ML101270490

WT-2009-10-Job Performance Measure - A1 Calculate Volume of Primary Makeup Water for Dilution to the Volume Control Tank

ML101270490
Person / Time
Site:	Waterford
Issue date:	10/05/2009
From:	Brian Larson Operations Branch IV
To:	Entergy Operations
References
50-382/09-301, WT-2009-10
Download: ML101270490 (300)

Text

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A1 Calculate Volume of Primary Makeup Water for Dilution to the Volume Control Tank Candidate:

Examiner:

JPM A1 JOB PERFORMANCE MEASURE DATA PAGE Task: Calculate Volume of Primary Makeup Water for Dilution to the Volume Control Tank Task Standard: Candidate correctly calculates the quantity of PMU to lower the RCS boron concentration from 1506 to 1500.

References:

OP-002-005, Chemical and Volume Control System Validation Time: 10 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-002-005, Chemical and Volume Control System, Attachment 11.7.

==

Description:==

This administrative task is performed in conjunction with simulator JPM S2. The candidate will calculate the amount of Primary Makeup Water needed to lower RCS boron concentration from 1506 ppm to 1500 ppm.

DIRECTION TO CANDIDATE:

Direct this JPM through JPM S2.

INITIAL CONDITIONS:

The Plant is in Mode 4 RCS boron concentration is 1506 ppm INITIATING CUES:

The CRS directs you to dilute the RCS to 1500 ppm.

Calculate the amount of PMU needed to achieve this concentration.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A1 TASK ELEMENT STANDARD Candidate completes OP-002-005, Attachment Attachment is completed according to key.

11.7.

Comment: CRITICAL STEP Evaluator: Direct the candidate to proceed with making the PMU addition based on what his calculation determined. Round the value of PMU to a factor of 10 (248 gallons, round to 250 gallons).

END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A1 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The Plant is in Mode 4 RCS boron concentration is 1506 ppm INITIATING CUES:

The CRS directs you to dilute the RCS to 1500 ppm.

Calculate the amount of PMU needed to achieve this concentration.

Waterford 3 Page 5 of 5 NRC Exam 2009

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 028 6.9 VCT MAKEUP USING THE DILUTE MAKEUP MODE (C)

CAUTION THIS SECTION AFFECTS REACTIVITY. THIS EVOLUTION SHOULD BE CROSS-CHECKED AND COMPLETED PRIOR TO LEAVING CP-4.

6.9.1 Inform SM/CRS that this Section is being performed.

NOTE When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD) 1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC ( 340 EFPD) 6.9.2 At SM/CRS discretion, calculate volume of Primary Makeup water to be added on Attachment 11.7, Calculation of Primary Makeup Water Volume for Direct Dilution or VCT Dilute Makeup Mode.

6.9.3 Set Primary Makeup Water Batch Counter to volume of Primary Makeup water desired.

6.9.4 Place Makeup Mode selector switch to DILUTE.

6.9.5 Open VCT Makeup Valve, CVC-510.

39

RCSi PW RCS RCSf

117 CALCULATON OF PRIMARY MAKEUP WATER VOLUME FOR DIRECT DILUTION OR VCT DILUTE MAKEUP MODE 1171 Check the appropriate RCS Volume (%) based on Reactor Power:

Reactor Power from 0 15%: 1 = 62126 GAL Reactor Power from 15 100%: fr 61 42i GAL Li 1 1 72 Record the foNowing data:

Key for JPM Al, to be Initial RCS Boron Concentration ((I ,L: PPM implemented with JPM S2 Desired Final ROS Boron Concentration ((*ct): I 5Q0 PPM 117,3 Calculate the volume (gallons) of Primary Makeup water to be added (C):

(,.

PPM GAL x In = GAL

)so PPM,J Performed by: Verified by C:

(S ig nature) (Date) (Signature)

C The independent verifier is responsible for verifying the accuracy of recorded data as well as calculations.

OP-002005 Revision 028 Attachment 11.7(1 of 1) 174

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A2 Complete OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Candidate:

Examiner:

JPM A2 JOB PERFORMANCE MEASURE DATA PAGE Task: Complete OP-903-001, Technical Specification Surveillance Logs, Attachment 11.18, Adjustment of CPC and Excore Nuclear Instrumentation Data Task Standard: Candidate correctly calculates new values for Core Protection Calculator constants KCAL, TCREF, and TPC and new PPS DVM reading.

References:

OP-903-001, Technical Specification Logs Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 6 NRC Exam 2009

JPM A2 Waterford 3 Page 3 of 6 NRC Exam 2009

JPM A2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-001, Technical Specification Logs, Attachment 11.18.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Instructions on candidates cue sheet Another operator has collected data on OP-903-001, Technical Specification Surveillance Logs, attachment 11.18, for adjustment of Channel A CPC and Excore Nuclear Instrumentation.

Complete the following calculations on Attachment 11.18:

step 11.10.5 for a new DVM reading step 11.10.6 for KCAL step 11.10.6 for TCREF step 11.10.7 for TPC This task is complete when you reach step 11.10.11.1.

Waterford 3 Page 4 of 6 NRC Exam 2009

JPM A2 TASK ELEMENT STANDARD Student must record values and make calculations Complete OP-903-001, Attachment 11.18 for 4 points. 1 transposition error is allowed; no according to key.

calculation errors are allowed.

Comment: CRITICAL STEP END OF TASK Waterford 3 Page 5 of 6 NRC Exam 2009

JPM A2 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

Another operator has collected data on OP-903-001, Technical Specification Surveillance Logs, attachment 11.18, for adjustment of Channel A CPC and Excore Nuclear Instrumentation.

Complete the following calculations on Attachment 11.18:

step 11.10.5 for a new DVM reading step 11.10.6 for KCAL step 11.10.6 for TCREF step 11.10.7 for TPC This task is complete when you reach step 11.10.11.1.

Waterford 3 Page 6 of 6 NRC Exam 2009

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE CHANNEL UNDER ADJUSTMENT: A B c DE a

11.10.3.1 CalcuLate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL-(PMC) © 0 PID-C24230 999( 9.9o ?? o 99.90 i/1/4 HI LINEAR POWER BISTABLE 1 VOLTS $ S o1S HIGH LINEAR POWER %

VOLT X 20 ioo.9o /r.9o /.92 Ic.(

PH ICAL.

(Calibrated Neutron Flux Power)

CPC PID 171 tø(.2.o /01.2.5 /el.2-l 1e4,:s IO/.Zo BDT (Static Thermal Power)

CPC PID 177 988 Calculations Performed by: Verified by:

nature 1gnature Refer to attachment 11.1 Note 9.1 to determine appropriate power indicaUon ii linear power is j 35% steady state. Document indication used in Remarks.

© H COLSS is Inoperable, then use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC CORE POWER is specified.

0 Adjusted is the average value plus 8.5% (8% tolO%) 11 adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 11.10.7 (refer to Attachment 11 .1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant)

CPC PID 064 KCAL (Neutron Flux Power Cal. Constant)

CPC PID 065 PCALIB (Secondary Calorimetric Power Used in Latest CPC Power Calibration) CPC RD 104 TC 1 (Loop 1 Cold Leg Temperature)

CPC PID 160 5Ef2,Z5 TC 2 (Loop 2 Cold Leg Temperature)

CPC PID 161 TCORF (Temp Shadowing Correction Factor)

CPC PID 180 0.

EXCORE LINEAR POWER CALIBRATE POTENTIOMETER POSITION ROM OP-903-001 Revision 036 Attachment 11.18(1 of 3) 142

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

DVM(new)=

Avg. Core Power (Step 11.10.3.1)* I 9?. c 20 20 Use the Average value from the table of step 11.10.3.1, not the Adjusted value.

DVM (new) = i(

?95 or Potentiometer position (new) =

Avg. Core Power % (Step 11 .10.3.1)* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4) d/A Use the Average value from the table of step 11.10.3.1, the Adjusted value.

Potentiometer position (new) =

Performed by: cZZ (Initials)

Verified by:

(Initials) 11.10.6.1 Calculate KCAL (new):

(Stepll.10,3.1)* (stepll.10.4) (stepll.10.4)

KCAL - Avg. Core Power (%) x KCAL x TCORF (new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL 99. 9o x x o. 9999/

(new) ,,o1zo KCAL (new) /0l2 KCAL (new) (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC1 (CPCPID16O)or TC2(CPCPID 161).

TCREF (new) (CPC PID 098)

Performed by: Verified by:

(Initials) (Initials)

OP-903-001 Revision 036 Attachment 11.18(2 of 3) 143

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (CONTD) 11.10.7 Calculate TPC (CPC PID 064):

Avg. Core Power%(Step 11.10.4)* XI TPC (Step 1110.5)

TPC(new)

Avg. BDT (Step 11.10.4)

An Adjusted value may be required. Refer to Note preceding step 11.101.

Q 9

99. x o.82-1z TPC (new)

TPC (new) o. 3oo (CPC PID 064).

Performed by: c___- Verified by:

(Initials) (Initials) 11.10.11.1 Record the following:

• Applicable CORE POWER PMC

• PCALIBCPCPID 104

• HI LINEAR POWER BISTABLE 1 VOLTS

• HI LINEAR POWER%VOLTSx2O

• PHICALCPCPID 171

• BDTCPCPID177 11.10.11,2.1 Record answers:

• HI LINEAR POWER %, VOLTS X 20 YES/NO

• PHICAL, CPC PID 171 YES/NO

• BDT, CPC PID 177 YES/NO 11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by:

SM/CRS Date/Time OP-903-001 Revision 036 Attachment 11.18 (3 of 3) 144

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA DATE CHANNEL UNDERADJUSTMENT:A BE CD DE a

11.10.3.1 Calculate and record the averages of each parameter in the space provided.

0 1 2 3 4 Average Adjusted BSCAL-(PMC) © / I PID-C24230 99*91 7i 99o 9o 99o Ahi?A. I HI LINEAR POWER BISTABLE 1 VOLTS $. s S ols esI HIGH LINEAR POWER %

VOLTX 20 iefc ioo.9o i&?o /cz.9o !of.92 ioo.(

PHICAL.

(Calibrated Neutron Flux Power)

CPC PID 171 tL2ø /ofZ$ 1&1.I /oi.7_t 1c4. g IO/.Zo 1

BDT (Static Thermal Power)

CPC PID 177 9tz Calculations Performed by: Verified by: .

nature ignature D Refer to attachment 11 .1 Note 9.1 to determine appropriate power indication if linear power is 35% steady state. Document indication used in Remarks.

© COLSS is Inoperable, jj use NE-005-201, Heat Balance Calculations, to determine Secondary Calorimetric Power substitute when PMC CORE POWER is specified.

© Adjusted is the average value plus 8.5% (8% tol 0%) It adjustments are being made to PHICAL and/or BDT as listed in Notes above steps 11.10.6 11.10.7 (refer to Attachment 11.1 Note 9.8). Otherwise N/A this block. Use the Average value, not the Adjusted value, for DVM calculation.

11.10.4 Record the following for channel under adjustment:

TPC (Thermal Power Calibration Constant)

CPC PID 064 KCAL (Neutron Flux Power Cal. Constant)

CPC PID 065 PCALIB (Secondary Calorimetric Power Used in Latest CPC Power Calibration) CPC PID 104 TC 1 (Loop 1 Cold Leg Temperature)

CPC PID 160 5V2.Z5 TC 2 (Loop 2 Cold Leg Temperature)

CPC PID 161 5q3./

TCORF (Temp Shadowing Correction Factor)

CPC PID 180 0.

EXCORE LINEAR POWER CALIBRATE POTENTIOMETER POSITION ROM

• 1 OP-903-001 Revision 036 Attachment 11.18 (1 of 3) 142

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (coNTD) 11.10.5 Calculate the new DVM reading or new potentiometer position as follows:

I Avg. Core Power (Step 11.10.3.1)*

DVM(new)=

20 20 Use the Average value from the table of step 1110.3.1, not the Adjusted value.

DVM(new)=

or Potentiometer position (new) =

Avg. Core Power % (Step 11.10.3.1 )* X Old Potentiometer Setting (Step 11.10.4)

Avg. Linear Power % (Step 11.10.4) d/A si Use the Average value from the table of step 11.10.3.1, the Adjusted value.

Potentiometer position (new)

Performed by: Verified by:

(Initials) (Initials) 11.10.6,1 Calculate KCAL (new):

KCAL I (Step 11.10.3.1)*

Avg. Core Power (%)

(step 11.10.4) (step 11.10.4)

- x KCAL x TCORF (new) Avg. PHICAL (step 11.10.3.1)

An Adjusted value may be required. Refer to Note preceding step 11.10.6.

KCAL - X X (new)

KCAL (new)

KCAL (new) = (CPC PID 065) 11.10.6.3 New TCREF (CPC PID 098) = Minimum TC from step 11.10.4:

TC 1 (CPC PID 160) or TC 2 (CPC PID 161).

TCREF (new) = (CPC PID 098)

Performed by: Verified by:

(Initials) (Initials)

OP-903-001 Revision 036 Attachment 11.18 (2 of 3) 143

11.18 ADJUSTMENT OF CPC AND EXCORE NUCLEAR INSTRUMENTATION DATA (C0NTD) 11.101 Calculate TPC (CPC PID 064):

Avg. Core Power % (Step 11.10.4)* xl TPC (Step 11.10.5)

TPC(new)=

Avg. BDT (Step 11.10.4)

An Adjusted value may be required. Refer to Note preceding step 11.10.7.

xl TPC(new) I TPC (new) (CPC PID 064).

Performed by: Verified by:

(Initials) (Initials) 11.10.11.1 Record the following:

• Applicable CORE POWER PMC

• PCALIBCPCPID1O4

• HI LINEAR POWER BISTABLE 1 VOLTS

• HI LINEAR POWER % VOLTS x 20

• PHICALCPCPID 171

• BDTCPCPID177 11.10.11.2.1 Record answers:

• HI LINEAR POWER %, VOLTS X 20 YES/NO

• PHICAL, CPC PID 171 YES/NO

• BDT, CPC PID 177 YES/NO 11.10.12 Performed by: Verified by:

(Initials) (Initials) 11.10.13 Reviewed by:

SM/CRS Date/Time OP-903-001 Revision 036 Attachment 11.18(3 of 3) 144

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A3 Complete Surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test, Attachment 10.11 for SI-329 A Candidate:

Examiner:

JPM A3 JOB PERFORMANCE MEASURE DATA PAGE Task: Complete Surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test, Attachment 10.11 for SI-329 A Task Standard: Candidate correctly calculates the leakage for SI-329 A using Attachment 10.11.

References:

OP-903-008, Reactor Coolant System Isolation Leakage Test Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-008, Attachment 10.11 OP-903-008, Section 7.11.8.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Instructions on candidates cue sheet OP-903-008, Reactor Coolant System Isolation Leakage Test, is in progress.

Currently, SI-329 A, Safety Injection Tank 1A Outlet Check, is being tested.

Steps 7.11.1 through 7.11.8.11 have been completed.

Another operator has recorded the data and entered it on Attachment 10.11, Safety Injection Tank Outlet Check Valves Leak Rate Data.

Complete step 7.11.8.11 on the provided Attachment 10.11.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A3 TASK ELEMENT STANDARD Complete OP-903-008, Attachment 10.11 Candidate correctly calculates the corrected leak according to key. rate for SI-329 A.

Comment: CRITICAL STEP END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A3 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

OP-903-008, Reactor Coolant System Isolation Leakage Test, is in progress.

Currently, SI-329 A, Safety Injection Tank 1A Outlet Check, is being tested.

Steps 7.11.1 through 7.11.8.11 have been completed.

Another operator has recorded the data and entered it on Attachment 10.11, Safety Injection Tank Outlet Check Valves Leak Rate Data.

Complete step 7.11.8.11 on the provided Attachment 10.11.

Waterford 3 Page 5 of 5 NRC Exam 2009

3K 10.11 SAFETY INJECTION TANK OUTLET CHECK VALVES LEAK RATE DATA Steo No.

7.11.2 Sl-329A Test ermission:

e 1 L 1 7 (SM/CRS SignWture) (Date/Time) 7.11.8.8 Initial SI Loop 1A Leak Detection Pressure (PMC PID A43006) a(D5g PSIG Initial SIT 1A Pressure (SI-IPl-0311, Sl-IPI-0312 orSI-IPI0313) 51(0 PSIG Initial millivolt value SIT 1A Level (PMC PID A44101 or A44102) (MV,)

Initial \P: I1Z PSID Start Time: 2.ooc 7.11.8.10.2 Stop Time: 2o2-o Elapsed Time (TE): 2-o mm Final SI Loop 1A Leak Detection Pressure (PMC PID A43006) j4P5 PSIG Final SIT 1A Pressure (SI-IPI-0311, SI-lPl-O3l2orSI-lPI-0313) 515 PSIG Final millivolt value SIT 1A Level (PMC PID A44101 or A44102) (MVf): 93s Final \P: ligo PSID Lowest of Initial AP and Final AP (APL): I (g2. PSID 7.11.8.11 Corrected Leak Rate Calculation for SI-329A:

GROSS LEAK RATE (RG):

RG=[.286 X(MVf-MVJ]ITE RG=[.286X( c,c?3S - )]/TE RG=[.286X( 11 RG[3.I(g ]/TE RG gal/ ,Zo mm RG GPM EXTRAPOLATION RATIO COEFFICIENT (CE): (N/A if RG = 0)

CE = [2235 PSID / APL]° 5

CE = [2235 PSID I 11 42- 5 PSID]° CE=[ i.q& jO.5 CE= I7 CORRECTED LEAK RATE (Rc):

RG X CE R GPMX R O2.. GPM 7.11.13 CVC-199 Locked Closed (may be N/A): Performed: Verified:

7.11.14 SI-219A and SI-219B Locked Open: Performed: Verified:

Calculations Performed by:

(Operator or STA Signature) (Date/Time)

Calculations Verified by:

(STA or Operator Signature) (Date/Time)

OP-903-008 Revision 7 111 Attachment 10.11(1 of 4)

10.11 SAFETY INJECTION TANK OUTLET CHECK VALVES LEAK RATE DATA Step No.

7.11.2 SI-329 Te Permission:

(SM/CRS Si ature) 7.11.8.8 Initial SI Loop 1A Leak Detection Pressure (PMC PIDA43006)

Initial SIT 1A Pressure (SI-IPI-0311, SI-IPI-0312 orSI-IPI-0313) PSIG Initial millivolt value SIT 1A Level (PMC PID A44101 or A441 02) (MV,)

Initial \P: PSID Start Time:

7.11.8.10.2 StopTime:

Elapsed Time (TE): mm Final SI Loop 1A Leak Detection Pressure (PMC PID A43006) PSIG Final SIT 1A Pressure (SI-IPI-0311, SI-IPI-0312 or SI-IPI-0313) 515 PSIG Final millivolt value SIT 1 A Level (PMC PID A441 01 or A441 02) (MVf): 935 Final \P: I i o PSID Lowest of Initial P and Final P (\PL): 11L17_ PSID 7.11.8.11 Corrected Leak Rate Calculation for SI-329A:

GROSS LEAK RATE (RG):

RG = [.286 X (MVf MV)] / TE RG=[.286X(

- )]/TE RG=[.286X( )]/TE R[ ]/TE RG gall mm RG= GPM EXTRAPOLATION RATIO COEFFICIENT (CE): (N/A if RG=0)

CE = [2235 PSID / APL]° 5

CE = [2235 PSID / 5 PSID]°

]O5 0

E [

CE =

CORRECTED LEAK RATE (Rc):

R = RG X CE R= GPMX R GPM Lk2a 7.11.13 CVC-199 Locked Closed (may be N/A): Performed: Verified:

7.11.14 SI-219A and SI-219B Locked Open: Performed: Verified:

Calculations Performed by:

(Operator or STA Signature) (Date/Time)

Calculations Verified by:

(STA or Operator Signature) (Date/Time)

OP-903-008 Revision 7 111 Attachment 10.11(1 of 4)

Surveillance Procedure OP-903-008 Reactor Coolant System Isolation Leakage Test Revision 7 Steps 7.11.8 through 7.11.11 may be performed concurrently and in any sequence. Steps 7.11.1 through 7.11.7 must be performed jgr to testing Safety Injection Tank 1A Outlet Check, SI-329A.

DURING THE LEAK TEST OF SAFETY INJECTION TANK IA OUTLET CHECK, SI-329A, SAFETY INJECTION TANK 1A LEVEL RISES 1%, THEN TECHNICAL SPECIFICATION 3.5.1 MUST BE COMPLIED WITH.

7.11.8 o test Safety Injection Tank 1A Outlet Check, SI-329A, perform the following:

7. . .1 Verify Safety Injection Tank 1A level is on scale.

7.1 . . erify the following valves Closed:

SI-139B LPSI Header To RC Loop 1A Flow Control SI-225B HPSI Header B To RC Loop 1A Flow Control 5If Hot Leg Injection Train A is being used, Open RC Loop 1 Hot Leg Injection Leakage Drain, SI-301.

.1 . if Hot Leg Injection Train B is being used, Open RC Loop 2 Hot Leg ction Leakage Drain, SI-302.

7 In leg injection is being used, then Open HPSI Header A to RC Loop 1A Flow Control, SI-225A.

Safety Injection Header 1A pressure is monitored on Leakage Pressure Cold Leg 1A (SI-IPI-0319) indicator on CP-8.

7.11. .

.f Hot Leg Injection Train A or B is being used, then Open Safety Injection Tank 1A Leakage Drain, SI-303A, monitor Safety Injection Header 1A pressure.

.11.8 . When Safety Injection Header 1A pressure is approximately equal to Pressurizer Pressure, then proceed to the next step.

77

Surveillance Procedure OP-903-008 Reactor Coolant System Isolation Leakage Test Revision 7 RCS PRESSURE MUST BE AT LEAST 50 PSIG GREATER THAN SAFETY INJECTION TANK 1A PRESSURE TO OPEN SAFETY INJECTION TANK 1A OUTLET ISOLATION, SI-331A.

Safety Injection Tank 1A Outlet Isolation, Sl-331A, Open.

Record the following on Attachment 10.11, Safety Injection Tank Outlet Check Valves Leak Rate Data:

(9 Initial SI Loop 1A Leak Detection Pressure (PMC PID A43006)

Initial SIT 1A Pressure (SI-IPI-031 1, SI-IPI-0312 or SI-IPl-031 3)

Initial millivolt value of SIT 1A Level (PMC PIDA44IO1 orA44102)

Initial \P (Initial SI Loop 1A Leak Detection Pressure minus Initial SIT 1A Pressure)

Start Time When either of the following criteria are met, then go to step 7.11.8.10:

Corrected Leak Rate will calculate to 1 GPM at any point between 20 minutes and 120 minutes elapsed time.

120 minutes have elapsed, or at any point beeen 20 minutes and 120 minutes elapsed time that it becomes apparent that Corrected Leak Rate will not calculate to 1 GPM by the time 120 minutes have elapsed 7.11.8.10 erform the following:

/ .11 . .1 Verify Safety Injection Tank 1A Leakage Drain, SI-303A, Closed.

7.11.8.10.2 ecord the following on Attachment 10.11, Safety Injection Tank Outlet Check Valves Leak Rate Data:

Stop Time (j/EIapsed Time Final SI Loop 1A Leak Detection Pressure (PMC PID A43006)

Final SIT 1A Pressure (SI-IPI-031 1, SI-IPI-031 2 or SI-IPI-031 3)

Final millivolt value of SIT 1A Level (PMC PID A44101 or A44102)

Final z\P (Final SI Loop 1A Leak Detection Pressure minus Final SIT 1A Pressure) 78

Surveillance Procedure OP-903-008 Reactor Coolant System Isolation Leakage Test Revision 7 7.11.8.11 Calculate Corrected Leak Rate for Safety Injection Tank 1A Outlet Check, Sl-329A, on Attachment 10.11 Safety Injection Tank Outlet Check Valves Leak Rate Data.

7.11.8.12 Verify Safety Injection Tank 1A OPERABLE in accordance with Technical Specification 3.5.1.

7.11.8.13 Record test results on Attachment 10.12, Results Summary Table.

7.11.8.14 Close HPSI Header A to RC Loop 1A Flow Control, Sl-225A.

NOTE Steps 7.11.1 through 7.11.7 must be performed jQrjtesting Safety Injection Tank lB Outlet Check, Sl-329B.

CAUTION DURING THE LEAK TEST OF SAFETY INJECTION TANK 1 B OUTLET CHECK, SI-329B, SAFETY INJECTION TANK lB LEVEL RISES 1%, THEN TECHNICAL SPECIFICATION 3.5.1 MUST BE COMPLIED WITH.

7.11.9 To test Safety Injection Tank 18 Outlet Check, SI-329B, perform the following:

7.11.9.1 Verify Safety Injection Tank lB level is on scale.

7.11.9.2 Verify the following valves Closed:

• SI-138B LPSI Header To RC Loop lB Flow Control

• SI-226B HPSI Header B To RC Loop lB Flow Control 7.11.9.3 If Hot Leg Injection Train A is being used, then Open RC Loop 1 Hot Leg Injection Leakage Drain, SI-301.

7.11.9.4 If Hot Leg Injection Train B is being used, Open RC Loop 2 Hot Leg Injection Leakage Drain, Sl-302.

7.11.9.5 cold leg injection is being used, then Open HPSI Header A to RC Loop 1 B Flow Control, SI-226A.

79

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A4 Calculate Stay Times Based on Dose Rates Candidate:

Examiner:

JPM A4 JOB PERFORMANCE MEASURE DATA PAGE Task: Calculate Stay Times Based on Dose Rates Task Standard: Candidate correctly calculates the allowed stay time to complete the described tagout without exceeding his yearly Waterford 3 administrative radiation dose limits.

References:

None Validation Time: 10 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A4 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None READ TO CANDIDATE DIRECTION TO CANDIDATE:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Instructions on candidates cue sheet You have been assigned to verify a tagout in the Regenerative Heat Exchanger Room.

Your yearly dose to date is 1750 mrem TEDE for the year.

The dose rate in the room is 650 mrem/hour.

Based on Waterford 3 yearly administrative limits, what is your stay time in the room?

Do all of your calculations on this sheet.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A4 TASK ELEMENT STANDARD Calculate stay time based on dose rate and Candidate correctly calculates the stay time as Waterford 3 yearly TEDE limits. 23.0 - 23.1 minutes.

Comment: CRITICAL STEP Waterford 3 administrative TEDE limit: 2000 mrem Dose for the year: 1750 mrem Remaining dose for the year: 250 mrem Time allowed in room: 250 mrem / 650 mrem/hour 0.3846 hour0.0445 days <br />1.068 hours <br />0.00636 weeks <br />0.00146 months <br /> or 23.0 - 23.1 minutes END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A4 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

You have been assigned to verify a tagout in the Regenerative Heat Exchanger Room.

Your yearly dose to date is 1750 mrem TEDE for the year.

The dose rate in the room is 650 mrem/hour.

Based on Waterford 3 yearly administrative limits, what is your stay time in the room?

Do all of your calculations on this sheet.

Waterford 3 Page 5 of 5 NRC Exam 2009

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A5 Verify Core Protection Calculator and Plant Protection System and Calorimetric Powers are within Limits Candidate:

Examiner:

JPM A5 JOB PERFORMANCE MEASURE DATA PAGE Task: Verify Core Protection Calculator and Plant Protection System and Calorimetric Powers are within Limits Task Standard: Candidate reviews completed OP-903-001 paperwork and concludes that PPS Channel D and CPC PID 171 Channel A must be adjusted for current power level and that all 4 channels of CPC PID 177 must be adjusted prior to exceeding 80% power.

References:

OP-903-001, Technical Specification Logs Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A5 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-001, Technical Specification Logs OP-010-004, Power Operations READ TO CANDIDATE DIRECTION TO CANDIDATE:

Each administrative JPM has a cue sheet with the instructions for that JPM. Each administrative JPM stands alone, and conditions from 1 JPM do not carry over to any other JPM. If you have any questions, raise your hand and I will come to your desk.

Provide all answers on the sheets provided.

Plant conditions are as follows:

68% power Raising power to 100% following mid-cycle outage Night Shift Tech Spec logs have been completed Review the completed OP-903-001, Technical Specification Logs.

1. Determine the status of Linear Power Channel Calibrations for the current plant conditions.

2. Determine what adjustments are necessary and at what power during power ascension to 100% power.

Write your conclusions on this sheet.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A5 TASK ELEMENT STANDARD Candidate determines that Channel D CP-10 DVM Determine the status of Linear Power Channel and CPC Channel A Cal Neutron Flux Power, CPC Calibrations for the current plant conditions PID 171, must be adjusted.

Comment: CRITICAL STEP The Channel D CP-10 DVM reading of 33.59 is more than -0.025 volts below Calculated Power volts at 3.386 volts.

CPC A Cal. Neutron Flux Power at 67.20 % is more than -0.5 below Calculated Power at 67.72 %.

TASK ELEMENT STANDARD Determine what adjustments are necessary and at Candidate determines that CPC Thermal Power, what power during power ascension to 100% CPC PID 177, on Channels A, B, C, and D must be power. adjusted prior to exceeding 80% power.

Comment: CRITICAL STEP All 4 Channels of CPC Thermal Power must be within 2% power of Calculated Power before power exceeds 80%, where the allowance changes from +10% to +2%.

Candidate could also state that power reading should be taken again as power nears 80% to see how far from Calculated Power they are.

END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A5 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

Plant conditions are as follows:

68% power Raising power to 100% following mid-cycle outage Night Shift Tech Spec logs have been completed Review the completed OP-903-001, Technical Specification Logs.

3. Determine the status of Linear Power Channel Calibrations for the current plant conditions.

4. Determine what adjustments are necessary and at what power during power ascension to 100% power.

Write your conclusions on this sheet.

Waterford 3 Page 5 of 5 NRC Exam 2009

DATE iiz4+/-

DESCRIPTION NOTE MODE T. S. COMP# I L I i I I 4.3.1.1 TbI Linear Power Channel 9.0 1, >15% 4.3-1 Calibration:

(2,9,10,14)

>15%

CORE AT PWR BDELT 9.1 C24104 N/A  %

<35%

BSCAL Power 9.1 35% C24230 N/A  %

Heat Balance Calculated 9.2 NE-5-201 N/A  % /1/4 Cal Power Calculated Excore Pwr 9.3 Cal Pwr:20% /V N/A Volts Chn A Actual Ex Pwr 9.4 CP-10 DVM N/A Volts 339 Chnl B Actual Ex Pwr 9.4 CP-10 DVM N/A Volts 3. %j Chnl C Actual Ex Pwr 9.4 CP-10 DVM N/A Volts Chnl D Actual Ex Pwr 9.4 CP-10 DVM N/A Volts 3S 9.5 80% N/A +/-0.1 Volts /, NA Compare Actual and Calculated Excore Pwr >15% -.025 Volts 99 N/A q, NA

< 80% +0.5 Volts CPC A Cal Ntn Flux Pwr 9.6 CPC PID 171 N/A  % 7. o CPCAThermal Pwr 9.6 CPC PID 177 N/A  %

CPC B Cal Ntn Flux Pwr 9.6 CPC PID 171 N/A  %

CPC B Thermal Pwr 9.6 CPC PID 177 N/A  %

CPC C Cal Ntn Flux Pwr 9.6 CPC PID 171 N/A  %

CPC C Thermal Pwr 9.6 CPC PID 177 N/A  % 7Q, CPC D Cal Ntn Flux Pwr 9.6 CPC PID 171 N/A  %

CPC D Thermal Pwr 9.6 CPC PID 177 N/A  %

9.7 80% N/A +/-2% J, NA Compare CPC and Calorimetric Power >15% -0.5% to 9.8 N/A NA

<80% +10%

REMARKS:

OP-903-001 Revision 036 Attachment 11.1 (14 of 37) 31

9.0 In Mode 1, >15% PWR, verify the values of plant power for j[four channels of CPC and Excore Nuclear Instrumentation agree within those values listed in notes 9.5, 9.7, 9.8, ic 9.9 of calorimetric power as calculated by COLSS NE-5-201, Heat Balance Calculations, if COLSS is Inoperable. If desired, use Attachment 11 .3, Linear Power Channel Calibration (Data Sheet) to record data for computing 5 minute averages, then attach documentation to this Attachment. NA all entries on Attachment 11 .1 (page 14 of 37) j.f.jQin Model, >15°/o pwr.

9.1 Use the following table for Linear Power Calibration for the specified conditions:

UFM in service and plant is at steady state Reactor Power 35% Secondary Calorimetric PMC PID 024230 Reactor Power <35% BDELT PMC PID C24l04 UFM not in service and plant is at steady state Reactor Power 95% MSBSRAW PMC PID C24631 Reactor Power <95% and 35% FWBSRAW PMC PID 024630 Reactor Power <35% BDELT PMC PID C24104 UFM in service and plant is at steady state Reactor Power 95% MSBSRAW PMC PID 024631 Reactor Power <95% and 40% USBSRAW PMC PID 024629 Reactor Power <40% and 35% FWBSRAW PMC PID 024630 Reactor Power <35% BDELT PMC PID 024104 NOTE: Steady State can be defined as .5% power change over the last 30 minutes as indicated by PMC PID B24007 indicating OFF.

9.2 If COLSS is Inoperable, then record Calculated Calorimetric Power as determined by applicable section of NE-5-201, Heat Balance Calculations. Attach documentation to this Attachment. NA if COLSS is Operable.

9.3 Determine the Calculated Excore Nuclear Power (in volts) by dividing the value of calorimetric power (secondary j calculated) by 20% I volt.

Calorimetric Power %

= volts 20% I volt OP-903-OOl Revision 036 Attachment 11.1 (21 of 37) 38

9.4 Record the value of Excore Nuclear Power (in volts) as indicated on the PPS Test Module for each Channel by performing one of the following:

a) Position the BISTABLE SELECT Switch to Position 1 b) Position the METER INPUT SELECT Switch to INPUT or c) Connect DVM to voltage output taps d) Position BISTABLE SELECT Switch to Position 1 e) Position the METER INPUT SELECT Switch to EXT DVM then a) Place BISTABLE SELECT Switch to OFF 9.5 If 80% RTP, then verify the value of Actual Excore Nuclear Power (in volts) is within 0.1 volts of the value of Calculated Excore Nuclear Power.

If >0.05 volts, then the applicable section of Attachment 11.10, Adjustment of CPC and Excore Nuclear Instrumentation should be performed.

1>0.1 volts, then perform applicable section of Attachment 11.10. Attach 11.10 to this Attachment.

9.6 Record the values of CPC Static Thermal Power (BDT) and CPC Calibrated Neutron Flux Power (PHICAL) for aM four CPC Channels, by displaying the applicable PID on the CPC ROM.

9.7 II 80% RTP, then verify the values BDT and PHICAL are within 2% RTP of the value of Secondary Calorimetric (COLSS or Manual Heat Balance) Power. If >1% RTP, then the applicable section of Attachment 11.10, should be performed.

If >2% RTP, then perform applicable section of Attachment 11.10. Attach Attachment 11 .10 to this Attachment.

OP-903-001 Revision 036 Attachment 11.1(22 of 37) 39

9.8 If >15% RTP (Rated Thermal Power) and <80% RTP, then verify the channel BDT and PHICAL values are within -0.5% RTP to +10% RTP of the value of the appropriate calorimetric (COLSS, Manual Heat Balance, or Primary ..\T) power.

If the channel is within the limit, then do not calibrate excert as required during initial power ascension following Refueling.

If the channel is cireater than the high limit during initial power ascension to <80%

following refueling, or anytime the channel is kss than the low limit, then perform applicable section of Attachment 11.10 to adjust affected signal(s) to within 0.0% RTP to +10% RTP of the Calorimetric power.

If the channel is greater than the high limit other than during initial power ascension to

<80% following refueling, then perform the following: [CR-WF3-2006-03726]

• Perform applicable section of Attachment 11.10 to adjust affected signal(s) to within +8.0% RTP to +10% RTP above Calorimetric power. Attach Attachment 11.10 to this Attachment.

• Initiate a Condition Report and request Reactor Engineering investigate the reason for the power difference being greater than +10%. This condition should be investigated resolved prior to power ascension to 80% power.

[NF-WTFD-06-48]

9.9 II >15% RTP and <80% RTP, then verify the value of Actual Excore Nuclear Power (in volts) is within -0.025 volts to +0.5 volts of the value of Calculated Excore Nuclear Power. If the channel is within the limit, then do not calibrate except as required during initial power ascension following Refueling. jf the value is outside the limit, tb..a perform applicable section of Attachment 11.10 to adjust affected signal(s) to within 0.0 volts to +0.5 volts of the value of Calculated Excore Nuclear Power. Attach Attachment 11.10 to this Attachment.

10.0 If Plant Monitoring Computer not available, then take readings locally.

11.0 Channel Check is between Primary Backup Towers at the 33 ft. level.

11.1 The 199 Primary Wind Speed Channel Check shall meet as a minimum >0.134 meters/second and not constant (jjy indicated change within 30 minutes). Zero wind speed or a seized instrument is indicated by a constant 0.134 meters/second.

12.0 Channel Check is between similar instruments on the Primary Tower.

1 3.0 Channel Check is between Primary nç Backup Towers at the 33 ft. level with either 33 ft. wind speed >2 meters/second.

13.1 If jy 33 ft wind speed sensor is >2 meters/second, then check Primary 199 ft. wind direction is within 60° of Primary jçj. Backup 33 ft. wind direction.

OP-903-001 Revision 036 Attachment 11.1(23 of 37) 40

(Initial/Date)

NOTE B )fl of instantaneous power during power maneuvers.

Additionally, once me srnuutiiing factor is applied at approximately 98% MSBSRAW (PMC PID 024631), BSCAL becomes a time weighted average of power recorded over approximately 20 minutes. The following tables list COLSS calculated powers available during power maneuvering to monitor instantaneous power: [CR-WF3-2005-03985]

UFM not in service Reactor Power 95% MSBSRAW PMC PID C24631 Reactor Power < 95% and 35% FWBSRAW PMC PID 024630 Reactor Power < 35% BDELT PMC PID 024104 UFM in service Reactor Power 95% MSBSRAW PMC PID C24631 Reactor Power < 95% and 40% USBSRAW PMC PID 024629 Reactor Power < 40% and 35% FWBSRAW PMC PID 024630 Reactor Power < 35% BDELT PMC PID 024104 9.1.61 When Reactor Power is greater than 40% pjj to /D/o exceeding 50%, place the UFM in service in accordance with OP-004-005, Core Operating Limits Supervisory System /

Operation. [TRM 3.3.5]

9.1.62 Prior to exceeding 50% Reactor Power, Start third CondensateqI / iicIo9 Pump in accordance with OP-003-003, Condensate.

i-ni, 9.1.63 Prior exceeding 50% power, verify CPC Calibrated Neutron 4Y / /fS/&9 Flux Power (CPC PID 171), CPC Thermal Power (CPC PID 177) and Actual Excore Nuclear Power (OP-la DVM)

/ I agree with Secondary Calorimeteric Power within limits specified in OP-903-00l, Technical Specification Surveillance Log.

OP-010-004 Revision 305 Attachment 9.1 (13 of 18) 31

(Initial/Date) 9.1.64 When Reactor power is 50%, then verify Feedwater Control is in Automatic and Start second Main Feedwater Pump in accordance with OP-003-033, Main Feedwater.

I //

9.1.65 When Reactor power exceeds 50%, th verify power escalation rate is within limits of Attachment 9.6, Fuel Preconditioning Guidelines.

9.1.66 Prior g exceeding 60% Reactor power, perform OP-904-010, Reactor Trip on Turbine Trip Channel Functional Test for all four PPS channels, ((not hi //

performed within the last 6 months.

CAUTION BOTH MAIN FEEDWATER PUMPS MUST BE OPERATING PRIOR TO PLACING REACTOR POWER CUTBACK IN SERVICE. [P.13902]

9.1.67 When Reactor power is 65%, then perform one of the following:

9.1 .67.1 If Reactor Power Cutback is to be placed in service, perform the following:

9.1.67.1.1 Verify both Main Feedwater Pumps are operating.

9.1.67.1.2 Align Reactor Power Cutback for service in accordance with OP-004-015, Reactor Power Cutback System.

9.1.67.2 [f Reactor Power Cutback is not to be placed in service, then perform the following:

9.1.67.2.1 Enable Loss of Turbine Loss of Load Trip functions in accordance with OP-004-01 5, Reactor Power Cutback System.

9.1.68 Prior to exceeding 70% reactor power, verify both ADV5 in AUTO with a setpoint of 990.0 psig to 992.0 psig as indicated by PMC PIDs A08223 and A08224. [T.S. 3.7.1.7]

9.1 .68.1 Document on the applicable Work Order for PMRQ 10845-01.

9.1.69 Prior exceeding 70% reactor power, verify Containment Temperature is in compliance with Technical Specification 3.6.1 .5.

OP-010-004 Revision 305 Attachment 9.1 (14 of 18) 32

(Initial/Date) 9.170 When Reactor power is 70%, i place Heater Drain Pumps in service in accordance with OP-003-034, Feed Heater Vents and Drains.

9.1.71 When Reactor Power is>70% power, jjjas soon as possible, but not to exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, verify RCS flow, as seen by each CPC channel, is <COLSS RCS flow rate in accordance with OP-903-00i, Technical Specification Surveillance Logs. [S.R. 4.3.1.1]

9.1 .72 Prior exceeding 80% power, verify the following agree with Secondary Calorimetric Power within limits specified in OP-903-001, Technical Specification Surveillance Log for power levels above 80%.

• CPC Calibrated Neutron Flux Power (CPC PID 171)

• CPC Thermal Power (CPC PID 177)

• Actual Excore Nuclear Power (OP-b DVM) 9.1.73 When Condensate Flow exceeds 21,000 gpm as indicated by PMC PID S02404, then verify OPEN CD-i 54, Gland Steam Condenser Bypass (PMC PID D02404).

9.1.74 Prior to exceeding 90% power, re-evaluate CEA Subgroups selected to drop on a Reactor Power Cutback event in accordance with OP-004-01 5, Reactor Power Cutback System.

NOTE The below control bands for Tc were selected to ensure that the Technical Specification limits for T are maintained in accordance with Technical Specification 3.2.6 while following TH reduction guidelines. Minor deviations to control bands are allowed at SM/CRS discretion.

9.1.75 At or above 92% power and at steady state conditions, control T in the following ranges:

• 0-30 EFPD 543 to 546°F

• 30-60 EFPD 543 to 545° F

• >6OEFPD 543to544°F 9.1.75.1 All other times, maintain T 536-549°F. ET.S. 3.2.6]

OP-010-004 Revision 305 Attachment 9.1 (15 of 18) 33

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A6 Review DNBR and LPD Limits Associated with OP-901-501, PMC or COLSS Malfunction Candidate:

Examiner:

JPM A6 JOB PERFORMANCE MEASURE DATA PAGE Task: Review DNBR and LPD Limits Associated with OP-901-501, PMC or COLSS Malfunction Task Standard: Candidate reviews completed OP-901-501 paperwork discovers errors on Attachments 2 and Attachment 3.

References:

OP-901-501, PMC or COLSS Malfunction Validation Time: 20 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A6 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

Instructions on candidates cue sheet Plant conditions are as follows:

100% power Both CEACs are operable 250 EFPD Containment temperature is 102 °F The Plant Monitoring Computer failed at 1855 on 10/5/2009.

You have entered OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction.

The BOP operator has completed Attachment 1, CPC DNBR Limit Calculation, , CPC LPD Limit Calculation , and Attachment 3, 15 Minute Log, of OP-901-501.

He has also informed you that the plant is not incompliance with Tech Spec 3.2.4 as a result of his calculations.

Review the attached paperwork and determine the necessary actions.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A6 TASK ELEMENT STANDARD Candidate discovers that all 4 CPC LPD limits were Corrects CPC LPD limits to 12.12, 11.91, 12.11, done with the calculation in step 5 vice step 4. and 13.41 for Channels A through D.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Candidate corrects the 4 CPC LPD limits on Each Channel LPD limit needs to be calculated to . the correct value.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Candidate discovers that the BOP operator filled in Channel A meets the limits of Tech Spec 3.2.4, No for DNBR within limits of Tech Spec 3.2.4 on using COLR figure 3/4-2-6.

any operable CPC Channel.

Comment: CRITICAL STEP END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A6 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

Plant conditions are as follows:

100% power Both CEACs are operable 250 EFPD Containment temperature is 100 °F The Plant Monitoring Computer failed at 1855 on 10/5/2009.

You have entered OP-901-501, PMC or Core Operating Limit Supervisory System Malfunction.

The BOP operator has completed Attachment 1, CPC DNBR Limit Calculation, , CPC LPD Limit Calculation , and Attachment 3, 15 Minute Log, of OP-901-501.

He has also informed you that the plant is not incompliance with Tech Spec 3.2.4 as a result of his calculations.

Review the attached paperwork and determine the necessary actions.

Waterford 3 Page 5 of 5 NRC Exam 2009

Off Normal Procedure OP-901-501 PMC or Core Operating Limit Supervisory System Malfunction Revision 011 E SUBSEQUENT OPERATOR ACTIONS 0

E GENERAL CAUTION CEA MOVEMENT SHOULD NOT BE ALLOWED FOR THE FIRST TWO HOURS AFTER COLSS BECOMES INOPERABLE BECAUSE THE ANALYSIS OF CPC/COLSS RATIO IS NOT LINEAR WHEN CEAS ARE MOVED.

PLACEKEEPER START DONE N/A 1 If CEA movement is necessary then within 15 minutes El Continuous verify LPD and DNBR are within the limits of Technical Specifications 3 2 1 and 3 2 4 on any operable CPC channel. [P-20634, P-13445, P-13446]

1.1 LPD or DNBR are outside the limits of 3.2.1 or 3.2.4, perform the following:

Li LI Li

• Within the next two hours, restore CPC LPD and DNBR to within limits of T.S. 3.2.1 and 3.2.4, or

• Reduce Thermal Power to less than or equal to 20 % of Rated Thermal Power within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 2 H entered due to a dropped or misaligned CEA, initiate a power reduction to comply with Technical LI LI LI Specification 3 1 3 1 jçj perform OP-901-102 CEA or CEDMS Malfunction concurrently with this procedure 3 If entered due to a PMC failure then g Loss of COLSS/PMC.

section E 1 LI LI

4. H entered due to COLSS failure only, then g j section E Loss of COLSS (PMC Available).

2 LI LI 6

Off Normal Procedure OP-901-501 PMC or Core Operating Limit Supervisory System Malfunction Revision 01 1 1

E Loss OF COLSS/PMC NOTE (1) CPC DNBR and LPD values are checked every 15 minutes to monitor for further degradation. An immediate power reduction is required if CPC DNBR is < CPC LPD is > the calculated limits on jjy operable CPC channel. LPD and DNBR values are acceptable if within limits of Technical Specifications 3.21 (LPD) and 3.2.4 (DNBR) on jy operable CPC channel, [P-20634]

(2> When a CEA drops into the core, the targeted CPC will trip. Recording data for the targeted CPC is required and should not be accomplished due to the possibility of improperly evaluating limits on the Attachments.

PLACEKEEPER START DONE N/A

1. K >20 % Reactor Power, perform the following:

1.1 Inform Shift Technical Advisor (STA) that his assistance is required in manually performing COLSS related LCO calculations.

1.2 iS a power reduction is not in progress, and no CEA movement has occurred, then within 15 minutes of LI LI loss of COLSS, perform the following: [P-20634]

1.2.1 Calculate CPC DNBR Limit on operable CPC U El channels by performing Attachment 1, CPC DNBR Limit Calculation, and document limits on Attachment 3, 15 Minute Log. [P-13446]

1.2.2 Calculate CPC LPD Limit on operable CPC channels by performing Attachment 2, CPC LPD Limit Calculation, and document limits on Attachment 3. [P-13445]

Contfnuous 1.2.3 Perform Attachment 3 to verify LPD, DNBR, U and ASI within limits on operable CPC channels, and every 15 minutes thereafter.

[P-13445, P.43446, P461]

7

Off Normal Procedure OP-901-501 PMC or Core Operating Limit Supervisory System Malfunction Revision 011 1

E Loss OF COLSS/PMC (coNTD)

PLACEKEEPER START DONE N/A 1 .3 jf a power reduction is in progress, CEA movement has occurred, ii2 verify the following on operable CPC channels within 15 minutes every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter: [P-20634]

• LPD within limits of Technical Specification 3.2.1 [P-13445]

• DNBR within limits of Technical Specification 3.2.4 [P-13446]

• ASI within limits of Technical Specification 3.2.7

[P-461]

1.4 If within two hours of COLSS inoperable LI DNBR drops below CPC DNBR Limit as calculated CPC El LI on Attachment 1, fi immediately commence a power reduction to restore CPC DNBR to limits of T.S. 3.2.4.

1.5 If within two hours of COLSS inoperable LI LPD exceeds the CPC LPD Limit as calculated on CPC El El Attachment 2, then immediately commence a power reduction to restore CPC LPD to limits of T.S. 3.2.1.

1.6 If COLSS is not restored within two hours and LPD LI or DNBR are outside the limits of Technical El El Specifications 3.2.1 3.2.4 as applicable, then perform the following:

• Within the next two hours, restore CPC LPD and DNBR to within limits of T.S. 3.2.1 and 3.2.4, or

• Reduce Thermal Power to less than or equal to 20 % of Rated Thermal Power within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, 8

Off Normal Procedure OP-901-501 PMC or Core Operating Limit Supervisory System Malfunction Revision 011 Page 1 of 1 ATrA0HMENT 1: CPC DNBR LIMIT CALcuLPm0N

1. Record CPC DNBR in column (1) and CPC PHICAL in column (3) using data from CPCs.

2. Record COLSS DNBR POL in column (2) using last valid value for COLSS DNBR POL. PMC PID C24560 as found on dedicated trend pen recorder.

3. Perform CPC DNBR Limit calculation for each CPC Channel with data from applicable column using the formula shown below. Record results in column (4).

[(2) - (3)] X 0.03 (4)

4. i_f available DNBR margin > 0.15, then perform the following calculation. Record results in column (5).

(1)-0.1 =(5)

5. if available DNBR margin 0.15, then perform the following calculation. Record results in column (5).

(1) - (4) = (5)

Available DNBR Limit (1) (2) (3) (4) (5)

CPC CPC DNBR COLSS CPC PHICAL Available CPC DNBR Channel PID 406 DNBR POL PID 171 Margin Limit A

2. 17 ID.7 O.Zf 2.07 8

2.13 (o.S o.tc C

2i9 f9 2.o7 D

11, Ioo. o.13 o3 Performed Date/Time /?/p is Verified Date/Time /?i3 SM/CRS Date/Time Transmit with OP-903-OO1, Technical Specification Surveillance Logs.

17

Off Normal Procedure OP-901-501 PMC or Core Operating Limit Supervisory System Malfunction Revision 011 Page 1 of 1 ATTAcHMENT 2: CPC LPD LIMIT CALcu1moN

1. Record CPC LPD in column (1) and CPC PHICAL in column (3) using data from CPCs

2. Record COLSS KW/FT POL (2) using last valid value for COLSS KW/FT POL, PMC PID C24561 as found on dedicated trend pen recorder.

3. Perform CPC KW/FT Limit calculation for each CPC Channel with data from applicable column using the formula shown below. Record results in column (4).

[(2)_(3)](4) 18

4. J available LPD margin > 0.4, then perform the following calculation. Record results in column (5).

(1) + 0.25 = (5)

5. II available LPD margin 0.4, il:n perform the following calculation. Record results in column (5).

(1) + (4) (5)

AIkI rvQflljtu aL..JcriI kA,,r,v. I rvI IVIQJ jfl I L11 I III.

(1) (2) (3) (4) (5)

CPC CPC LPD COLSS CPC PHICAL Available CPC LPD Channel PID 179 KW/FT POL PID 171 Margin Limit A /13 o.s B

1(3 35 C

1i3 D

113 1X.

Performed Date/Time________

Verified 4d Date/Time_________

SM/CRS Date/Time__________

Transmit with OP-903-OO1, Technical Specification Surveillance Logs.

18

Ott Normal Procedure OP-901 -501 PMC or Core Operating Limit Supervisory System Malfunction Revision 011 Page 1 of 1 ATTACHMENT 3: 15 MINuTE LOG

[P-461, P-13445, P.13446, P-206341 TIME: 113o GPO Channel A Limits LPD PID 179 (I. 7 (I 7 LPD= 12.5S DNBRPID4O6 2-.17 2.i DNBR = 2.o7 ASI PID 268 0. c2-5 O.ol..5 GPO Channel B Limits LPD PID 179 1/. (,4 Ii. (,(

LPD= iZ.3S DNBRPID4O6 1.13 2.17_

DNBR= 2.o3 ASIPID268 Oct o.ot GPO Channel 0 Limits LPD PID 179 Ij. 11. 5 LPD= IZ.55 DNBRPID4O6 2-.o 2.o4 DNBR 2- °7 ASI PlO 268

=

o.o24 0. OtLl GPO Channel D Limits LPD PID 179 f3.I 13.

LPD= j33 DNBRPID4O6 2.o7 DNBR= Z.o3 ASIPID268 .&.D1L 0.0(0 GPO LPD < GPO LPD LIMIT ON OPERABLE GPO CHANNELS (YIN)

LPD WITHIN LIMITS OF T.S. 3.2.1 ON ANY OPERABLE CPC CHANNEL (Y/N)

• GPO DNBR > GPO DNBR LIMIT ON J OPERABLE GPO CHANNELS (Y/N)

DNBR WITHIN LIMITS OF T.S. 3.2.4 ON ANY OPERABLE OPC CHANNEL (YIN) **

ASI AGGEPTABLE (Y/N)

PERFORMED BY (INITIALS) (7.4 IF LPD is not within the limits of T.S. 311 on aerable GPO Channel, then enter Technical Specification 3.2.1.

If complying with the applicable action of Technical Specification 3.6.1 .5, then reduce the OPC LPD Limit of T.S. 3.2.1 as follows:

Average Containment Temperature Action to take to comply with Technical (TAVG CNTMT) as determined by OP-903-O01. Specification 3.6.1.5.

86 TAVG CNTMT < 96 Reduce CPC LPD Limit by 0.2 kW/ft 76 TAvGGNTMT < 86 Reduce CPC LPD Limit by 0.4 kW/ft

[f DNBR is not within the limits of T.S. 3.2.4 on operable OPC Channel, then enter Technical Specification 3.2.4.

ASI is acceptable IF wit in limits of Technical Specification 3.2.7.

Performed by:

4) nature) (Oat)

SM/CRS Review:

I (Signature) (Date/Time)

Transmit with OP-903-OO1, Technical Specification Surveillance Logs.

19

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A7 Review Surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test Candidate:

Examiner:

JPM A7 JOB PERFORMANCE MEASURE DATA PAGE Task: Review Surveillance OP-903-008, Reactor Coolant System Isolation Leakage Test Task Standard: Candidate reviews completed OP-903-008 paperwork and discovers errors on Attachments 10.11.

References:

OP-903-008, Reactor Coolant System Isolation Leakage Test Technical Specifications Tech Spec 3.4.5.2 Validation Time: 30 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A7 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-008, Attachment 10.11 and 10.16 Tech Specs READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

INITIAL CONDITIONS:

The plant is in Mode 4.

OP-903-008 is in progress.

The data has been recorded for SI-329 A and ready for review.

INITIATING CUES:

Review OP-903-008 and verify acceptance criteria.

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A7 TASK ELEMENT STANDARD The candidate must identify that the calculation of Extrapolation Ratio Coefficient, CE, is incorrect.

This error, after corrected, results in the Corrected Candidate reviews OP-903-008, Attachment 10.11. Leak Rate RC going from 0.512 gpm to 1.02 gpm.

The correct leak rate does not meet the acceptance criteria. Tech Spec 3.4.5.2 should be entered.

Comment: CRITICAL STEP See attached key.

END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A7 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is in Mode 4.

OP-903-008 is in progress.

The data has been recorded for SI-329 A and ready for review.

INITIATING CUES:

Review OP-903-008 and verify acceptance criteria.

Waterford 3 Page 5 of 5 NRC Exam 2009

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1 0. 1 6 AccEPTANcE CRITERIA AND TEST AccEPTANcE INITIALS 1 0.1 6.1 For all cneck valves tested. test is acceptable if leakage is less than or ecual to 1 GPM.

10.162 For Power Operated valves tested. test is acceptable if leakage is:

1) Less than or equal to 1 GPM OR 2l Greater than 1 GPM but less than or equal to 5 GPM. f the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between the previous measured leakage rate and the maximum permissible rate of 5 GPM by 50°c or more.

10,16,3 Send a copy of Attachment 10.12, Result Summary Table, to the Programs 1ST Coordinator.

REMARKS:

Performed by: Verified by:

(RO or SRO Signaturm (Date/Time) (RO or SRO Signature (Date/Time)

(AO Signature) (Date/Time) (AO Signature) (Date/Time)

(STA Signature) (Date/Time) (STA Signature) (Date Time)

All data revewec and test accepted by:

(SMiCRS SIcIna1ure (DateTime)

OP903008 Revision 7 [LAST PAGE] Attachment 1016 (1 of 1) 119

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.5.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 75 gallons per day primary to secondary leakage through any one steam generator (SG),

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

e. 1 gpm leakage at a Reactor Coolant System pressure of 2250 +/- 20 psia from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

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

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

c. With any Reactor Coolant System pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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

4.4.5.2.1 Reactor Coolant System leakages, except for primary to secondary leakage, shall be demonstrated to be within each of the above limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

4.4.5.2.2 Primary to secondary leakage shall be verified to be < 75 gallons per day through any one SG at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

WATERFORD - UNIT 3 3/4 4-18 AMENDMENT NO. 197, 199, 204

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.2.3 Each Reactor Coolant System pressure isolation valve specified in Table 3.4-1, Section A and Section B, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

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

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

d. Following valve actuation for valves in Section B due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

4.4.5.2.4 Each Reactor Coolant System pressure isolation valve power-operated valve specified in Table 3.4-1, Section C, shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months, and

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

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

WATERFORD - UNIT 3 3/4 4-19 AMENDMENT NO. 96, 197, 204

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A8 Plan Work and Assign Workers Based on Dose Rates and Shielding Candidate:

Examiner:

JPM A8 JOB PERFORMANCE MEASURE DATA PAGE Task: Plan Work and Assign Workers Based on Dose Rates and Shielding Task Standard: Candidate calculates dose with and without shielding, and with 1 or 2 workers, and directs job to achieve the lowest dose.

References:

None Validation Time: 30 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A8 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

Refuel 16 is in progress and Safety Injection Train A is being aligned from its Shutdown Cooling alignment to its Safety Injection alignment. You are the Work Management Center SRO and have been assigned to coordinate the venting of Safety Injection Train A.

The dose rates in Safeguards Room A are 400 mrem/hour unshielded.

Installing shielding will reduce the dose rate to 110 mrem/hour.

It will take 2 workers 25 minutes to install the shielding (25 minutes each worker).

It will take 1 person 90 minutes to complete the venting if he works alone.

It will take 2 people 36 minutes to complete the venting (36 minutes each worker).

How will you direct the execution of the Safety Injection System venting to allow the least amount of total worker dose?

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A8 TASK ELEMENT STANDARD Candidate correctly concludes that installing the Calculate work assignment to provide for the shielding and using 2 workers results in the lowest lowest total dose.

total dose.

Comment: CRITICAL STEP Unshielded, the job will result in 600 mrem to 1 worker or 480 mrem total dose for 2 workers.

Installing the shielding will result in 333 mrem total dose.

Performing the work with 1 worker will result in 165 mrem + 333 mrem = total dose of 498 mrem Performing the job with 2 workers will result in 132 mrem + 333 mrem = total dose of 465 mrem Perform the job by installing the shielding and completing the job with both operators.

END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A8 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

Refuel 16 is in progress and Safety Injection Train A is being aligned from its Shutdown Cooling alignment to its Safety Injection alignment. You are the Work Management Center SRO and have been assigned to coordinate the venting of Safety Injection Train A.

The dose rates in Safeguards Room A are 400 mrem/hour unshielded.

Installing shielding will reduce the dose rate to 110 mrem/hour.

It will take 2 workers 25 minutes to install the shielding (25 minutes each worker).

It will take 1 person 90 minutes to complete the venting if he works alone.

It will take 2 people 36 minutes to complete the venting (36 minutes each worker).

How will you direct the execution of the Safety Injection System venting to allow the least amount of total worker dose?

Waterford 3 Page 5 of 5 NRC Exam 2009

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE A9 Classify Emergency Plan Entry Level Candidate:

Examiner:

JPM A9 JOB PERFORMANCE MEASURE DATA PAGE Task: Classify Emergency Plan Entry Level Task Standard: Candidate determines correct E Plan Emergency Action Level..

References:

EP-001-001 Recognition and Classification of Emergency Conditions Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Waterford 3 Page 2 of 5 NRC Exam 2009

JPM A9 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

EP-001-001, Recognition and Classification of Emergency Conditions READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Ensure you indicate to me when you understand your assigned task. To indicate that you have completed your assigned task return the handout sheet I provided you.

Instructions on candidates cue sheet Plant status is as follows:

The plant is at 100%.

EDG A is tagged out for an oil change.

You are the Emergency Coordinator.

At 1045, the plant tripped due to a loss of offsite power.

Emergency Diesel Generator B started, but the output breaker will not close. The EDG B Output Breaker has been identified as the problem.

At 1100, PME personnel commenced efforts to replace the Emergency Diesel Generator B output breaker with the Emergency Diesel Generator A output breaker.

1. Based on these conditions, determine which Emergency Plan Action Level should be entered, if any.

2. How does your classification changes, if at all, if during this event, RCS pressure was 1800 PSIA and dropping, with indications of a Pressurizer Relief valve lifted?

Waterford 3 Page 3 of 5 NRC Exam 2009

JPM A9 TASK ELEMENT STANDARD Enter a Site Area Emergency based on IC SS1, Determine the correct EAL for plant conditions. Loss of all off site power and loss of all onsite AC power to essential busses.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Upgrade to SG1, Prolonged loss of all offsite Determine upgrade based on changing conditions power and prolonged loss of all onsite AC power to listed as part 2.

essential busses.

Comment: CRITICAL STEP END OF TASK Waterford 3 Page 4 of 5 NRC Exam 2009

JPM A9 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

Plant status is as follows:

The plant is at 100%.

EDG A is tagged out for an oil change.

You are the Emergency Coordinator.

At 1045, the plant tripped due to a loss of offsite power.

Emergency Diesel Generator B started, but the output breaker will not close. The EDG B Output Breaker has been identified as the problem.

At 1100, PME personnel commenced efforts to replace the Emergency Diesel Generator B output breaker with the Emergency Diesel Generator A output breaker.

3. Based on these conditions, determine which Emergency Plan Action Level should be entered.

4. How does your classification changes, if at all, if during this event, RCS pressure was 1800 PSIA and dropping, with indications of a Pressurizer Relief valve lifted?

Provide answers to part 1 and part 2 on this sheet.

Waterford 3 Page 5 of 5 NRC Exam 2009

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT FISSION PRODUCT BARRIER DEGRADATION EP-001-001 Revision 022 13 Attachment 7.1 (8 of 16)

Fuel Clad Barrier EALs RCS Barrier EALs Containment Barrier EALs EP-001-001 Revision 022 14 Attachment 7.1 (9 of 16)

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTION Loss of AC Power Failure of Reactor Protection System EP-001-001 Revision 022 19 Attachment 7.1 (14 of 16)

SYSTEM MALFUNCTION SS1 Initiating Condition -- SITE AREA EMERGENCY Loss of all offsite power and loss of all onsite AC power to essential busses.

Operating Mode Applicability: Power Operations (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level(s):

1. Loss of power to all unit auxiliary and startup transformers AND Failure of the and emergency diesel generators to supply power to emergency busses AND Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including Shutdown Cooling, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency.

Escalation to General Emergency is via Fission Product Barrier Degradation (F) or SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power."

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to essential busses. Even though an essential bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus, then the bus should not be considered operable for this IC. If this bus was the only energized bus, then a Site Area Emergency in accordance with SS1 should be declared.

EP-001-001 Revision 022 139 Attachment 7.2 (118 of 128)

SYSTEM MALFUNCTION SS1 Loss of all offsite power varies depending on the plant mode and source transformers.

If the unit is back feeding via the unit Auxiliary Transformers and offsite power is lost in conjunction with loss of onsite AC power from the emergency diesel generators, then declaration of a Site Area Emergency is warranted.

When temporary emergency diesels (TEDs) are used to supplement onsite AC power for essential busses in the event diesels are lost, they are credited in this EAL. The EAL condition does not apply unless the TED also failed, provided the TED powers necessary loads as described above.

EP-001-001 Revision 022 140 Attachment 7.2 (119 of 128)

SYSTEM MALFUNCTION SG1 Initiating Condition -- GENERAL EMERGENCY Prolonged loss of all offsite power and prolonged loss of all onsite AC power to essential busses.

Operating Mode Applicability: Power Operations (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level(s):

1. Loss of power to all unit auxiliary and startup transformers.

AND Failure of both and emergency diesel generators to supply power to emergency busses.

AND Either of the following: (a or b)

a. Restoration of at least one emergency bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely OR

b. FA1 entry conditions met.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including Shutdown Cooling, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment.

This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory. The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore AC power is based on the site blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout.

EP-001-001 Revision 022 146 Attachment 7.2 (125 of 128)

SYSTEM MALFUNCTION SG1 Appropriate allowance for offsite emergency response, including evacuation of surrounding areas has been considered. Although this EAL may be viewed as redundant to the Fission Product Barrier Degradation (FG1) EALs, its inclusion is necessary to better assure timely recognition and emergency response.

When temporary emergency diesels (TEDs) are used to supplement onsite AC power for essential busses in the event diesels are lost, they are credited in this EAL.

In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Coordinator/EOF Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of Fission Product Barriers is imminent?

2. If there are no present indications of such core cooling degradation, then how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with particular emphasis on Emergency Coordinator/EOF Director judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers using the barrier indicators in section F of the EALs.

EP-001-001 Revision 022 147 Attachment 7.2 (126 of 128)

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE P1 Transfer EFW Pump Suctions to the Wet Cooling Tower after Condensate Storage Pool Depletion Candidate:

Examiner:

JPM P1 JOB PERFORMANCE MEASURE DATA PAGE Task: Transfer EFW Pump Suctions to the Wet Cooling Tower after Condensate Storage Pool Depletion Task Standard: Candidate aligns EFW Pump suction to the Auxiliary Component Cooling Water system.

References:

OP-902-009, Standard Appendices Appendix 10, Transferring EFW Pump Suction Validation Time: 10 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 6 NRC Exam 2009

JPM P1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-902-009, Appendix 10

==

Description:==

This task is performed on the -35 level. The candidate must align EFW Pumps Suction to Auxiliary Component Cooling Water. The procedure directs this to be accomplished on only 1 train. This will be performed on Train A. Step 1.2 is a 30 minute hold step.

The evaluator will tell the candidate that 30 minutes has elapsed at this point.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

A loss of Main Feedwater event is in progress.

Condensate Storage Pool Level is 25% and lowering.

INITIATING CUES:

The CRS directs you to transferring EFW Pump suction to Auxiliary Component Cooling Water on Train A in accordance with OP-902-009, Appendix 10.

Revision 0 Page 3 of 6 NRC Exam 2009

JPM P1 TASK ELEMENT STANDARD Verify Auxiliary Component Cooling Water Pump A This should be done by simulating a call to the operating. Control Room.

Comment: CRITICAL STEP Evaluator: Respond as the CRS, and reply that ACCW Pump A is running.

TASK ELEMENT STANDARD Close ACC 115A, Auxiliary Component Cooling Valve is closed.

Header A to Emergency Feedwater Drain.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Unlock and open the following valves:

ACC 116A, Auxiliary Component Cooling Header A to Emergency Feedwater Isolation Both valves are open.

ACC 114A, Auxiliary Component Cooling Header A Supply to EFW Header Isolation Comment: CRITICAL STEP Revision 0 Page 4 of 6 NRC Exam 2009

JPM P1 TASK ELEMENT STANDARD WHEN 30 minutes has elapsed, THEN close and lock the valves for the Train aligned in step 1:

Both valves are closed.

ACC 116A ACC 114A Comment: CRITICAL STEP Evaluator: Prompt the candidate that 30 minutes have elapsed when this step is reached.

TASK ELEMENT STANDARD Open Auxiliary Component Cooling Header to Emergency Feedwater Drain for the Train aligned Valve is open.

in step 1:

ACC 115A Comment: CRITICAL STEP Evaluator: Prompt the candidate that 30 minutes have elapsed when this step is reached.

END OF TASK Revision 0 Page 5 of 6 NRC Exam 2009

JPM P1 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

A loss of Main Feedwater event is in progress.

Condensate Storage Pool Level is 25% and lowering.

INITIATING CUES:

The CRS directs you to transferring EFW Pump suction to Auxiliary Component Cooling Water on Train A in accordance with OP-902-009, Appendix 10.

Revision 0 Page 6 of 6 NRC Exam 2009

WATERFORD 3 SES OP-902-009 Revision 301 Page 94 of 195 STANDARD APPENDICES Appendix 10 Page 1 of 4 Transferring EFW Pump Suction INSTRUCTIONS CONTINGENCY ACTIONS

---

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

CSP Indicated level will be lower than actual when drawing suction from the CSP.

CSP Indicated level will be higher than actual when drawing suction from the ACCW system. When EFW suction is drawn from the CSP, consideration should be given to reducing flow to less than 500 gpm to read CSP level.

---

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

Transfer of EFW Pump suction should be completed by a CSP level of 11% to prevent cavitation of EFW Pumps.

WATERFORD 3 SES OP-902-009 Revision 301 Page 95 of 195 STANDARD APPENDICES Appendix 10 Page 2 of 4 INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.1 Transfer Emergency Feedwater Pump suction to ONE side of the Auxiliary Component Cooling System as follows:

Train A

a. Verify Auxiliary Component Cooling Water Pump A operating.

b. Close ACC 115A, Auxiliary Component Cooling Header A to Emergency Feedwater Drain.

c. Unlock and open the following valves:

ACC 116A, Auxiliary Component Cooling Header A to Emergency Feedwater Isolation ACC 114A, Auxiliary Component Cooling Header A Supply to EFW Header Isolation (continue)

WATERFORD 3 SES OP-902-009 Revision 301 Page 96 of 195 STANDARD APPENDICES Appendix 10 Page 3 of 4 INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.1 (continued)

Train B

a. Verify Auxiliary Component Cooling Water Pump B operating.

b. Close ACC 115B, Auxiliary Component Cooling Header B to Emergency Feedwater Drain.

c. Unlock and open the following valves:

ACC 116B, Auxiliary Component Cooling Header B to Emergency Feedwater Isolation ACC 114B, Auxiliary Component Cooling Header B Supply to EFW Header Isolation

____ 1.2 WHEN 30 minutes has elapsed, THEN close and lock the valves for the Train aligned in step 1:

ACC 116A ACC 116B ACC 114A ACC 114B

WATERFORD 3 SES OP-902-009 Revision 301 Page 97 of 195 STANDARD APPENDICES Appendix 10 Page 4 of 4 INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.3 Open Auxiliary Component Cooling Header to Emergency Feedwater Drain for the Train aligned in step 1:

ACC 115A ACC 115B End of Appendix 10

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE P2 Reset emergency Diesel Generator A following an Overspeed Trip with a LOOP Candidate:

Examiner:

JPM P2 JOB PERFORMANCE MEASURE DATA PAGE Task: Reset emergency Diesel Generator A following an Overspeed Trip with a LOOP Task Standard: Candidate resets Emergency Diesel Generator A.

References:

OP-009-002, Emergency Diesel Generator Validation Time: 10 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 6 NRC Exam 2009

JPM P2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-009-002, Emergency Diesel Generator, section 8.8.

==

Description:==

This task is performed on the +21 level in Emergency Diesel Generator Room A. The candidate will simulate all actions in the EDG Room A. Manipulations 1 and 2 take place on the upper level of EDG A. The last manipulation takes place at the EDG A control panel.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The plant has experienced a loss of off site power.

Emergency Diesel Generator A tripped on overspeed.

INITIATING CUES:

The CRS has directed you to reset Emergency Diesel Generator A.

Revision 0 Page 3 of 6 NRC Exam 2009

JPM P2 TASK ELEMENT STANDARD (1) If the EDG was running in Emergency Mode and the signal for the EDG to Start still exists, then the EDG will automatically start when Steps 8.8.1 & 8.8.2 are completed. Step 8.8.3 should still be performed to prevent an EDG trip when the engine goes from Emergency Mode to None Test Mode during paralleling operations.

(2) Resetting the Combustion Air Intake Butterfly valve may take up to 30 seconds.

Comment:

Evaluator: This is a note at the start of the section. The cue states that there is a loss of off site power, so the candidate should conclude his actions will cause the EDG A to start. If asked, both EDG A Starting Air Receivers pressure are as they are currently indicated.

TASK ELEMENT STANDARD Reset the Turbocharger Butterfly Valve by performing one of the following:

Depress and hold the EDG A Combustion Air Overspeed Trip Reset, EGA-418 A, pushbutton on the Governor until the Combustion Air Intake Butterfly Valve is reset. (pushbutton is located below the overspeed trip Butterfly valve is reset.

plunger on the side of the Overspeed Trip Block)

Or Manually at the Combustion Air Intake Butterfly Valve.

Comment: CRITICAL STEP Evaluator: When the reset push-button is depressed, the linkage will move and the sound of air flowing will be heard. This reset can take as long as 30 seconds, as is prompted in the note. The button is on the upper level of EDG A.

Revision 0 Page 4 of 6 NRC Exam 2009

JPM P2 TASK ELEMENT STANDARD Reset the Fuel Oil Overspeed Trip by pushing in the plunger on the Plunger is pushed in.

Governor Overspeed Trip Block.

Comment: CRITICAL STEP Evaluator: The reset plunger is on the upper level of EDG A. After this is reset, EDG A will crank and start. If the candidate failed to accomplish the preceding step, then EDG A will not start.

TASK ELEMENT STANDARD Push the System Reset pushbutton on the Emergency Diesel Generator Reset button is pressed.

A Control Panel.

Comment: CRITICAL STEP END OF TASK Revision 0 Page 5 of 6 NRC Exam 2009

JPM P2 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant has experienced a loss of off site power.

Emergency Diesel Generator A tripped on overspeed.

INITIATING CUES:

The CRS has directed you to reset Emergency Diesel Generator A.

Revision 0 Page 6 of 6 NRC Exam 2009

System Operating Procedure OP-009-002 Emergency Diesel Generator Revision 310 8.8 RESETTING THE EMERGENCY DIESEL GENERATOR AFTER AN OVERSPEED TRIP NOTE (1) If the EDG was running in Emergency Mode and the signal for the EDG to Start still exists, then the EDG will automatically start when Steps 8.8.1 & 8.8.2 are completed.

Step 8.8.3 should still be performed to prevent an EDG trip when the engine goes from Emergency Mode to Test Mode during paralleling operations.

(2) Resetting the Combustion Air Intake Butterfly valve may take up to 30 seconds.

8.8.1 Reset the Turbocharger Butterfly Valve by performing one of the following:

Depress and hold the EG A(B) Combustion Air Overspeed Trip Reset, EGA-418A(B), pushbutton on the Governor until the Combustion Air Intake Butterfly Valve is reset. (pushbutton is located below the overspeed trip plunger on the side of the Overspeed Trip Block) or Manually at the Combustion Air Intake Butterfly Valve.

8.8.2 Reset the Fuel Oil Overspeed Trip by pushing in the plunger on the Governor Overspeed Trip Block.

NOTE Depressing the System Reset pushbutton before the EDG has come to a complete stop may cause the unit to attempt to crank. [CR-WF3-2005-00807]

8.8.3 When Emergency Diesel Generator A(B) has come to a complete stop, then push the System Reset pushbutton on the Emergency Diesel Generator A(B) Control Panel.

46

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE P3 Align High Pressure Safety Injection Pump AB for Performance of OP-903-030, Safety Injection Pump Operability Check Candidate:

Examiner:

JPM P3 JOB PERFORMANCE MEASURE DATA PAGE Task: Align High Pressure Safety Injection Pump AB for Performance of OP-903-030, Safety Injection Pump Operability Check Task Standard: During alignment of HPSI Pump AB, candidate recognizes that SI-208 A, HPSI Pump A Discharge Isolation, reach rod malfunctions and informs the Control Room.

References:

OP-903-030, Safety Injection Pump Operability Check OP-903-011, High Pressure Safety Injection Pump Pre-service Operability Check EN-OP-115, Conduct of Operations Validation Time: 15 minutes Time Critical: No Alternate Path: Yes Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Revision 0 Page 2 of 7 NRC Exam 2009

JPM P3 Signature Revision 0 Page 3 of 7 NRC Exam 2009

JPM P3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-011, High Pressure Safety Injection Pump Pre-service Operability Check

==

Description:==

This task is performed on the -15 level. The candidate will begin to align High Pressure Safety Injection Pump AB to replace HPSI Pump A. The fourth valve in the task, SI-208 A, a reach rod operated valve, will malfunction. The reach rod pin will bottom out prior to the valve reaching full closed position. The candidate should stop at this point and get shift management involved in the situation. The evaluator will ask the candidate what actions are necessary with the given indications. The candidate must be able to recall at least 3 items that are required to pass this task.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

HPSI Pump A is aligned in its normal alignment.

HPSI Pump AB operability surveillance, OP-903-030, is scheduled to be completed.

INITIATING CUES:

The CRS directs you to align HPSI Pump AB to replace A in accordance with OP-903-011, High Pressure Safety Injection Pump Pre-service Operability Check, section 7.1.

Steps 10.1.1 through 10.1.6 have already been completed.

Revision 0 Page 4 of 7 NRC Exam 2009

JPM P3 TASK ELEMENT STANDARD Complete the following valve lineup:

Lock open SI-202 A, HPSI Pump AB Suction From Candidate simulates opening SI-202 A.

HPSI A Isolation Comment: CRITICAL STEP Evaluator: Provide indication that SI-202 A is unlocked and travels open. If the candidate asks, the position indicating pin moves freely side to side with the valve open.

TASK ELEMENT STANDARD Complete the following valve lineup:

Lock open SI-212 A, HPSI pump AB Discharge To Candidate simulates opening SI-212 A.

HPSI A Isolation Comment: CRITICAL STEP Evaluator: Provide indication that SI-212 A is unlocked and travels open. If the candidate asks, the position indicating pin moves freely side to side with the valve open.

TASK ELEMENT STANDARD Complete the following valve lineup:

Lock closed SI-203 A, HPSI Pump A Suction Candidate simulates closing SI-203 A.

Isolation Comment: CRITICAL STEP Evaluator: Provide indication that SI-203 A is unlocked and travels closed. If the candidate asks, the position indicating pin moves freely side to side with the valve closed.

Revision 0 Page 5 of 7 NRC Exam 2009

JPM P3 TASK ELEMENT STANDARD Complete the following valve lineup:

Candidate discovers that the position indicating pin Lock closed SI-208 A, HPSI Pump A Discharge for SI-208 A bottoms out before the valve is closed.

Isolation Comment: CRITICAL STEP Evaluator: Provide indication that SI-208 A is unlocked and travels closed. Provide cue that the position indicating pin hits the bottom of its travel while the candidate is simulating closing SI-208 A.

TASK ELEMENT STANDARD From EN-OP-115, Conduct of Operations:

If a valve position indicator pin bottoms out or tops out during repositioning or a dial indicator is not indicating correctly, then perform the following:

1. Verify valve position by at least one of the following methods:

a. Local position indication Ask candidate requirements due to these b. Computer point indication indications. c. Verification of system parameters (flow, pressure, etc.)

2. Generate a Caution Tag for the valve and hang the tag on the remote hand wheel.

3. Generate a Work Request on the reach rod post indicator.

4. Classify the work request as a Work Around or Burden as appropriate on the valve position verification.

Comment: CRITICAL STEP Evaluator: Candidate must be able to recall at least 3 of these requirements.

END OF TASK Revision 0 Page 6 of 7 NRC Exam 2009

JPM P3 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

HPSI Pump A is aligned in its normal alignment.

HPSI Pump AB operability surveillance, OP-903-030, is scheduled to be completed.

INITIATING CUES:

The CRS directs you to align HPSI Pump AB to replace A in accordance with OP-903-011, High Pressure Safety Injection Pump Pre-service Operability Check, section 7.1.

Steps 10.1.1 through 10.1.6 have already been completed.

Revision 0 Page 7 of 7 NRC Exam 2009

Surveillance Procedure OP-903-011 High Pressure Safety Injection Pump Preservice Operability Check Revision 9 7.0 PROCEDURE 7.1 REPLACEMENT OF HPSI PUMP A BY HPSI PUMP AB 7.1.1. Obtain permission to perform Section 7.1 from SM/CRS and document on Attachment 10.1, HPSI Pump AB Replacing HPSI Pump A Data Sheet.

7.1.2 Perform replacement of HPSI Pump A by HPSI Pump AB in accordance with Attachment 10.1, HPSI Pump AB Replacing HPSI Pump A Data Sheet.

9

10.1 HPSI PUMP AB REPLACING HPSI PUMP A DATA SHEET 10.1.1 Obtain permission to replace HPSI Pump A with HPSI Pump AB. ________________/__________

SM/CRS (Signature) (Date/Time)

Performed Verified (initials) (Initials) 10.1.2 Verify AB Electrical Busses lined up from A Train. _________

10.1.3 Place HPSI Pump A control switch in the Off position. _________

10.1.4 Open HPSI Pump A Breaker knifeswitch. _________

10.1.5 Rack down HPSI Pump A Breaker, SI-EBKR-3A-4. _________

10.1.6 Verify PMI has isolated HPSI Pump AB Bearing Cooler CCW Low Flow Indicator Switch, CC-IFIS-7850C. _________

10.1.7 Complete the following valve lineup:

Valve Description Req. Position SI-202A HPSI Pump AB Suction From HPSI A locked open Isolation _________ _________

SI-212A HPSI pump AB Discharge To HPSI A locked open Isolation _________ _________

SI-203A HPSI Pump A Suction Isolation locked closed SI-208A HPSI Pump A Discharge Isolation locked closed CC-930A HPSI Pump A To AB CCW Supply locked open Cross-Connect _________ _________

CC-931A HPSI Pump AB To A CCW Supply locked open Cross-Connect _________ _________

CC-945A HPSI Pump AB To A CCW Return locked open Cross-Connect _________ _________

CC-944A HPSI Pump A To AB CCW Return locked open Cross-Connect _________ _________

CC-934A HPSI Pump A CCW Inlet Isolation locked closed CC-942A HPSI Pump A CCW Outlet Isolation locked closed SI-205AB HPSI Pump AB Min Flow To Recirc locked open Line A Stop Check _________ _________

SI-245 HPSI Pump AB Min Flow To Recirc locked closed Line B Stop Check _________ _________

OP-903-011 Revision 9 Attachment 10.1 (1 of 4) 16

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S1 ATC Immediate Operator Actions on 2 Dropped CEAs Candidate:

Examiner:

JPM S1 JOB PERFORMANCE MEASURE DATA PAGE Task: ATC Immediate Operator Actions on 2 Dropped CEAs Task Standard: Candidate trips reactor using 32 A and 32 B breakers.

References:

OP-901-102, CEA or CEDMCS Malfunction OP-902-000, Standard Post Trip Actions Validation Time: 5 minutes Time Critical: No Alternate Path: Yes Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 6 NRC Exam 2009

JPM S1 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

None

==

Description:==

Candidate will position himself as the ATC operator at CP-2. CEAs 3 and 14 will drop into the core. The candidate should notice the condition, announce the condition, and trip the reactor without direction. The normal reactor trip pushbuttons will not function.

The candidate should move to the first contingency and use the Diverse Reactor Trip pushbuttons. One of these buttons is faulted. The DRTS alarms will come in, but the CEA MG set load contactors will not open. The candidate should then move to the second contingency, and open both 32 Bus Feeder breakers, and reclose them 5 seconds later. The task should be stopped after the 32 Bus Feeder breakers have been reclosed.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The plant is at 100% power.

INITIATING CUES:

Respond to conditions as required.

Revision 0 Page 3 of 6 NRC Exam 2009

JPM S1 Simulator Operator: Initiate trigger 1 on direction for examiner.

TASK ELEMENT STANDARD Determines 2 CEAs have dropped, attempts to trip Pushes both reactor trip pushbuttons on CP-2.

the reactor from CP-2.

Comment: CRITICAL STEP Trip pushbuttons are faulted and Reactor Trip Circuit Breakers will not open.

TASK ELEMENT STANDARD Attempts to trip reactor using DRTS pushbuttons Pushes both DRTS pushbuttons on CP-2.

on CP-2.

Comment: CRITICAL STEP 1 DRTS pushbutton is faulted and CEA MG set load contactors will not open.

TASK ELEMENT STANDARD Open BOTH the following breakers for 5 seconds Opens SST A32 FEEDER and SST B32 FEEDER and close:

breakers for 5 seconds and then re-closes both

• SST A32 FEEDER breakers.

• SST B32 FEEDER Comment: CRITICAL STEP Evaluator: Inform candidate that the task is complete after both 32 Feeder breakers are re-closed.

END OF TASK Revision 0 Page 4 of 6 NRC Exam 2009

JPM S1 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is at 100% power. You are the ATC operator.

INITIATING CUES:

Monitor conditions as the ATC operator.

Revision 0 Page 5 of 6 NRC Exam 2009

JPM S1 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-97 Verify the following Malfunctions:

No Trigger o rp01a, RPS manual pushbutton a o rp01b, RPS manual pushbutton b o rp01c, RPS manual pushbutton c o rp01d, RPS manual pushbutton d Trigger 1:

o rd02a03, drop CEA 3 o rd02a14, drop CEA 14 Verify the following Overrides:

No Trigger o di-02a06s02-1, DRT pushbutton 1 of 2 Revision 0 Page 6 of 6 NRC Exam 2009

Off Normal Procedure OP-901-102 CEA or CEDMCS Malfunction Revision 5 D IMMEDIATE OPERATOR ACTIONS

1. If in Mode 1 and two or more Control Element Assemblies drop or are misaligned by >19 inches, then manually trip the Reactor and go to OP-902-000, Standard Post Trip Actions.

9

WATERFORD 3 SES OP-902-000 Revision 10 Page 5 of 15 STANDARD POST TRIP ACTIONS 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS Verify Reactivity Control

___1. Determine Reactivity Control acceptance criteria are met:

___ a. Check reactor power is dropping. a.1 Perform the following as necessary to insert CEAs:

1) Manually trip the reactor.

2) Manually initiate DIVERSE REACTOR TRIP.

3) Open BOTH the following breakers for 5 seconds and close:

SST A32 FEEDER SST B32 FEEDER

___ b. Check startup rate is negative.

___ c. Check less than TWO CEAs are c.1 Commence emergency boration.

NOT fully inserted.

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S2 Volume Control Tank Makeup Using the Dilute Makeup Mode Candidate:

Examiner:

JPM S2 JOB PERFORMANCE MEASURE DATA PAGE Task: Volume Control Tank Makeup Using the Dilute Makeup Mode Task Standard: Candidate adds Primary Makeup to the VCT and secures PMU flow to the VCT.

References:

OP-002-005, Chemical and Volume Control Validation Time: 15 minutes Time Critical: No Alternate Path: Yes Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 9 NRC Exam 2009

JPM S2 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-002-005, Chemical and Volume Control

==

Description:==

This task is performed at CP-4. The candidate will makeup to the Volume Control Tank in the dilute mode. Step 6.9.2 direct the operator to calculate volume of Primary Makeup water to be added. This calculation is reactor operator administrative JPM A1.

The fault in this JPM is on the automatic termination feature of the PMU Batch Counter. When the required amount of PMU has been added, the PMU will continue flowing. The candidate will be required to manually secure PMU flow.

This task is typically peer checked by another operator and observed by a SRO.

Inform the candidate that all required personnel are present.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The Plant is in Mode 4 RCS boron concentration is 1506 ppm INITIATING CUES:

The CRS directs you to dilute the RCS to 1500 ppm.

Revision 0 Page 3 of 9 NRC Exam 2009

JPM S2 TASK ELEMENT STANDARD Inform SM/CRS that this Section is being Notification is made.

performed.

Comment:

Evaluator: Acknowledge that PMU addition is about to commence.

TASK ELEMENT STANDARD At SM/CRS discretion, calculate volume of Primary Makeup water to be added on Attachment 11.7, Candidate completes calculation Calculation of Primary Makeup Water Volume for Direct Dilution or VCT Dilute Makeup Mode.

Comment:

Evaluator: Inform the candidate that the CRS has directed him to perform this step. Provide the blank 1.7 from JPM A1 After candidate has completed the calculation, direct him to add the amount of PMU he calculated, rounded to the nearest factor of 10 gallons.

TASK ELEMENT STANDARD Set Primary Makeup Water Batch Counter to 250 PMU counter is set to 25 (or volume calculated) gallons (or other volume calculated)

Comment: CRITICAL STEP Value of PMU entered into the counter is multiplied by 10.

TASK ELEMENT STANDARD Place Makeup Mode selector switch to DILUTE. Makeup Mode selector switch is in DILUTE Comment: CRITICAL STEP Revision 0 Page 4 of 9 NRC Exam 2009

JPM S2 TASK ELEMENT STANDARD Open VCT Makeup Valve, CVC-510. CVC-510 is open Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X is placed in Manual PMU-IFIC-0210X, in Manual.

Comment: CRITICAL STEP This step allows for controlling flow in manual or automatic. If asked, direct manual mode.

TASK ELEMENT STANDARD Adjust Primary Makeup Water Flow controller, PMU flow is > 10 gpm PMU-IFIC-0210X, output to > 10 GPM flow rate.

Comment: CRITICAL STEP Evaluator: The note for this step in the procedure informs the operator that the Dilution Flow Totalizer will not register below 5 GPM. The Dilution Flow Totalizer is most accurate at > 10 GPM.

TASK ELEMENT STANDARD Verify Primary Makeup Water Control Valve, PMU-PMU-144 is intermediate or open 144, Intermediate or Open.

Comment:

TASK ELEMENT STANDARD Observe Primary Makeup water flow rate for Observes flow indication proper indication.

Comment:

Revision 0 Page 5 of 9 NRC Exam 2009

JPM S2 TASK ELEMENT STANDARD Operate VCT Inlet/Bypass to Holdup Tanks, CVC-169 Control Switch to BMS/Auto positions as CVC-169 is operated as necessary necessary to maintain VCT pressure and level within normal operating bands.

Comment:

Based on VCT level in this IC, this step should be N/A. There would be no concern if the candidate chooses to operate CVC-169 to maintain VCT level and pressure constant.

TASK ELEMENT STANDARD When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary PMU flow is stopped Makeup Water Control Valve, PMU-144, Closed.

Comment: CRITICAL STEP This is the faulted step. When the counter counts to 0, flow will continue and the counter will begin to register negative numbers. The candidate can secure flow in 2 ways. 1 method would involve taking the controller for PMU-144 to 0 output. Another method would be to close CVC-510. Either method is acceptable.

TASK ELEMENT STANDARD Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X is in manual PMU-IFIC-0210X, in Manual.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify both Primary Makeup Water Flow controller, PMU-IFIC-0210X, output and setpoint Both are set to 0 potentiometer set to zero.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Close VCT Makeup Valve, CVC-510. CVC-510 is closed Revision 0 Page 6 of 9 NRC Exam 2009

JPM S2 Comment: CRITICAL STEP TASK ELEMENT STANDARD Place Makeup Mode selector switch to MANUAL. Makeup Mode Select switch is in manual Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify VCT Inlet/Bypass To Holdup Tanks, CVC-169, aligned to the VCT and control switch in CVC-169 is in auto AUTO.

Comment: CRITICAL STEP END OF TASK Revision 0 Page 7 of 9 NRC Exam 2009

JPM S2 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The Plant is in Mode 4 RCS boron concentration is 1506 ppm INITIATING CUES:

The CRS directs you to dilute the RCS to 1500 ppm.

Revision 0 Page 8 of 9 NRC Exam 2009

JPM S2 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-95 Verify the following Malfunctions:

cv35a for PMU batch counter Revision 0 Page 9 of 9 NRC Exam 2009

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 028 6.9 VCT MAKEUP USING THE DILUTE MAKEUP MODE (C)

CAUTION THIS SECTION AFFECTS REACTIVITY. THIS EVOLUTION SHOULD BE CROSS-CHECKED AND COMPLETED PRIOR TO LEAVING CP-4.

6.9.1 Inform SM/CRS that this Section is being performed.

NOTE When performing a Plant down power where final RCS Boron Concentration needs to be determined, the following Plant Data Book figure(s) will assist the Operator in determining the required RCS Boron PPM change.

1.2.1.1 Power Defect Vs Power Level 1.4.3.1 Inverse Boron Worth Vs. Tmod at BOC (<30 EFPD) 1.4.4.1 Inverse Boron Worth Vs. Tmod at Peak Boron (30 EFPD up to 170 EFPD) 1.4.5.1 Inverse Boron Worth Vs. Tmod at MOC (170 EFPD up to 340 EFPD) 1.4.6.1 Inverse Boron Worth Vs. Tmod at EOC ( 340 EFPD) 6.9.2 At SM/CRS discretion, calculate volume of Primary Makeup water to be added on Attachment 11.7, Calculation of Primary Makeup Water Volume for Direct Dilution or VCT Dilute Makeup Mode.

6.9.3 Set Primary Makeup Water Batch Counter to volume of Primary Makeup water desired.

6.9.4 Place Makeup Mode selector switch to DILUTE.

6.9.5 Open VCT Makeup Valve, CVC-510.

39

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 028 NOTE The Dilution Flow Totalizer will not register below 5 GPM. The Dilution Flow Totalizer is most accurate at > 10 GPM.

CAUTION DILUTION SHALL IMMEDIATELY BE STOPPED IF PRE-POWER DEPENDENT INSERTION LIMIT (H-9, CABINET H) ALARM IS INITIATED OR ANY UNEXPECTED REACTIVITY CHANGE OCCURS.

6.9.6 If manual control of Primary Makeup Water flow is desired, then perform the following:

6.9.6.1 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Manual.

6.9.6.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, output to

> 5 GPM flow rate.

6.9.7 If automatic control of Primary Makeup Water flow is desired, then perform the following:

6.9.7.1 Place Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Auto.

6.9.7.2 Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, setpoint potentiometer to > 5 GPM flow rate.

6.9.8 Verify Primary Makeup Water Control Valve, PMU-144, Intermediate or Open.

6.9.9 Observe Primary Makeup water flow rate for proper indication.

6.9.10 Operate VCT Inlet/Bypass to Holdup Tanks, CVC-169 Control Switch to BMS/Auto positions as necessary to maintain VCT pressure and level within normal operating bands.

6.9.11 When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU-144, Closed.

40

System Operating Procedure OP-002-005 Chemical and Volume Control Revision 028 NOTE Step 6.9.12 may be repeated as necessary to achieve desired total Primary Makeup Water addition for plant conditions.

6.9.12 If additional Primary Makeup Water addition is required and with SM/CRS permission, then perform the following:

6.9.12.1 Reset Primary Makeup Water Batch Counter.

6.9.12.2 Verify Primary Makeup Water Control Valve, PMU-144, Intermediate or Open.

6.9.12.3 Observe Primary Makeup water flow rate for proper indication.

6.9.12.4 When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU-144, Closed.

6.9.13 Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Manual.

6.9.14 Verify both Primary Makeup Water Flow controller, PMU-IFIC-0210X, output and setpoint potentiometer set to zero.

6.9.15 Close VCT Makeup Valve, CVC-510.

6.9.16 Place Makeup Mode selector switch to MANUAL.

6.9.17 Verify VCT Inlet/Bypass To Holdup Tanks, CVC-169, aligned to the VCT and control switch in AUTO.

41

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S3 BOP Operator Actions on RAS Actuation Candidate:

Examiner:

JPM S3 JOB PERFORMANCE MEASURE DATA PAGE Task: BOP Operator Actions on RAS Actuation Task Standard: Candidate completes actions required post RAS actuation.

References:

OP-902-002, Loss of Coolant Accident Recovery Validation Time: 15 minutes Time Critical: Yes Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 7 NRC Exam 2009

JPM S3 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-902-002, Loss of Coolant Accident Recovery

==

Description:==

This task is performed at CP-8 and CP4. The candidate performs required manipulations after Recirculation Actuation is initiated. After the simulator is taken out of freeze, the RAS will initiate in about 2 minutes.

This JPM has time critical elements. SI-120 A and B and SI-121 A and B must be closed within 2 minutes of RAS initiation to prevent recirculating SI sump water to the RWSP.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

A loss of coolant accident is in progress.

RAS initiation is imminent.

INITIATING CUES:

The CRS has directed you to perform step 42 of OP-902-002 after RAS initiates.

Revision 0 Page 3 of 7 NRC Exam 2009

JPM S3 Time of RAS:____________

TASK ELEMENT STANDARD Verifies RAS by either ROM indication at CP-7 or Verify the RAS is initiated. RAS Train A and B Logic Initiated annunciators on Panel K.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify that BOTH LPSI pumps are stopped. Verifies both LPSI Pumps at CP-8 are off.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify that ESF PUMPS SUCTION SI SUMP valves are open:

SI-602 A and B at CP-8 are open.

• SI 602A

• SI 602B Comment: CRITICAL STEP TASK ELEMENT STANDARD Close the SI PUMPS RECIRC ISOL VALVES within two minutes of receipt of RAS:

• SI 120A SI-120 A and B and SI-121 A and B at CP-8 are

• SI 120B closed within 2 minutes of RAS actuation.

• SI 121A

• SI 121B Comment: CRITICAL STEP Time completed:

TASK ELEMENT STANDARD Close the ESF PUMPS SUCTION RWSP:

• SI 106A SI-106 A and B at CP-8 are closed.

• SI 106B Comment: CRITICAL STEP Revision 0 Page 4 of 7 NRC Exam 2009

JPM S3 TASK ELEMENT STANDARD All Charging Pumps control switches at CP-4 are in Place ALL charging pumps in "OFF."

OFF.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Close CVC 209, CHARGING HEADER CVC-209 at CP-4 is closed.

ISOLATION.

Comment: CRITICAL STEP END OF TASK Revision 0 Page 5 of 7 NRC Exam 2009

JPM S3 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

A loss of coolant accident is in progress.

RAS initiation is imminent.

INITIATING CUES:

The CRS has directed you to perform step 42 of OP-902-002 after RAS initiates.

Revision 0 Page 6 of 7 NRC Exam 2009

JPM S3 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-96 There are no malfunctions or overrides necessary for this JPM.

Revision 0 Page 7 of 7 NRC Exam 2009

WATERFORD 3 SES OP-902-002 Revision 012 Page 31 of 67 LOSS OF COOLANT ACCIDENT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

---

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

SI 120A(B) and SI 121A(B), SI PUMPS RECIRC ISOL, should be closed within two minutes of receipt of RAS to prevent recirculating SI sump water to the RWSP.

RAS Initiation Criteria

• 42. IF the break is inside containment, AND RWSP level is less than 10%,

THEN:

a. Verify the RAS is initiated.

b. Verify that BOTH LPSI pumps are stopped.

c. Verify that ESF PUMPS SUCTION SI SUMP valves are open:

SI 602A SI 602B

d. Close the SI PUMPS RECIRC ISOL VALVES within two minutes of receipt of RAS:

SI 120A SI 120B SI 121A SI 121B (continue)

WATERFORD 3 SES OP-902-002 Revision 012 Page 32 of 67 LOSS OF COOLANT ACCIDENT RECOVERY INSTRUCTIONS CONTINGENCY ACTIONS

• 42. (continued)

e. Close the ESF PUMPS SUCTION RWSP:

SI 106A SI 106B

f. Place ALL charging pumps in "OFF."

g. Close CVC 209, CHARGING HEADER ISOLATION.

SI-602 Override

• 43. IF in the opinion of the Emergency Coordinator, closing SI 602A(B) Safety Injection System Sump Isolation valve to stop Emergency Core Cooling System leakage is in the best interest of protecting the public health and safety, THEN REFER TO Appendix 29, "SI-602 Override" and isolate SI 602A(B).

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S4 Place Shutdown Cooling Train A in Service Candidate:

Examiner:

JPM S4 JOB PERFORMANCE MEASURE DATA PAGE Task: Place Shutdown Cooling Train A in Service Task Standard: Candidate places Shutdown Cooling Train A in service.

References:

OP-009-005, Shutdown Cooling OP-901-131, Shutdown Cooling Malfunction Validation Time: 30 minutes Time Critical: No Alternate Path: Yes Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 8 NRC Exam 2009

JPM S4 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-009-005

==

Description:==

This task is performed at CP-8. The candidate must place Shutdown Cooling Train A in service. The fault in this task is that SI-405 A, RC Loop 2 SDC Suction Inside Containment Isol, will fail closed, requiring the candidate to secure Low Pressure Safety Injection Pump A. The task can be stopped after LPSI Pump A is secured.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The plant is in Mode 4 RCS temperature is 280 °F RCS pressure is 340 PSIA Shutdown Cooling Train A has been placed in Standby in accordance with OP-009-005, Section 5.3.

INITIATING CUES:

The CRS has directed you to place Shutdown Cooling Train A in service.

Revision 0 Page 3 of 8 NRC Exam 2009

JPM S4 TASK ELEMENT STANDARD Verify Shutdown Cooling Train A has been aligned to Standby condition in accordance with Section Cue sheet lists this as complete.

5.3, Alignment of Shutdown Cooling Train A to Standby Condition.

Comment:

TASK ELEMENT STANDARD Verify sufficient number of Dry Cooling Tower Fans running to accept increased heat load on CCW Any fans started must be started in FAST.

System.

Comment: Candidate may ask the CRS how man fans he wants running. If asked, respond to keep the fans in automatic and verify they cycle on as required. Any number of fans started is acceptable.

TASK ELEMENT STANDARD Place Shutdown HX A CCW Flow Control, CC-963 A is opened.

CC-963A, control switch to Open.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Unlock and Open RC Loop 2 SDC Suction Outside SI-407 A is opened.

Containment Isol, SI-407A.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Notify Radiation Protection Department that Shutdown Cooling Train A is being placed in Call is made.

service.

Comment:

Revision 0 Page 4 of 8 NRC Exam 2009

JPM S4 TASK ELEMENT STANDARD Start LPSI Pump A. LPSI Pump A is started.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Raise Shutdown Cooling flow by Manually adjusting LPSI Header Flow controller 2A/2B, SI-IFIC-0307, output until Shutdown Cooling Header Flow is raised to 4100 gpm.

A Flow indicates 4100 GPM, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Adjust LPSI Header Flow Controller 2A/2B, SI-Setpoint potentiometer is set to 73%, and controller IFIC-0307, setpoint potentiometer to 73%, and is placed in AUTO place controller to AUTO.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify LPSI Header Flow Controller 2A/2B, SI-IFIC-0307, is maintaining 4100 GPM Shutdown Cooling Flow is verified.

Header A flow, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

Comment:

TASK ELEMENT STANDARD At SM/CRS discretion, direct Chemistry Department to sample Shutdown Cooling Train A None for boron concentration.

Comment:

Evaluator: When requested provide information to candidate that all required Chemistry requirements are met.

Revision 0 Page 5 of 8 NRC Exam 2009

JPM S4 TASK ELEMENT STANDARD Open the following valves:

SI-139A LPSI Header to RC Loop 2A Flow Control SI-139 A and SI-138 A are open.

SI-138A LPSI Header to RC Loop 2B Flow Control Comment: CRITICAL STEP TASK ELEMENT STANDARD Throttle Closed RC Loop 2 Shdn Cooling Warmup, SI-135A, until one of the following is within 100°F of Shutdown Cooling Train A temperature, as indicated by LPSI Pump A Discharge Header Temperature Indicator, SI-ITI-0351X:

Temperature is within 100 °F Hot Leg 2 temperature, as indicated by RC Loop 2 Hot Leg Temperature Indicator, RC-ITI-0122-HA or Hot Leg 1 temperature, as indicated by RC Loop 1 Hot Leg Temperature Indicator, RC-ITI-0112-HB Comment:

TASK ELEMENT STANDARD Close RC Loop 2 Shdn Cooling Warmup, SI-135 A is closed SI-135 A.

Comment: CRITICAL STEP Simulator Operator: Initiate Trigger 1.

TASK ELEMENT STANDARD Secure LPSI Pump A LPSI Pump A is off.

Comment: CRITICAL STEP END OF TASK Revision 0 Page 6 of 8 NRC Exam 2009

JPM S4 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is in Mode 4 RCS temperature is 280 °F RCS pressure is 340 PSIA Shutdown Cooling Train A has been placed in Standby in accordance with OP-009-005, Section 5.3.

INITIATING CUES:

The CRS has directed you to place Shutdown Cooling Train A in service.

Revision 0 Page 7 of 8 NRC Exam 2009

JPM S4 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-95 Verify the following Malfunctions:

Si23a for SI-405 A Revision 0 Page 8 of 8 NRC Exam 2009

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 6.0 NORMAL OPERATIONS 6.1 PLACING SHUTDOWN COOLING TRAIN A IN SERVICE NOTE The Shutdown Cooling Train placed in service should be on the Protected Train.

CAUTION IF A DESIGN BASIS TORNADO EVENT OCCURS, THEN SHUTDOWN COOLING (SDC)

SHOULD NOT BE INITIATED UNTIL 23 HOURS POST-EVENT TO ENSURE CCW HEAT EXCHANGER (CCW HX) OUTLET CCW TEMPERATURE REMAINS 130 F. WHEN SDC IS INITIATED FOR THIS EVENT, THEN MAINTAIN CCW HX OUTLET TEMPERATURE 130 F.

6.1.1 Verify Shutdown Cooling Train A has been aligned to Standby condition in accordance with Section 5.3, Alignment of Shutdown Cooling Train A to Standby Condition.

6.1.2 Verify sufficient number of Dry Cooling Tower Fans running to accept increased heat load on CCW System.

CAUTION (1) A DESIGN BASIS TORNADO EVENT OCCURS, THEN THE CCW HEAT EXCHANGER OUTLET CCW TEMPERATURE SHOULD BE MAINTAINED 130 F.

THE TEMPERATURE OF THE CCW RETURN LINE FROM SHUTDOWN COOLING HEAT EXCHANGER A CAN BE MONITORED ON PMC PID A43806.

(2) CC-963A IS REQUIRED TO BE MAINTAINED OPEN WHILE IN MODE 4 TO PRESERVE THE DESIGN TEMPERATURE BASIS OF PIPING AND ASSOCIATED COMPONENTS AT THE CCW OUTLET OF SHUTDOWN COOLING HEAT EXCHANGER A. WITH CC-963A OPEN, FLOW THROUGH SHUTDOWN COOLING HEAT EXCHANGER A WILL BE MAINTAINED ABOVE 2305 GPM. [EC-738]

6.1.3 Place Shutdown HX A CCW Flow Control, CC-963A, control switch to Open.

18

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 CAUTION (1) THE FOLLOWING REACTOR COOLANT SYSTEM LIMITS SHALL BE MET FOR SHUTDOWN COOLING ENTRY:

RCS TEMPERATURE LIMIT: <350 F RCS PRESSURE LIMIT: <392 PSIA (2) IF CONTAINMENT SPRAY HEADER A ISOLATION, CS-125A, IS OPEN WHILE SHUTDOWN COOLING TRAIN A IS OPERATING, THEN CONTAINMENT SPRAY A RISER MAY FILL AND POSSIBLY SPRAY WATER INTO CONTAINMENT, DUE TO LEAKAGE PAST CONTAINMENT SPRAY PUMP A DISCHARGE STOP CHECK, CS-117A.

6.1.4 Unlock and Open RC Loop 2 SDC Suction Outside Containment Isol, SI-407A.

6.1.5 Notify Radiation Protection Department that Shutdown Cooling Train A is being placed in service.

6.1.6 Start LPSI Pump A.

6.1.7 Raise Shutdown Cooling flow by Manually adjusting LPSI Header Flow controller 2A/2B, SI-IFIC-0307, output until Shutdown Cooling Header A Flow indicates 4100 GPM, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

6.1.8 Adjust LPSI Header Flow Controller 2A/2B, SI-IFIC-0307, setpoint potentiometer to 73%, and place controller to AUTO.

6.1.9 Verify LPSI Header Flow Controller 2A/2B, SI-IFIC-0307, is maintaining 4100 GPM Shutdown Cooling Header A flow, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

NOTE If a sample was drawn prior to shutdown and no interim shutdown has occurred where SDC was placed in service and boron concentration could have been reduced, then sampling is not required.

6.1.10 At SM/CRS discretion, direct Chemistry Department to sample Shutdown Cooling Train A for boron concentration.

6.1.10.1 When Chemical Analysis results indicate that Shutdown Cooling Train A boron concentration is greater than Reactor Coolant boron concentration or 2050 PPM (required for Mode 6), then proceed to next step.

19

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 NOTE Shutdown Cooling Train A requires one operable Low Pressure Safety Injection Flow Control Valve for the train to be operable.

CAUTION THE REACTOR COOLANT SYSTEM SHALL NOT EXCEED THE 100 F PER HOUR COOLDOWN RATE OF TECHNICAL SPECIFICATION 3.4.8.1.

6.1.11 Raise Shutdown Cooling Train A temperature to within 100 F of Reactor Coolant Hot Leg temperature as follows:

6.1.11.1 Open the following valves:

SI-139A LPSI Header to RC Loop 2A Flow Control SI-138A LPSI Header to RC Loop 2B Flow Control 6.1.11.2 Throttle Closed RC Loop 2 Shdn Cooling Warmup, SI-135A, until one of the following is within 100 F of Shutdown Cooling Train A temperature, as indicated by LPSI Pump A Discharge Header Temperature Indicator, SI-ITI-0351X: [P-23174]

Hot Leg 2 temperature, as indicated by RC Loop 2 Hot Leg Temperature Indicator, RC-ITI-0122-HA or Hot Leg 1 temperature, as indicated by RC Loop 1 Hot Leg Temperature Indicator, RC-ITI-0112-HB 6.1.11.3 Close RC Loop 2 Shdn Cooling Warmup, SI-135A.

20

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 CAUTION THE FOLLOWING APPLIES TO SHUTDOWN COOLING FLOW:

(1) A TOTAL MINIMUM SHUTDOWN COOLING FLOW NECESSARY TO REMOVE DECAY HEAT AND PREVENT BORON STRATIFICATION SHOULD BE MAINTAINED AT ALL TIMES.

(2) WHEN CONSIDERING THE MINIMUM SHUTDOWN COOLING FLOW REQUIRED TO ADEQUATELY REMOVE DECAY HEAT AND PREVENT BORON STRATIFICATION, THE FLOW OF BOTH OPERATING SHUTDOWN COOLING TRAINS MAY BE USED.

(3) THE REQUIRED MINIMUM SHUTDOWN COOLING FLOW FOR MODES 5 AND 6 ARE AS FOLLOWS:

TIME AFTER SHUTDOWN (HOURS) REQUIRED FLOW (GPM)

<175 HOURS 4000 GPM 175 HOURS 3000 GPM 375 HOURS 2000 GPM IF THE REACTOR HAS BEEN SHUTDOWN <175 HOURS, THEN SHUTDOWN COOLING FLOW MAY BE REDUCED TO 3000 GPM IF RCS TEMPERATURE IS VERIFIED TO BE <135 F AT LEAST ONCE PER HOUR.

6.1.12 Adjust LPSI Header Flow Controller 2A/2B, SI-IFIC-0307, setpoint potentiometer to obtain desired Shutdown Cooling Train A flow, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

21

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 CAUTION (1) THE REACTOR COOLANT SYSTEM SHALL NOT EXCEED THE 100 F PER HOUR COOLDOWN RATE OF TECHNICAL SPECIFICATION 3.4.8.1.

(2) IF A DESIGN BASIS TORNADO EVENT OCCURS, THEN THE CCW HEAT EXCHANGER OUTLET CCW TEMPERATURE SHOULD BE MAINTAINED 130 F.

THE TEMPERATURE OF THE CCW RETURN LINE FROM SHUTDOWN COOLING HEAT EXCHANGER A CAN BE MONITORED ON PMC PID A43806.

(3) CC-963A IS REQUIRED TO BE MAINTAINED OPEN WHILE IN MODE 4 TO PRESERVE THE DESIGN TEMPERATURE BASIS OF PIPING AND ASSOCIATED COMPONENTS AT THE CCW OUTLET OF SHUTDOWN COOLING HEAT EXCHANGER A. WITH CC-963A OPEN, FLOW THROUGH SHUTDOWN COOLING HEAT EXCHANGER A WILL BE MAINTAINED ABOVE 2305 GPM. [EC-738]

6.1.13 Maintain RCS temperature control as follows:

6.1.13.1 Throttle Open Shutdown Cooling HX A Temperature Control, SI-415A, as required.

6.1.13.2 Place Shutdown HX A CCW Flow Control, CC-963A, to Open or Setpoint, as required.

NOTE Once activated the SHUTDOWN COOLING TROUBLE annunciator (Window H-18 on cabinet N) will alarm since the Low Flow setpoints are initially failed High.

6.1.14 Verify Computer Point PID B43800, SDCS Alarm Processing, set to ACTIVE state in accordance with OP-004-012, Plant Computer System.

6.1.15 Verify Computer Point PID K43101, SDCS/LPSI PMP A LOW FLOW LIM, set to approximately 200 gpm below the established Shutdown Cooling Train A flow, as indicated by RC Loop 2 Shdn Line Flow Indicator, SI-IFI-1307-A1.

6.1.16 If Shutdown Cooling Train B is not in service, then set Computer Point PID K43201 SDCS/LPSI PMP B LOW FLOW LIM, to Zero in accordance with OP-004-012, Plant Computer System.

6.1.17 If CETs from QSPDS Channel 1(2) are not available, then set PID C26417(C26510), TRCET Representative CET, to zero.

6.1.18 Verify SHUTDOWN COOLING TROUBLE annunciator (WINDOW H-18 ON CABINET N) is Clear.

22

System Operating Procedure OP-009-005 Shutdown Cooling Revision 022 NOTE Due to thermal expansion, the Shutdown Cooling Heat Exchanger A Outlet Stop Check, CS-117A, must be re-tightened in the Closed direction approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after Shutdown Cooling Train A is placed in service.

6.1.19 Verify Closed, Shutdown Cooling Heat Exchanger A Stop Check, CS-117A, approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after Shutdown Cooling Train A is placed in service.

6.1.20 If splitting of CCW Trains is necessary as directed by the SM/CRS, then go to Section 6.13, Splitting Out CCW Trains When on Shutdown Cooling.

23

Off Normal Procedure OP-901-131 Shutdown Cooling Malfunction Revision 2 D. IMMEDIATE OPERATOR ACTIONS

1. IF ANY of the following Shutdown Cooling Loop Suction Isolation valves close on the operating Shutdown Cooling train, THEN secure LPSI Pump:

For LPSI Pump A:

SDCS LOOP 2 SUCTION ISOL UPSTREAM INSIDE (SI 401A)

SDCS LOOP 2 SUCTION ISOL DOWNSTREAM INSIDE (SI 405A)

SDCS LOOP 2 SUCTION ISOL DOWNSTREAM OUTSIDE (SI 407A)

For LPSI Pump B:

SDCS LOOP 1 SUCTION ISOL UPSTREAM INSIDE (SI 401B)

SDCS LOOP 1 SUCTION ISOL DOWNSTREAM INSIDE (SI 405B)

SDCS LOOP 1 SUCTION ISOL DOWNSTREAM OUTSIDE (SI 407B).

6

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S5 Perform OP-903-037, Containment Cooling Fans Operability Check Candidate:

Examiner:

JPM S5 JOB PERFORMANCE MEASURE DATA PAGE Task: Perform OP-903-037, Containment Cooling Fans Operability Check Task Standard: Candidate completes OP-903-037 with CCS Fans A, B, and C running.

References:

OP-903-037, Containment Cooling Fans Operability Check OP-008-003, Containment Cooling System Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 8 NRC Exam 2009

JPM S5 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-903-037, Containment Cooling Fans Operability Check OP-008-003, Containment Cooling System

==

Description:==

This task is performed at CP-18. The candidate must perform surveillance OP-903-037, which will require logging differential pressure for the 3 running fans. The candidate will then have to secure a running Containment Cooling Fan and start Containment Cooling Fan D, at which time the data for CCS Fan D can be recorded.

The candidate should then leave the CCS Fans in an alignment with A, B, and D running, as specified inOP-903-037.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

Plant is in Mode 1 INITIATING CUES:

The CRS has directed you to perform OP-903-037, Containment Cooling Fans Operability Check Revision 0 Page 3 of 8 NRC Exam 2009

JPM S5 TASK ELEMENT STANDARD Record differential pressure (DP) and CCW flow rate for operating CFC units on Attachment 10.1, Correct values are recorded.

CFC Data Sheet.

Comment: CRITICAL STEP TASK ELEMENT STANDARD 0.2, Run Time Equalization Schedule Determine correct alignment for the month of Sheet, should be referenced in determining which October.

CFC operating unit(s) is secured.

Comment:

TASK ELEMENT STANDARD Stop Containment Fan Cooler C, from CP-18, by CCS Fan C is off.

placing Fan Cooler C control switch to Stop.

Comment: CRITICAL STEP Evaluator: This is written for the candidate to secure CCS Fan C, which is the most reasonable fan for him to stop. It is acceptable for the candidate to secure CCS Fan A or B at this point. This would require the candidate to maneuver fans later in the task to leave fans A, B, and D running.

TASK ELEMENT STANDARD Verify at CP-18, that CC-807 A and CC-823 A, Component Cooling Water Inlet and Outlet Valves Verification is complete.

for Containment Fan Cooler C are Closed.

Comment:

Evaluator: Statement for step above also applies to this step.

TASK ELEMENT STANDARD Start Containment Fan Cooler D from CP-18, by CCS Fan D is running.

placing Fan Cooler D control switch to Start/ Fast.

Comment: CRITICAL STEP Revision 0 Page 4 of 8 NRC Exam 2009

JPM S5 TASK ELEMENT STANDARD Verify at CP-18 that CC-808 B and CC-822 B, Component Cooling Water Inlet and Outlet Valves Verification complete.

for CCS Fan D, are Open Comment:

TASK ELEMENT STANDARD Record CFC D start time and CCW flow rate on Time and flow is recorded. 0.1, CFC Data Sheet.

Comment: CRITICAL STEP TASK ELEMENT STANDARD When CFC D has operated for > 15 minutes, then record differential pressure on Attachment 10.1, Differential pressure is recorded.

CFC Data Sheet.

Comment: CRITICAL STEP Evaluator: Prompt candidate that 15 minutes has passed at step 7.4.

TASK ELEMENT STANDARD Verify all four CFC units were operated and data Attachment updated.

recorded on Attachment 10.1, CFC Data Sheet.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify each CFC Unit CCW flow rate is > 625 GPM. Attachment updated.

Comment: CRITICAL STEP Revision 0 Page 5 of 8 NRC Exam 2009

JPM S5 TASK ELEMENT STANDARD Refer to Attachment 10.2, Run Time Equalization Candidate determines the correct alignment is Schedule Sheet, to obtain CFC unit alignment for CCS Fans A, B, and D running.

the upcoming month.

Comment:

TASK ELEMENT STANDARD Verify that CFC units are aligned as required on Attachment updated. 0.1, CFC Data Sheet.

Comment: CRITICAL STEP END OF TASK Revision 0 Page 6 of 8 NRC Exam 2009

JPM S5 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

Plant is in Mode 1 INITIATING CUES:

The CRS has directed you to perform OP-903-037, Containment Cooling Fans Operability Check Revision 0 Page 7 of 8 NRC Exam 2009

JPM S5 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-97 There are no malfunctions or overrides for this JPM.

Revision 0 Page 8 of 8 NRC Exam 2009

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 TABLE OF CONTENTS 1.0 PURPOSE .....................................................................................................................2 2.0 PREREQUISITES..........................................................................................................3 3.0 PRECAUTIONS AND LIMITATIONS ............................................................................4 3.1 PRECAUTIONS...............................................................................................................4 3.2 LIMITATIONS .................................................................................................................4 4.0 INITIAL CONDITIONS...................................................................................................5 5.0 MATERIAL AND TEST EQUIPMENT ...........................................................................6 6.0 ACCEPTANCE CRITERIA ............................................................................................7 7.0 PROCEDURE................................................................................................................8 8.0 AUTOMATIC FUNCTIONS ...........................................................................................9

9.0 REFERENCES

............................................................................................................10 9.1 USE ...........................................................................................................................10 9.2 SOURCE .....................................................................................................................10 10.0 ATTACHMENTS .........................................................................................................11 10.1 CFC DATA SHEET ......................................................................................................12 10.2 RUN TIME EQUALIZATION SCHEDULE SHEET.................................................................13 LIST OF EFFECTIVE PAGES Revision 5 1 - 13 CONTINUOUS USE 1

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 1.0 PURPOSE 1.1 Provide instructions for verifying operability of each group of Containment Fan Coolers (CFC).

1.2 Perform the following Technical Specifications Surveillance Requirements:

[Commitment P-1487]

4.6.2.2.a.1 15 minute operation of each group 4.6.2.2.a.2 > 625 gpm of CCW flow rate to each cooler 1.3 Running time equalization of Containment Fan Cooling unit motors ensures Containment Fan Coolers are maintained in an environmentally qualified state.

2

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 2.0 PREREQUISITES NONE 3

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 3.0 PRECAUTIONS AND LIMITATIONS 3.1 PRECAUTIONS NONE 3.2 LIMITATIONS 3.2.1 Inform the SM/CRS if the conditions of Section 6.0, Acceptance Criteria, cannot be met.

4

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 4.0 INITIAL CONDITIONS 4.1 Obtain SM/CRS permission to perform this test and document on Attachment 10.1, CFC Data Sheet.

4.2 Containment Cooling System in service in accordance with OP-008-003, Containment Cooling System Operating Procedure.

5

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 5.0 MATERIAL AND TEST EQUIPMENT NONE 6

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 6.0 ACCEPTANCE CRITERIA 6.1 Each component tested on Attachment 10.1, CFC Data Sheet, shall meet the following criteria.

6.1.1 Each CFC unit not already in operation is started from Control Room and operates for > 15 minutes.

6.1.2 Each CFC unit CCW flow rate is > 625 GPM.

7

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 7.0 PROCEDURE 7.1 Record differential pressure (DP) and CCW flow rate for operating CFC units on Attachment 10.1, CFC Data Sheet.

NOTE 0.2, Run Time Equalization Schedule Sheet, should be referenced in determining which CFC operating unit(s) is secured .

CAUTION TO PREVENT VIBRATION ALARMS, AND DAMAGE TO CONTAINMENT COOLING UNIT DUCT WORK, LIMIT CONFIGURATION TO ONLY THREE (3) OF FOUR (4) UNITS OPERATING AT A TIME.

7.2 Adjust CFC operating unit configuration to operate idle CFC unit(s).

7.3 Record CFC unit(s) start time and CCW flow rate, for unit(s) started in Step 7.2 on Attachment 10.1, CFC Data Sheet.

7.4 When CFC unit(s) have operated for > 15 minutes, then record CFC unit(s) differential pressure for unit(s) started in Step 2, on Attachment 10.1, CFC Data Sheet.

7.5 Verify all four CFC units were operated and data recorded on Attachment 10.1, CFC Data Sheet.

7.6 Verify each CFC Unit CCW flow rate is > 625 GPM.

7.7 Refer to Attachment 10.2, Run Time Equalization Schedule Sheet, to obtain CFC unit alignment for the upcoming month.

7.8 Verify that CFC units are aligned as required on Attachment 10.1, CFC Data Sheet.

8

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 8.0 AUTOMATIC FUNCTIONS NONE 9

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5

9.0 REFERENCES

9.1 USE 9.1.1. OP-008-003, Containment Cooling System Operating Procedure.

9.2 SOURCE 9.2.1. Technical Specification 4.6.2.2.a, Items 1 and 2 9.2.2. LOU-1564-G-853, Sheet 2 9.2.3. Letter W3B90-0308, Run Time Equalization of Containment Cooling System Fan Motors 9.2.4 Commitments P-1487 Containment Cooling System - Fan Operability/Cooling Water Flow Verification P-4806 Containment System Equipment Tests 10

Surveillance Procedure OP-903-037 Containment Cooling Fan Operability Verification Revision 5 10.0 ATTACHMENTS 10.1 CFC Data Sheet 10.2 Run Time Equalization Schedule 11

10.1 CFC DATA SHEET Test Permission: _____________________/_____________

SM/CRS (Signature) Date/Time CFC CCS-IDPR-5154AS __________ INWC CC-IFI-7570A2S __________ GPM Circle one: OPERATING STARTED Start time: __________ (if required)

CFC CCS-IDPR-5154BS __________ INWC CC-IFI-7570B2S __________ GPM Circle one: OPERATING STARTED Start time: __________ (if required)

CFC CCS-IDPR-5154AS __________ INWC CC-IFI-7570A1S __________ GPM Circle one: OPERATING STARTED Start time: __________ (if required)

CFC CCS-IDPR-5154BS __________ INWC CC-IFI-7570B1S __________ GPM Circle one: OPERATING STARTED Start time: __________ (if required) initials 10.1.1 Verify each CFC Unit CCW flow rate is >625 GPM and all CFC units started have been operated at least 15 minutes before recording unit DP. _________

10.1.2 Verify CFC units aligned, if possible, per Attachment 10.2, Run Time Equalization Schedule Sheet. _________

REMARKS:_________________________________________________________________

Performed By: ________________________________ ________/________

Operator (Signature) (Date/Time)

Reviewed By: ________________________________ ________/________

SM/CRS (Signature) (Date/Time)

OP-903-037 Revision 5 Attachment 10.1 (1 of 1) 12

10.2 RUN TIME EQUALIZATION SCHEDULE SHEET NOTE CFC units should be aligned in accordance with monthly schedule to equalize run times. If conditions do not allow running CFCs in accordance with this schedule or the System Engineer requests a different alignment, note change in remarks section of Attachment 10.1, CFC Data Sheet.

Month CFC Alignment January A, B, C February A, B, D March B, C, D April A, C, D May A, B, C June A, B, D July B, C, D August A, C, D September A, B, C October A, B, D November B, C, D December A, C, D OP-903-037 Revision 5 [LAST PAGE] Attachment 10.2 (1 of 1) 13

System Operating Procedure OP-008-003 Containment Cooling System Revision 6 6.0 NORMAL OPERATIONS 6.1 STARTING CONTAINMENT FAN COOLERS NOTE Normal Containment Cooling System Configuration in modes 1-4 is three (3) Containment Fan Coolers operating and one (1) in standby.

CAUTION TO PREVENT VIBRATION ALARMS AND DAMAGE TO CONTAINMENT COOLING UNIT DUCT WORK, LIMIT CONFIGURATION TO ONLY THREE (3) OF FOUR (4)

CONTAINMENT FAN COOLERS OPERATING AT ONE TIME WHEN IN FAST SPEED.

6.1.1. Start desired Containment Fan Coolers (CFC), from CP-18, by placing Fan Cooler A (B, C, D) control switch to Start/ Fast.

6.1.2. Verify at CP-18 that the Component Cooling Water Inlet and Outlet Valves Open for the in-service coolers:

CFC VALVES A Inlet CC-808A Outlet CC-822A B Inlet CC-807B Outlet CC-823B C Inlet CC-807A Outlet CC-823A D Inlet CC-808B Outlet CC-822B NOTE If Containment Fan Coolers are being started with slow speed jumpers installed for Containment cooling with Temporary Chilled Water, then steps 6.1.3 and 6.1.4 are not applicable.

6.1.3. Check in-service Containment Fan Coolers A (B, C, D) differential pressures within expected range of 5.0 INWC to 8.0 INWC, as indicated on CCS-IDPR-5154A(B).

6.1.4. Check Component Cooling Water flow is >625 gpm as indicated on CC-IFI-7570A(B).

7

System Operating Procedure OP-008-003 Containment Cooling System Revision 6 7.0 SYSTEM SHUTDOWN 7.1 SECURING CONTAINMENT FAN COOLERS 7.1.1. Stop desired Containment Fan Coolers (CFC), from CP-18, by placing Fan Cooler A (B, C, D) control switch to Stop.

7.1.2. Verify at CP-18, that Component Cooling Water Inlet and Outlet Valves for all secured Containment Fan Coolers are Closed.

CFC VALVES A Inlet CC-808A Outlet CC-822A B Inlet CC-807B Outlet CC-823B C Inlet CC-807A Outlet CC-823A D Inlet CC-808B Outlet CC-822B 8

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S6 Energize 4.16 KV Safety Bus from Offsite Power Candidate:

Examiner:

JPM S6 JOB PERFORMANCE MEASURE DATA PAGE Task: Energize 4.16 KV Safety Bus from Offsite Power Task Standard: Candidate synchronizes EDG B with in-coming power and secures EDG B.

References:

OP-902-009, Standard Appendices Attachment 12-C: Transfer 4.16 KV Safety Bus from EDG to Offsite Power Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 7 NRC Exam 2009

JPM S6 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-902-009, Standard Appendices, Attachment 12-C: Transfer 4.16 KV Safety Bus from EDG to Offsite Power

==

Description:==

This task is performed at CP-1. The candidate must synchronize Emergency Diesel Generator B across the 3B to 2B Tie Breaker and secure EDG B.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

A loss of off site power has occurred Emergency Diesel Generator B is running in Emergency Mode 2B Bus has been restored INITIATING CUES:

The CRS directs you to energize the 3B Safety Bus from Offsite Power using OP-902-009, Attachment 12-C.

Revision 0 Page 3 of 7 NRC Exam 2009

JPM S6 TASK ELEMENT STANDARD Verify BUS B3S TO B2 TIE BKR open. Verification complete.

Comment:

TASK ELEMENT STANDARD Close BUS B2 TO B3S TIE Breaker. Breaker is closed.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Place the SYCHRONIZER key switch in BUS TIE. Key switch is in BUS TIE.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Adjust EDG voltage to equal system voltage. Voltages approximately matched.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Adjust engine speed until synchroscope rotates Synchroscope is rotating slowly in the clockwise slowly clockwise. direction.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Close BUS B3S TO B2 TIE Breaker at 5 minutes to Breaker is closed.

twelve position on the synchroscope.

Comment: CRITICAL STEP Revision 0 Page 4 of 7 NRC Exam 2009

JPM S6 TASK ELEMENT STANDARD Place SYNCHRONIZER key switch in OFF. Key switch is in OFF.

Comment:

TASK ELEMENT STANDARD Reduce load on EDG A to 0.1 MW and 0.1 Load is reduced.

MVARS.

Comment:

TASK ELEMENT STANDARD Open EDG B Output Breaker. Breaker is open.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Stop EDG B. EDG B is off.

Comment: CRITICAL STEP Evaluator: After taking the EDG B control switch to OFF, both the red and green lights will go off on the control switch. The EDG is in cooldown mode at this point. This continues for 5 minutes. After cooldown is complete, the green light comes on. This task can be terminated when the EDG B control switch is takes to OFF, or after the cooldown cycle is complete, at the examiners discretion.

END OF TASK Revision 0 Page 5 of 7 NRC Exam 2009

JPM S6 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

A loss of off site power has occurred Emergency Diesel Generator B is running in Emergency Mode 2B Bus has been restored INITIATING CUES:

The CRS directs you to energize the 3B Safety Bus from Offsite Power using OP-902-009, Attachment 12-C.

Revision 0 Page 6 of 7 NRC Exam 2009

JPM S6 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-95 There are no malfunctions or overrides for this JPM.

Revision 0 Page 7 of 7 NRC Exam 2009

WATERFORD 3 SES OP-902-009 Revision 301 Page 106 of 195 STANDARD APPENDICES Appendix 12 Page 5 of 14 Electrical Restoration Attachment 12-C: Transfer 4.16 KV Safety Bus from EDG to Offsite Power INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.1 IF 4.16 KV nonsafety bus A2 is energized AND the 4.16 KV safety bus A3 is energized from the associated EDG, THEN perform the following:

a. Verify BUS A3S TO A2 TIE BKR open.

b. Close BUS A2 TO A3S TIE BKR.

c. Place the SYCHRONIZER keyswitch in "BUS TIE."

d. Parallel EDG A as follows:

1) Adjust EDG voltage to equal system voltage.

2) Adjust engine speed until synchroscope rotates slowly clockwise.

3) Close BUS A3S TO A2 TIE BKR at 5 minutes to twelve position on the synchroscope.

4) Place SYNCHRONIZER keyswitch in "OFF."

(continue)

WATERFORD 3 SES OP-902-009 Revision 301 Page 107 of 195 STANDARD APPENDICES Appendix 12 Page 6 of 14 INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.1 (continued)

e. IF SIAS is reset, THEN unload EDG A as follows:

1) Reduce load on EDG A to 0.1 MW and 0.1 MVARS.

2) Open GEN BREAKER.

3) Stop EDG A.

____ 1.2 IF 4.16 KV nonsafety bus B2 is energized AND the 4.16 KV safety bus B3 is energized from the associated EDG, THEN perform the following:

a. Verify BUS B3S TO B2 TIE BKR open.

b. Close BUS B2 TO B3S TIE BKR.

c. Place the SYCHRONIZER keyswitch in "BUS TIE."

(continue)

WATERFORD 3 SES OP-902-009 Revision 301 Page 108 of 195 STANDARD APPENDICES Appendix 12 Page 7 of 14 INSTRUCTIONS CONTINGENCY ACTIONS

____ 1.2 (continued)

d. Parallel EDG B as follows:

1) Adjust EDG voltage to equal system voltage.

2) Adjust engine speed until synchroscope rotates slowly clockwise.

3) Close BUS B3S TO B2 TIE BKR at 5 minutes to twelve position on the sychroscope.

4) Place SYNCHRONIZER keyswitch in "OFF."

e. IF SIAS is reset, THEN unload EDG B as follows:

1) Reduce load on EDG B to 0.1 MW and 0.1 MVARS.

2) Open GEN BREAKER.

3) Stop EDG B.

End of Attachment 12-C

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S7 Place Reactor Power Cutback in Service Candidate:

Examiner:

JPM S7 JOB PERFORMANCE MEASURE DATA PAGE Task: Place Reactor Power Cutback in Service Task Standard: A Reactor Power Cutback will be generated when the candidate places RXC in service. The candidate will be required to take the immediate operator actions for Reactor Power Cutback.

References:

OP-004-015, Reactor Power Cutback System OP-901-101, Reactor Power Cutback off normal procedure Validation Time: 20 minutes Time Critical: No Alternate Path: Yes Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 10 NRC Exam 2009

JPM S7 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-004-015, Reactor Power Cutback System

==

Description:==

This task is performed at CP-2. The candidate must perform OP-004-015 section 6.1 and place Reactor Power Cutback in service. At step 6.1.10, when RXC is placed in service, a RXC signal will be generated. The candidate will be required to take the immediate operator action for RXC, and place the CEA Mode Select switch in Auto-Sequential (AS). The examiner can stop the task after the candidate completes this action.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The plant is at 100% power.

Reactor trip on Turbine trip is in service.

OP-004-015, section 5.1, Reactor Power Cutback System Standby Alignment, has been completed.

Attachment 11.1, Manual CEA Subgroup Selection has been completed and reviewed.

Subgroups 5 and 11 are both required for current plant conditions.

INITIATING CUES:

The CRS directs you to place Reactor Power Cutback in service.

Revision 0 Page 3 of 10 NRC Exam 2009

JPM S7 TASK ELEMENT STANDARD Verify Section 5.1, Reactor Power Cutback System Included in cue sheet.

Standby Alignment, completed.

Comment:

TASK ELEMENT STANDARD Depress LAMP TEST pushbutton and verify all Lamps tested.

pushbuttons on panel illuminate.

Comment:

TASK ELEMENT STANDARD Verify AUTO ACTUATE OUT OF SERVICE Verification complete.

pushbutton Illuminated.

Comment:

TASK ELEMENT STANDARD Verify Reactor Pwr Cutback Single Chnl Trouble (L-5, Cabinet H) annunciator Clear.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Verify MANUAL SELECT Illuminated on AUTO Verification complete.

SELECT /MANUAL SELECT pushbutton.

Comment:

Revision 0 Page 4 of 10 NRC Exam 2009

JPM S7 TASK ELEMENT STANDARD Determine the appropriate CEA subgroup selection by performing Attachment 11.1, Manual CEA This Attachment has already been completed.

Subgroup Selection.

Comment:

Evaluator: If asked, inform the candidate that Attachment 11.1 concluded that subgroups 5 and 11 were necessary for both RXC events.

TASK ELEMENT STANDARD Depress ENTER MANUAL SUBGRPS SELECT Manipulation completed.

pushbutton and verify pushbutton Illuminates.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and Selections made.

verifying each selected pushbutton Illuminates.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Depress LARGE LOAD REJECT pushbutton and Manipulation completed.

verify pushbutton Illuminates.

Comment: CRITICAL STEP Revision 0 Page 5 of 10 NRC Exam 2009

JPM S7 TASK ELEMENT STANDARD When the SUBGROUP SELECT and LARGE LOAD REJECT pushbuttons have Extinguished (after approximately 60 seconds), then depress Manipulation completed.

DISPLAY SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

Comment:

TASK ELEMENT STANDARD Depress LARGE LOAD REJECT pushbutton and Manipulation completed.

verify pushbutton Illuminates.

Comment:

TASK ELEMENT STANDARD Verify correct CEA subgroup pattern is displayed. Verification complete.

Comment:

TASK ELEMENT STANDARD Depress ENTER MANUAL SUBGRPS SELECT Manipulation completed.

pushbutton and verify pushbutton Illuminates.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and Manipulation completed.

verifying each selected pushbutton Illuminates.

Comment: CRITICAL STEP Revision 0 Page 6 of 10 NRC Exam 2009

JPM S7 TASK ELEMENT STANDARD Depress LOSS OF FEED PUMP pushbutton and Manipulation completed.

verify pushbutton Illuminates.

Comment: CRITICAL STEP TASK ELEMENT STANDARD When the SUBGROUP SELECT and LOSS OF FEED PUMP pushbuttons have Extinguished (after approximately 60 seconds), then depress DISPLAY Manipulation completed.

SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

Comment:

TASK ELEMENT STANDARD Depress LOSS OF FEED PUMP pushbutton and Manipulation completed.

verify pushbutton Illuminates.

Comment:

TASK ELEMENT STANDARD Verify correct CEA subgroup pattern is displayed. Verification complete.

Comment:

TASK ELEMENT STANDARD Verify both Main Feedwater Pumps operating. Verification complete.

Comment:

Revision 0 Page 7 of 10 NRC Exam 2009

JPM S7 TASK ELEMENT STANDARD Depress AUTO ACTUATE OUT OF SERVICE Manipulation complete.

pushbutton and verify pushbutton Extinguishes.

Comment: CRITICAL STEP Evaluator: The malfunction is tied to the manipulation of this button. After this button is pressed, the RXC signal will be generated.

TASK ELEMENT STANDARD Place Control Element Drive Mechanism Mode Manipulation completed.

Select switch to AS.

Comment: CRITICAL STEP Evaluator: This is an immediate action step for a Reactor Power Cutback.

TASK ELEMENT STANDARD Verify selected subgroups dropped. Verification complete.

Comment: CRITICAL STEP Evaluator: This is an immediate action step for a Reactor Power Cutback.

Evaluator: Task should be terminated after this step is completed.

END OF TASK Revision 0 Page 8 of 10 NRC Exam 2009

JPM S7 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is at 100% power.

Reactor trip on Turbine trip is in service.

OP-004-015, section 5.1, Reactor Power Cutback System Standby Alignment, has been completed.

Attachment 11.1, Manual CEA Subgroup Selection has been completed and reviewed.

Subgroups 5 and 11 are both required for current plant conditions.

INITIATING CUES:

The CRS directs you to place Reactor Power Cutback in service.

Revision 0 Page 9 of 10 NRC Exam 2009

JPM S7 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-99 Verify the following under the Event Trigger prompt:

zdipwautoact.EQ.1 is listed for trigger 1 Keys required to setup this scenario:

Keys 179 and 180 for CP-2 Keys 173 - 176 for CP-7 Revision 0 Page 10 of 10 NRC Exam 2009

System Operating Procedure OP-004-015 Reactor Power Cutback System Revision 9 6.0 NORMAL OPERATIONS 6.1 ALIGNING REACTOR POWER CUTBACK FOR MANUAL CEA SUBGROUP SELECTION 6.1.1 Verify Section 5.1, Reactor Power Cutback System Standby Alignment, completed.

6.1.1.1 Depress LAMP TEST pushbutton and verify all pushbuttons on panel illuminate.

6.1.1.2 Release LAMP TEST pushbutton 6.1.2 Verify AUTO ACTUATE OUT OF SERVICE pushbutton Illuminated.

6.1.3 If the TEST RESET pushbutton is illuminated, then depress the TEST RESET pushbutton and verify pushbutton extinguishes.

6.1.4 Verify Reactor Pwr Cutback Single Chnl Trouble (L-5, Cabinet H) annunciator Clear.

6.1.4.1 If Reactor Pwr Cutback Single Chnl Trouble (L-5, Cabinet H) annunciator is not Clear, then realign Reactor Power Cutback in accordance with Section 5.1.

6.1.5 Verify MANUAL SELECT Illuminated on AUTO SELECT /MANUAL SELECT pushbutton.

6.1.6 Determine the appropriate CEA subgroup selection by performing Attachment 11.1, Manual CEA Subgroup Selection.

6.1.7 As determined from Attachment 11.1, manually align CEA subgroups for a Large Load Reject as follows:

6.1.7.1 Depress ENTER MANUAL SUBGRPS SELECT pushbutton and verify pushbutton Illuminates.

6.1.7.2 Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and verifying each selected pushbutton Illuminates.

6.1.7.3 Depress LARGE LOAD REJECT pushbutton and verify pushbutton Illuminates.

6.1.7.4 When the SUBGROUP SELECT and LARGE LOAD REJECT pushbuttons have Extinguished (after approximately 60 seconds), then perform the following:

6.1.7.4.1 Depress DISPLAY SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

8

System Operating Procedure OP-004-015 Reactor Power Cutback System Revision 9 6.1.7.4.2 Depress LARGE LOAD REJECT pushbutton and verify pushbutton Illuminates.

6.1.7.4.3 Verify correct CEA subgroup pattern is displayed.

6.1.8 As determined from Attachment 11.1, manually align CEA subgroups for a Loss of Feed Pump as follows:

6.1.8.1 Depress ENTER MANUAL SUBGRPS SELECT pushbutton and verify pushbutton Illuminates.

6.1.8.2 Establish CEA subgroup pattern by Depressing desired SUBGROUP SELECT pushbuttons and verifying each selected pushbutton Illuminates.

6.1.8.3 Depress LOSS OF FEED PUMP pushbutton and verify pushbutton Illuminates.

6.1.8.4 When the SUBGROUP SELECT and LOSS OF FEED PUMP pushbuttons have Extinguished (after approximately 60 seconds), then perform the following:

6. 1.8.4.1 Depress DISPLAY SUBGRP SELECT pushbutton and verify pushbutton Illuminates.

6.1.8.4.2 Depress LOSS OF FEED PUMP pushbutton and verify pushbutton Illuminates.

6.1.8.4.3 Verify correct CEA subgroup pattern is displayed.

NOTE Turbine DEH System Program has a minimum floor of 20% power. A Reactor Cutback rod configuration should not be selected that would drop Reactor Power below 20% in the event of a Reactor Power Cutback.

6.1.9 Verify both Main Feedwater Pumps operating.

6.1.10 Depress AUTO ACTUATE OUT OF SERVICE pushbutton and verify pushbutton Extinguishes.

6.1.11 At SM/CRS discretion, remove Reactor Trip on Turbine Trip from service as follows:

6.1.11.1 On CP-2, place LOSS OF LOAD keyswitch to RPC.

6.1.11.2 On CP-7, place all four LOSS OF TURB BYPASS keyswitches to BYPASS and verify and all four red BYPASS lamps Illuminate.

6.1.11.3 On CP-2, place LOSS OF TURBINE TRIP keyswitch to DISABLE.

9

System Operating Procedure OP-004-015 Reactor Power Cutback System Revision 9 6.1.12 As Reactor Power and Core EFPD change, reevaluate manual CEA subgroup selection and change as necessary in accordance with Section 6.3, Changing Manual CEA Subgroup Selection. [P-21931]

10

Off Normal Procedure OP-901-101 Reactor Power Cutback Revision 6 D. IMMEDIATE OPERATOR ACTIONS

1. Place Control Element Drive Mechanism Mode Select switch to AS.

2. Verify selected subgroups dropped.

7

Waterford 3 2009 NRC Exam JOB PERFORMANCE MEASURE S8 Secure Containment Purge Candidate:

Examiner:

JPM S8 JOB PERFORMANCE MEASURE DATA PAGE Task: Secure Containment Purge Task Standard: Candidate secures Containment Purge and Containment Airborne Radiation Removal System.

References:

OP-002-010, Reactor Auxiliary Building HVAC and Containment Purge OP-008-009, Airborne Radioactivity Removal OP903-001, Technical Specification Logs Validation Time: 15 minutes Time Critical: No Alternate Path: No Candidate:

Time Start: Time Finish:

Performance Time: minutes Performance Rating: SAT UNSAT Comments:

Examiner: Date:

Signature Revision 0 Page 2 of 7 NRC Exam 2009

JPM S8 EXAMINER COPY ONLY Tools/Equipment/Procedures Needed:

OP-002-010, Reactor Auxiliary Building HVAC and Containment Purge OP-008-009, Airborne Radioactivity Removal OP903-001, Technical Specification Logs

==

Description:==

This task is performed at CP-18. The candidate must secure Containment Purge. He must evaluate if Airborne Radioactivity Removal Fan A must be secured based on Containment iodine level. After securing Containment Purge and Airborne Radioactivity Removal Fan A, the candidate must then update the Containment Purge Cumulative Hour log using OP-903-001, Technical Specification Logs.

READ TO CANDIDATE DIRECTION TO CANDIDATE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM, including any required communications. I will provide initiating cues and reports on other actions when directed by you. Indicate to me when you understand your assigned task.

INITIAL CONDITIONS:

The plant is in Mode 1.

Containment Purge was started at 2245 yesterday.

INITIATING CUES:

The CRS directs you to secure Containment Purge.

Revision 0 Page 3 of 7 NRC Exam 2009

JPM S8 TASK ELEMENT STANDARD Place RAB Vent Mode selector switch to Normal. Switch is in Normal.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Secure Containment Airborne Radioactivity Candidate uses RM-11 to verify Containment Removal System in accordance with OP-008-009, iodine is < 2.5 x 10-9 ci/ml.

Airborne Radioactivity Removal.

Comment: CRITICAL STEP Note from procedure states: Airborne Radioactivity Removal Units should be secured when Containment Atmosphere Radiation Monitor (PRM-IRE-0100-AS) Iodine activity is < 2.5 x 10-9 ci/ml or as directed by SM/CRS.

TASK ELEMENT STANDARD Momentarily place control switch for Airborne Radioactivity Removal Units A, ARR-0002A, to Fan is off.

Stop.

Comment: CRITICAL STEP TASK ELEMENT STANDARD If plant is in Modes 1-4, then enter Containment Purge Hours in Cumulative Hours Tracking Log in N/A accordance with OP-903-001, Technical Specification Surveillance Logs.

Comment:

This step directs the candidate to update the Containment Purge Cumulative tracking Log.

At this step, provided the candidate with the prepared log sheet.

TASK ELEMENT STANDARD Enter the date and mode in the column Data entered.

Date/Mode.

Comment: CRITICAL STEP Revision 0 Page 4 of 7 NRC Exam 2009

JPM S8 TASK ELEMENT STANDARD Enter the time the Purge was Initiated in the column Time Initiated. If the Purge was in service Data entered.

at 0000, then enter 0000.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Enter the time the Purge was terminated in the column Time Terminated. If the Purge was Data entered.

continued at 2400, then enter 2400.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Subtract column Time Initiated from column Time Terminated and enter the result in the column Data entered.

Time.

Comment: CRITICAL STEP TASK ELEMENT STANDARD Initial for performance of all of the entries in Data entered.

column Performed.

Comment:

END OF TASK Revision 0 Page 5 of 7 NRC Exam 2009

JPM S8 CANDIDATE CUE SHEET (TO BE RETURNED TO EXAMINER TO UPON COMPLETION OF TASK)

INITIAL CONDITIONS:

The plant is in Mode 1.

Containment Purge was started at 2245 yesterday.

INITIATING CUES:

The CRS directs you to secure Containment Purge.

Revision 0 Page 6 of 7 NRC Exam 2009

JPM S8 SIMULATOR OPERATOR INSTRUCTIONS Reset to IC-99 There are no malfunctions or overrides for this JPM.

Revision 0 Page 7 of 7 NRC Exam 2009

System Operating Procedure OP-002-010 Reactor Auxiliary Building HVAC and Containment Purge Revision 303 6.6 STOPPING CONTAINMENT PURGE W HILE MAINTAINING NORMAL VENTILATION [C]

NOTE Containment Purge should remain in service as much as practical during outage activities (e.g., Steam Generator primary side activities, Reactor Head disassembly, Reactor Cavity decontamination, Reactor Head inspections, etc.)

6.6.1 Notify Radiation Protection that Containment Purge will be secured.

6.6.2 Place RAB Vent Mode selector switch to Normal.

NOTE HVR-112 is failed to a fixed purge position. The closed and open limit switches are made up.

The PMC will show this damper in the intermediate position.

6.6.3 Verify following valves and dampers align as indicated:

HVR-112 (D52031) ........................................................................ OPEN HVR-115 (D52033) ........................................................................ OPEN HVR-116 (D52029) ........................................................................ OPEN CAP-202 (D51014) ........................................................................ OPEN CAP-203 (D51009) .................................................................... CLOSED CAP-204 (D51011) .................................................................... CLOSED CAP-205 (D51013) .................................................................... CLOSED CAP-101 (D51019) .................................................................... CLOSED CAP-102 (D51003) .................................................................... CLOSED CAP-103 (D51005) .................................................................... CLOSED CAP-104 (D51007) .................................................................... CLOSED 6.6.4 Secure Containment Airborne Radioactivity Removal System in accordance with OP-008-009, Airborne Radioactivity Removal.

6.6.5 If plant is in Modes 1-4, then enter Containment Purge Hours in Cumulative Hours Tracking Log in accordance with OP-903-001, Technical Specification Surveillance Logs.

22

System Operating Procedure OP-008-009 Airborne Radioactivity Removal Revision 8 7.0 SYSTEM SHUTDOWN 7.1 AIRBORNE RADIOACTIVITY REMOVAL SYSTEM SHUTDOWN NOTE Airborne Radioactivity Removal Units should be secured when Containment Atmosphere 9

Radiation Monitor (PRM-IRE-0100-AS) Iodine activity is < 2.5 x 10- ci/ml or as directed by SM/CRS.

7.1.1 Momentarily place control switch for Airborne Radioactivity Removal Units A(B),

ARR-0002A(B), to Stop.

8

11.6 CONTAINMENT PURGE CUMULATIVE HOURS CALCULATION 11.6.1 If a Purge is performed in Modes 1-4, then perform the following on the Containment Purge Tracking Hours Calculation Data Sheet:

11.6.1.1 Enter the date and mode in the column Date/Mode.

11.6.1.2 Enter the time the Purge was Initiated in the column Time Initiated. If the Purge was in service at 0000, then enter 0000.

11.6.1.3 Enter the time the Purge was Terminated in the column Time Terminated. If the Purge was continued at 2400, then enter 2400.

11.6.1.4 Subtract column Time Initiated from column Time Terminated and enter the result in the column Time.

11.6.1.5 Initial for performance of all of the entries in column Performed.

11.6.1.6 Verifier check all entries and initial column Verified.

11.6.2 If a Purge was performed with the Plant in Modes 1-4, then complete the following on the Containment Purge Cumulative Hours/Daily Calculation Data Sheet, for the previous day, on the 19-07 shift:

11.6.2.1 Enter the date and mode in the column Date/Mode.

11.6.2.2 From the Containment Purge Cumulative Hours Tracking Data Sheet add all the entries in the column Time for the previous day and record in the column Purge Time.

11.6.2.3 Add column Purge Time to column Accumulated Purge Time and enter the result in column Sub Total.

11.6.2.4 Find the date 365 days past and record the value from column Purge Time from that date in column Purge Time Last YRS Date for the present Date.

11.6.2.5 Subtract column Purge Time Last YRS Date from column Sub Total and record this value in column Accumulated Purge Time.

11.6.2.6 Verify column Accumulated Purge Time result is <90 hours and record YES/NO in column <90 HRS.

11.6.2.7 Initial for performance of all of the entries in column Performed.

11.6.2.8 Verifier check all entries and initials in column Verified.

OP-903-001 Revision 036 Attachment 11.6 (1 of 4) 94

11.6.3 If a Containment Purge was not performed on the previous day, or the Plant was in Mode 5 or 6 on the previous day, then perform the following on the Containment Purge Cumulative Hours/Daily Calculation Data Sheet, on the 19-07 shift.

11.6.3.1 Enter the previous date and mode in the column Date/Mode.

11.6.3.2 Enter 0 in column Purge Time.

11.6.3.3 Add column Purge Time to the previous value in column Accumulated Purge Time and enter the results in column Sub Total.

11.6.3.4 Find the date 365 days and record the value from column Purge Time from that date in column Purge Time Last YRS Date for the present Date.

11.6.3.5 Subtract column Purge Time Last YRS Date from column Sub Total and record this value in column Accumulated Purge Time.

11.6.3.6 Verify column Accumulated Purge Time result is <90 hours and record YES/NO in column <90 HRS.

11.6.3.7 Initial for performance of all the entries in column Performed.

11.6.3.8 Verifier check all entries and initials in column Verified.

OP-903-001 Revision 036 Attachment 11.6 (2 of 4) 95

Containment Purge Cumulative Hours Tracking Data Sheet (Typical)

TIME TIME PERFORMED VERIFIED DATE/MODE TIME INITIATED TERMINATED (Initial) (Initial)

OP-903-001 Revision 036 Attachment 11.6 (3 of 4) 96

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 1 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 100%, MOC, AB buses aligned to B side.

Protected Train is B Turnover: Maintain 100 % power Continue with Surveillance OP-903-094, section 7.20 Event Malf. No. Event Event No. Type* Description 1 Di08A04S08-1 N - BOP Perform surveillance OP-903-094, section 7.20.

TS - SRO BD-103B fails to close.

2 RC15-A1 I - ATC Pressurizer level instrument RC-ILI-0110-X fails I - SRO high TS-SRO 3 CC12-E2 I - BOP Component Cooling Water Surge Tank level TS - SRO instrument CC-ILS-7013A fails low 4 FW21-A R- ATC Main Condenser leak with lowering Main N-BOP Condenser vacuum requiring a Rapid Plant Power Reduction N-SRO 5 RC23B M-All Small Break LOCA, SIAS and CIAS CC12-E2 C-ATC Secure Reactor Coolant Pumps due to the C - SRO combination of event 3 and event 5.

6 SI02 C - BOP Low Pressure Safety Injection Pump A fails to auto start on SIAS requiring manual start 7 CS01-A C-BOP Containment Spray Pump A trip, OP-902-008, C-SRO Safety Function Recovery Procedure Alignment of LPSI Pump A to replace CS Pump A

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 1 Rev 1 Scenario Event Description NRC Scenario 1 The crew assumes the shift at 100% power with instructions to maintain 100% power.

Surveillance procedure OP-903-094, ESFAS Subgroup Relay Test - Operating, is in progress. The previous crew stopped at section 7.20, Train A Position 44, Relay K310 (BD-103B). This crew should resume testing. The BOP will secure Blowdown flow for Steam Generator #2 and test BD-103B, which will fail to close. The SRO should enter Tech Spec 3.6.3.

After briefing the failure, Pressurizer level instrument RC-ILI-0110X fails high. Due to the failure, Letdown flow goes to maximum flow of approximately 125 gpm and all Pressurizer Heaters energize. The SRO should enter OP-901-110, Pressurizer Level Control Malfunction. The crew should utilize sub section E1, Pressurizer Level Control Channel Malfunction. The ATC should take manual control of Pressurizer level and select the non-faulted channel. Using Tech Specs and OP-903-013, Monthly Channel Checks, the SRO should enter Tech Spec 3.3.3.5, a 7 day action requirement, and determine Tech Spec 3.3.3.6 entry is not required since QSPDS is operable and meeting the Pressurizer level channel check. SPDS indication of Pressurizer level is affected by this failure.

After the non-faulted channel is selected and Tech Specs are addressed, Component Cooling Water Surge Tank level instrument CC-ILS-7013A fails low. CCW Dry Cooling Tower A will bypass due to the failure. CCW Headers A and B will split, and CCW Loop AB supply and return from the A Header will close. The SRO should enter OP-901-510, Component Cooling Water System Malfunction. The BOP should use Attachment 1 to diagnose which instrument is failed. The crew should verify Auxiliary Component Cooling Water Pump A starts and control CCW system temperature with ACC-126 A.

CCW Train A should be declared inoperable and 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action Tech Spec 3.7.3 entered as well as cascading Tech Specs. The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b. They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

After the crew has addressed Tech Specs, a leak in the Main Condenser develops and Main Condenser vacuum begins to drop. Off normal procedure OP-901-220, Loss of Condenser Vacuum, should be entered. Main Condenser vacuum will drop below 25 inches, requiring a rapid plant power reduction. The SRO should enter OP-901-212, Rapid Plant Power Reduction. Vacuum will drop below 25 inches but remain above 20 inches, the procedure trigger for tripping the Reactor. For the power reduction, the ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the controls to reduce Main Turbine load.

Scenario 1 Rev 1 Scenario Event Description NRC Scenario 1 Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, a small break loss of coolant accident will occur. The crew should diagnose Pressurizer level dropping with all available Charging Pumps operating, trip the Reactor, and initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS). Because of the earlier CCW level instrument failure, all CCW flow will be lost to the Reactor Coolant Pumps; the pumps must be manually secured within 3 minutes of the loss of CCW flow. When Containment Spray is actuated, either manually or automatically, CS-125B will fail to automatically open and will not open using the control switch. This does not create a need for action at this time, but Containment Spray flow will only be provided from Train A with CS-125B failed closed. Low Pressure Safety Injection Pump A will fail to automatically start on SIAS, requiring the BOP operator to manually start LPSI Pump A.

After the crew completes OP-902-000, Standard Post Trip Actions and diagnoses into OP-902-002, Loss of Coolant Accident Recovery, Containment Spray Pump A will trip, resulting in no Containment Spray flow. The crew should recognize that they are not meeting the Safety Function Status Checklist of OP-902-002 and transition to OP-902-008, Safety function Recovery Procedure.

Prioritization in OP-902-008 should result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2. The crew should address Containment Isolation by overriding CS-125A closed. The crew should address Containment Temperature and pressure Control by aligning Low Pressure Safety Injection Pump A to replace the failed Containment Spray Pump A. It is acceptable to pursue these tasks in parallel, since establishing flow with LPSI A to the Containment Spray header will also satisfy Containment Isolation concerns.

The scenario can be terminated after Low Pressure Safety Injection Pump A is aligned for Containment Spray, or after the CRS gives the order to perform that alignment, at the lead examiners discretion.

Scenario 1 Rev 1 Scenario Event Description NRC Scenario 1 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow. This task is set up by the failure of CC-ILS-7013 A. The required task becomes applicable after SIAS is initiated following event 5. The time requirement of 3 minutes is based on the RCP operating limit of 3 minutes without CCW cooling.

2. Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump A to replace CS Pump A prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008. This task becomes applicable following the failure of Containment Spray Pump A in event 7. The Functional Recovery procedure utilized following this failure will direct this activity to satisfy the Containment Pressure and Temperature Control safety function.

Scenario 1 Rev 1 NRC Scenario 1 Scenario Notes:

A. Reset Simulator to IC-91.

B. Verify the following Scenario Malfunctions are loaded:

1. rc15a1 for Pressurizer level instrument RC-ILI-0110 X

2. cc12e2 for CCW Surge Tank level instrument failure, CC-ILS-7013 A

3. fw21a for Main Condenser vacuum leak

4. rc23b for SBLOCA

5. cs01a for Containment Spray Pump A trip

6. cs04b for CS-125 B fail to auto open

7. si02d for LPSI Pump A fail to auto start C. Verify the following remotes

1. egr26 for EDG A local alarm acknowledgement

2. egr27 for EDG B local alarm acknowledgement D. Verify the following overrides

1. di-08a07s22-1 for CS-125 B over-ride closed

2. di-08a04s08-1 for BD-103 B over-ride closed E. Ensure Protected Train B sign is placed in SM office window.

F. Verify EOOS is 10.0 Green G. Complete the simulator setup checklist.

Scenario 1 Rev 1 NRC Scenario 1 Simulator Booth Instructions Event 1 BD-103 B Fails to Close during OP-903-094

1. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

2. If called to check BD-103 B locally, report BD-103 B stem appears bent.

Event 2 Pressurizer Level Instrument RC-ILI-0110X Fails High

3. On Lead Examiner's cue, initiate Event Trigger 2.

4. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

5. If called to check any Charging Pumps, report requested Charging Pump is ready for a start or is running normal, which ever is applicable.

Event 3 CCW Level Switch CC-ILS-7013A Fails Low

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If called as RCA watch to check ACCW Pump A, report pump looked good on its start and is running normal.

Event 4 Main Condenser Leak, Rapid Power Reduction

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. If called as TGB watch report all Air Evacuation Pumps look normal, no indications of a leak.

3. Approximately 5 minutes after being called to investigate, TGB watch should report finding a non-isolable leak up-stream of AE-401 A, Condenser Vacuum Breaker A. Location of failure is preventing any successful repair efforts.

4. If called as other watch standers to assist, respond that you are going to the TGB to assist.

5. If Work Week Manager is called, inform the caller that a team will be sent to the Turbine Building to assist.

Scenario 1 Rev 1 NRC Scenario 1 Event 5 Small Break LOCA Inside Containment

1. On Lead Examiner's cue, initiate Event Trigger 5.

2. If called as RCA watch report CS-125 B appears to be mechanically bound, the stem looks bent.

3. If called as RAB watch to check the Emergency Diesel Generators, initiate Trigger 10, EDG A & B Trouble alarms clear, report they are running satisfactorily.

4. If the Duty Plant Manager is called, inform the caller that he will make the necessary calls.

Event 6 Low Pressure Safety Injection Pump A fails to start

1. If called to check the LPSI Pump A breaker, report all indications are normal.

2. If called to check the LPSI Pump A locally, report all indications are normal.

Event 7 Containment Spray Pump A Trips

3. After the crew has entered OP-902-002 and on the Lead Examiner's cue, initiate Event Trigger 6.

4. If called to check the Containment Spray Pump A breaker, report over-current flags are picked up on all 3 phases.

5. If called to check the Containment Spray Pump A, report that there are visible charring on the motor with an acrid smell, but no indications of a fire or smoke.

Scenario 1 Rev 1 NRC Scenario 1 Scenario Timeline:

Ramp Time Event Malfunction Severity Delay Trigger HH:MM:SS (Min) 1 di-08a04s08-1 Open N/A N/A N/A BD-103 B Fails to close with MSIS signal 2 RC15 A1 N/A N/A N/A 2 Pressurizer level instrument RC-ILI-0110 X fails high 3 CC12 E2 0% N/A NA 3 CCW Surge Tank level instrument CC-ILS-7013 A fails low 4 FW21 A 20% 00:03:00 N/A 4 Main Condenser vacuum leak 5 RC23B .3% 00:02:00 N/A 5 Small Break LOCA ramping to 160 GPM to Containment 5 CS04 B N/A N/A N/A N/A CS-125 B fails to auto-open 6 SI02 D N/A N/A N/A LPSI Pump A fails to auto start on SIAS 7 CS01 A TRUE N/A N/A 6 Containment Spray Pump A trips.

Scenario 1 Rev 1 NRC Scenario 1

REFERENCES:

Event Procedures 1 OP-903-094, ESFAS Subgroup Relay Test - Operating Tech Spec 3.6.3 2 OP-901-110, Pressurizer Level Control Malfunction OP-903-013, Monthly Channel Checks Tech Spec 3.3.3.5 3 OP-901-510, Component Cooling Water System Malfunction OP-100-014, Technical Specification and Technical Requirements Compliance OP-903-066, Electrical Breaker Alignment Checks Tech Spec 3.7.3 and cascading Tech Specs to include 3.8.1.1 4 OP-901-220, Loss of Condenser Vacuum OP-901-212, Rapid Plant Power Reduction Tech Spec 3.1.3.6 Regulating and Group P CEA Insertion Limits 5 OP-902-000, Standard Post Trip Actions OP-902-002, Loss of Coolant Accident Recovery Procedure OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart 7 OP-902-008, Safety Function Recovery Procedure OP-902-009, Standard Appendices, Appendix 28, Aligning LPSI to Replace CS Scenario 1 Rev 1 Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 1 Page 1 of 14 Event

Description:

During surveillance OP-903-094, BD-103B fails to close Time Position Applicants Actions or Behavior BOP Lower Blowdown Flow SG2, BD-IHIC-0104-B, to zero percent output.

Both the Blowdown flow controller and Blowdown flow indication are located on CP-1.

BOP Align ESFAS Test Module A to test position 44.

Located on CP-33 BOP Press and Hold the INITIATE ACTUATION pushbutton.

Located on CP-33 CRS Upon notification of BD-103B failure to auto close, evaluate Tech Specs.

Correct Tech Spec is 3.6.3. Blowdown is identified as a closed system, so the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action applies to the failure. The CRS may discuss the need to verify that the component is failed and not the test relay. If asked, the report from the field will assist in that decision.

Examiner Note This event is complete after Tech Spec 3.6.3 has been addressed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 2 of 14 Event

Description:

Pressurizer Level Instrument RC-ILI-0110-X Fails High Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

PRESSURIZER LEVEL HI/LO (Cabinet H, B-1)

PRESSURIZER LEVEL HI-HI (Cabinet H, A-1)

LETDOWN FLOW HI/LO (Cabinet G, C-1)

Indications Mismatch between Charging (CVC-IFI-0212) AND Letdown (CVC-IFI-0202) flow indications High level indicated on Pressurizer level indicator RC-ILI-0110 X Deviation between actual level AND programmed level as indicated on Pressurizer level recorder (RC-ILR-0110)

CRS Enter and direct the implementation of OP-901-110, Pressurizer Level Control Malfunction.

CRS In section E0, General, of OP-901-110, direct use of sub-section E1.

CRS There is a note at the start of sub-section E1 that reads:

Selecting the non-faulted channel may cause automatic actions to occur if actual level is not at program level.

The CRS should evaluate this note and either wait for Pressurizer level to return to setpoint before selecting Channel Y or acknowledge that additional Charging Pumps may start as a result of selecting Channel Y.

If he chooses to wait, it will take several minutes for Pressurizer level to return to setpoint. In this case, consider moving to malfunction 3 during the wait.

All ATC manipulations are located on CP-2 ATC Place Pressurizer Level Controller (RC-ILIC-0110) in MAN AND adjust OUTPUT to slowly adjust letdown flow to restore Pressurizer level.

ATC Transfer Pressurizer Level Control CHANNEL SELECT switch to Channel Y.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 2 Page 3 of 14 Event

Description:

Pressurizer Level Instrument RC-ILI-0110-X Fails High Time Position Applicants Actions or Behavior ATC Transfer Pressurizer CHANNEL SELECT LO LEVEL HEATER CUTOFF switch to Channel Y.

ATC / CRS Verify desired backup Charging pumps in AUTO ATC Place Pressurizer Level Controller (RC-ILIC-0110) in AUTO and verify Pressurizer Level is being restored to setpoint.

ATC / CRS Verify Pressurizer level controlling at program setpoint in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

CRS Refer to Technical Specifications 3.3.3.5 and 3.3.3.6 for Remote Shutdown and Accident Monitoring operability determination.

OP-903-013, Monthly Channel Checks should be used to implement tech Specs 3.3.3.5 and 3.3.3.6.

CRS Enter Tech Spec 3.3.3.5 Since QSPDS meets the channel check requirement of OP-903-013, entry into Tech Spec 3.3.3.6 is not appropriate.

There will be a RCS pressure rise during this malfunction. The ATC may discuss the reactivity effects associated with this pressure rise.

Waterford 3 does have an ODMI (Operations Decision Making Issue) with a trigger associated with a 50 PSIA change in RCS Pressure. This trigger should not be reached, but the CRS may discuss the ODMI and the actions should the trigger be reached.

Examiner Note This event is complete when Channel Y is selected for Pressurizer Level Control and for Pressurizer Lo Level Heater Cutout and after tech Specs 3.3.3.5 and 3.3.3.6 have been addressed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 4 of 14 Event

Description:

Component Cooling Water Surge Tank Level Instrument CC-ILS-7013 A Fails Low Time Position Applicants Actions or Behavior BOP Recognize and report indications of failed Component Cooling Water level switch.

Alarms CCW A Surge Tank Level Lost (Cabinet SA, B-5)

Dry Tower A Isolated (Cabinet B, N-9)

Aux CCW Sys A Press Lo Jockey Pump Trip/Trouble (Cabinet B, K-9)

CCW Makeup Pump A Running / Power Lost (Cabinet M, G-2)

Shutdown HX A CCW Flow Lo (Cabinet M, H-2)

Indications Component Cooling Water Makeup Pump A Starts, if in Auto, and runs for 3 minutes. Further Operation will require manually starting and stopping the pump.

Auxiliary Component Cooling Water Pump A auto-starts.

CC-134A, CCW A Dry Cooling Tower Bypass, Opens CC-135A, CCW A Dry Cooling Tower Isolation, Closes CC-126A/CC-114A, CCW Suct & Disch Header Tie Valves AB To A, Close CC-126B/CC-114B, CCW Suct & Disch Header Tie Valves AB To B, Close CC-200A/CC-727, CCW Suct & Disch Header Tie Valves A To AB, Close CC-620, Fuel Pool Heat Exchs Temperature Control, Closes CRS Enter and direct the implementation of OP-901-510, Component Cooling Water System Malfunction.

BOP Verify Auxiliary Component Cooling Water Pump A automatically starts and controls CCW temperature at setpoint.

CRS In section E0, General, of OP-901-510, direct use of Attachment 1 to determine which instrument is failed.

BOP Report that the failed level switch is CC-ILS-7013 A.

CRS Enter tech Spec 3.7.3 and cascading Tech Specs.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 3 Page 5 of 14 Event

Description:

Component Cooling Water Surge Tank Level Instrument CC-ILS-7013 A Fails Low Time Position Applicants Actions or Behavior CRS Direct the performance of surveillance procedure OP-903-066, Electrical Breaker Alignment Check. This surveillance must be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b.

CRS Direct the verification of all B Train safety components that rely on EDG B are operable. Direct the verification that EFW Pump AB is operable. These verifications must be completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to comply with Tech Spec 3.8.1.1.d.

BOP At CP-8, verify Train B components and EFW Pump AB are operable.

Verification includes verifying switch indications are normal and that no Train B annunciators are in alarm.

Examiner Note This event is complete when Tech Specs 3.7.3 and cascading Tech Specs have been addressed Or As directed by the Lead Evaluator This failure requires additional actions by the ATC after the initiation of SIAS.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 6 of 14 Event

Description:

Main Condenser Leak, Lowering Main Condenser Vacuum, Rapid Plant Power Reduction Time Position Applicants Actions or Behavior BOP Recognize and report indications of lowering Main Condenser vacuum.

Alarms VACUUM PUMP A AUTO START (Cabinet E, E-1)

VACUUM PUMP B AUTO START (Cabinet E, E-2)

VACUUM PUMP C AUTO START (Cabinet E, E-3)

Indications Condenser Vacuum dropping on any of the following:

o PMC alarms A01103 and A10203 o Wide Range Condenser Vacuum (CD-IPI-1902-B2) o Narrow Range Condenser Vacuum (CD-IPI-1901-B) o Condenser Vacuum recorder (CD-IPR-1902-A)

At 26 INHG, standby Condenser Vacuum Pump(s) start CRS Enter and direct the implementation of OP-901-220, Loss of Condenser Vacuum.

There are several steps to look for the source of the vacuum problem.

The CRS will direct the verification of these items. There will be no problems indicated on any of these parameters.

CRS Enter and direct the implementation of OP-901-212, Rapid Plant Power Reduction.

CRS Establish a reactor trip criteria of > 20 Hg vacuum.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 7 of 14 Event

Description:

Main Condenser Leak, Lowering Main Condenser Vacuum, Rapid Plant Power Reduction Time Position Applicants Actions or Behavior ATC Begin RCS Boration by either of the following methods as directed by the CRS:

Direct Boration Emergency Boration using one Charging Pump ATC Steps for Direct Boration:

Set Boric Acid Makeup Batch Counter to volume of Boric Acid desired.

Place Direct Boration Valve, BAM-143, control switch to AUTO.

Place Makeup Mode selector switch to BORATE.

Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

Steps for Emergency Boration:

Place Makeup Mode selector switch to MANUAL.

Align borated water source by performing one of the following:

o Initiate Emergency Boration using Boric Acid Pump as follows:

o Open Emergency Boration Valve, BAM-133.

o Start one Boric Acid Pump.

o Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A BAM-126B Boric Acid Makeup Pump Recirc Valve B OR o Initiate Emergency Boration using Gravity Feed as follows:

o Open the following Boric Acid Makeup Gravity Feed valves:

o BAM-113A Boric Acid Makeup Gravity Feed Valve A o BAM-113B Boric Acid Makeup Gravity Feed Valve B o Close VCT Disch Valve, CVC-183.

Verify at least one Charging Pump operating and Charging Header flow 40 GPM.

This manipulation is performed at CP-4. The ATC should use the Reactivity Worksheet to recommend a boron quantity to the CRS.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 8 of 14 Event

Description:

Main Condenser Leak, Lowering Main Condenser Vacuum, Rapid Plant Power Reduction Time Position Applicants Actions or Behavior ATC Perform Boron Equalization as follows:

Place available Pressurizer Pressure Backup Heater Control Switches to ON.

Reduce Pressurizer Spray Valve Controller (RC-IHIC-0100) setpoint potentiometer to establish spray flow and maintain RCS pressure 2250 PSIA (2175 - 2265).

This manipulation is performed at CP-2.

ATC Operate CEAs to maintain ASI using CEA Reg. Group 6 or Group P Control Element Assemblies.

Operate CEAs in Manual Group mode as follows:

Position Group Select switch to desired group.

Place Mode Select switch to MG.

Operate CEA Manual Shim switch to INSERT CEA Group P should be used first to a low limit of 120 inches, followed by CEA Group 6 to a low limit of 120 inches to comply with Tech Spec 3.1.3.6.

This manipulation is performed at CP-2.

Crew Maintain RCS Cold Leg Temperature 536°F to 549°F.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 4 Page 9 of 14 Event

Description:

Main Condenser Leak, Lowering Main Condenser Vacuum, Rapid Plant Power Reduction Time Position Applicants Actions or Behavior BOP Commence Turbine load reduction by performing the following:

Depress LOAD RATE MW/MIN pushbutton.

Set selected rate in Display Demand Window.

Depress ENTER pushbutton.

Depress REFERENCE pushbutton.

Set desired load in Reference Demand Window.

Depress ENTER pushbutton.

Depress GO pushbutton.

This manipulation is performed at CP-1. The BOP will set up the Main Turbine controls. The ATC will direct the BOP when to commence unloading the Main Turbine based on the drop in RCS Cold Leg temperature.

Examiner Note This event is complete when the desired power reduction has been accomplished Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5/6 Page 10 of 14 Event

Description:

Small Break LOCA, CCW lost to Reactor Coolant Pumps, CS-125 B Stuck Closed, Low Pressure Safety Injection Pump A Fails to Auto Start on SIAS.

Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of RCS Leak.

Alarms Containment Water Leakage Hi (Cabinet N, L-20)

Containment Water Leakage Hi-Hi (Cabinet N, K-20)

Class 1E Rad Monitoring Sys Activity Hi-Hi (Cabinet SA, K-4)

Indications Lowering Pressurizer level.

Lowering Pressurizer pressure.

Backup Charging Pumps auto-start until all 3 Charging Pumps are running.

ATC If directed by CRS, trip Reactor using 2 Reactor Trip pushbuttons at CP-2.

ATC If directed by CRS, initiate Safety Injection Actuation (SIAS) and Containment Isolation Actuation (CIAS) at CP-7.

Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow.

ATC Following initiation of SIAS (auto or manual) secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop.

This is done at CP-2. This step is required here because of the earlier failure of CCW level switch CC-ILS-7013A. At this point, the RCPs will be operating with no CCW flow.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 5/6 Page 11 of 14 Event

Description:

Small Break LOCA, CCW lost to Reactor Coolant Pumps, CS-125 B Stuck Closed, Low Pressure Safety Injection Pump A Fails to Auto Start on SIAS.

Time Position Applicants Actions or Behavior The ATC and BOP operators will perform the required post trip verifications. Specific actions required during these activities will be specifically listed.

BOP Reset Moisture Separator Reheaters by pushing the RESET button located on CP-1.

BOP Start Low Pressure Safety Injection Pump A at CP-8.

BOP Secure AH-12 A or B on CRS direction after initiation of SIAS at CP-18.

BOP After Containment Spray is initiated (CSAS), attempt to open CS-125 B.

Valve will not open when this action is taken.

Verify > 1750 GPM Containment Spray flow on Train B.

This action is taken at CP-8.

CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-002, Loss of Coolant Accident Recovery Crew When Containment Temperature rises above 200 F, update crew on need to use bracketed parameters due to harsh environment in Containment.

CRS During brief in OP-902-002, should discuss necessary strategy of using Steam Generators to cool RCS.

Examiner Note This event is complete after entry into OP-902-002. It is not necessary to allow a brief to occur at this point. SIAS initiation verification and CSAS verification should occur before moving forward. ATC should have secured RCPs and the BOP should have started LPSI Pump A.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 12 of 14 Event

Description:

Containment Spray Pump A Trips, Entry into OP-902-008, Safety Function Recovery Procedure Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of Containment Spray Pump A trip.

Alarms CNTMT Spray Pump A Unavailable (Cabinet M, A-4)

CNTMT Spray Pump A Trip/Trouble (Cabinet M, B-4)

CNTMT Spray Hdr A Flow Lo (Cabinet M, F-4)

Indications Amber light on Containment Spray Pump A control switch.

No Containment Spray flow indicated on CS-IFI-.7122 A.

CRS Recognize the Containment Temperature and Pressure Control safety function is not met. Exit OP-902-002 and enter OP-902-008, Functional Recovery procedure.

BOP Place Hydrogen Analyzers in service as follows:

Train A o Place Train A H2 ANALYZER CNTMT ISOL VALVE keyswitch to OPEN.

o Place H2 ANALYZER A POWER to ON.

Train B o Place Train B H2 ANALYZER CNTMT ISOL VALVE keyswitch to OPEN.

o Place H2 ANALYZER B POWER to ON.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 13 of 14 Event

Description:

Containment Spray Pump A Trips, Entry into OP-902-008, Safety Function Recovery Procedure Time Position Applicants Actions or Behavior CRS Identify success paths to be used and prioritize Safety Functions.

Proper prioritization will result in Containment Isolation being priority 1 and Containment Temperature and Pressure Control being priority 2.

The CRS should address Containment Isolation by overriding CS-125 A closed. OP-902-008, section CI-1 will direct this to be accomplished in accordance with OP-902-009, Standard Appendices, Attachment 21-A.

The CRS should address Containment Temperature and Pressure Control by aligning Low Pressure Safety Injection Pump A to replace the failed Containment Spray Pump A. OP-902-008, section CTPC, Continuing Actions will direct this to be accomplished in accordance with OP-902-009, Standard Appendices, Attachment 28.

These actions should be pursued in parallel. The CR may choose to prepare for, but not close, CS-125 A, pending attempts to accomplish aligning LPSI Pump A.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 1 Event # 7 Page 14 of 14 Event

Description:

Containment Spray Pump A Trips, Entry into OP-902-008, Safety Function Recovery Procedure Time Position Applicants Actions or Behavior Critical Task Establish Containment temperature and pressure control.

This task is satisfied by aligning LPSI Pump A to replace CS Pump A prior to exiting the Containment Temperature and Pressure Control safety function in OP-902-008.

BOP Align LPSI Pump A to replace CS Pump A as follows:

Verify LPSI Pump A control switch in OFF.

Verify Containment Spray Pump A control switch in OFF.

Place SI-129A, LPSI FLOW CONTROL VALVE to AUTO.

Place SI-IFIC-0307, LPSI FLOW CONTROLLERS HEADER 2A/2B in MAN.

Adjust SI-IFIC-0307, LPSI FLOW CONTROLLERS HEADER 2A/2B to 0% output.

Verify the following valves Closed:

o SI-415A, LPSI SHUTDOWN TEMP CONTROL valve.

o SI-138A, LPSI FLOW CONTROL COLD LEG 2B.

o SI-139A, LPSI FLOW CONTROL COLD LEG 2A.

Open SI-125A/SI-412A, SHDN HX A ISOL valves.

Verify CS-125A, CNTMT SPRAY HEADER ISOL valve open.

Start LPSI Pump A.

Examiner Note This event is complete after aligning LPSI Pump A to replace CS Pump A Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 2 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 100%, MOC, AB buses aligned to B side.

Protected Train is B Emergency Diesel Generator A is tagged out for planned maintenance.

Turnover: Maintain 100 % power Event Malf. No. Event Type* Event No. Description 1 C- ATC Swap Charging Pump using OP-002-005.

C - SRO Charging Pump B develops oil leak.

TS - SRO 2 RC22 B1 I - BOP Pressurizer narrow range safety pressure I - SRO instrument RC-IPI-0101 B fails high TS - SRO 3 SG05 B I - BOP Steam Generator #2 level instrument, I - SRO SG ILR1106, fails low.

4 TPR13, 14 R - ATC Main Generator Stator Coil Water temperature N - BOP high, normal plant downpower N - SRO 5 TU01A, D, M - All Main Turbine High Vibration and Reactor Trip R

6 RD11A-10 C-ATC 2 CEAs stuck out requiring Emergency Boration RD11A-22 C - SRO 7 ED01 M-All Loss of Off Site Power A, B, C, D 8 EG08B C- BOP EDG B fails to auto-start C - SRO

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 2 Rev 0 Scenario Event Description NRC Scenario 2 The crew assumes the shift at 100% power with instructions to maintain 100% power.

OP-903-003, Charging Pump Operability Check is scheduled for night shift.

The shift manager has instructed the control room supervisor to swap Charging Pumps leaving Charging Pump B running and Charging Pump AB secured and in auto. After starting Charging Pump B, the watchstander will call and report an oil leak, recommending Charging Pump B be secured. With Charging Pump B control switch in off and inoperable and with Emergency Diesel Generator A tagged out, the SRO should recognize that Tech Spec 3.8.1.1.d is no longer met. Additionally, Tech Spec and TRM 3.1.2.4 must be entered. Tech Spec 3.1.2.4 is a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action and TRM 3.1.2.4 is a 7 day action. By aligning Charging Pump AB to replace Charging Pump B, Tech Spec 3.8.1.1.d will be satisfied and Tech Spec 3.1.2.4 can be exited. The CRS should stay in TRM 3.1.2.4.

After the ATC aligns Charging Pump AB or at the lead examiners direction, Pressurizer narrow range safety pressure instrument RC-IPI-0101 B fails high. After identifying the failure, the SRO should enter Tech Spec 3.3.1. The BOP should be directed to bypass the PPS bistables for High Pressurizer Pressure, Low DNBR, and High LPD within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After the crew bypasses the appropriate bistables, Steam Generator #2 level instrument, SG ILR1106, Steam Generator 2 Downcomer Level (red pen), fails low.

The controllers for Main Feedwater Regulating Valve 2, Startup Feedwater Regulating Valve 2, and Main Feedwater Pump B transfer to manual. The crew should enter OP-901-201, Steam Generator Level Control Malfunction. No Tech Spec entries are required and no actions by the Balance of Plant operator are necessary at this time.

After the crew has completed their brief, PMC alarms will come in for Main Generator Stator Coil Water hose temperatures. The crew should enter OP-901-211, Generator Malfunction. Using Attachment 1, SCW High Temperature, the crew will determine the need to commence a normal plant shutdown in accordance with OP-010-005. Due to the earlier Steam Generator level instrument failure, the BOP operator will have to control Steam Generator level in manual for Steam Generator #2. The ATC will perform direct Boration to the RCS as well as ASI control with CEAs and Pressurizer boron equalization. The BOP will manipulate the Main Turbine controls to reduce load.

Scenario 2 Rev 0 Scenario Event Description NRC Scenario 2 Once the crew has commenced the power reduction and lowered power to ~ 90%, or at the lead examiners discretion, high vibration alarms will come in on the Main Turbine.

Using annunciator response procedure OP-500-001, Control Room Cabinet A, and OP-901-210, Turbine Trip, the SRO should direct a Reactor trip. The crew should enter OP-902-000, Standard Post Trip Actions, and work this procedure concurrent with the Turbine Trip off normal procedure. OP-901-210 will direct breaking Main Condenser vacuum. On the Reactor Trip, 2 CEAs will stick out, requiring the ATC operator to Emergency Borate. The BOP will have to establish Feedwater Control Reactor Trip Override conditions manually on Steam Generator #2 due to the earlier level instrument failure.

The SRO should direct the BOP to continue with the actions to break Main Condenser vacuum. The crew should diagnose into OP-902-006, Loss of Main Feedwater Recovery, and secure 2 Reactor Coolant Pumps. After 2 RCPs are secured and the BOP has commenced breaking vacuum, a loss of off site power occurs. Emergency Diesel Generator B will fail to auto-start on the LOOP and the BOP will be required to start EDG B. The crew will transition to OP-902-003, Loss of Off Site Power/Loss of Forced Flow Recovery procedure. During the scenario, environmental conditions will have rain occurring. After the LOOP, the high level alarms will come in for Dry Cooling Tower 1 and 2 Sumps. The CRS will direct the performance of OP-902-009, Standard Appendices, Appendix 20, Operation of DCT Sump Pumps.

The scenario can be terminated after the CRS orders the performance of OP-902-009 Appendix 20 or at the lead examiners discretion.

Scenario 2 Rev 0 Scenario Event Description NRC Scenario 2 Critical Tasks

1. Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification. The required task becomes applicable after the Reactor is tripped and 2 CEAs remain stuck out following event 5.

2. Energize at least one vital electrical AC bus.

This task is satisfied by starting Emergency Diesel Generator B. This task becomes applicable following the loss of off site power triggered in event 7.

Scenario 2 Rev 0 NRC Scenario 2 Scenario Notes:

A. Reset Simulator to IC-92.

B. Verify the following Scenario Malfunctions:

1. rc22b1 for Pressurizer pressure narrow range instrument RC-IPI-0101 B

2. sg05b for S/G #2 level instrument SG-ILR1106

3. tu01a for Main Turbine bearing 1X vibration

4. tu01d for Main Turbine bearing 4X vibration

5. tu01r for Main Turbine bearing 7Y vibration

6. ed01a for Off Site Feeder Breaker 7172

7. ed01b for Off Site Feeder Breaker 7176

8. ed01c for Off Site Feeder Breaker 7182

9. ed01d for Off Site Feeder Breaker 7186

10. rd11a22 for CEA 22 stuck

11. rd11a10 for CEA 10 stuck

12. eg10a for EDG A overspeed device

13. eg08b for EDG B fail to auto start C. Verify the following Remotes

1. epr09a for precipitation at 1.5 inches per hour

2. egr29a for EDG A Output breaker racked out

3. tpr14 for TCW to SCW to Main Generator closed

4. tpr13 for TCW to SCW bypass for Main Generator 30% open D. Verify the following Annunciators

1. b_e07 for Dry Cooling Tower 1 level high

2. b_e08 for Dry Cooling Tower 2 level high E. Verify the following Control Board Conditions:

1. Danger tag placed on EDG A control switch

2. Danger tag placed on EDG A Output Breaker F. Ensure Protected Train B sign is placed in SM office window.

G. Verify EOOS is 9.8 Green H. Complete the simulator setup checklist.

Scenario 2 Rev 0 NRC Scenario 2 Simulator Booth Instructions Event 1 Charging Pump B Develops an Oil Leak after Start

1. If the RCA watch is called, report that the oil leak is coming from a crack in the oil filter and that he recommends securing Charging Pump B.

2. If Work Week Manager is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 Pressurizer pressure instrument RC-IPI-0101 B fails high

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Steam Generator #2 level instrument, SG ILR1106, fails low

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Main Generator Stator Coil Water temperature high, normal plant downpower

1. On Lead Examiner's cue, initiate Event Trigger 4.

2. After crew enters OP-901-211, Generator Malfunction, change the severity of TPR13 to 20%.

3. If called as TGB watch report all conditions are normal at the Stator Coil Water skid, except temperatures are elevated.

4. If called as TGB watch regarding TC-332, SCW Temperature Control Valve, report that you cant determine if the control valve is moving. The manual bypass valve TC-335 is stuck shut.

5. If called as TGB watch to check the Generator Condition Monitor, report that all indications are normal and there are no alarms.

6. If the Work Week Manager or Duty Plant Manager is called, inform the caller that they will make the necessary calls.

Scenario 2 Rev 0 NRC Scenario 2 Event 5 Main Turbine high vibration and Reactor Trip

1. After the crew has reduced power to ~90% or on the Lead Examiner's cue, initiate Event Trigger 5.

2. If called as TGB watch report elevated vibrations in the Turbine Building.

Event 6 2 CEAs stuck out requiring Emergency Boration

1. No calls for this malfunction should occur.

Event 7 Loss of Off Site Power with rain conditions

1. After the crew has secured 2 Reactor Coolant Pumps and the BOP has commenced breaking Main Condenser vacuum, and on the Lead Examiner's cue, initiate Event Trigger 6.

2. If called as OSW watch report that a steady rain has been falling all shift.

3. If called to come to the Control Room to get a copy of OP-902-009, Appendix 20 for aligning the DCT Sump Pumps, report that you have a copy at the 314 Bus.

Event 8 Emergency Diesel Generator B fails to start

1. If called as RAB watch to check EDG B, initiate Trigger 9, and when the EDG B Trouble alarm is clear, report that it is running satisfactorily.

Scenario 2 Rev 0 NRC Scenario 2 Scenario Timeline:

Ramp Time Event Malfunction Severity Delay Trigger HH:MM:SS (Min)

EPR09 A 1.5 N/A N/A Precipitation at 1.5 inches per hour EGR29 A N/A NA NA EDG A output breaker racked out EG10 A N/A NA NA EDG A overspeed device pulled 1 N/A N/A N/A N/A N/A N/A N/A Charging Pump B oil leak 2 RC22 B1 N/A 00:00:00 N/A 2 Pressurizer narrow range safety pressure instrument RC-IPI-0101 B fails high 3 SG05 B 0% N/A N/A 3 Steam Generator #2 level instrument, SG-ILR-1106, fails low 4 TPR14 Closed 00:00:00 N/A 4 TPR13 30% 00:00:00 Main Generator Stator Coil Water temperature high 5 TU01a 9 mils 00:00:10 00:00:10 5 TU01d 9 mils 00:00:10 00:00:20 TU01r 12 mils 00:01:00 00:00:30 Main Turbine High Vibration and Reactor Trip 6 RD11A 10 N/A N/A N/A N/A CEA 10 stuck out after reactor trip 6 RD11A 22 N/A N/A N/A N/A CEA 22 stuck out after trip 7 ED01 A, B, N/A N/A N/A 6 C, D Loss of Off Site Power 8 EG08 B N/A N/A N/A N/A EDG B Fail to Auto-start 9 B_E07 N/A N/A 00:02:30 6 B_E08 00:04:00 Dry Cooling Tower 1 & 2 high level alarm Scenario 2 Rev 0 NRC Scenario 2

REFERENCES:

Event Procedures 1 OP-002-005, Chemical and Volume Control Tech Spec 3.1.2.4 TRM 3.1.2.4 2 OP-903-013, Monthly Channel Checks Tech Spec 3.3.1 3 OP-901-201, Steam Generator Level Control Malfunction 4 OP-901-211, Generator Malfunction OP-010-005, Plant Shutdown 5 OP-500-001, Control Room Cabinet A OP-901-210 Turbine Trip OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-006, Loss of Main Feedwater Recovery Procedure 6 OP-902-000, Standard Post Trip Actions 7 OP-902-003, Loss of Off Site Power/Loss of Forced Flow OP-902-009, Standard Appendices, Appendix 20, Operation of Dry Cooling Tower Sump Pumps 8 OP-902-000, Standard Post Trip Actions OP-902-003, Loss of Off Site Power/Loss of Forced Flow Scenario 2 Rev 0 Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 1 of 11 Event

Description:

Swap Charging Pumps using OP-002-005. Charging Pump develops an oil leak.

Time Position Applicants Actions or Behavior ATC If a Pressurizer Backup Heater Bank is operating, then secure the Pressurizer Backup Heater Bank by placing control switch in AUTO.

This step is at the CRS discretion.

Backup Heater Bank 3 will be operating.

Control switch is located on CP-2.

ATC Start Charging Pump B by placing the control switch to ON.

Located on CP-4.

ATC Secure Charging Pump AB by placing control switch to OFF Located on CP-4.

ATC Place Standby Charging Pump A control switch to OFF Located on CP-4.

ATC Place Standby Charging Pumps selector switch to AB - A.

Located on CP-4.

ATC Place control switches for Standby Charging Pumps A and AB to AUTO.

Located on CP-4.

Booth Operator calls as NAO and reports oil leak, recommends securing Charging Pump B.

CRS Direct ATC to start Charging Pump AB and secure Charging Pump B.

ATC Start Charging Pump AB by placing the control switch to ON.

Located on CP-4.

ATC Secure Charging Pump B by placing control switch to OFF Located on CP-4.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 1 Page 2 of 11 Event

Description:

Swap Charging Pumps using OP-002-005. Charging Pump develops an oil leak.

Time Position Applicants Actions or Behavior ATC Place Standby Charging Pump A control switch to OFF Located on CP-4.

ATC Place Standby Charging Pumps selector switch to A - B.

Located on CP-4.

ATC Place control switch for Standby Charging Pump A to AUTO.

Located on CP-4.

CRS Declare Charging Pump B inoperable. With EDG A already inoperable, Tech Spec action 3.8.1.1.d is no longer met. Enter Tech Spec and TRM 3.1.2.4.

Tech Spec action 3.8.1.1.d and Tech Spec 3.1.2.4 will be satisfied when Charging Pump AB Assignment switch is placed in the B position.

The CRS should direct this manipulation.

ATC Place the Charging Pump AB Assignment switch to B.

Located on CP-4.

Examiner Note This event is complete after Tech Spec 3.8.1.1.d and Tech Spec and TRM 3.1.2.4 have been addressed with Charging Pump AB assigned to replace B Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 2 Page 3 of 11 Event

Description:

Pressurizer narrow range safety pressure instrument RC-IPI-0101 B fails high Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS CHANNEL TRIP LOCAL PWR DENSITY HI (Cabinet K, A-11)

RPS CHANNEL TRIP DNBR LO (Cabinet K, A-12)

RPS CHANNEL TRIP PZR PRESSURE HI (Cabinet K, A-15)

PZR PRESSURE HI PRETRIP B/D (Cabinet K, C-15)

RPS CHANNEL B TROUBLE (Cabinet K, F-18)

Indications Pressurizer pressure instrument RC-IPI-0101 B on CP-7 indicates pegged high.

CRS Review Tech Specs based on the failed instrument.

Enter Tech Spec 3.3.1 Direct bypassing Channel B bistables 3, 4, and 5 for Hi Pressurizer Pressure, Low DNBR, and Hi local Power. This is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action.

BOP Bypass Channel B bistables 3, 4, and 5 for Hi Pressurizer Pressure, Low DNBR, and Hi local Power Located on CP-10 B (rear panel).

Examiner Note This event is complete when the proper bistables have been bypassed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 3 Page 4 of 11 Event

Description:

Steam Generator #2 level instrument, SG ILR1106, fails low.

Time Position Applicants Actions or Behavior BOP Recognize and report indications of failed Steam Generator level instrument.

Alarms SG 2 FW Contl Lvl Signal Dev/Pwr Lost (Cabinet F, T-19)

SG 2 Level Hi/Lo (Cabinet F, U-18)

Indications Controllers for Steam Generator #2 shift to MANUAL.

o Main Feedwater Regulating Valve B Controller.

o Startup Feedwater Regulating Valve B Controller o Main Feedwater Pump B Speed Controller CRS Enter and direct the implementation of OP-901-201, Steam Generator Level Control Malfunction.

Coverage of the flow chart in Attachment 1 should conclude that the failed instrument is the problem.

The CRS should discuss with the BOP necessary contingency actions necessary with the listed controllers in MANUAL. This should include actions on a Reactor trip or on Steam Generator High Level Override.

Examiner Note This event is complete when the flow chart in Attachment 1 has been completed and the contingencies have been discussed.

Or As directed by the Lead Evaluator This failure requires additional actions by the BOP during the power reduction and after the Reactor trip later in the scenario.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 5 of 11 Event

Description:

Main Generator Stator Coil Water temperature high, normal plant downpower Time Position Applicants Actions or Behavior ATC/BOP Recognize and report indications of rising SCW temperature.

Alarms PMC alarms for various Stator Coil Water Hose temperatures GEN AUXILIARY SYS TROUBLE (Cabinet D, G-8)

Indications PMC point for Stator Coil Water Hose temperature rising.

CRS Enter and direct the implementation of OP-901-211, Generator Malfunction.

Proper use of section E0, General should result in the use of Attachment 1, SCW High Temperature.

CRS Based on use of Attachment 1, direct a power reduction using OP-010-005, Plant Shutdown.

Direct lowering Main Generator reactive load to approximately 0.

BOP Lower Main Generator reactive load to approximately 0 using the Main Generator Voltage Adjust Regulator control switch on CP-1.

ATC Perform Boron Equalization as follows:

Place available Pressurizer Pressure Backup Heater Control Switches to ON.

Reduce Pressurizer Spray Valve Controller (RC-IHIC-0100) setpoint potentiometer to establish spray flow and maintain RCS pressure 2250 PSIA (2175 - 2265).

This manipulation is performed at CP-2.

ATC Begin Direct Boration as follows:

Set Boric Acid Makeup Batch Counter to volume of Boric Acid desired.

Place Direct Boration Valve, BAM-143, control switch to AUTO.

Place Makeup Mode selector switch to BORATE.

Adjust Boric Acid Flow controller, BAM-IFIC-0210Y, output to >3 GPM flow rate.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 6 of 11 Event

Description:

Main Generator Stator Coil Water temperature high, normal plant downpower Time Position Applicants Actions or Behavior This manipulation is performed at CP-4. The ATC should use the Reactivity Worksheet to recommend a boron quantity to the CRS.

ATC Operate CEAs to maintain ASI using CEA Reg. Group 6 or Group P Control Element Assemblies.

Operate CEAs in Manual Group mode as follows:

Position Group Select switch to desired group.

Place Mode Select switch to MG.

Operate CEA Manual Shim switch to INSERT CEA Group P should be used first to a low limit of 120 inches, followed by CEA Group 6 to a low limit of 120 inches to comply with Tech Spec 3.1.3.6.

This manipulation is performed at CP-2.

Crew Maintain RCS Cold Leg Temperature 536°F to 549°F.

BOP Commence Turbine load reduction by performing the following:

Depress LOAD RATE MW/MIN pushbutton.

Set selected rate in Display Demand Window.

Depress ENTER pushbutton.

Depress REFERENCE pushbutton.

Set desired load in Reference Demand Window.

Depress ENTER pushbutton.

Depress GO pushbutton.

This manipulation is performed at CP-1. The BOP will set up the Main Turbine controls. The ATC will direct the BOP when to commence unloading the Main Turbine based on the drop in RCS Cold Leg temperature.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 4 Page 7 of 11 Event

Description:

Main Generator Stator Coil Water temperature high, normal plant downpower Time Position Applicants Actions or Behavior BOP Because of the earlier failure of SG-ILI-1106, Main Feedwater controls for Steam Generator #2 must be controlled in MANUAL.

The BOP operator will adjust controller FW-IHIC-1121 for Main Feed Regulating Valve #2 as power is reduced to maintain Steam Generator #2 level 50-70% narrow range.

This manipulation is on CP-1.

Examiner Note This event is complete when the desired power reduction has been accomplished Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5/6 Page 8 of 11 Event

Description:

Main Turbine High Vibration and Reactor Trip with 2 CEAs stuck out Time Position Applicants Actions or Behavior ATC/BOP Recognize and report indications of rising Main Turbine vibration..

Alarms Turbine Rotor Vibration High (Cabinet A, L-6)

DIFF EXP/VIBR HI TURBINE TRIP DISABLED (Cabinet E, D-6)

Indications Rising vibration indication on the PMC and on recorder TUR-IUR-4302.

CRS Direct trip of the Main Turbine and enter OP-901-210, Turbine Trip.

CRS could go directly to reactor trip.

BOP Push the THINK and TURBINE TRIP pushbuttons on CP-1.

CRS Direct Reactor Trip and direct ATC and BOP to carry out Standard Post trip Actions.

ATC Trip Reactor using 2 Reactor Trip pushbuttons at CP-2.

Critical Task Establish reactivity control.

This task is satisfied by establishing Emergency Boration prior to completing Standard Post Trip Actions Reactivity Control verification.

ATC Determine 2 CEAs are stuck out and commence Emergency Boration:

Place Makeup Mode selector switch to MANUAL.

Align borated water source by performing one of the following:

o Initiate Emergency Boration using Boric Acid Pump as follows:

o Open Emergency Boration Valve, BAM-133.

o Start one Boric Acid Pump.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5/6 Page 9 of 11 Event

Description:

Main Turbine High Vibration and Reactor Trip with 2 CEAs stuck out Time Position Applicants Actions or Behavior o Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A BAM-126B Boric Acid Makeup Pump Recirc Valve B OR o Initiate Emergency Boration using Gravity Feed as follows:

o Open the following Boric Acid Makeup Gravity Feed valves:

o BAM-113A Boric Acid Makeup Gravity Feed Valve A o BAM-113B Boric Acid Makeup Gravity Feed Valve B Close VCT Disch Valve, CVC-183.

Verify at least one Charging Pump operating and Charging Header flow 40 GPM.

BOP Establish Reactor Trip Override on Feedwater Control System for Steam Generator #2:

Close Main Feedwater Regulating Valve #2.

Throttle Startup Feedwater Regulating Valve #2 to 13-21% open.

This is done at CP-1.

BOP Reset Moisture Separator Reheaters by pushing the RESET button located on CP-1.

CRS Direct BOP to continue with step 5 of OP-901-210 and break Main Condenser Vacuum.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 5/6 Page 10 of 11 Event

Description:

Main Turbine High Vibration and Reactor Trip with 2 CEAs stuck out Time Position Applicants Actions or Behavior BOP Carryout the actions in OP-901-210 to break Main Condenser Vacuum:

Place STEAM BYPASS MASTER controller, MS-IPIC-1010 in MAN.

Slowly lower output of STEAM BYPASS MASTER controller to zero.

Close BOTH Main Steam Isolation Valves (MS 124A AND MS 124B)

Open Condenser Vacuum Breaker valves by simultaneously depressing THINK push button AND placing CNDSR VAC BKR control switch to OPEN.

WHEN the following annunciators alarm, THEN secure respective Condenser Vacuum pumps by placing the CONDENSER VACUUM PUMP control switch to STOP for 5 seconds:

o Condenser Vacuum Pump A AUTO START (Cabinet E, E-1) o Condenser Vacuum Pump B AUTO START (Cabinet E, E-2) o Condenser Vacuum Pump C AUTO START (cabinet E, E-3).

All steps are performed at CP-1 except closing the Main Steam Isolation Valves, which is done at CP-8.

CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-006, Loss of Main Feedwater Recovery CRS Direct ATC to secure RCP 1A and 2A.

ATC When directed, secure Reactor Coolant Pumps 1A and 2A as follows:

Place each RCP control switch on CP-2 to stop.

Examiner Note This event is complete after OP-902-006 has been entered and 2 Reactor Coolant Pumps have been secured.

Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 2 Event # 7 Page 11 of 11 Event

Description:

Loss of Off Site Power Time Position Applicants Actions or Behavior Alarms Multiple alarms associated with LOOP Indications Numerous alarms come in, all lights go off except for Control Room emergency lighting..

Emergency Diesel Generator B control switch remains green.

Critical Task Energize at least one vital electrical AC bus.

This task is satisfied by starting Emergency Diesel Generator B with 15 minutes of the LOOP.

BOP Start Emergency Diesel Generator B at CP-1.

CRS Return to OP-902-009, Attachment 1, Diagnostic Flowchart.

Proper use of chart will result in use of OP-902-003, Loss of Off Site Power/Loss of Forced Flow Recovery.

CRS Upon review of step 4 combined with alarms DRY CLNG TOWER SUMP 1 LEVEL HI and DRY CLNG TOWER SUMP 2 LEVEL HI on panel B, then direct performance of OP-902-009, Appendix 20, Operation of DCT Sump Pumps.

Examiner Note This event is complete after directing performance of Appendix 20 Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 3 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 1.2 % Power Power ascension is being held pending Main Feedwater Pumps governor adjustment Preparations are being made to start Main Feedwater Pump A AB Bus is aligned to the A side Turnover: OP-903-052 for CVAS Train A will go late this shift. Complete OP-903-052, section 10.1.

OP-007-004, attachment 11.4 is in the field to discharge Waste Condensate Tank A.

Event Malf. No. Event Type* Event No. Description 1 DI-18A4s27-1 N - BOP During performance of OP-903-052, CVAS N - SRO Fan A will fail to start.

TS - SRO 2 AO-04A3a12c-1 C - ATC Waste Condensate Tank A flow controller C - SRO LWM-IFIC-0647 output fails high 3 CH08-A1 I - BOP Containment pressure instrument I - SRO CB-IPT-6701-SMC fails high TS - SRO 4 RX14-A I - ATC Pressurizer pressure instrument I - SRO RC-IPR-0100 X fails low 5 RX06-D1 C - BOP Steam Bypass Valve MS-320 A fails open C - SRO 6 FW38-B M - ALL Main Feedwater line break in Containment.

7 RP08G C - BOP Main Feedwater Isolation Valve #1 FW-184A C - SRO fails to automatically close on MSIS.

8 CV34a1 C - ATC CVC-109 fails to auto close on CIAS.

C - SRO

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 3 Rev 1 Scenario Event Description NRC Scenario 3 The crew assumes the shift at 1.2 % power. Reactor Engineer has completed Low Power Physics Testing. I&C Techs are making adjustments to Main Feedwater Pumps A & B governors based on vendor recommendations. The estimated time to completion is less than 60 minutes. When this is complete, Main Feedwater Pump A will be started and power ascension will commence.

Last shift, it was discovered that OP-903-052, CVAS Operability Test, will exceed its Tech Spec late date this shift. You have been directed to start CVAS Train A in accordance with OP-903-052. This surveillance will have the BOP operator secure RAB Normal Supply and Normal Exhaust Fans A and start CVAS Fan A. After securing both normal ventilation fans, CVAS Fan A will fail to start. This will require entering Tech Spec 3.7.7, a 7 day action requirement. RAB Normal Supply and Normal Exhaust Fans A will have to be re-started.

After the failure of CVAS Fan A, the RCA watch will call and report that he has completed his lineup and is ready for the ATC to perform his actions to discharge Waste Condensate Tank A and is ready for the ATC to continue with step 6.10.7. When the ATC initiates flow on step 6.10.10, LWM-IFIC-0647 will fail high, raising flow in excess of 50 gpm, the discharge permit limit. The ATC should close LWM-441 and LWM-442 from CP-4 to terminate the release.

After the release is secured, Containment pressure instrument CB-IPT-6701 SMC fails high. The SRO should enter Tech Spec 3.3.1 and 3.3.2 and the BOP should bypass PPS bistables 13 and 16.

After the appropriate bistables are bypassed, Pressurizer pressure instrument RC-IPR-0100 X fails low. This causes Pressurizer Backup and Proportional heaters to energize.

The SRO should enter OP-901-120, Pressurizer Pressure Malfunction. The ATC will select the non-faulted Pressurizer pressure channel.

After the Pressurizer Pressure Control Channel Y is selected, Steam Bypass valve MS-320A controller will begin to fail high. The crew should respond to the cooldown and reactivity effects by taking manual control of MS-320 A and closing it.

After MS-320 A is closed, a Main Feedwater line break occurs in Containment. If the control room supervisor directs a reactor trip based on the un-controlled power rise, then the trigger for the Main Feedwater line break will be inserted while the crew is performing their standard post trip actions. After the malfunction is inserted, the Main Feedwater Isolation Valve #1 fails to automatically close on the MSIS and must be closed manually by the BOP operator. CVC-109 fails to automatically close on the CIAS and must be manually closed by the ATC operator. The crew should enter OP-902-004, Excess Steam Demand Recovery.

Scenario 3 Rev 1 Scenario Event Description NRC Scenario 3 Actions to address pressurized thermal shock should be taken when CET temperature and Pressurizer pressure start to rise. This can be accomplished using OP-902-009, Appendix 13 or with OP-902-004, based on whether or not the crew has diagnosed into OP-902-004 before those parameters start to rise. The scenario can be terminated after PTS actions have been accomplished or at the lead examiners discretion.

The conditions in this scenario do not warrant declaration of any Emergency Plan Classification.

Scenario 3 Rev 1 Scenario Event Description NRC Scenario 3 Critical Tasks

1. Establish Containment Isolation This task is satisfied by taking action to close CVC-109. This task becomes applicable after the CIAS signal has been initiated.

2. Establish Containment Isolation This task is satisfied by taking action to close FW-184 A.

3. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of CSAS initiation.

4. Establish RCS temperature control This task is satisfied by taking action to stabilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1. Action to address this task should commence within 10 minutes after the applicable parameters begin to rise.

5. Establish RCS pressure control This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to address this task should commence within 10 minutes after the applicable parameters begin to rise.

Scenario 3 Rev 1 Scenario Event Description NRC Scenario 3 Scenario Notes:

A. Reset Simulator to IC-93.

B. Verify the following Scenario Malfunctions:

1. cho8a1 for Containment pressure CB-IPT-6701 SMC

2. rx14-A for Pressurizer pressure instrument RC-IPT-0100 X

3. fw38b for Main Feedwater line break.

4. rp08G for MFIV #1 failing to auto close

5. cv34a1 for CVC-109 failing to auto close C. Verify the following Remotes:

1. anr04h for EDG A local alarm acknowledgement.

2. anr04i for EDG B local alarm acknowledgement.

D. Verify the following Overrides:

1. di-18a4s27-1 for CVAS Fan A

2. di-04a3a12c-1 for LWM-IFIC-0647 E. Ensure Protected Train B sign is placed in SM office window.

F. Verify EOOS is 10.0 Green G. Complete the simulator setup checklist.

Scenario 3 Rev 1 Simulator Booth Instructions Event 1 CVAS Fan A Fails to Start

1. If Work Week Manager or Electrical Maintenance is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 2 LWM-IFIC-0647 Fails High during WCT A Discharge

1. After RAB Normal Ventilation is running, call the Control Room and report that you are the RCA Watch and you are standing by to discharge Waste Condensate Tank A. You have completed step 6.10.6 and you are ready for step 6.10.7.

2. After the ATC initiates flow, if called, report that you heard flow, then you heard the isolation valves close. All other indications were normal.

3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 Containment pressure instrument CB-IPT-6701 SMC fails high

4. After the release is secured, or on Lead Examiner's cue, initiate Event Trigger 3.

5. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 4 Pressurizer pressure instrument RC-IPR-0100 X fails low

1. After Tech Specs for Event 3 are addressed, or on Lead Examiner's cue, initiate Event Trigger 4.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 5 MS-320A Fails Open

1. After Pressurizer Pressure Channel Y is selected and on the Lead Examiner's cue, initiate Event Trigger 5.

2. If called as the watchstander and sent to MS-320A, report valve is stroking open.

3. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Scenario 3 Rev 1 Event 6 Main Feedwater Line Break in Containment

1. After MS-320A is closed or, if the CRS directs a reactor trip, during performance of SPTAs and on the Lead Examiner's cue, initiate Event Trigger 6.

2. If called as RAB watch to check EDG A & B, initiate Trigger 10, and when the EDG A & B Trouble alarms are clear, report that they are running satisfactorily.

Event 7 MFIV #1 Fails to Auto Close

1. If called to check MFIV #1 locally, report no visible problems locally.

Event 8 CVC-109 Fails to Auto Close

1. No communications should occur for this malfunction.

Scenario 3 Rev 1 Scenario Timeline:

Ramp Time Event Malfunction Severity Delay Trigger HH:MM:SS (Min) 1 di-18a4s27-1 N/A N/A N/A N/A CVAS Fan A fails to start 2 di-04A3a12c-1 N/A N/A N/A N/A LWM-IFIC-0647 Fails High during WCT A Discharge 3 CH08 A1 N/A N/A N/A 3 Containment pressure instrument CB-IPT-6701 SMC fails high 4 RX14 A 0% N/A N/A 4 Pressurizer pressure instrument RC-IPR-0100 X fails low 5 RX06 D1 N/A N/A N/A 5 MS-320 A fails open 6 FW38 B 50 % 00:01:00 N/A 6 Main Feedwater line break inside Containment 7 RP08G N/A N/A N/A N/A Main Feedwater Isolation Valve #1 fails to auto close on MSIS 8 CV34a1 N/A N/A N/A N/A CVC-109 fails to auto open on CIAS Scenario 3 Rev 1

REFERENCES:

Event Procedures 1 OP-903-052, Controlled Ventilation Area System Operability Check OP-002-010, Reactor Auxiliary Building HVAC and Containment Purge Tech Spec 3.7.7 2 OP-007-004, Liquid Waste Management 3 Tech Spec 3.3.1 and 3.3.2 4 Tech Spec 3.1.1.4 5 OP-901-120, Pressurizer Pressure Control Malfunction Tech Spec 3.2.8 6 OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart OP-902-004, Excess Steam Demand Recovery 7 OP-902-000, Standard Post Trip Actions 8 OP-902-000, Standard Post Trip Actions Scenario 3 Rev 1 Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 1 Page 1 of 10 Event

Description:

During performance of OP-903-052, CVAS Fan A will fail to start.

Time Position Applicants Actions or Behavior CRS Direct BOP to perform OP-903-052 and test CVAS Train A.

BOP Secure the RAB Normal Supply Fan, HVR0002A.

Secure the RAB Normal Exhaust Fan, HVR0009A.

Start Controlled Vent Area Exh Fan A, HVR0021A.

All control switches are located on CP-18.

Controlled Vent Area Exh Fan A will not start when the control switch is taken to start.

CRS Declare CVAS Train A inoperable and enter Tech Spec 3.7.7.

If the CRS does not direct the BOP operator to re-start RAB Normal Ventilation, the booth operator will give this prompt from the shift manager.

BOP Start RAB Normal Exhaust Fan, HVR0009A.

Verify >69,000 scfm as indicated on PID S52432.

Start RAB Normal Supply Fan, HVR0002A.

Examiner Note This event is complete after Tech Spec 3.7.7 has been addressed and RAB Normal Ventilation is running Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 2 Page 2 of 10 Event

Description:

Waste Condensate Tank A flow controller LWM-IFIC-0647 fails high Time Position Applicants Actions or Behavior ATC On CP-4, Reset Liquid Waste Discharge Flow Integrator to zero.

ATC Record WCT level and Liquid Waste Discharge Integrator reading on Liquid Release Permit.

ATC Position Liquid Waste Condensate Flow Control handswitch, LWM-441 and LWM-442, to Open.

ATC Adjust flow using Liquid Waste Condensate Flow Controller, LWM-IFIC-0647, not to exceed value specified on Liquid Release Permit.

Indications at execution of above step:

Discharge flow rate rises above 50 gpm after ATC releases the raise pushbutton.

Using the lower pushbutton does not lower flow.

ATC Position Liquid Waste Condensate Flow Control handswitch, LWM-441 and LWM-442, to close to stop release.

Examiner Note This event is complete when LWM-441 and LWM-442 are closed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 3 Page 3 of 10 Event

Description:

Containment pressure instrument CB-IPT-6701-SMC fails high Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS CHANNEL TRIP CNTMT PRESSURE HI (Cabinet K, A-17)

CNTMT PRESSURE HI PRETRIP A/C (Cabinet K, B-17)

ESFAS CHANNEL TRIP CNTMT PRESSURE HI (Cabinet K, L-17)

CNTMT PRESSURE HI ESFAS PRETRIP A/C (Cabinet K, M-17)

RPS CHANNEL C TROUBLE (Cabinet K, G-18)

Indications Containment pressure instrument CB-IPI-6701 SMC on CP-7 indicates pegged high.

CRS Review Tech Specs based on the failed instrument.

Enter Tech Spec 3.3.1 and 3.3.2.

Direct bypassing Channel B bistables 13 and 16 for Hi Containment Pressure, RPS and ESFAS. This is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action.

BOP Bypass Channel B bistables 13 and 16 for Hi Containment Pressure, RPS and ESFAS.

Located on CP-10 C (rear panel).

Examiner Note This event is complete when Tech Specs 3.3.1 and 3.3.2 have been addressed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 4 Page 4 of 10 Event

Description:

Pressurizer pressure instrument RC-IPR-0100 X fails low Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed pressure instrument.

Alarms PRESSURIZER PRESSURE HI/LO (Cabinet H, E-1)

PRESSURIZER PRESS SIGNAL DEVIATION (Cabinet H, F-1)

Indications Recorder RC-IPR-0100 red pen fails lo.

Controller RC-IPIC-0100 process fails lo.

All Pressurizer Backup Heaters energize.

Pressurizer Proportional Heaters go to full fire at 200 amps.

Pressurizer pressure rises to 2270 psia at which time all Pressurizer Backup and Proportional Heaters de-energize. Pressurizer pressure begins to drop from 2270 psia.

CRS Enter and direct the implementation of OP-901-120, Pressurizer Pressure Malfunction, and use sub-section E1, Pressurizer Pressure Control Channel Instrument Failure.

ATC Transfer Pressurizer pressure control to channel Y using Pressurizer Pressure Channel Selector control switch.

CRS Refer to Technical Specification 3.2.8.

Entry required if RCS pressure exceeded 2275 PSIA Pressure should not exceed 2270 PSIA.

Examiner Note This event is complete when Pressurizer pressure control has been transferred to Y.

Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 5 of 10 Event

Description:

Steam Bypass Valve MS-320A fails open Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of RCS cooldown.

Indications MS-319A modulates closed.

Lowering RCS Cold Leg temperature.

Rising Reactor power.

Permissives for Steam Bypass Control System clear.

As the RCS cools, Letdown flow goes to minimum and Charging Pump A and AB start.

There are 2 acceptable paths that the CRS may take on this malfunction.

One option would be to direct a Reactor trip based on power rising and the RCS cooldown. Another option would be to close MS-320A. Steps are included for either option.

BOP If directed by CRS, attempt to close MS-320A.

If attempted, controller MS-IHIC-320A will not close MS-320A If attempted, MS-320A control switch will close MS-320A if placed in OFF.

MS-320A controls are located on CP-1.

ATC If directed by CRS, trip Reactor using 2 Reactor Trip pushbuttons at CP-2.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

If the cause of RCS cooldown is not discovered and addressed, the CRS may direct initiation of MSIS and Emergency Boration.

ATC/BOP Initiate MSIS using 2 pushbuttons on either CP-7 or CP-8.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 5 Page 6 of 10 Event

Description:

Steam Bypass Valve MS-320A fails open Time Position Applicants Actions or Behavior ATC If directed to perform Emergency Boration:

Place Makeup Mode selector switch to MANUAL.

Align borated water source by performing one of the following:

o Initiate Emergency Boration using Boric Acid Pump as follows:

o Open Emergency Boration Valve, BAM-133.

o Start one Boric Acid Pump.

o Close recirc valve for Boric Acid Pump started:

BAM-126A Boric Acid Makeup Pump Recirc Valve A BAM-126B Boric Acid Makeup Pump Recirc Valve B OR o Initiate Emergency Boration using Gravity Feed as follows:

o Open the following Boric Acid Makeup Gravity Feed valves:

o BAM-113A Boric Acid Makeup Gravity Feed Valve A o BAM-113B Boric Acid Makeup Gravity Feed Valve B Close VCT Disch Valve, CVC-183.

Verify at least one Charging Pump operating and Charging Header flow 40 GPM.

Examiner Note This event is complete when MS-320A is closed Or The Reactor has been tripped with MSIS initiated Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6/7/8 Page 7 of 10 Event

Description:

Main Feedwater line break in Containment, FW-184 A fails to auto close, CVC-109 fails to auto close.

Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications.

Alarms Containment Water Leakage Hi (Cabinet N, L-20)

Containment Water Leakage Hi-Hi (Cabinet N, K-20)

Containment Pressure Hi/Lo (Cabinet M, H-4 and N, H-14)

Containment Fan Cooler B Disch Air Temp Hi (Cabinet SB, B-6)

Containment Fan Cooler D Disch Air Temp Hi (Cabinet SB, C-6)

Indications Rising Containment Pressure Lowering level Steam Generator #2 ATC If directed by CRS, initiate Safety Injection Actuation (SIAS), Main Steam Isolation (MSIS) and Containment Isolation Actuation (CIAS) at CP-7.

Critical Task Establish Containment Isolation This task is satisfied by taking action to close CVC-109. This task becomes applicable after the CIAS signal has been initiated.

ATC Close CVC-109 at CP-4.

This step is applicable after CIAS has been initiated.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6/7/8 Page 8 of 10 Event

Description:

Main Feedwater line break in Containment, FW-184 A fails to auto close, CVC-109 fails to auto close.

Time Position Applicants Actions or Behavior Critical Task Establish Containment Isolation This task is satisfied by taking action to close FW-184 A.

BOP Close FW-184 A, MFIV #1 at CP-8.

This step is applicable after MSIS has been initiated.

Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of CSAS initiation.

ATC Following initiation of CSAS secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop at CP-2 BOP Secure AH-12 A or B on CRS direction after initiation of SIAS at CP-18.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6/7/8 Page 9 of 10 Event

Description:

Main Feedwater line break in Containment, FW-184 A fails to auto close, CVC-109 fails to auto close.

Time Position Applicants Actions or Behavior CRS After Excess Steam Demand is identified, direct ATC and BOP to monitor for the trigger points of OP-902-009, Appendix 13, Stabilize RCS Temperature.

Critical parameters are Pressurizer pressure rising and RCS Representative CET temperature rising.

Steps for stabilizing RCS temperature following an excess steam demand are contained in 2 procedures.

Appendix 13 is used if the critical parameters are both rising before the CRS has entered OP-902-004, Excess Steam Demand Recovery.

Step 16 of OP-902-004 is used if both parameters start rising after the crew has entered OP-902-004.

CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-004, Excess Steam Demand Recovery Critical Task Establish RCS temperature control This task is satisfied by taking action to stabilize RCS temperature within the limits of the RCS P/T curve using ADV #1 and establishing EFW flow to Steam Generator #1. Action to address this task should commence within 10 minutes after the applicable parameters begin to rise.

BOP When directed by the CRS to take action to stabilize RCS temperature:

Place the ADV for Steam Generator #1 to manual and fully open the ADV #1.

Manually initiate EFAS for Steam Generator #1.

Place EFW Flow Control Valve to manual and commence feeding Steam Generator #1.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 3 Event # 6/7/8 Page 10 of 10 Event

Description:

Main Feedwater line break in Containment, FW-184 A fails to auto close, CVC-109 fails to auto close.

Time Position Applicants Actions or Behavior Critical Task Establish RCS pressure control This task is satisfied by taking action to stabilize RCS pressure within the limits of the RCS P/T curve and additionally maintain RCS pressure within 1500-1600 psid of the faulted steam generator. Action to address this task should commence within 10 minutes after the actions to establish TCS temperature control have been commenced.

ATC When directed by the CRS to take action to stabilize RCS temperature:

IF RCS pressure is 1500 psia, THEN stabilize RCS pressure at a value not to exceed 1600 psid between the RCS and the lowest SG pressure.

IF RCS pressure is < 1500 psia, THEN stabilize RCS pressure at >

HPSI shutoff head (1500-1600 psia).

Examiner Note This event is complete after RCS temperature and pressure have been stabilized Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Scenario Outline Form ES-D-1 Facility: WATERFORD 3 Scenario No.: 4 Op Test No.: NRC Examiners: Operators:

Initial Conditions: 30% Power on RCS chemistry hold Main Feedwater Pump B is running Turnover: Hold power until directed by plant management Start ACCW Pump A for chemical mixing Event Malf. No. Event Event No. Type* Description 1 DI-33A04S36-1 C - BOP Start Auxiliary Component Cooling Water Pump C - SRO A for chemical mixing. ACC-110A will fails to open.

TS - SRO 2 SG11C I - BOP Steam Generator Level #2 level instrument SG-TS - SRO ILT1123 C fails low 3 CV05B2 C - ATC Letdown Backpressure Control Valve CVC-C - SRO 123B, fails closed 4 R - ATC Direction given to raise power to < 50% using N - BOP OP-010-004, Power Operations.

N - SRO 5 RP04A3 I - BOP Inadvertent Containment Spray Actuation Signal, RP04B3 I - SRO secure Containment Spray Pumps 6 M - All Manual Reactor trip 7 DI-08A07S26-1 I - ATC CC-641 will fail to reopen, Secure all Reactor I - SRO Coolant Pumps on loss of Component Cooling Water flow 8 SG01A M - All Steam Generator #1 tube rupture ramps in over 3 minute period following reactor trip 8 C - BOP Isolate Steam Generator #1 when < 520 °F hot C - SRO leg temperature 8 C - ATC Reduce RCS pressure using Auxiliary Spray C - SRO while maintaining sub-cooled margin.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Scenario 4 Rev 1 Scenario Event Description NRC Scenario 4 The crew assumes the shift at 30% power with instructions to maintain power.

The crew is directed to start Auxiliary Component Cooling Water Pump A for basin chemical mixing. During the start, ACC-110A will not auto open. If the BOP operator takes the control switch to open, ACC-110A will remain closed. The SRO should declare ACC-110A inoperable and enter a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action for Tech Spec 3.7.3 as well as cascading Tech Specs.

The SRO should address the need to accomplish surveillance OP-903-066, Electrical Breaker Alignment Checks, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to comply with Tech Spec 3.8.1.1.b. They must also address the need to accomplish the requirements of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

After Tech Specs have been addressed, Steam Generator #2 Level instrument SG-ILT-1123 C fails low. The SRO should enter Tech Spec 3.3.1 and 3.3.2. PPS bistables for Channel C Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #2 should be placed in bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

After the appropriate bistables have been bypassed, CVC-123B, Chemical and Volume Control Backpressure Control Valve B will fail closed. Letdown flow will go to 0 gpm. The SRO should enter OP-901-112, Charging or Letdown Malfunction, and transition to sub-section E2, Letdown Malfunction. The ATC operator should place the standby Letdown Backpressure Control Valve in service and restore Letdown flow.

After Letdown has been restored, the SRO will be given direction to raise power to 50% for placing Main Feedwater Pump A in service. The SRO should use OP-010-004, Power Operations to direct the power ascension. The ATC operator will add Primary Makeup Water to the Volume Control Tank and the BOP operator will raise Main Turbine load.

At the direction of the lead examiner, an inadvertent Containment Spray Actuation Signal will be generated. The SRO should enter OP-901-504, Inadvertent ESFAS Actuation. The BOP should be directed to secure Containment Spray Pumps A and B. The BOP operator will be directed to restore CCW flow to the Reactor Coolant Pumps. CC-641, RCP Inlet Outside Isolation will fail to reopen when attempted by the BOP operator. The SRO should direct the ATC to trip the reactor and secure Reactor Coolant Pumps.

A Steam Generator tube rupture ramps in for Steam Generator #1 during the Containment Spray actions. The crew should diagnose into OP-902-007, Steam Generator Tube Rupture Recovery. The SRO will direct a rapid RCS cooldown to < 520 °F hot leg temperature.

Following the rapid cooldown, the BOP should be directed to isolate Steam Generator #1 and the ATC operator should be directed to lower RCS pressure using Auxiliary Spray within the RCS temperature and pressure limits.

The scenario can be terminated after Steam Generator #1 is isolated and the crew has taken action to reduce RCS pressure.

Scenario 4 Rev 1 Scenario Event Description NRC Scenario 4 Critical Tasks

1. Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of CSAS initiation.

2. Prevent opening the Main Steam Safety valves on Steam Generator #1 This task is satisfied by taking action lower RCS pressure to < 945 PSIA.

3. Isolate Steam Generator #1 This task is satisfied by isolating Steam Generator #1 in accordance with step 17 after RCS THOT is reduced below 520 °F.

Scenario 4 Rev 1 Scenario Event Description NRC Scenario 4 Scenario Notes:

A. Reset Simulator to IC-94.

B. Verify the following Scenario Malfunctions:

1. sg11c for SG-ILT-1123 C failing low

2. cv05b2 for CVC-123 B failing closed

3. sg01a for Steam Generator #1 tube rupture C. Verify the following Remotes:

1. cvr03 for CVC-121 A manipulations

2. cvr04 for CVC-121 B manipulations D. Verify the following overrides

1. di-08a07s26-1 for CC-641 failing closed E. Verify the following item entered under Event Triggers

1. zdiccaccecs752as(2).eq.0 F. Ensure Protected Train B sign is placed in SM office window.

G. Verify EOOS is 10.0 Green H. Complete the simulator setup checklist.

Scenario 4 Rev 1 Simulator Booth Instructions Event 1 ACC-110A fails to open

1. Ensure trigger 1 is set after coming out of freeze but prior to BOP starting ACCW Pump A.

2. If Work Week Manager, PME or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If called as the RCA watch, report ACCW Pump A looks good when started.

4. If called as RCA watch to check ACC-110A, report that the valve is closed with no abnormal indications.

Event 2 SG-ILT-1123 C fails low

1. On Lead Examiner's cue, initiate Event Trigger 2.

2. If Work Week Manager or I&C is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Event 3 CVC-123 B fails closed

1. On Lead Examiner's cue, initiate Event Trigger 3.

2. If Work Week Manager or PMM is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

3. If called as the RCA watch, report standing by on the -4 for valve manipulations.

4. If called regarding CVC-123 B, report signs of air leaking behind shield wall, but will have to investigate further.

Event 4 Power Ascension to 50%

1. If called as TGB watch to monitor Condensate Polishers, report standing by on station.

Event 5 Inadvertent CSAS

1. On the Lead Examiner's cue, initiate Event Trigger 5.

2. If Work Week Manager or PMM is called, inform the caller that a work package will be assembled and a team will be sent to the Control Room.

Scenario 4 Rev 1 Event 6 Manual Reactor Trip

1. No communications should occur for this event.

Event 7 CC-641 fails to reopen, secure Reactor Coolant Pump on loss of CCW

1. If called as RCP system engineer, respond that you will check out the RCP parameters and get back with you.

2. If called as RCA watch to check CC-641, report no abnormal indications.

Event 8 SGTR on Steam Generator #1

1. If called as the shift chemist to carry out the actions of UNT-005-032, simply confirm request; no additional report is necessary.

Scenario 4 Rev 1 Scenario Timeline:

Ramp Time Event Malfunction Severity Delay Trigger HH:MM:SS (Min) 1 Zdiccaccecs752 N/A N/A N/A 1 as(2).eq.0 ACC-110 A fails to auto open 2 SG11C 0% N/A N/A 2 SG-ILT-1123 C fails to 0%

3 CV05B2 N/A N/A N/A 3 CVC-123 B fails closed 5 RP04A3 N/A N/A N/A 5 RP04B3 Inadvertent CSAS 7 DI-08A07s26-1 N/A N/A N/A N/A CC-641 fails to reopen 8 SG01A 3.0 3:00 1:00 5 Steam Generator #1 Tube Rupture (ramps in during CSAS event)

Scenario 4 Rev 1

REFERENCES:

Event Procedures 1 OP-002-001, Auxiliary Component Cooling Water OP-100-014, Technical Specification and Technical Requirements Compliance Tech Spec 3.7.3 and 3.8.1.1 2 OP-903-013, Monthly Channel Checks Tech Spec 3.3.1 and 3.3.2 3 OP-901-112, Charging or Letdown Malfunction 4 OP-010-004, Power Operations OP-002-005, Chemical and Volume Control 5 OP-901-504, Inadvertent ESFAS Actuation OP-902-000, Standard Post Trip Actions OP-902-009, Standard Appendices, Appendix 1, Diagnostic Flow Chart 6 OP-901-504, Inadvertent ESFAS Actuation OP-902-000, Standard Post Trip Actions 7 OP-901-504, Inadvertent ESFAS Actuation 8 OP-902-007, Steam Generator Tube Rupture Recovery Scenario 4 Rev 1 Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 1 Page 1 of 12 Event

Description:

Start Auxiliary Component Cooling Water Pump A for chemical mixing.

ACC-110A fails to open.

Time Position Applicants Actions or Behavior All manipulations for this event are located on CP-33 BOP Direct local watchstander to check the following:

Verify a visible clearance exists between piping and jack stand.

Verify ACCW Pump A bearing oil level is between the OFF level marks.

Verify Locked Open ACC Pump A Recirc Line Isolation, ACC-107A.

BOP Close ACC Header A CCW HX Outlet Temp Control Valve, ACC-126A, using Component Cooling Water Header A Temperature Indicating Controller, CC-ITIC-7070A, in Manual.

BOP Start ACC Pump A BOP Verify Open ACC Pump A Discharge Line Isolation, ACC-110A.

ACC-110A will not auto open. If the BOP takes the control switch to open, the valve will not open.

Stop ACC Pump A by taking the control switch to STOP.

Verify ACC Jockey Pump A Starts.

BOP Restore Component Cooling Water Header A Temperature Indicating Controller, CC-ITIC-7070A, setpoint to 95 F or as otherwise directed by the SM/CRS.

BOP Verify Component Cooling Water Header A Temperature Indicating Controller, CC-ITIC-7070A, is in AUTO.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 1 Page 2 of 12 Event

Description:

Start Auxiliary Component Cooling Water Pump A for chemical mixing.

ACC-110A fails to open.

Time Position Applicants Actions or Behavior CRS Upon notification of ACC-110 A failure to open, evaluate Tech Specs.

Correct Tech Spec is 3.7.3 and cascading Tech Specs. OP-903-066, Electrical Breaker Alignment Check, is required within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

verification of Tech Spec 3.8.1.1.d within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The CRS may choose to have the BOP operator secure ACCW Pump A following the failure of ACC-110A. This course of action is acceptable.

Examiner Note This event is complete after Tech Spec 3.7.3 and cascading Tech Specs have been addressed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 2 Page 3 of 12 Event

Description:

Steam Generator Level #2 level instrument SG-ILT1123 C fails low Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed channel.

Alarms:

RPS Channel Trip SG 2 Level Lo (Cabinet H, E-17)

SG 2 Level Lo Pretrip A/C (Cabinet H, F-14)

RPS Channel C Trouble (Cabinet K, G-18)

Indications SG-ILI-1123 C indicates 0% narrow Range on CP-8 LO SG-2 Level trip and pre-trip on Channel C CRS Review Tech Specs based on the failed instrument.

Enter Tech Spec 3.3.1 and 3.3.2 Direct bypassing Channel C bistables 8, 10, and 20 for Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #2. This is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action.

BOP Bypass Channel C bistables 8, 10, and 20 for Steam Generator Level Low, Steam Generator Level High, and Steam Generator Differential Pressure for Steam Generator #2.

Located on CP-10 C (rear panel).

Examiner Note This event is complete when the proper bistables have been bypassed Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 3 Page 4 of 12 Event

Description:

Letdown Backpressure Control Valve CVC-123B, fails closed Time Position Applicants Actions or Behavior ATC Recognize and report indications of failed Backpressure Control Valve.

Alarms Letdown Flow Hi/Lo (Cabinet G, C-1)

Letdown HX Outlet Pressure Hi (Cabinet G, A-2)

Indications Letdown Backpressure Regulating valve not controlling at setpoint.

Letdown flow 0 gpm CRS Enter and direct the implementation of OP-901-112, Charging or Letdown Malfunction, subsection E2, Letdown Malfunction.

CRS Direct the ATC operator to operate Charging Pumps as necessary to maintain Pressurizer level in accordance with Attachment 1, Pressurizer Level Versus Tave Curve.

This curve does not allow continued operation with a Pressurizer level of 17% or lower.

All manipulations for this event will be at CP-4.

ATC Secure necessary Charging Pumps by placing the control switch to STOP.

CRS Direct the ATC to place the standby Backpressure control valve in service in accordance with step 8.

ATC Place CVC-IPIC-0201, Letdown Backpressure Controller, in MAN AND make necessary adjustments to maintain pressure during transfer.

With CVC-123 B failed closed, the ATC will not be able to establish Letdown flow at this point.

ATC Place Letdown Backpressure Control Valve Selector switch to BOTH.

ATC Direct the RCA watchstander to make the following manipulations:

Verify open standby Letdown Back PCV A Outlet Isolation CVC 125 A.

Slowly open standby Letdown Backpressure Control Valve Inlet Isolation CVC 121 A.

Slowly close in service Letdown Backpressure Control Valve Inlet Isolation CVC-121 B.

Close in service Letdown Back PCV B Outlet Isolation CVC 125 B.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 3 Page 5 of 12 Event

Description:

Letdown Backpressure Control Valve CVC-123B, fails closed Time Position Applicants Actions or Behavior ATC Position Letdown Backpressure Control Valve Selector switch to select the A Backpressure Control valve.

ATC Adjust setpoint of CVC-IPIC-0201, Letdown Backpressure Controller, to 460 PSIG.

ATC Place CVC-IPIC-0201, Letdown Backpressure Controller in AUTO.

ATC IF the LETDOWN FLOW CONTROL controller, RC-IHIC-0110, has been placed in manual to control Pressurizer level, THEN match controller process with its output and place in AUTO Examiner Note This event is complete when CVC-123 A has been placed in service Or As directed by the Lead Evaluator Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 4 Page 6 of 12 Event

Description:

Direction given to raise power to < 50% using OP-010-004, Power Operations.

Time Position Applicants Actions or Behavior CRS Receives direction to raise power to 45-50%.

ATC Set Primary Makeup Water Batch Counter to volume of Primary Makeup water desired.

ATC Place Makeup Mode selector switch to DILUTE.

ATC Open VCT Makeup Valve, CVC-510.

ATC Verify Primary Makeup Water Flow controller, PMU-IFIC-0210X, in Manual.

ATC Adjust Primary Makeup Water Flow controller, PMU-IFIC-0210X, output to >

5 GPM flow rate.

ATC Operate VCT Inlet/Bypass to Holdup Tanks, CVC-169 Control Switch to BMS/Auto positions as necessary to maintain VCT pressure and level within normal operating bands.

ATC When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU-144, Closed.

ATC If additional Primary Makeup Water addition is required and with SM/CRS permission, then perform the following:

ATC Reset Primary Makeup Water Batch Counter.

ATC Verify Primary Makeup Water Control Valve, PMU-144, Intermediate or Open.

ATC Observe Primary Makeup water flow rate for proper indication.

ATC When Primary Makeup Water Batch Counter has counted down to desired value, then verify Primary Makeup Water Control Valve, PMU-144, Closed.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 4 Page 7 of 12 Event

Description:

Direction given to raise power to < 50% using OP-010-004, Power Operations.

Time Position Applicants Actions or Behavior BOP Commence raising Main Turbine load by performing the following:

Depress LOAD RATE MW/MIN pushbutton.

Set selected rate in Display Demand Window.

Depress ENTER pushbutton.

Depress REFERENCE pushbutton.

Set desired load in Reference Demand Window.

Depress ENTER pushbutton.

Depress GO pushbutton.

This manipulation is performed at CP-1. The BOP will set up the Main Turbine controls. The ATC will direct the BOP when to commence unloading the Main Turbine based on the rise in RCS Cold Leg temperature.

Crew Maintain RCS Cold Leg Temperature 536°F to 549°F.

ATC Operate CEAs to maintain ASI using CEA Reg. Group 6 or Group P Control Element Assemblies.

Operate CEAs in Manual Group mode as follows:

Position Group Select switch to desired group.

Place Mode Select switch to MG.

Operate CEA Manual Shim switch to WITHDRAW Examiner Note This event is complete when the desired power ascension has been accomplished Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7 Page 8 of 12 Event

Description:

Inadvertent Containment Spray Actuation Signal, secure Containment Spray Pumps, CC-641 fails to reopen, Reactor trip, secure all RCPs Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of a Containment Spray Actuation.

Alarms Numerous alarms on Panels H, SA, and SB for RCP low CCW flow Numerous alarms on Panel M for Containment Spray Pump A and Train A isolation valve Numerous alarms on Panel M for Containment Spray Pump B and Train B isolation valve Indications Containment Spray Pumps A and B running.

CS-125 A and B open.

Containment Spray Header flow on Trains A and B.

CC-710, CC-713, and CC-641 indicate closed.

CRS Enter and direct the implementation of OP-901-504, Inadvertent ESFAS Actuation, subsection E2, Inadvertent CSAS.

BOP Secure BOTH Containment Spray Pumps by placing each control switch to OFF.

BOP Within 3 minutes restore CCW flow to Reactor Coolant Pumps as follows:

Open the following valves:

CC 710 RCP OUTLET INSIDE ISOL.

CC 641 RCP INLET OUTSIDE ISOL CC 713 RCP OUTLET OUTSIDE ISOL CC-641 will not reopen when the control switch is taken to open.

CRS On the report that CC-641 will not reopen, direct the actions of step 3, trip the reactor and secure all RCPs.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 5/6/7 Page 9 of 12 Event

Description:

Inadvertent Containment Spray Actuation Signal, secure Containment Spray Pumps, CC-641 fails to reopen, Reactor trip, secure all RCPs Time Position Applicants Actions or Behavior ATC Trip the reactor on direction from the CRS.

Critical Task Trip any RCP not satisfying RCP operating limits.

This task is satisfied by securing all RCPs within 3 minutes of loss of CCW flow.

ATC Following tripping of the reactor, secure all running Reactor Coolant Pumps as follows:

Place each RCP control switch to stop.

This is done at CP-2. At this point, the RCPs will be operating with no CCW flow since the actuation of Containment Spray.

This condition is only allowed for 3 minutes.

CRS Direct ATC and BOP to carry out Standard Post trip Actions.

BOP Reset Moisture Separator Reheaters by pushing the RESET button located on CP-1.

Examiner Note This event is complete after the Reactor has been tripped and all RCPs are secured.

Or As directed by the Lead Evaluator.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 7 Page 10 of 12 Event

Description:

Steam Generator #1 tube rupture ramps in over 3 minute period following reactor trip Time Position Applicants Actions or Behavior ATC / BOP Recognize and report indications of activity in Steam Generator #1.

Alarms Rad Monitoring System Activity Hi Hi (Cabinet L, A-8)

Vacuum Pump Exhaust Activity Hi/Monitor Trouble (Cabinet E, C-3)

Indications Rising activity in S/G #1.

Pressurizer level and pressure lowering CRS After review of Standard Post Trip Actions, use Diagnostic Flow Chart of OP-902-009 to select appropriate optimal recovery procedure.

Proper use of chart will result in use of OP-902-007, Steam Generator Tube Rupture Recovery.

CRS Direct implementation of OP-902-007.

BOP Commence a rapid RCS cooldown to less than 520 °F THOT using Steam Bypass valves.

CRS IF MSIS is NOT present, THEN lower the automatic initiation setpoints as the cooldown and depressurization proceed for MSIS (low SG Pressure).

The CRS should direct the ATC to perform this action during the rapid cooldown to < 520 THOT.

ATC Reset MSIS setpoints on all 4 channels at CP-7.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 7 Page 11 of 12 Event

Description:

Steam Generator #1 tube rupture ramps in over 3 minute period following reactor trip Time Position Applicants Actions or Behavior Critical Task Prevent opening the Main Steam Safety valves on Steam Generator #1 This task is satisfied by taking action lower RCS pressure to < 945 PSIA.

ATC Depressurize the RCS:

Maintain pressurizer pressure within ALL of the following criteria:

Within Appendix 2A-D, "RCS Pressure and Temperature Limits" Less than 945 psia Within 50 psi of the most affected steam generator pressure Operate Main or Auxiliary Pressurizer spray.

The ATC operator should receive direction from the CRS to perform this step. He should evaluate plant conditions and decide on a minimum RCS pressure. The critical task is satisfied when the candidate takes action to start reducing RCS pressure (< 945 PSIA does not need to be reached in the scenario).

BOP / ATC IF HPSI throttle criteria are met, THEN perform ANY of the following:

Control charging and letdown flow Throttle HPSI flow HPSI Throttle criteria includes.

ALL of the following conditions are satisfied:

RCS subcooling is greater than or equal to 28 ºF Pressurizer level is greater than 7% and controlled ALL steam generators capable of steaming are being maintained or restored to within the following level:

o 50% to 70% NR using MFW or EFW in auto or manual RVLMS indicates level higher than Hot Leg by at least one of the following:

o QSPDS REACTOR VESSEL LEVEL 5 NOT voided o VESSEL LEVEL PLENUM greater than or equal to 80%

The conditions for HPSI Throttle criteria should be met during the scenario.

Scenario 1, Revision 0

Appendix D Required Operator Actions Form ES-D-2 Op Test No.: NRC Scenario # 4 Event # 7 Page 12 of 12 Event

Description:

Steam Generator #1 tube rupture ramps in over 3 minute period following reactor trip Time Position Applicants Actions or Behavior Critical Task Isolate Steam Generator #1 This task is satisfied by isolating Steam Generator #1 in accordance with step 17 after RCS THOT is reduced below 520 °F.

BOP When the RCS TH is less than 520°F, THEN isolate Steam Generator #1:

Place the ADV setpoint to 980 psig and verify the controller in AUTO.

Verify the MSIV is closed.

Verify the MFIV is closed.

IF EFAS-1 is NOT initiated, THEN close EFW Isolation Valves:

EFW 228A SG 1 PRIMARY EFW 229A SG 1 BACKUP Place EFW Flow Control Valves in MAN and close:

EFW 224A SG 1 PRIMARY EFW 223A SG 1 BACKUP Close MS 401A, PUMP AB TURB STM SUPPLY SG 1 Close Main Steam Line 1 Drains:

MS 120A NORMAL MS 119A BYPASS Close Steam Generator Blowdown isolation valves:

BD 103A STM GEN 1 (OUT)

BD 102A STM GEN 1 (IN)

Check the Main Steam Safety valves are closed.

Examiner Note This event is complete after Steam Generator #1 is isolated Or As directed by the Lead Evaluator.

Scenario 1, Revision 0