90°F is the maximum KC temperature for shutdown.

C. Incorrect: The intended F§AR bctions cannot be performed above

80 OF - so it cannot be operable above 80 O F

Piausibie: B train of KC can perform FSAR functions until lake

water temp exceeds 80 OF. The FSAR design iiniit for Lake water

temperature is 90 OF - and 50me candidates may become confused

between these temperatures

D. Correct: B train KC meets the requirements ofNSD 203.7 - the

intended FSAR function can be satisfied by the presence of certain

temporary conditions (Le. lake water remaining below 80 O F )

Level: SRO Only 10CFR55.43@)2

K A APE 026 G2.2.25 (2.513.7)

Lesson Plan Objective: KC Obj: 13

Source: Mod Catawba NRC 1999

Level of knowledge: comprehension

References:

1.OP-CN-PSS-kc page 13

2. Tech Spec 4.7.7 PROVIDED

~

3. Tech Spec Bases 3.7.7 -PROVIDED

DUKE POWER .....

..... .-., = .,..-I.- ...CA TA

I.. WBA OPERA TlONS TRAlNlNG

.-.......

z

P

Objective T

R

Q

State the purpose of the KC System.

Describe how the KC System is cooled.

Describe the normal flowpath of the KC System, including X

each header and the type of loads serviced by each.

Explain what happens in the KC System during: X

- Safety Injection (Ss)

~ Phase A Containment Isolation (Stj

- Phase 4 Containment Isolation (Sp)

- Blackout

- Low Low KC Surge Tank Level

_s

Given appropriate pfant conditions, apply limits and X

precautions associated with OP/l(2)/A64OO/QQ5

~ComoonentCooling Water System)

State the typical values of the KC pump discharge X

pressure, KC Hx outlet temperature and KC pump flow.

State the basic actions required of an NLO for a loss of

Component Cooiina Water and why.

Describe KC system makeup.

Draw a block diagram of the KC system per the KC System

Simplified Drawing.

Explain when the Chemistry group is to be notified X

concernina the KC svstem.

Describe the purpose of the EMF'Sassociated with the KC X

System and what is indicated by a high level radiation

alarm.

List the instrumentation available in the control room for the

KC System.

When given a set of plant conditions and access to X

reference materials, determine the actions necessary to

comply with Tech SpedSLC's.

Discuss the supplementary actions for the loss of KC AP. X

__

QP-CN-BSS-KC FOR TRAINING PURPOSES ONLY REV. 42

Page 3 of 26

DUKE BOW%!? CATAWEA OPERATlONS TRAINING

b) NCDT and Excess Letdown Hxs

I ) Flow controlled

2 ) Containment isolation for excess letdown is controlled ft-01~1 the

NV board.

H. KC Drain header and Drain Sump

1. Containment drain isolations will ciose on St

2. KC Drain Header Loop Seal

a) Located outside containment downstream of drain header

containment penetration.

b) Allows drain header to be aligned during all modes of operation.

3. One 500 gal. steel lined covered sump per unit.

-

a) 2 pumps per sump Aux. Bldg 522

b) Able to discharge to:

1) NR Chiller Surge Tank

2) Other Units Sump

3) KC Surge Tank

4) Mixing and Seffling Tank

2.2 Operation

A. Technical Specifications and Selected Licensee Commitments (8BJ. #13)

1. Refer to Technical Specification 3.7.7(Component Cooling Water (CCW)

System) and Bass.

2. Refer to Selected Licensee Commitments 16.7-10 (Radiation Monitoring

For Plant Operations)

B. KC System Limits and Precautions and Special Lineups.

I Review Limits and Precautions per OP/I/A/6400/05 (OBJ. #5j

~

2. KC System Alignment fer KC Heat Exchanger Cleaning (O-C95-?69)

a) Alignment is used to maintain the availability of all essential heat

loads associated with the KC Train having its heat exchanger

cleaned.

b) The KC Train containing the Heat Exchanger which is not being

cleaned supplies all Train A and B component loads.

QP-CN-PSS-KC FOR TRAINIMG PURPQSES ONLY R N . 42

Page 13 of 26

CCW System

I3 3.7.7

B 3.7 PLANT SYSTEMS

5 3.7.7 Component Cooling Water (CCW) System

BASES

BACKGROUND The CCW System provides a heat sink for the removal of process and

operating heat from safety related components during a Design Basis

Accident (DBA) or transient. During normal operation, the CCW System

also provides this function for various nonessential components, as well

as the spent fuel storage poo!. The CCW System serves as a barrier to

the release of radioactive byproducts between potentially radioactive

systems and the Nuclear Service Water System (NSWS), and thus to the

environment.

The CCW System is arranged as two independent, full capacity cooling

loops, and has isolatable nonsafety related components. Each safety

related train includes two 50% capacity pumps, surge tank, heat

exchanger, piping, valves, and instrumentation. Each safety related train

is powered from a separate bus. An open surge tank in the system

provides sufficient inventory to protect the pumps from a lack of net

positive suction head available (NPSHA) due to a moderate energy line

break. The pumps have sufficient NPSHA with the surge tank empty

provided the piping up to the tank is filled. The pumps on each train are

automatically started on receipt of a safety injection signal, and all

nonessential components are isolated.

Additional information on the design and operation of the system, along

with a list of the components served, is presented in the UFSAR,

Section 9.2 (Ref. I). The principal safety related function of the CCW

System is the removal of decay heat from the reactor via the Residual

Heat Removal (RHR) System. This may be during a normal or post

accident cooldown and shutdown.

APPLICABLE The safety related design basis function of the CCW System is to remove

SAFETY ANALYSES waste heat from various components essential in mitigating design basis

events which require Emergency Core Cooling System ( E M S )

operation. The CCW System is also used to support normal operation.

The normal temperature of the CCW is 9 0 T , and, during unit cooldown to

MODE 5 (Tdd -= 200°F). a maximum temperature of 120T is

Catawba Units 1 and 2 B 3.7.7-4 Revision No. 0

CCW System

B 3.7.7

BASES

APPLICABLE SAFETY ANALYSES (continued)

assumed (Ref. 1). This 1 2 0 F limit is to prevent thermal degradation of

the large pump motors supplied with cooling water from the CCW

System.

The CCW System is designed to perform its function with a single failure

of any active component, assuming a loss of offsite power.

The CCW System also functions to cool the unit from RHR entry

conditions (Tcold6 35OoF),to MODE 5 (Tmu c 2OO0F), during normal and

post accident operations. The time required to cool from 35QVto 20PF

is a function of the number of CCW and RHR trains operating. One CCW

train is sufficient to remove decay heat during subsequent operations with

TWlde 200°F. This assumes a maximum service water temperature of

10B°F occurring simultaneously with the maximum heat loads on the

system.

The CCW System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2).

LCQ The CCW trains are independent of each other to the degree that each

has separate controls and power supplies and the operation of one does

not depend on the other. In the event of a DBA, one CCW train is

required to provide the minimum heat removal capabllity assumed in the

safety analysis for the systems to which it supplies cooling water. To

ensure this requirement is met, two trains of CCW must be OPERABLE.

At least one CCW train will operate assuming the worst case single active

failure occurs coincident with a loss of offsite power.

A CCW train Is considered OPERABLE when:

a. Both pumps and associated surge tank are OPERABLE; and

b. The associated piping, valves, heat exchanger, and instrumentation

and controls required to perform the safety related function are

The isoiation of CCW from other components or systems not required for

safety may render those components or systems inoperable but does not

affect the OPERABILITY of the C6W System.

Catawba Units 1 and 2 E) 3.7.7-2 Revision No. 0

CCW System

B 3.7.7

APPLICABILITY In MODES 1,2, 3, and 4,the CCW System is a normally operating

system, which must be prepared to perform its post accident safety

functions, primarily RCS heat removal, which is achieved by cooling the

RHR he& exchanger.

In MODE 5 or 6, the requirem6nts of the CCW System are determined by

the systems it supports.

ACTIONS a_a

Required Action A.Z is modified by a Note indicating that the applicable

Conditions and Required Actions of LCO 3.4.6, "RCS L ~ Q ~ s - M B D E 4,"

be entered if an inoperable CCW train results in an inoperable RHR loop.

This is an exception to LCO 3.0.6and ensures the proper actions are

taken for these components.

If one CCW train is inoperabie, action must be taken to restore

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining

OPERABLE CCW train is adequate to perform the heat removal function.

The 2 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months Completion Time is reasonable, based on the redundant

'

capabilities afforded by the OPERABLE train, and the low probability of a

DBA occurring during this period.

R.landR.2

If the CCW train cannot be restored to OPERABLE status within the

associated Completion Time, the unit must be placed in a MODE in which

the LCO does not apply. To achieve this status, the unit must be placed

in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The

allowed Completion Times are reasonable, based on operating

experience, to reach the required unit Conditions from full power

conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE 5B2JI-l

REQUIREMENTS

This S R is modified by a Note indicating that the isolation of the CCW

flow to individual components may render those components inoperable

but does not affect the QPERABILITY of the CCW System.

Verifying the correct alignment for manual, power operated, and

automatic valves in the CCW flow path to safety related equipment

provides assurance that the proper flow paths exist for CCW operation.

Catawba Units 1 and 2 B 3.7.7-3 Revision No. 0

CCW System

R 3.7.7

BASES

SURVEILLANCE REQUIREMENTS (continued)

Phis SR does not apply to valves that are locked, sealed, or otherwise

secured in position, since these valves are verified to be in the correct

position prior to locking, seaiing, or securing. This SR also does not

apply to valves that cannot be inadvertently misaligned, such as check

valves. This Surveillance does not require any testing or valve

manipulation; rather, it involves verification that those valves capable of

being mispositioned are in the correct position.

The 31 day Frequency is based on engineering judgment, is consistent

with the procedural controls governing valve operation, and ensures

correct valve positions.

2x3Lz.2

This SR verifies proper automatic operation of the CGW valves on an

actual or simulated actuation safety injection, Phase A Isolation, or

Phase B Isolation signal. The CCW System is a normaily operating

system that cannot be fully actuated as part of routine testing during

normal operation. This Surveillance is not required for valves that are

locked, sealed, or otherwise secured in the required position under

administrative controls. The 18 month Frequency is based on the need to

perform this Surveillance under the conditions that apply during a unit

outage and the potential for an unplanned transient if the Surveillance

were performed with the reactor at power. Operating experience has

shown that these components usually pass the Surveillance when

performed at the 18 month Frequency. Therefore, the Frequency is

acceptable from a reliability standpoint.

sR-3LZ2

This SR verities proper automatic operation of the CCW pumps on an

actual or simulated actuation signal. The CCW System is a normally

operating system that cannot be fully actuated as part of routine testing

during normal operation. The 18 month Frequency is based on the need

to perform this Surveillance under the conditions that apply during a unit

outage and the potential for an unplanned transient if the Surveillance

were performed with the reactor at power. Operating experience has

shown that these components usually pass the Surveiilance when

performed at the 18 month Frequency. Therefore, the Frequency is

acceptable from a reliability standpoint.

Gatawba Units 1 and 2 8 3.7.7-4 Revision No. 0

CCW System

B 3.7.7

REFERENCES 1. UFSAR, Section 9.2.

2. 10 CFR 50.36, Bechnlcal Specifications, (c)(2)(ii)

Catawba Units 1 and 2 E3 3.7.7-5 Revision No. 0

CCW System

3.7.7

3.7 PLANT SYSTEMS

3.7.7Component Cooling Water (CCW) System

LCO 3.7.7 Two CCW trains shall be OPERABLE*.

APPLICABILITY: MODES 1 2, 3,and 4.

I

ACTIONS -

CONDITION REQUIRED ACTION COMPLETION TIME

A. One CCW train A.1 __ss __

_____-- NOTE--------------

inoperable. Enter applicable

Conditions and Required

Actions of LCO 3.4.6,

"RCS LooPs-MODE 4,"

for residual heat removal

loops made inoperable by

ccw.

Restore CCW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months '

OPERABLE status.

B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

associated Completion

Time of Condition A not NJ

.p

l

met.

8.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

For each CCW :rain on Unit 2 , the Completion Time that one CCW train can be inoperable as specified by Required

Action A.1 may he extended beyond the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months up to 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months as part of the NSWS system upgrades. System

upgrades include maintenance and modification activities associated with the NSWS piping. valves, arid branch 1hnos.

necessary repairs and!or replacement, and replacement of portions of the NSWS piping to the AFW syster.. Upon

completion of the cleaning. upgrades, and system restoratior in refueling outage 1 EOC12. this footnote is nci longer

applicable.

Catawba Unlts 1 and 2 3.7.7-1 Amendment Nos. 189/182

CCW System

3.7.7

SURVEILLANCE REQUIREMENTS

SURVEILLANCE

Verify each CCW manual, power operated, and 32 days

automatic valve in the flow path servicing safety related

equipment, that is not locked, sealed, or otherwise

secured in position, is in the correct position.

SR 3.7.7.2 Verify each CCW automatic valve in the flow path 18 months

servicing safety related equipment that is not locked,

sealed, or otherwise secured in position, actuates to the

correct position on an actual or simulated actuation

signal.

SI? 3.9.9.3 Verify each CCW pump starts automatically on an actual 18 months

or simulated actuation signal.

Catawba Units Iand 2 3.7.7-2 Amendment Nos. 1731165

1 Pt(s) Unit 1 was shutdown in mode 6. Engineering reported that a recent test on

the w1T heat exchanger showed that fouling had reduced its heat transfcr

capability.

The heat exchanger was now incapable of meeting the design delta

temperature with a maximum (FSAR) lake water temperature of 90 O F .

The design delta temperature could be met with lake water with a

maximum temperature of 80 O F .

Current lake temperature is 68 "P

Lake temperature is not expected to reach 80 "F for another 30 days.

The plant is expected to remain in mode 6 for another 30 days

Which of the following statements correctly characterizes the operability of

the RN heat exchanger?

REFEREACES PROVIDED: Tech Spec 3.7.8 & Bases

A. Operable for 30 days

B. Operable but degraded as long as Lake Wylie temperature

remains below 80 OF.

C. OperabIe but degraded as long as Lake Wylie temperature

remains below 90 O F .

D. Inoperable

Distracter Analysis:

A. Incorrect: It is operable BUT in a degraded condition because it

cannot perform its intended safety function - withim the WAR limits.

Plausible: If lake temperature will not exceed it's delta temp limits

for 30 days, it could be considered operable until temp reaches the

limit. In addition, the RN system is not required to be operable in

mode 6 so some may think that they can call it operable if it is not

required in the mode that they are in.

B. Correct answer meets the requirements of NSD 203.7 - the intended

S A R function can be satisfied by the presence of certain temporary

conditions @.e.lake water remaining below 80 O F )

C. Incorrect: The intended S A R functions cannot be performed above

80 "F - so it cannot be operable above 80 O F

Plausible: Can perfom S A R functions until lake water temp

exceeds 80 OF. The FSAR design limit for Lake water temperature is

90 "F - and some candidates may become confused between these

temperatures

D. Incorrect: The SAR function can be performed for the existing

situation - can temporarily satisfy the SAR requirements

Plausible: if the candidate does not know the definition of

"operable" and "operable but degraded".

Bank Question: 483.f Answer: D

1 Pt(s) Unit 1 is in mode 6 and refueling operations are in progcss. Given the

following Conditions and events:

The Fuel Handling Manipulator Crane Operator (FHMCO) has indexed

the mast over the location where fuel assembly N-8 \vi11 be inserted.

All conditions and indications on the Fuel Handling Manipulator Crane

are satisfied for inserting the fuel assenibly.

Which one of the fallowing statements describes the responsibility of the

Fuel Handling SRO associated with inserting the fuel assembIy?

A. Must give his permission to the OAT6 prior to inserting the

assembly. Operates the Latch/Unlatch switch. Verifies the fuel

handlers are documenting satisfactory completion of each rod

latch.

B. Must obtain permission from the OATC prior to inserting the

assembly. Qperates the Latchflinlatch switch. Documents

satisfactory completion of each assembly insertion.

C. Supervises reactivity management and verifies maintenance

technicians are continuously monitoring the load cell. Must

obtain permission from the OATC prior to inserting the

assembly.

D. Supervises reactivity management and personally monitors the

fuel insertion operation. Must give his permission to the

FHM60 prior to unlatching the assembly.

__--

Distrarter Analysis:

A. Incorrect: The OATC is not in the approval chain - FHMCO

operates the latching mechanism.

Plausible: matches with elements in other distractew and sounds

supervisory.

R. Incorrect: The OATC is not in the approval chain - FIIMCO

operates the latching mechanism.

Plausible: if the candidate feels that more control is needed.

C. Incorrect: The OATC is not in the approval chain

Plausible: If the candidate thinks that the CR is in charge.

D. Correct:

Level: SRO Only lOCFR55.43@)6

KA: C 2.2.27 (2.6 / 3.5)

Lesson Plan Objective: FL 0b.j: 5

Source: Bank

Level of howledge: memory

References:

1. OP-CN-FH-FL pages IO, I4

Qua-483.1 .doc

-~

DUKE POWER --. CA TA..WBA OPERATIONS TRAINING

-__ i__ _ .--.

.

. _ . ~._m____

Objective

Zxplain the purpose of the Containment Purge System.

Iescribe the normal flowpath of the VP System and the refueling

lowpath.

Sxplain the importance of a proper flow balance during VP System

3perations.

Zxplain the purpose and use of local controls.

Explain the functions and locations of the VP System controls and

ndications.

Iescribe the startup, monitoring, and shutdown of the VP System per

.he OP.

lescribe the automatic actions that occur to the VP System in the

?vent of an SH signal, an alarm on EMF-39, or fan trig signal.

3ven the appropriate plant conditions, apply Limits and Precautions

3ssociated with related station procedures.

-

sxplain how to complete a purge release form after a purge is

omplete.

lescribe stattup, monitoring and shutdown of the lncore Instrument

Soom PurQeSystem per the OP.

3ven a set of plant conditions and access to reference materials,

letermine the actions necessary to comply with Tech SpecdSLCs.

State the system designator and nomenclature for major components

OP-WCNT-VP FOR TRAINING PURPOSES ONLY REV. 23

sag% 3 of 29

DUKE BOWER CATAWBA OPEaATIONS TRAINING

D. The rheostats associated with the supply and exhaust dampers will be

positioned fully clockwise to ensure the suppiy and exhaust dampers are

closed and recirc dampers are open (refer to Figure 9, 10).

E. Set EMF-39 setpoints to the specified values and setup the associated

chart recorder.

F. Enter release initiation information on the Release Record

1. DateTTime release initiated

2. Initial integrator reading

6. Start VP supply and exhaust fans. (refer to Figure 6 )

H. Verify containment isolation valves open.

I. Balance supply and exhaust flows to prevent pressurization or vacuum

inside containment (refer to Figure 5).

J. Notify RP that Containment Purge has been initiated

3.3 Shutdown of Containment Purge (Obj. #6)

A. Verify Initial Conditions.

B. Notify appropriate personnel that that the VP release will be terminated.

C. Place the Containment Purge Fan Units Control Switch to "OFF" (refer to

Figure 6).

1 . Verify all fans stop and all containment isolation valves close (refer to

Figure 6).

D. The rheostats associated with the supply and exhaust dampers will be

positioned fully clockwise to ensure the supply and exhaust dampers are

closed and recirc dampers are open (refer to Figures 9, 10).

E. Ensure mode selector switch is in the "NORM" position prior to the reactor

vessel missile shieid being put into place. This will prevent

overpressurizing upper containment (refer to Figure 5).

F. Place the key operated valves "Enable" switches to the "BLK CLSD"

position (refer to Figures 7, 8).

G. Position "Enable" switches for valves and fans to "BLOCK' (refer to

Figures 7, 8).

H. Stamp EMF chart recorder.

I. Notify RP and enter release termination information on the release record.

(Obj. #9)

1. DateTTime release terminated

2. Final integrator reading and volume released

3. Highest EMF reading

OP-CN-CNT-VP FOR TRAlNl" PURPOSES ONLY REV. 23

Page 15 of 29

O P i k B i 6 100!010X

PANEL: lRAD-l Page 4 of 37

1EMF-39 CONTAINMENT GAS HI RAD A/2

SETPOINT: Per IIP/O/B/1000/010 (Determination of Radiation Monitor Setpoints).

ORIGIN: 1EMF-39 beta scintiilation detector (low range).

PROBABLE Radioactive spill/leak inside containment,

CAUSE:

AUTOMATIC I. -

IF below P-6. the containment evacuation alarm is actuated.

ACTIONS: 2. -

IF at least one train of SSPS is NOT in test, the containment

ventilation isolation signal is actuated (SF,).

3. The Containment Purge System (VP) is isolated.

IMMEDIATE: 1. Verify that the Containment Purge System (VP) has isolated.

ACTIONS: 2. -

IF at least one train of SSPS is NOT in test, verify that the

Containment Air Reiease and Addition System has isolated.

3. Ensure all personnel are evacuated from containment.

4. LJse "SAMPLE FLOW SELECT' module to determine if alarm is

from upper containment. lower containment or incore instrument

room.

5. Refer to .4Pil!Ai5500ilO (Reactor Coolant Leak).

SlJPPLEMENTAHY 1. Notify Radiation Protection personnel of this alarm

ACTIONS:

I

_-_

-. -

1

~

NOTE: Ifthe EMF is reset prior to secnrmg VP per the following step, the system will restart

automatically.

2. Secure VP per OP/I!A/6450iOlS (Containment Purge System).

3. E actuated, reset the containment ventilation isolation signal when

this alarm clears.

NOTE: If annunciator alarm is due to an actual hi rad signal, a new GWR will be required hefore

reinitiating VP or VQ. I

4. Manually reinitiate the Containment Purge System (VP) or

Containment Air Release and Addition System (VQ) as needed

5. Refer to Tech Specs 3.4.13,3.4.14 and 3.4.15.

CONTINUED ON THE NEXT PAGE

Bank Question:479.3 Answer: B

1 Pt(s) Unit 2 is conducting a containment purge in accordance with

OP/2/A/6450/01S(C0ntainment Purge System). Given the following

conditions provided on the GWR permit:

Most restrictive release rate = 16000 CFM

0 Recommended release rate = 16000 CFM

2EMF-39(Lj trip 1 setpoint = 1.0E5 CPM

2EMF-39(L) trip 2 = 2.OE5 CPM

1EMF-36(L) is in service

Time 0215 0230 &?g

Release rate (CFM) 15750 16500 17500 18500

EMF-39 (CPM) 1.8E5 2.2E5 2.1E5 3.2E5

If the operators restart the VP purge whenever allowed by procedure, what is

the earliest time (if any) that the operators are reauired to terminate the

gaseous release and obtain a revised GWW?

A. 0200

B. 0215

C. 0230

D. 0245

~~~~~~~~~~~~~~~~~~~~~~~~ ~ ~

Distracter Analysis: OP/l/A/64501015 allows the operators to reinitiate a

containment release IF 1EMF-39(L) spikes. This question does not

indicate that there is a spike - but rather that the maximum release

rate has been exceeded due to high flow rates through the

containment purge line.

A. Incorrect: - no reason to terminate at 0200.

Plausible: - exceeds trip I on IEMF-39 and close to recommended

release rate.

B. Correct answer - exceeded recommended release rate, most

restrictive release rate and reachcd trip 2 on lEMF-39(L). This

clearly exceeds the release limits and there is no indication that the

EMF spiked.

C. Incorrect: -already should have terminated.

Plausible: - Exceeds recommended release rate - must terminate.

EMF-39 tripped VP for the 2"*time - if the candidate thinks you can

reset once and continue as allowed for a spike in the EMF

D. Incorrect: - should already be terminated.

. ......

. .. . .

.................... .

Plausible: - This is the answer is the release rate was not exceeded.

Level: SRO Only 10CF;R55.43(b)4

KA: G 2.3.9 (2.5b.4)

Lesson Plan Objective: CNT-VP Obj:7/9

Source: Mod Ques-479. I McGuirc 22002

Level of knowledge: comprehension

i . OP-CN-ChTT-VP page 15

2.0P/2/M6450//15page 2

3,OP/I/A/6450 End 4.1 page 4

3.OPlliI3/6100/010X I U D - 1 Ai2

4. HP/O/B/1004/005 pages 1-5

5 . WO/B/1004/034 End 5.2 (GWR)

Ques-479.3.doc

DUKE POWER

a

.<

.- ......-

...... CA TA.....WBA

... .._-.OPERATlONS TRAINING

Objective

Explain the purpose of the Containment Purge System.

Describe the normal flowpath of the VP System and the refueling

flowpath.

Explain the importance of a proper flow balance during VP System

Operations.

Explain the purpose and use of local controls.

Explain the functions and locations of the VP System controls and

indications.

Describe the startup, monitoring, and shutdown of the VP System per

the OQ.

Describe the automatic actions that occur to the VP System in the

event of an SH signal, an alarm on EMF-39, or fan trip signal.

Given the appropriate

. . . plant conditions, apply Limits and Precautions

associated with related station procedures.

Explain how to complete a purge release form after a purge is

comolete.

Describe startup, monitoring and shutdown of the Incore Instrument

Room Purge System per the OP.

Given a set of plant conditions and access to reference materials,

determine the

.....

actions necessary to comply

.. .................. .........

with Tech Specs/SLCs.

State the system designator and nomenclature for major components

09-CN-CNT-VP FOR TRAlNlNG PURPOSES ONLY REV. 23

Page 3 of 29

OP/1/N6450/015

Page 2 of 3

Containment Purge System

1. Purpose

The purpose of this procedure is to outline the operation of the Containment Purge System (VP).

2. Limits and Precautions

2.1 After refueling, ensure the "FUEL-NORM" switch is returned to "NORM" prior to

putting the reactor vessel missile shield into place. This prevents pressurizing upper

containment.

2.2 A new Gaseous Waste Release (GWR) sample is required if:

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has elapsed since the last sample.

NOTE: If actuation is due to an EMF spike, the release may be re-attempted twice before a new

sample is required. I

VP release is automatically stopped due to a controIling EMF actuatiun.

2.3 Any time initial entry into lower containment is desired, the incore instrument room shall

be sampled by Radiation Protection (RP) and purged unless the Operations Shift

Manager (OSM) deems purging U M ~ C C ~ S X Y .

2.4 Do reset containment ventilation isolation until spurious containment radiation

level signals OK any associated EMF alarms are properly cleared.

2.5 Any sudden increase or decrease in pressure across any filter bank shall be investigatcd

immediately.

2.6 For two train operation, the VP Pre-filters shall be replaced when the combined Pre-filter

and upstream HEPA differential pressure reaches 2" HzQ.

2.1 For single train operation, the VP Pre-filters shall be replaced when the combined Fre-

filter and upstream HEPA differential pressure reaches 3" H20.

2.8 If IEMF-37 or IEMF-40 has reached the Trip 1 setpoint, RP shall be notified to change

the CdItridge before a release is attempted.

2.9 Do NQT initiate purge of containment with 1EMF-39 inoperable. Once initiated, VP can

continue with IEMF-39L inoperable under the limitations presented in Enclosure 4.9

(Actions for EMF-39 Inoperability With VP in Service).

Enclosure 4.1 OPll/A/6450/0 I5

containment Purge System Startup Page 4 of 4

-2.9.4 Verify the following valves located on IRE%-ECP-3(AH-594, MM-52) open:

61 lVPOOlB (Upper Cont Purge Supply Outside Hsol)

0 1VP003B (Upper Cont Purge Supply Outside isol)

0 1VP006B (Lower Cont Purge Supply Outside Isol)

e3 IVP008B (Lower Cont Purge Supply Outside Isol)

66 iVPOl1B (Upper Cont Purge Exhaust Outside Isol)

lVP013B (Upper Cont Purge Exhaust Outside Isol)

0 lVP016B (Lower Cont Purge Exhaust Outside Isol)

CAUTION: *

_ _ - 2.10

e

Supply and exhaust air flow rates should be identical to prevent pressurization OF

vacuum inside containment.

Bo NOT exceed the Recommended Release Rate on the GWR Permit Report.

Establish Containment Purge supply and exhaust air flow at IRE%-CP-1(AB-534, LL52, Rm

1

500) at a rate greater than 10,000 CFM and Less than or equal to the "Recommended Release

Rate (cfm)" specified on the GWR Permit Report as follows:

2.10.1 Maintain supply and exhaust air flow rates q u a i as indicated on the following

gauges while adjusting flow in the foliowing step:

"lVPP5150 (CPS-AFMD-1) C O W . PURGE SUPPLY AIR FLOW"

"1VPP5200 (CPE-AFMD-I) CONT. PURGE EXHAUST AIR FLOW"

2.10.2 Simultaneously adjust the following minimum position switches in the

"DECREASE" (counter-clockwise) direction to increase supply and exhaust air

flow rates to achieve the desired flow rate while maintaining air flow rates

balanced:

"CONTAINMENT AREA PURGE SUPPLY"

"CONTAINMENT AREA FILTER EXHAUST"

2.1 1 Not@ RP shift personnel that containment purge has been initiated.

Person notified

2.12 File this enclosure in the Control Copy folder ofthis procedure.

O P i l i S 6 I oo/o lox

PANEL: 1RAD-1 Page 4 of 37

1EMF-39 CONTAINMENT GAS HI RAD AI2

SETPOINT: Per HP!O~Hi1000!010 (Determination of Radiation Monitor Setpoints).

ORIGIN: IEMF-39 beta scintillation detector (low range).

PROBABLE Radioactive spillileak inside containment.

CAUSE:

AUTOMATIC 1. E below P-6, the containment evacuation alarm is actuated.

ACTIONS: 2. -

IF at least one train of SSPS is in test, the containment

ventilation isolation signal is actuated (SH).

3. The Containment Purge System (VP) is isolated.

IMMEDIATE 1, Verify that the Containment Purge System (W)has isolated.

ACTIONS: 2. at least one train of SSPS is NOT in test, verify that the

Containment Air KeIease and Addition System has isolated.

3. Ensure all personnel are evacuated from containment.

4. Use "SAMPLE FL.OW SE.LECT" module to determine if alann is

from upper containment, lower containment or incore instrument

room.

5. Refer to APiUN5500110 (Reactor Coolant Leak).

SUPPLEMENTARY I. Notify Radiation Protection personnel of this alarm.

ACTIONS:

NOTE:

automatically.

2. Secure VP per OP!l/A/6450/015 (Containment Purge System).

A

If the EMF is reset prior to securing VP per the following step, the system will restart

3. E actuated, reset the containment ventilation isolation signal when

this alarm clears.

~~

NOTE: If annunciator alarm is due to an actual hi rad signal, a new GWR will be required before

reinitiating VP or VQ.

Enclosure 5.1

Gaseous Waste Release ( W R ) Record

Shift Supervisor or Designee authoriring GUX release (signahlre required) - Datflime I GWR#_

V Q release monitored by EMF 39(L) This Copy has been compared with the Control Copy and Verified Correct Initial Date Time

(I' DateiTime Initial Final ("EMF39L Operable EW39L

VQ Release Integrator Integrator Checked and setpoints

Initiated Reading Reading Verification (1.V.) reset to

WQtify w) EMF Trip Setpoints non-releasc

release value

Total volume released ft3

Note 1 Notify RP Compliance of each VQ release start and stop time. Ensure EMF chart recorder is stamped at start of release and at completion of each release.

Note 2 E consecutive VQ releases are made ensure GWR record is updated for each release. EMF 39(L) is removed from service and EMF 36(L) is used to monitor the

release, request a new GWR from Rp Compliance. Not Applicable (N/A) may be used on this GWR record.

Note 3 Reset EMF 39 (I-) Trip 1 and 2 ~ ~ t p o i ntot snon-release setpoints when VQ System is NOT in service. Update Control Room EMF Setpoint Log.

Note 4 Volume for each VQ release =z Final Integator Reading x 10

Termination of GWR release acknowledged by Shift Supenisor or Designee (signature required) __ DateEime I

Enclosure 5.1 a~/Qi~/1oo4/oos

Gaseous Waste Release (CWR) Record Page 2 of 6

Shift Supervisor or Designee authorizing GWR release (signature required) - DakfTime / GWR #

EMF 39(L.) was inoperable in TSAIL on: Date Time (OH)Initial

VQ release monitored by EMF 3 6 0 y been compared with the Control Copy and Verified Correct Initial

This C ~ p has Date Tie

(I' Datefhne Initial Final u1EMF36L Operable and Source Highe# EMF36L Datemime (" VQ votme = (2) Control ROO^ operator

VQ Rclease Integrator integrator Checked and Independent EMF36L Setpoints Release Final Integrator (signature required)

Initiated Reading Reading Verification (1.Q.) required for reading per OPS Supended OK reading x 10

(Notify RP) l3W Trip Setpoints during Setpoint Terminated

release (Notify RP)

(LV.)

l1.V.)

(LV.)

1I.V.)

(1.V.)

0.V.)

Total volme releawk -

Note 1 Notify RP Compliance of each V Q release start and stop time. Ensure EMF chart recorder is stamped at start of release and at M n a t i o n of each release.

Note 2 E consecutive releases arc made ensure GWR record is updated for each release. E EMF 36(L) is removed from service. request a new GWR froni W Compliance

Enme EMF inoperable date and time is consistent with TSAIL. Not Applicable (N/A) m y be used on this GWR record.

Note 3 EMF 3 6 0 T ~ 1pand 2 setpoints are per OPS Setpoint Log. During normal VQ operation, EMF 36&) setpoints are NOT required to be reset.

Note 4 Volume for each VQ release = Final Integrator Reading x 10

Termination of GWR releax acknowledged by Shift Supervisor or Designee (si-maturc required) DateRime i

Enclosure 5.1 HP/QE3/1004/005

Gaseous Waste Release (GWB) Record Page 4 of 6

Shift Supervisor or Designee authorizing GWR release (signature required) DateiTime i GWRii

EMF 39&) was inoperable in TSAIL on: Date - Time (QPS) Initial

\T release madtared by E M F 36(L) This Copy has been compared With the Control Copy and Verified Correct Initial Date Time

VPDatemime

Release

2Control Room Operator

(signature required)

Initiated

I I I J I I I I E

Total volume released , A

Note 1 Noti& Rp Compliance of each VP release start and stop time. Ensure EMF chart recorder is stamped at start of release and at termination of each release.

Note 2 E consecutive releases are made, ensme the GWR record is updated for each release. @EMF 39(L) is removed fromservice and EMF 36(L) is used to monitor the

release, ensure EMF inoperable date and time on GWR record is consistent with TSAIUI,. Not Applicable (NIA) may be used on this GWR record.

Note 3 EMF 36(L) Trip 1 and 2 setpoints are per OPS Setpoint Log. During normal VP operation, EMF 36(L)setpoints are NOT required to be reset.

Note 4 Volume for each VP release = Final Integrator Reading x 1000

Termination of GWR release acknowledged by Shift Supervisor or Designee (signature required) - Date,Time !

Enclosure 5.1

Gaseous Waste Release (GWR)Record

Shift Supervisor or Designee authorizing G W R release (signature requuired) __ Datflime I---.- GWR #

IP release monitored by EMF 39(Jd) This Copy has been compared With the Control Copy and Verified Correct Initial Date __Time

(I) DateEime "EMFvIF39L Operable and Source Highhe@ EMF39L DatcKi (4) IP vol~me = "'Control Room Operator

IP Release Checked and Independent EMF39L setpoints Release Final Integrator (signature required)

Initiated Verification (LV.) required for reading resetto Suspended or reading x 100

(Notify RP) EMF Trip Sgoints during non-release Terminated

release value (Now w

Total. volume released: ft3

Note 1 Notify $3Compliance of each IP release start and stop time. E n m e EMF chart recorder is stamped at start of release and at completion of each release.

Note 2 E consecutive releases are on the same GVIU ensure record is updated for each release. E EMF 39&) is removed from service and EMF 3 6 0 is used to monitor the

release, ensure RP Compliance is notified. Not Applicable @#A) may be used on this record.

-

Note 3 Reset EMF 3%) Trip 1 and 2 setpoints to non-release setpoints when 1P System is NOT in service. Update Control Room EMF Setpoint Log.

Note 4 Volume for each IP release Final Integrator Reading x 100

Termination of GWR release acknowledgedby Shift Supenisor or Designee (signature required) _I

.- DateKime I

Enclosure 5.1

Gaseous Waste Release (GWQ Record

ShiA Supervisor or Designee authorizing GWK release (signamre required) ~

.DateiTime I ._ GWR #

E M F 3 9 0 was Inoperable in TSAIL on: Date Time (OPS) Initial

IF' release monitored lay EMF 3qL) This Copy has been compared wth the Control Cmpy and Verified Correct. Initial Date Time

(')Datet%ne Initial Final si EMF36L Operable and Source Highesp) EW36L DaterTim IP Volume = (2iControlRoom operator I

Ii

IP Release Integrator Integrator Checked and Independent EMF36L Setpoints Release Final Integrator (signature required)

Initiated Reading Reading Verification (1.V.) required for reading per 0PS Suspended or reading x 100

o\lotify Rp) E.W Trip Setpoints during Setpoint Terminated

I

release Log wo'rotifyw

Total volume released: ff

Note 1 Notify RP Cmmpliance of each IP release start and stop time. Ensure BMMF c h i recorder is stamped at start of release and at completion of each release.

Note 2 E consecutive relaqes are on the same GWR, ensure record is updated for each release. E EMF 3 9 0 is removed from service and EMF 36(L) is used to monitor the

release, ensure EMF inoperable date and time on CiWR record is consistent with TSAIL. Not Applicable (N/A) may be used on this word.

Note 3 EMF 3 6 0 Trip 1 and 2 setpoints are per OPS Setpoint Log. During n o m 1 IP operation EMF 36&) setpoints are NOT required to be reset.

Note 4 Volume for each IP release * Find Integrator Reading x 100

Tennination of GWR release aclohowledged by Shift Supervisor or Designee (signature required) Date/Time /

Bank Question: 479. I Answer: B

1 Pt(s) Unit 1 is in the process of making a radioactive gaseous waste release h m

the waste gas decay tank in accordance with OP/O/A/6200/18 (Waste Gm

Qperation). Given the following conditions:

MRIRR-31CFM

MOSRR=40CFM

EMF-50 (WASTE GASDZSCH) trip 1 setpoint = 2.OE5 CPM

IEMF-50 trip 2 = 3.0E5 CPM

1EW-34 @INKT VENTGAS) is in scrvicc

Time 0215 0230 0245

Release rate (CFM) 30 32 41 27

EMF-50 (CPM) 2.8E5 3.2E5 3.1E5 4.255

If the operators reset 1EMF-50 whenever allowed by procedure, what is the

earliest time that the operators are rewired to terminate (and not

immediately restart) the gaseous release?

A. 0200

B. 0215

c. 0230

D. 0245

Distracter Analysis:

A. Incorrect: - neither Trip 2 nor MRIRR exceeded at 0200.

Plausible: If candidate thinks Trip 1 is sufficient to terminate.

B. Correct: the release rate (32 CFM) MRIRR (3 I CFM) (most

restrictive instantaneous release rate)

C. Incorrect: - exceeded MRIRR at 0215.

Plausible: exceeded MOSRR (maximum observed system release

~

rate) - if the candidate thinks he/& can reset EMF-50 once before

being required to terminate the release - this is the 2"' time EMF-50

has reached trip 2.

D. Incorrect: - exceeded MRIWR at 0215

Plausible: exceeded trip 2 on EMF-50 for the 3rdtime - allowed to

~

reset this trip 2 times before terminating release

Lev& SRO Only; 10CFR55.43(B)(4)

KA: SYS 073A4.01(3.9/3.9)

Lesson Pian Objective: WE-KGR SEQ 5

Source: Mod; Ques-479, McGuire NRC 2000

Level of knowledge: analysis

References:

1. QP-MC-WE-RGR page 15

Ques-479.1.doc

Enclosure 5.2 HP/O/B/1004/034

Sample of RETDAS Gaseous Waste Release Page 2 of 2

Permit Reports

RETDAS <DPCCNS R.v.O.O> VSST

GASEOUS PRE-RELEASE P E W I T REPORT

- - - _ - _ _ _ - - _ - - _ s _ _ s s - _ _ _ _ _ _ s s _ _ _ _ _

QWR Number:

Releaae ID: Waste Gas Decay Tank "C"

-PI RECOKMENDED FLOWRPITES (cfm) I ~ ~ ~ ~ ~ ~ ~ ~ P P - = ~ P ~ - ~ - ~ I I I I P I / I I D P I I . c I P ~ ~ ~ ~ ~ ~ ~ - ~ - -

1 Units 2 Unite

Releasing Releasing

2 1 2 Station 1 / 2 Staticn

Limit Limit

(U.1) (U-2)

_ _ _ - _ - s s _ _ _ _ _ s _ _ _ _ _ _ _ _

Total body dose release rate (cfm) ................... 4.52B+05 2.26E+05

Skin and Gamma air dose release rate (cfm)..... ....... 1.12E905 5.58E+04

Food. Qround, Inhalation dose release rata (cfm).... ..5.00E+01 2.50E+01

- - - - - - _ - - - - s - - _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ s s - ~ ~ - - - - - - - ~ ~ - - ~ - - - - - - - - - - - - - - - - ~ - - - - - - ~ ~

Float restrictiva releaee rate (cfm) .................................. 1.50Ec01

Recommended release rate ............................................. 2.50E+01

==- l6L5TIPLE RELE?.SB CALCULATION s s i i ~ i i i - n n n / ~ i i - i ~ i ~ ~ - ~ ~ - = ~ ~ ~ ~

Sum o f calculated release rats divided by

allowable releaeo rate for a l l ccncurrent r e l e a e s e . . . . . .............. 6.3SB-03

I-D SETPOINT DATA I I I D P I I I I D I P P P _ P ~ D ~ ~ ~ ~ - ~ ~ ~ ~ ~ ~ ~ ~ - - - ~ - ~ ~ ~ ~ ~ - -

KMFSOL monitor Operable? ............................................ Yes

BIPSOL Entered Background ( c p )...................................... 4.00Ec03

-SOL Expected (cpm)................................................ 3.12E+04

W 3 6 L Monitor Operable. ............................................. NA

KUF3BL Enterad Background (cpm) ...................................... NA

m 3 S L Expectad (cpm) ................................................ NA

X e - 1 3 3 Bquivalonce (uCi/cc) ......................................... 3.12E-04

T r i p 1 Satpoint (Cpm) ................................................ 9.13E+04

T r i p 2 Setpoint (epm)................................................ 1.30B+05

Duke Power Company 'rocedurc No.

Catawba Nuclear Station tIPfiIB/I004/034

Levision No.

007

RADIOACTIVE WASTE GAS ( W S ) SYSTEM

RELEASE

HP/O/B/L004/034

Page 2 of 6

Radioactive Waste Gas (WG) System Relgase

-

1. Purpose

To provide a method of preparing a gaseous waste release package and to describe the

method of calculating the gaseous waste release rate to be used when making a Gaseous

Waste Release (GWR) from the WG system during normal plant conditions.

0 To provide actions for EMFSOL upon notification of inoperability/operability.

2. References

2.1 HP/OiE/lOOO/OIO - Determination of Radiation Monitor Setpoints

2.2 HP/OW/1001/01S - RP Compliance Sampling

2.3 SN/O/B/2001/004Investigation of Unusual Radiological Occurrences

2.4 Catawba Nuclear Station, Liquid and Gaseous Radwaste Computer Program User

Documentation

3. Limits and Precautions

None.

4. Procedure

4.1 Use of Procedure

NOTE: Staff Support Scientist or designated Qualified Reviewer may authorize operation

outside the scope and acceptance criteria stated in this procedure provided the

technical basis and impact to existing procedure 10CFRS0.59 evaluation is clearly

documented an applicable paperwork.

ation is affected, another evalu

4. I. 1 Notify Operations Shift Manager immediately of any inadvertent

(uncontrolled) gaseous effluent release.

4.1.2 Refer to SH/0/B/2001/004 (Investigation of Unusual Radiological

Occurrences) for notification process following any inadvertent (uncontrolled)

gaseous waste release.

HPIOIBI I0041034

Page 3 of 6

4.2 Sampling of WGDT-C

4.2.1 WHEN notified by Chemistry, collect samples per HP/O/W!1001!018 (W

Compliance Sampling).

4.2.2 Review completed WGDT-C sample results

4.2.3 Maintain results for use with GWR process.

4.3 Determination of Release Monitor for GWR

4.3.1 E EMF 50L is operable, obtain EMF 50L background reading (existing

reading).

Ensure background reading (existing reading) is less than 20,000 cpm.

4.3.2 E monitor background (existing reading) is 2 20,000 cpm, declare EMFSOL

0 Generate work request for IAE IO decontaminate EMF.

Notify Radwaste Chemistry personnel of actions taken.

Complete Section 4.7

4.3.3 E EMFSQLis inoperabie, use lEhaF36L as controlling monitor for release.

A. Ensure 1EMF36L background reading (existing reading) is Less than I000

cpm.

-

IF 1EMF36L background (existing reading) is 2 1000 cpni,

discontinue release process.

/

B. Obtain IEIvlF36L Trip 2 and Trip 1 setpoints.

4.4 Generate GWR Permit Report (Pre-Release)

4.4.1 Start RETDAS Launcher for data entry.

4.4.2 Verify GWR number assigned by RETDAS is next consecutive number.

4.4.3 Input assigned GWR number in LWWGWR Logbook.

4.4.4 Compiete remaining required entries.

Refer as nccessary to RETDAS Computer P K O ~ EUser U ~ Documentation.

HP&B/l 004!034

Page 4 of 6

NOTE: "Multiple Release Calculation" (MRC) is performed by RETDAS to account for multiple

releases in progress and to ensure no release limits are exceeded. MRC uses data from all

open GWR Pre-Release packages. MRC sums the "recommended release rate" for each

open GWR and ratios to the allowable release rate of the GWR Pre-Release currently

being created. An information "box" opens to notify the user when the ratio is >0.9.

4.4.5 -

IF MRC is > 0.9, notify RP Staff Support and/or RP Supervision before

continuing GWR Pre-Release process.

4.4.6 Select "Knstmctions" on RETDAS to print information automatically on GWR

Pre-Release Permit report.

4.4.1 Ensure release data is correct on screen

4.4.8 Print report.

4.4.9 Compiete "Performed by" signature and "Date" blocks.

4.4.10 Perform verification of information by another qualified technician.

Complete "Verified by" signa(ure/date block.

4.4.1 1 Make copy of GWR Pre-Release Report.

Maintain with original Count Room sample analysis results in Compliance

Lab while GWR is active.

4.4.12 Ensure GWR package contains the following:

Sample analysis results

m Working copy of Chemistry procedure

Reports forms generated by RETDAS

0 Additional information as needed for release

4.4.13 Deliver GWR package to Control Room.

4.4.14 WHEN notified by Chemistry, record initiation, suspension and/or

termination of WGDT GWR in RP Shift Gomplknce Logbook.

IIP/O/B/lO04/034

Page 5 of 6

4.5 RP Response to Trip 3 Alarms

4.5.1 notified of Trip 2 Alarm, ensure setpoints are correct.

4.5.2 Notify RP Supervision for approval to re-start release.

4.5.3 release is not re-started, go to Section 4.6.

4.6 G\YR Ciosed Permit

I

~ ~~

NOTE: Performance of RETDAS Closed Permit shail be accomplished by RP Staff Support or

qualified RP Compliance Technicians.

4.6.1 Review GWR package.

A. Verify isotope narne and concentration of computer inputs.

B. Verify background value is acceptable per Section 4.3.

C. Trip 2 alann(s) occurred during release, determine actual duration

time:

1. Add total time between initiations and suspensions of release.

2. Determine calculated stop time of release by adding actual duration

time of release to initial start time of release:

Example: 9-1-99/2i03 plus 80 minutes gives

calculated stop time of 9-1-99/2223.

3. Enter calculated stop on RETDAS Closed Permit screen.

4.6.2 Perform find review of GWR package.

4.6.3 File GWR in RP Satellite master file.

4.6.4 Discard duplicate copy of GWK maintained in RP Compliance Lab.

4. RF' Ac --Ins for Inoperabie EMF50L

4.7.1 WHEN EMF5OL is declared inoperabie, complete applicable sections of

EMF 50L Inoperable Status Log Sheet (Enclosure 5.1).

e Place Jnoperabie Status Log Sheet in EMF Status Logbook.

HP/Q/B/1004/034

Page 6 of 6

4.7.2 -EMF 50L is declared operable, perform the following:

A. Detemine/verify EMF SOL setpoints per HP/O/B/1000;010

(Determination of Radiation Monitor Setpoints).

B. Document setpoints on Enclosure 5.1.

C. Complete "Notified Operable" section.

4.8 Record Retention

4.8.1 Maintain the following documents in RP Satellite Master File:

0 GWR Release Papenvork

Enclosure 5.1

5. Enclosures

5.1 EMF SOL Inoperable Status Log Sheet

5.2 Sample of RE'TDAS Gaseous Waste Release Permit Reports

Enclosure 5.1 IIPIoIB/10041034

EMF 5QLInoperable Status Log Sheet Page 1 of 1

UNIT #: 1 EMF #: 50L Waste Gas Dischk NOTIFIED INOPERABLE PER

LOCATION m..594. Col.JJ-KK,.49-50 DATE/TIME:

WORK REQUEST #: RIP TECHNICIAN NOTIFIED:

REASON FOR INOPERABLITY:

NOTIFIED OPERABLE PER: DATEXIME I

DETERMINED SETPOINTS (CPM): TRIP 2 = TRIPl=

SETPOINTS PROVIDED TO: GROUPISECTION:

COMPLETED BY: DATEEIME: I

RP TECKNICIAN

Enclosure 5.2 NP/OiB/ 1OO4/034

Sample of RETDAS Gaseous Waste Release Page i of 2

Permit Reports

RETDAS cDPCCNS Rev.0.0r VBST

GASEOUS PRE-RELEASE PERMIT REPORT

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ s s _ _ _ _ _ _ _ _ _ _ _ s _ _

(IIYR Number:

Release ID: Waste Gas Decay Tank " C =

Release nods: Batch

Pernit Statnm: P - Bre-Rslease

Bank Question: 4?8. I Answer: A

1 Pt(s) Unit 1 is responding to a faulted steam generator from 100% power. Given

the following events at their respective times:

0200 - a reactor trip occurred, the crew entered E-0, (Reactor Trip or

Sufeg Injection)

0220- tmnsitioned to E-3, (Steam Generutor Tube Rupture)

0215 - entered FR-Z. 1 (Responseto High Contuimient Pressure), on a

valid orange path

0220- safety injection was actuated manually after a failure of auto S/I

0225 - completed FR-Z. 1

If the SRO determined that they were responding to the event in the wrong

procedure, which one of the following statements is correct regarding the

time the SRQ could transition to ES-0.0, (Rediagnosis)?

A. The SRO can enter ES-0.0 at 0211.

B. The SRO can enter ES-0.0 at 0216.

C. The SRO can enter ESO.0 at 0221.

D. The SRO can enter ES-0.0 at 0226.

Distracter Analysis: A faulted S/G has caused containment pressure to rise

to 3 psig. Safety injection has failed to x h t e in auto. The SRQ

implements E-3 instead of E-2. He recognizes that E-3 is not the

right procedure - so he wants to go to ES-0.0 (Rediugnosis). ES-0.0

does not apply if safety injection has not actuated -but

administratively, you can enter ES-0.0. Step 2 will verify that SI has

actuated and kick you out of ES-0.0 if SI has not yet actmted.

A. Correct: ES-0.0 can be enteEd after completion of E-0.

B. Incorrect: - Can enter ES-0.0 after completion of E-0 -also cannot

enter ES-0.0 until FR-Z. 1 has been completed.

Plausible: - if candidate notes SI has not actuated and thinks that

hdshe ES-0.0 cannot be entered until SI has actuated.

C. Incorrect: - ES-0.0 can be entered any time after completion of E-0.

Plausible: - if candidate notes that ES-0.0 is applicable only after SI

actuation and does not know that FR-2. 2 must be completed.

D. Incorrect: - ES-0.0 can be entered any time after completion of E-0.

Plausible: - if candidate thinks that ES-0.0 is not applicable until

after SI has actuated and cannot leave FR-Z. 1 until completed.

Level: SRQ Only 10CFR55.43@)5

KA. WEOt EA2.qA3.2/4.0)

Lesson Plan Objective: CSF Obj: 1, EP1 Obj: 19

Source: Mod Catawba Audit Exam 2000

Level of knowledge: comprehension

Referenccs:

1. OP-CN-EP-EPI page 6

2. ES-0.0 Background Document step 2 page 2

3. O W 1-7 page 11

4. ES-0.0 step 2 page 2

DUKE POWER CATAWBA OPERA TlONS TRAlNI ~

OBJECTIVES

Objective

correct procedure flowpath and necessary actions

DYKE POWER CATAWBA OPERATIONS TRAINING

Objective

(Natural Circulation Cooldown with Steam Void in Vessel (wlth

Given a set of specific plant conditions and all required

OP-CN-P-Pl FOR TRAINING PURPOSES ONLY REV. 7

Page 4 of 8

DUKE POWER CATAWBA OPERATlONS TRAlNlNG

I.EP/I/N5000/E-O (Reactor Trip or Safety Injection)

1.I Major Action Step Summary

A. Verify automatic actions as initiated by the Protection and Safeguards

Systems - Following entry into this procedure, the operators verify reactor

trip, turbine trip, and essential power available. If S/I was not required, a

transfer is made to E§-0.1. If §/I was required, the operators proceed to

verify proper equipment alignments. Additional safeguards equipment is

verified as the system setpoint is reached.

B. identify appropriate optimal recovery guideline - This begins the first

attempt to identify an appropriate recovery guideline.

C. Shutdown unnecessary equipment and continue trying to identify

appropriate recovery guideline - If SI1 cannot be terminated and diagnosis

has not been determined, the operators continue evaluating plant conditions.

1.2 Use the most current retype to explain the purpose, symptoms, and immediate

actions of E-0.

1.3 Use the "Enhanced Background Document" to explain the bases for all steps,

notes, cautions, and Enclosure 1 actions.

2. EP/I/N5000/ES-0.0 (Rediagnosis)

2.1 Major Action Step Summary

A. Determine if any SlGs are not faulted -Attempts to diagnose a faulted S/G.

3. Determine if any S/G is faulted and if it was isolated - Identify any faulted

S/G and see if it's isolated.

C. Determine if there is a SGTR - If faulted or faulted and isolated, attempt

to identify any S/G ruptured. If not, it must be a LQCA.

2.2 Use the most current retype to explain the purpose and symptoms E.S-0.0.

2.3 Use the "Enhanced Background Document" to explain the bases for all steps,

notes, cautions, and Enclosure 1 actions.

3. E$/1/N5000/ES-0.1 (Reactor Trip Response)

3.1 Major Action Step Summary

A. Ensure the primary system stabilizes at no-load conditions The ~

operators verify NC temperature returns to no-load via steam dump

operation. Verify, and if necessary, establish adequate shutdown margin.

Check Pnr level and pressure responding correctly.

B. Ensure the secondary system stabilizes at no-load conditions -Verifies

the operation of the steam dump and feed systems. S/Gs are returned to no-

load level values.

OP-CN-EP-EPI FOR TRAINING PURPOSES ONLY REV. 7

Page 6 of 8

STEP 2: V e r i f y S / I - HAS ACTUATED

PURPOSE :

To remind t h e operator t h a t E S - 0 . O only app'lies when S/I i s i n service o r i s

requi red.

APPLICASLE ERG BASIS:

The p a r t i c b l a r sequence o f steps i n t h i s procedure was based on the assumption

t h a t S/I i s i n service o r skolild be i n service. Therefore. t h i s procedure

should be used only i f S / i s i n service o r is required and E - 0 (Reactor T r i p

Or Safety I n j e c t i o n ) has been completed.

PLANT SPECIFIC INFORTATION:

KNDWLEDGE/ABILITY:

Page 2 o f 8 Revision 4

Operations Management Procedure 1-7 Page 11 of 25

0 If a valid orange path is encountered, the

operator is expected to scan all of the remaining

trees, and then, if no red path is encountered, to

promptly implement the corresponding EP. If

during the performance of an orange path

procedure, any red condition or higher priority

orange condition arises, then the red or higher

priority orange condition shall be addressed

first, and the original orange path procedure

suspended.

e Once a procedure is entered due to a valid red or

orange condition, that procedure shall be

performed to completion unless preempted by

some higher priority condition. It is expected

that the actions in the procedure will clear the

red or orange condition before all the operator

actions are complete. However, these

procedures shall be performed to the point of

the defined transition to a specific procedure.

At this point, any lower priority red or orange

paths currently indicating or previously started

but completed shall be addressed.

e If a CSF procedure directs the operator to return

to the procedure and step in effect and the

Corresponding status tree continues to display

the off normal condition, then the corresponding

CSF procedure does havc to be implemented

again since all recovery actions have already

bcen completed. However, if the same status

tree subsequently changes to a valid higher

priority condition, then the corresponding CSF

procedure shall be implemented as required by

its priority.

0 Certain CSF procedures are used to address both

orange and red path conditions for the same

parameters. If the procedure is already in

progress due to the orange path condition, it is

required to return to the first step if the

condition becomes red. Also, at the completion

of the procedure, the procedure does have to

be implemented again, since all recovery

actions have already been implemented.

REDIAGNOSIS

EP/l/N5000/ES-0.0

ACTIONlEXPECTED RESPONSE I RESPONSE NOT OBTAINED I

C. Operator A c m

- 1. Monitor Enclosure I (Foldout Page).

- 2. -

Verify SI1 HAS ACTUATED. - RETURN TO procedure and step in

effect.

3. E any of the following procedures are in

effect, THEN RETURN Ip procedure and

step in effect:

0 EP/I/N5000/E-0 (Reactor Trip Or

Safety Injection)

- 0 EP/l/N5000/ECA-0.0 (LOSS Of All AC

Power)

- 0 P/l/N5000/ECA-0.1 (LOSS Of All AC

Power Recovery Without §/I Required)

- e EP/l/N5000/ECA-0.2 (LOSSOf All AC

Power Recovery With SI1 Required)

1

Bank Question: 478 Answer: D

1 Pt(s) Unit 1 is rcsponding to a reactor trip from 100% power. Given the following

events and conditions:

0200 - a reactor trip oecurred, SRO entered E-0

0210- transition4 to ES-0.1, Reactor Trip Recovery

0215- cntered FR-Z.2, Containment Flooding, on a valid orange path

0216 - safety injection actuated

0220 - completed FR-Z.2

If the SRO determined that thcy were responding to the event in the wrong

procedure, which one afthe following statements is correct regarding the

appropriate action to take to transition to ES-0.0, Rediagnnsis?

A. The SRO can enter to ES0.O at any time during the event based

solely upon SRO judgement

B. The SRO can enter ES-0.0 between 0210 and 0215 or after 0220

C. The SRO can enter ES-0.0 after 0216

D. The SRO can enter ES-O.0 after 0220

Distracter Analysis:

A. Incorrect: - cannot enter ES-0.0 while in E-0 or while in FR-2.2.

ES-0.0 is onLy applicable after SI actuates

Plausible: - if candidate does not know restrictions and applicability

of ES-0.0

B. Incorrect: - cannot enter ES-0.0 while in ES-0.1 because ES-0.0 is

not applicable until after SI actuation

Plausible: - if candidate does not h o w that ES-0.0 is applicable

only after SI actuation

C. Incorrect: - cannot enter ES-0.0 while in PR-Z.2

Plausible: - if candidate notes SI has actuated and does not know

that he/she must complete FR-Z.2 before ES-0.0 can be entered

D. Correct answer

Bank Question: 357 Answer: C

iiu'd#7S& 1

1 Pt(s) &&+is respondinggte a main steam line break i&w containment, iq t'~ A.

A

Given the following events and conditions:

The operators completed E-0 (Reactor Tn'p and Sa&@ Injection) arid

transitioned to E-2 (Faulted Steam GeneratorIsolation)

A RED PATH on Containment Integrity occurred and the opentors

transitioned to FR-Z. 1 (Raponse to High Containment Pressure) at step 8

of E-2.

A RED PATH on NC Integrity occurred and the operators transitioned to

FR-P. 1 (RRfsponseto Imminent Pressurized Thermal Shock Condition)

fiornstep4ofFR-2.1.

The operators performed ail required actions in FR-P. 1

Upon completion of FR-P.l, the STA reports that all CSFs are now

GREEN (including Containment Integrity).

Which one of the following describes the correct procedure flow path?

A. Return to E-2 step 1and continue.

B. Return to E-2 step 8 and continue.

C. Return to FR-Z.l step 4 and complete the procedure, then return

to E-2 step 8.

D. Enter ES-0.0 (Rediagnosis) and rediagnose the situation.

Distracter Analysis:

A. Incorrect: Must address FR-2.1 first

Plausible: since all CSFs are green, the candidate may feel a r e m to

step 1 is appropriate

B. Incorrect: Must address FR-2.1 first

Plausible: this would be correct if Z. 1 had been completed.

6. Correct: per OMP 1-4, once a procedure is entered due to a valid red

or orange condition, that procedure shall be performed M completion

unless preempted by some higher priority condition. It is expected that

the actions in the procedure will clear the red or orange condition

before all the operator actions are complete. However, these

prokedures shall be performed to the point of the defined transition to a

specific procedure. At this point, any lower priority red or orange paths

currently indicating or previously started but not completed shall bs

addressed.

D. Incorrect: Have specific guidance to return to FR-Z. 1

Plausible: the operator can use rediagnosis at any time, hut the CSP

would not be addrcssed as required.

Level: SRO Only lOCFR55.43@)2

KA: APE 069 (32.4.4(4.CV4.3)

Lesson Plan Objective: FRZ Ohj: 5

Source: Bank

Levei of knowledge: comprehension

References:

1. O m 1-7 page 13

Ques-357.doc

DUK POWER CATAWBA OPERAPlONS TRAlNl"

OBJECTIVES

Objective

I I

3 Explain the Bases for the Major Actions of each of the Function

Restoration ~rocedures:EP/l/A/5000/FR-ZSeries - Containment

I I

4 Explain the Bases for all steps in each of the Function Restoration

procedures: EP/1/A/5OQQ/FR-2 Series Containment

~

5 Given a set of specific plant conditions and required procedures,

apply the rules of usage and outstanding PPRBs tQidentify the

correct procedure flowpath and necessaw actions

OP-CN-EP-FRZ FOR TRAINING PURPOSES ONLY REV. QI

Page 3 of 8

Operations !v4atiagcmentProcedure 1-7 Page 13 of 25

0 Yellow path procedures are to be performed

concurrent with the non-critical safety function

EP in effect when the yellow path is

implemented. While performing the actions of

the yellow path, continuous actions or foldout

page items of the non-critical safety function EP

in effect are still applicable and shall be

monitored by the operator. (DW-95-043)

0 If a red or orange condition indicates and then

clsars prior to implementation of the

corresponding procedure, the procedure shall

not be performed. The CSF procedure is

considered to be implenicnted when the

procedure reader reads the first step to the crew.

0 The STA shall keep the Operations Shift

Manager informed of all off normal CSFs. The

Operations Shift Manager shall ensure the crew

is updated as appropriate, typically by ailocating

time during updates for the STA. (SOER 94- 1)

F. Normally, the condition of the CSF Status Trees is

continuously displayed by SPBS on the OAC. Control room

indications shall be used to validate any off normal alami and

to determine which procedure to implement. Once status tree

monitoring is initiated, the STA should periodically monitor

the status trees and compare against control board indications

to ensure SPDS is functioning properly. Status tree monitoring

shall be continuous if an orange or red condition exists.

Otherwise, monitoring frequency shall be every 10 to 20

minutes. (SOER 94-1)

Bank Question: 339.2 Answer: D

I Pt(s) 5 3 (Stem Generator Tube Rupture) step 18 reads as follows:

WHEN P-IT PZR S/I BLOCK

PEMISSWE status light (ISI-18) is lit,

E N :

__a. Depress ECCS Steam Pressure ~ ~ L 0 C K p u ~ ~ h b u t t o i i ~

Pressurizer pressure is 1985 psig when the SRQ !mmitions to ECA-3.1.

After the transition into ECA-3.1, which one ofthe following statements is

correct with regard to this step?

A. The step is applicable only while in E-3. The SRO shall not

return to the step when notified that permissive status light is

LIT.

B. The step is applicable only while in E-3. The RO shall

immediately perform the actions of sub step 18a, only after

reentering E-3 upon completion of ECA-3.1.

C. The step is applicable while in E-3 and after transition to

ECA-3.1 until alternative gnidance is provided. The RO shall

immediately perform the actions of sub step 18a when the

permissive status light is LIT and report this action to the SRO.

D. The step is applicable while in E-3 and after transition to

ECA-3.1 until alternative guidance is provided. The RQ shall

notify the SRO that the permissive status light is LIT and the

SRO will return to step 18a and direct the action.

Distracter Analysis:

A. Incorrect: the step is applicable in E-3 atid after transition out of E-3

until alternative guidance is provided

Plausible: partially correct - page-keeping requirement is correct.

B. Incorrect: the step is applicable in E-3 and aftcr transition out of E-3

until alternative guidance is provided - RO must notify the SRO to

return to the step and receive direction to implement sub-step 18a.

Plausible: Ifthe candidate does not know the process for conditional

steps, this is reasonable and similar to foldout page actions.

C. t : notify the SRQ to retain to the step and receive

I n c ~ ~ rR~8 must

direction to implement sub-step 18a.

Ques-333.2.dos

Plausible: this is reasonable and similar to foldout page actions.

D. Correct Answer:

Level: SRO 0nly tOCFRSS.43@)5

MA:APE 038 G 2.4.6 (3.U4.0)

Lesson Plan Objective: EP-INTRO Obj: 2,3

Source: Mo& Qucs-339, Catawba NRC 2000

Level of howledge: memory

References:

1. OP-CN-EF-INTRO page 8

2.0MP 1-7 page 9-8

3. E-3 step 18 page 17

DUKE POWER - -.....................................

CATAWBA OPERATlONS

.. TRAINING

OBJECTIVES

=

L

0 bjective P

R

1 I

4 State when ACC conditions are used.

1 5 I Explain why ACC values are used.

1 6 1 Explain the purpose of the red film on the monitor light panel

I 7 I Explain how Enclosure 1 (F~IC-IOU~page) is used

1 1

13 Explain the General Statements of Philosophy.

State when a unit trip must be investigated and any exceptions to

NSD 505.

I 5 State who is responsible for notifying station supervision of a

reactor trip.

1 16 I State the requirements that must be met prior to plant restart

for completion of time critical operator

actions. ............ .-

actions identify those that are time

critical in nature.

OP-CN-EP-iff TRO FOR TRAlfflffG PURPOSES ONLY REV. 02

Page 3 of 10

DUKE BOWER CATAWBA OPERA 7YQNS TRAiNlNG

4. EPIAP IMPLEMENTATION

4.1 OMP 1-7 (EIneFgenCy/AbnQrmalProcedure Implementation Guidelines) provides

the rules for use and implementation for EPs and AQs.

4.2 PPRBs:

A. PPRBs may also provide general and specific guidance for use and

implementation of EPs and APs.

B. The PPRB process is described in OMP 4-10 (Validation Process for

Operations Procedures).

5. REACTOR TRIP INVESTIGATION

Refer to current revision of NSD 505 (Reactor Trip Investigation).

6. TIME CRITICAL OPERATOR ACTIONS

6.1 Refer to PT/0/A/4700/061 (Time Critical Operator Action Review).

6.2 Expectations for performance of time critical actions (Obj. #17)

A. Operators are expected to perform time critical tasks in accordance with all

established procedures and work practices. The time critical nature of the

task does not preclude the need to perform the task properly and safely.

B. Inability to complete these time critical tasks within the time allotted may

indicate a need to evaluate procedures and work practices or to reanalyze

the event to include the actual performance time.

C. See the following table for the specific NLO tasks and consequences.

~

OP-CN-EP-INTRO FOR TRAINING PURPOSES ONLY REV. 02

Psge 8 Qf 70

Operations Management Procedure 1-7 Page 7 of 25

6. If desired. conditional steps (E ..., THEN...)can be

evaluated by the procedure reader for applicability prior

to giving a verbal command. The command should

then be constructed from the statement(s) after the word

T".

Example:

Step: IF Containment pressure has exceeded 3 psig,

TIIEN verify Phase B Isolation actuated.

Evaluation: Based on a previous response. or a

request for verification of containment

pressure status, the procedure reader

determines that the step is applicable.

Command: Verify Phase B Isolation actuated.

7. Konscquential steps (WHEN ..., THEN ...) require that

an action be performed after a particular event has

occurred. These steps should be handled as follows:

a. The procedure reader gives the RO a command

to notify lrini when the particular event occurs.

b. The procedure reader can mark the page at the

nonsequential step with a "tape flag" and

continue in the procedure.

C. Upon observing the event. the KO notifies the

procedure reader.

d. The procedure reader returns to the flagged step

and gives the command. The command should

be constructed from the statement(s) after the

word "THEN".

e. 'Ihe procedure reader removes the "tape flag"

and checks off the step.

Prior to branching to another procedure, the procedure

in effect should be reviewed for remaining "tape flags"

to ensure all required actions have been taken. Those

not completed should be reviewed to dctemine whether

_ I

they are progressing satisfactorily or if they are even

pertinent to the remainder of the recoven. effort.

Operations Management Procedure 1-7 Page 8 of 25

J. The following rules of usage shall be applied to continuous

action steps:

1. "IF AT ANY TIME" steps that require returning to

and performing a particular step of a prcxedure shall

only be applicable while that procedure is in progress.

2. "IF A 1 ANY TIME " steps that do E t require

performance of a particular step shall apply until the

current or subsequent procedures provide alternate

guidance.

3. "WHEN" steps shall apply continuously unless they

are determined to not be pertinent to the recovery effort.

4. All other continuous action steps (Le.*control, monitor,

maintain, etc.) shall apply until the current or

subsequent procedures provide alternate guidance.

K. Parameter values listed in parentheses with the acronym

"AGC." (Advwx. Containment Conditions) shall be used when

containment pressure has exceeded 3.0 PSIG. Once

containment pressure has exceeded 3.0 PSIG, ACX values shall

be used during the remainder of the accident even if pressure

subsequently decreases below 3.0 PSIG. This second setpoint

is used to account for the additional error in the setpoint due to

the containment environment following a high energy line

break.

L. The monitor light panels are used in the EPs to verify safety

components in their proper position for various phases of an

accident. Certain indicating lights are misplaced and do fit

the rules of use. These lights are marked with red film to

indicate an exception to the rules of use 'and shall be evaluated

to determine if the component is in the proper position.

M. Enclosure 1 (Foldout Page) shall be delegated to at least one

RO for the purpose of monitoring and reacting to various plant

parameters independent of the procedure reader's guidance.

Additional copies should be distributed to other crew members

to aid in the monitoring.

N. General statements of philosophy for EP/AIPusage are listed on

Attachment 1 1. 1 (General Statements of Philosophy).

Operations Management Procedure 1-7 Page 9 of 25

7.2. Optimal Recovery Procedures

A. This group of event related emergency procedures (EPs) covers

the diagnostic, mitigating and recoven actions for the

following accidents:

0 Loss of coolant accident (LOCK)

Steam/feed line break

0 Steam generator tube rupture (SGTR)

0 Loss of all AC power

R. Entrance into this EP series is limited to the following two

procedures:

0 E-0 (Reactor Trip or Safety Injection) shall be entered

when:

0 A reactor trip occurs or is required above P-1 1.

8 A safely injection occurs or is required above

P-l I , or below P-11 with autornatic. S/I

blocked.

A safety injection occurs below P-l 1 with

automatic SI blocked, and entry is dictated by

applicable procedure.

8 ECA-0.0 (1.0~sof All AC Iower) shall be entered if a

complete loss of power on both emergency buses

occurs. This includes any time during the performance

of any other emergency procedulr.

I CNS STEAM GENERATOR TUBE RUPTURE PAGE NO.

EP/l/N5000/E-3 17of84

Revision 23

RESPONSE NOT OBTAINED

16. Verify the following valves on all Perform the following:

ruptured S/G(s) CLOSED:

~

a. Verify the following valves on at least

- 0 MSlV one intact S/G - CLOSED:

-0 MSlV bypass valves.

- MSlV

- MSlV bypass valve.

- b. E at least one intact §E cannot be

isolated from all ruptured S/G(sj, THEN

_

GO _TO EP/l/N5000/ECA-3.1 ( S G F

With boss Of Reactor Coolant -

Subcooled Recovery Desired).

- 17. Verify at least one NC pump ON. -

CAUTION NC T-Cold indication in

the ruptured loop may

cause an invalid Integrity

Status Tree condition.

- Disregard NC T-Cold indication in the

ruptured loop, until directed by this EP

or until this EP is exited.

18. "P-I 1 BZR Sll BLOCK

PERMISSIVE" status light (IS-18) is lit,

m:

a. Depress ECCS steam pressure

"BLOCK" pUshbUttons.

- b. Verify main steam isolation blocked

status lights (1SI-13) - LIT.

c. Maintain NC pressure less than

1955 PSlG using one of the following:

- Pzr spray

Technical Specification 5.6.3.

- PzrPORV.

I Pt(s) E-3, (Steam Generator Tube Rupture), step #21.b reads as follows:

IFAT ANY TIME ruptured S/G(s) pressure is decreasing...,

E N p e v f o r m Step 21.

Which onc of the following statements is correct with regards to this step?

A. The step is applicable continuously unless it is determined not to

be pertinent to the recovery effort.

B. The step Is applicable while in E 3 and after transition to

subsequent procedures until alternative guidance is provided.

C. The step is only applicable until another continuous action step is

reached in E-3.

D. The step is only applicable while in E-3.

Distracter Analysis:

A. Incorrect: step applicable only inE-4

Plausible: this is the construct for When ... then actions

B. Incorrect: step applicable only id!-3

Plausible: this is an alternate construct for If at any time actions

C. Ineorrect: step applicable only inE-3

Plausible: this is the construct for generalized continuous action

steps.

D. Correct:

Level: RO&SRB

KA: 62.4.19 (2.7 13.7)

Lesson Plan Objective: ADM-OP SEQ 21

Source: NRC Catawba Exam 97 Ques-339

Level of knowledge: meniory

References:

1. OP-CN-ADM-QP page 10

2. Qh4P 1-7 page 7

3. EP/l/A/SOOO/E-3 page 22

Bank QW?St/OR: 323.1 Answer: 19

1 Pt(s) Unit 1 is shutdown, in mode 5 , following a S/G tube rupture event.

Maintenance is dewatering the main condenser into a portable holdup tank

outside the turbine building (for disposal offsite). Upon completion of the

dewatering and piping flushes, the tank is 3f4 hull. Given the following

radiochemistry analysis of the tank contents:

e Total tank activity = 18 Ci with a combined half life of 8 days

0 Tritium activity = 1 Ci with a half life of 12.6 years

Noble gas activity = 6 Ci with a half life of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

Which one of the following action(s) (if any) meets plant requirements for

these conditions, with the minimum risk of further contarnination?

REFERENCESBROVIDED: SLC16-11.17

A. Immediately move the tank into the turbine building.

B. Stop all additions of radioactive niaterial into the tank and allow

the contents to decay to within limits.

6. Reduce the tank contents by transferring radioactive material

back to the condenser.

D. No action is required at this time.

~~~~~~~~~~~~~~~~~~~

Distractcr Analysis: SEC 16-11.17 requires a maximum of 10 Ci in an

outside tank, exclusive of tritium and noble gases. This tank has 11 Ci, but

the non-tritiudnoble gas contribution is decaying at a rate of more than

1Ci/24 brs. Thus after 48 h o w , the total activity = 18-1-7= 11-2=9 Ci

A. Incorrect: Moving the tank does not provide spill/ovefflow

protection and risks further contamination by spillage.

Plausible: Moving the tank indoors technically meets the SLC

requirement.

B. Correct: Within 48 hrs the tank will be within the SLC limit.

C. Incorrect: Transferring water back to the condenser would

recontaminate it and risk spillage.

Plausible: This is a possible option to meet the SLC requirement, and

could be chosen if the candidate cannot detemiine the decay rate of

the tank.

D. Incorrect: Must at least suspend additions to the tank.

Plausible: If the candidate miscalculates the non-tritiwdnoble gas

contribution.

Level: SRO Only 10CFR55.43(b)4

d K A : G2.3.3 (1.8/2.9)

Lesson Plan Objective: WE-WL. Obj: 16

Source: Bank

Level of knowledge: analysis

References:

1. OP-CN-WE-WL page 22

2.SLC 16.11-17pages 1-2-PROVpDED

-. DUKE POWER

...-. .. .......-.. CATAWBA OPERATIONS TRAINING

. ._._- . ......

OP-CN-WE-WL FOR TRAINING PURPOSES ONLY REV. I I

Page 4 Of 23

16.iz RADlObOGlCAL EFFLUENTS CONTROLS

16.31-17 LIQUID HOLDUP TANKS

COMMITMENT:

The quantity of radioactive material contained in each temporary unprotected

outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and

dissolved or entrained noble gases.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With the quantity of radioactive material in any of the above tanks exceeding the

above limit, immediately suspend all additions of radioactive material to the tank,

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events

leading to this condition in the next Radioactive Effluent Release Report. pursuant to

TESTING REQUIREMENTS:

The quantity of radioactive material contained in each of the above tanks shall be

determined to be within the above limit by analyzing a representative sample of the

tank's contents at least once per 7 days when radioactive materials are being added

to the tank.

1, Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical

Specifications Amendments for Catawba, September 30, 1998.

2. Technical Specification 5 5 4 2 7Explosive Gas and Storage Tank

Radioactivity Monitoring Program.

BASES:

The tanks induded in this COMMITMENT are all those outdoor radwaste tanks that

are not surrounded by liners, dikes or walls capable of holding the tank contents and

Chapter 16.11-17 Page 1 of 2 01/16/99

-

16.11 RADIOLOGICAL EFFLUENTS CONTROLS

16.11-17 LlQUlD HOLDUP TANKS

COMMITMENT:

The quantity of radioactive material contained in each temporary unprotected

outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and

dissolved or entrained noble gases.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With the quantity of radioactive material in any of the above tanks exceeding the

above limit, immediately suspend all additions of radioactive material to the tank,

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events

leading to this condition In the next Radioactive Efnuent Release Report, pursuant to

Technical Specification 5.6.3.

TESTING REQUIREMENTS:

The quantity of radioactive material contained in each of the above tanks shall be

determined to be within the above limit by analyzing a representative sample of the

tanks contents at least once per 7 days when radioactive materials are being added

to the tank.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical

Specifications Amendments for Catawba, September 30, 1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank

Radioactivity Monitoring Program.

BASES:

The tanks included in this COMMITMENT are all those outdoor radwaste tanks that

are not surrounded by liners, dikes or walls capable of holding the tank contents and

Chapter 18.13-17 Page 1 of 2 01/16/99

BASES (con't)

that do not have tank overflows and surrounding area drains connected to the Liquid

Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks

provides assurance that in the event of an uncontrolled release of the tank's

contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix 5, Table II, Column 2, at the nearest potable water supply and the

nearest surface water supply in an UNRESTRICTED AREA.

Chapter 16.$1-57 Page 2 o f 2 O i l 1 6/99

-- DUKE POWER CATAWBA OPERATIONS Tf?A!Mh!G

D. S.L.6. 46.11-2- Radioactive liquid effluent monitoring instrumentation.

E. S.L.C. 16.11-3 Dose - Limits quarterly and yearly dose to members of the

Public at all times.

1. Quarterly

a) Less than or equal to 1.5 mrem whole body

b) Less than or equal to 5 mrem any organ

2. Yearly

a) Less than or equal to 3 mrem whole body

b) Less than or equal to 10 mrem any organ

F. S.L.6. 16.11-4 Liquid Radwaste Treatment System shall be operable when

the liquid effluent, from each unit, to unrestricted areas would exceed 0.06

mrem whole body or 0.2 mrem to any organ in a 31 day period.

G. S.L.C. 16.11-7 Radioactive Gaseous Effluent Monitoring Instrumentation.

Table 16.11-5.

H. S.L.C. 16.11-17 Liquid Holdup Tanks -The quantity of radio-active material

contained in each temporary unprotected outdoor tank shall be limited to

less than or equal to 10 Curies, excluding tritium and dissolved or entrained

noble gases at all times.

I. Design Basis Specifications - Wb

1. NDINS Sump Pumps and level switches are safety related, and are

required to be operable. At least 2 pumps shall be operable at all times.

2. CAPT Sump Pumps (?Aand 2A) are also safety related. If either of

these sump pumps are inoperable, then the CAPT is declared

inoperable and the SSF is placed in degrade.

2.6 Annunciator Response (Obi. # ) - Refer to actual Annunciator Response for

II

current and complete actions and information.

2.7 Review Limits and Precautions (Obj. #17).

A. 0P/0/B/6500/014 (Operations Liquid Waste Release)

B. OP/?/N6500/009(Miscelianeous Equipment Brains to WL System)

C. OP/l/Ai6500/014 (Operations Controlled Liquid Waste Systems)

3. Summary

3.1 Review Objectives

OP-CN-WE-WL FOR TRAINING PURPOSES ONLY REV. f l

Page 22 of 23

Bank Question: 096. f Answer: B

1 Pt(s) Unit 1 is operating in mode 3 preparing for a reactor startup following a

refueling outage. Given the following events and conditions:

NC Pump IC is running.

Reactor trip breakers are tagged open.

Maintenance determincs that the MOV test data from the outage

indicates that the torque switches for 1ND-6SB (NO TR" I B IIOT

LEG INJISQL) have been set too low.

The SMW request3 OSM approval to tag closed 1ND-6SB for repairs

Which one of the following statements correctly describes the operating

restrictions and implications of tagging closed 1ND-6SB?

REFERENCES PROWDED: - Tech Spec's w/Bases 3.4.5,3.4.6,3.5.2

A. 1ND-65B may be tagged closed for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , if the steam

generator in the running NC loop is operable.

B. 1ND-65B may not be tagged closed because this would make both

trains of MD inoperable.

C. 1ND-65B may not be tagged closed, unless two NCPs are running

with operable steam generators.

D. 1ND-65B may be tagged closed, if 1ND65B is restored to

operation prior to transitioning to mode 2.

Distracter Analysis:

A. Incorrect: Both trains of hl)will be inoperable.

Plausible: If the candidate assumes 1 S i 6 and the A ND loop.

B. Correct: ND-65 prevents ND flow to all 4 loops.

6. Incorrect: Both trains of ND will be inoperable.

Plausible: If the candidate focuses only on decay heat removal.

D. Incorrect: Both trains of ND will be inoperable.

Plausible: Ifthe candidate assumes that one ND train is sufficient In

mode 3.

Level: SRO Only IOCFR55.43@)2

KA: G 2.2.24(2.6/3.8)

Lesson Plan Objective: PS-NB SEQ 11

Source: Bank Catawba NRC 2000

Level of ktiowledge: comprehension

References:

1. OP-CN-PS-ND pages 12.18

2. Tech Spec & Bases 3.4.5 -PROVIDED

3. Tech Spec & Bases 3.4.6 -PROVIDED

4. Tech Spec &Bases 3.5.2 -PROVIDED

DUKE POWER..........................

-....... CATAWBA OPERATIONS

.........

TRAINING

....... ..

-

P

Objective T

R

Q

I

X

Describe ND system startup

Describe NB system operation in parallel mode

Describe establishing pressurizer spray from the ND system

-

X

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

-

Time: 2.0 hours0 days 0 hours 0 weeks 0 months

QP-CN-PS-ND FOR TRAINING PURPOSES ONLY REV.29

Page 5 of I 9

DUKE POWER _c_ - CATAWBA OPERATIONS TRAINING

c) Local Temperature indication downstream of ND Hx (NDT 5020,

5030)

5. Flow Indications (Obj. #8)

a) Downstream of ND Hx's (NDFT 5180.5190)

1) Indication in CIR (MC-11) for Cold legs flow A&B (C&B)

2) Used for auto control of NB Hx Bypass valves

I. NV System letdown Isolation ND-24A f58B)

1. Motor-operated, normally closed

2. Supplies letdown from NB to NV

3. Operated from MC-11 (Obj. #8)

J. Discharge Paths

1. ND Injection to the Cold legs. A & 8 (C & D)

a) Train cross ties. ND-3% (65B) (Obj. #$)

1) Motor operated gate valves

2) Normally open, to ensure flow to all four cold legs (FSAR

Commitment) even if one train fails

3) Close during recirculation phases foBlowing a LOCA.

b) Containment Isolation valves NI-173A (1788) with 3 position select

switch. (Obj. %)

1) Normally opened with power removed to ensure injection Row;

DISCONNECT positron.

2) Throttle position was added for use as a means of controlling

ND flow in the event normal flow control is lost (loss of VI)

during partial drain operations which would cause ND-26 & SO

to fail open resulting in vortexing.

3) In the ENABLE position the valves will open or close without

throttle capability.

c) Flow limiting orifices in discharge line to ea& Cold Leg to balance

Row during injection.

d) Flow indication provided in Cia.

e) Annunciator, AD-9, for "ND TKN A(B) to NC C-LEGS LOOPS C-

D(A-B) LO FLOW set at 1050 gpm and ND purn'p running (time

delay keeps alarm functional two (2) minutes following breaker trip.

9 Discharge relief set at 600 psig relieves to the Recycle Holdup Tank.

1) To protect against backleakage from the NC system through the

discharge flowpath check valves.

OP-CN-FS-ND FOR TRAINING FURFOSES ONLY REV.29

Page ?2of 19

DUKE POWER

.............................. ......_ _ .....

CATAWBA

. .

OPERAT/ONS TRAINING

. nxmm,irrr

2) Design analysis has shown that ND flow through the hot leg

injection header is not required to dilute the boron concentration

at the top of the core. One NI pump injecting to its respective

hot legs provides sufficient flow to perform this task.

3) The shift to hot leg recirculation for the NI train will ensure that

heat removal from this upper area of the core is accomplished

and the boron concentration is diluted...

4) The ND system will be aligned for hot leg recirculation only if

flow cannot be established from at least one NI train.

5) The NV pumps continue to discharge to the cold legs to ensure

suficient core cooling in the event of a hot leg break.

b) Hot leg recirc. transfer is made with the NI or ND pumps running

c) Procedure Summary

1) Nl Pumps are aligned for Hot Leg Recirculation

2) If flow is verified from at least one train of the NI system the

operators are directed to return to the procedure and step in

effect.

3) If flow is not verified from at least one train of the NI system, the

ND pumps will be aligned for Hot Leg Recirculation as follows:

(a) Verify the ND loop suction isolations and the crossover

isolations to the cold legs are closed.

(b) Open the hot leg injection isolation

(c) For any ND train not aligned for auxiliaty containment

spray, dose the cold leg isolation and open the

associated crossover isolation valve.

(d) Ensure monitor light panel in correct alignment for hot leg

recirculation and return to procedure and step in effect.

d) Hot leg recirculation flow is to NC loops " B and "C" from the ND

system.

e) The ND pumps will continue to supply the NV and NI pump suction.

2.4 Technical Specifications (Obj. #I 1 & #12)

A. 3.4.6 RCS Loops - Mode 4

B. 3.4.7 RCS Loops - Mode 5, Loops Filled

C. 3.4.8 RCS Loops - Mode 5, Loops Not Filled

D. 3.5.2 ECCS-Operating

E. 3.5.3 ECCS-Shutdown

QP-CN-PS-ND FQR TRAINING PURPOSES ONLY REV.29

Page 18 of 19

ECCS .-Operating

3.5.2

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

3.5.2 ECCS-Operating

LCO 3.5.2 Two ECCS trains shall be OPERABLE'.

APPLICABILITY MODES 1 ,2, and 3,

In MODE 3, both safety injection (SI) pump flow paths may be isolated by

closing the isolation valves for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to perform pressure isolation

valve testing per SR 3.4.14.1

ACTIONS

CONDITION REQUIRED ACTION COMtXETION TIME

A. One or more trains A.l Restore train($ to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ' I

inoperable. OPERABLE status.

At least iOO% of the

ECCS how equivalent to

a single OPERABLE

ECCS tnin available.

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

1 6.2 Be in MODE 4 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

'For eaee ECCS train on Unit 2. the Completion Time that one ECCS Uain can be inoperable a8 specikd by

Required Actian A.1 may be extended beyond the 72 hour8 up io 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months as pati Of ihe NSWS system>Upgrades.

System upgrades include maintenance and modification activilles 89MCialed wilh cleaning OF NSWS poing. va!ves

and branch line?, neressav repairs a~,bior r s p t a c m t , vahe repair andior repla.lacement.and replacement ai

pCltBn~of the NSWS pip!ng io the AFW sy8tenl. Upon completior of the cleaning. up~radees.and sys!em

restoration in refueling oiltage 1 EOCIZ. this footnote is no longer applicable.

Catawba Units 1 and 2 3.5.2-1 Amendment Nos. 189!182

RCS Loops - MODES 4

3.4.6

3.4.6 RCS LOODS-MODE 4

LCO 3 4 6 Two loops consisting of any combination of RCS loops and residual heat

removal (RHR) loops shall be OPERABLE, and one loop shall be in

operation

1. All reactor coolant pumps (RCPs) and RHR pumps may be

de-energized f o r 5 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause reduction of the

RCS boron concentration: and

b. Core outlet temperature is maintained st least 10°F below

saturation temperature.

2. No RCP shall be started with any RCS cold leg tsmperature 5 285°F

unless the secondary side water temperature of each stearn

generator (SG) is 5 50°F above each of the RCS cold leg

temperatures.

APPLICABILITY MODE 4

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. OneRCSloop A.l Initiate action to restore a Immediately

OPERABLE. second loop to

OPERABLE status.

A Q

Two RHR [oops

C a t a d a Units 1 and 2 3.4.6-1 Amendment Nos. 1731165

RCS LoODs - MODES 4

3.4.6

COMPLETION TIME

8. One RHR loop B1 Be in MODE 5 . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

-

AND

ALL RCS loops

ineperable.

C Both required RCS or C1 Suspend all operations Immediately

RHR loops inoperable involving a reductlon of

KCS boron eollcentratlon

w

-

No RCS or RHK loop in

weratin C.2 lnniate action to restore Immediately

I one bop io OPERABLE

status and operation.

S U R V E I U N C E REQUIREMENTS

I

SURVEILLANCE FREQUENCY

SR 3.4.6 1 Verify one RHR or RCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

range for required RCS loops.

Catawba Units I and 2 3.4.6-2 Amendment Nos. 1931165

RCS Loops - MODES 3

3.4.5