Text

Initial Exam 2012-301 Final Administrative JPMs
	ML120720356
Person / Time
Site:	Sequoyah
Issue date:	02/21/2012
From:	; NRC/RGN-II
To:	;
	Gerald Laska
References
	Download: ML120720356 (128)
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SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC RO ADMIN A.1.a

1201 NRC RO ADMIN A.1.a Page 2 of 7 SRO JOB PERFORMANCE MEASURE Task:

Determine Main Turbine Acceleration and Loading Rates Task#:

0450020101 Task Standard:

Given plant data during a plant startup, the exam inee will determine the following during a Main Turbine startup using TI-28, Plant Curve Book.

1.

Determine the minimum allowable period for turbine roll-up from turning gear speed to synchronous speed of 25 (22.5 to 27.5) minutes.

2.

Determine the maximum ACCELERATION RATE for turbine roll 72 (65.5 to 80) rpm/mm.

3.

Determine time to hold at 5% turbine load of 30 (25 to 35) minutes.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.1.25 (3.9/4.2)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name! Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

17 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC RO ADMIN A.1.a Page 3 of 7 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/EquipmentlProcedures Needed:

1.

0-GO-4 Power Ascension From Less than 5% Reactor Power to 30% Reactor Power, section 5.3 step 31 (page 39) 2.

TI-28, Curve Book Figure 6 3.

Calculator

References:

Reference Title Rev No.

1.

0-GO-4 Power Ascension From Less than 5% Reactor 75 Power to 30% Reactor Power 2.

Tl-28, Figure 6 Curve Book 0244 Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1.

A Plant Startup is in progress in accordance with 0-GO-4, Power Ascension from Less than 5%

Reactor Power to 30% Reactor Power.

2.

The crew is preparing to roll the turbine.

3.

Turbine Impulse Metal Temperature on 1-TR-47-1 Pt. 1 indicates 150 degrees F.

INITIATING CUES:

1.

0-GO-4, section 5.3, Turbine Roll, step 31 you have been directed to determine the following:

a)

Determine the minimum allowable period for turbine roll-up from turning gear speed to synchronous speed.

b)

Determine the maximum ACCELERATION RATE for turbine roll.

2.

Then from 0-GO-4, section 5.4, Placing Main Generator In Service, step 38, you have been directed to time to hold at 5% turbine load.

3.

Record your answers on the JPM briefing sheet.

4.

Notify the Examiner when you are complete.

1201 NRC RO ADMIN A.1.a Page 4 of 7 Start Time Obtain a copy of 0-GO-4 Power Ascension From Less than 5% Reactor STEP 1

Power to 30% Reactor Power, section 5.3 step 31 (page 39), section 5.4 SAT step 39 (page 64) and TI-28, Curve Book Figure 6.

UNSAT Copy of 0-GO-4 Power Ascension From Less than 5% Reactor Power to Standard:

30% Reactor Power, section 5.3 step 31 (page 39), section 5.4 step 39 (page 64) and TI-28, Curve Book Figure 6 are obtained.

Provide a copy of 0-GO-4 Power Ascension From Less than 5% Reactor Cue Power to 30% Reactor Power, section 5.3 step 31 (page 39), section 5.4 step 39 (page 64) and Tl-28, Curve Book Figure 6.

Comment NOTE The minimum allowable period for turbine roll-up from turning gear speed to synchronous speed is determined by reading the first stage metal temperature then plotting a vertical line on Tl-28 figure 6 at that temperature. The allowable minimum time Is read in minutes on the vertical axis where the vertical line crosses the bottom curve of figure 6. The initial first stage metal temperature is also used to determine the holding time at minimum load and for raising load to full capacity.

5.3 Turbine Roll (continued)

STEP 2

[31]

DETERMINE the maximum ACCELERATION RATE for turbine roll using SAT recorder 1,2-TR-47-1 (Impulse Chamber Metal Temp., pt. #1) and TI-28 Figure 6 Recommended Start-up and Loading Times.

by performing the UNSAT following:

[31.1]

RECORD the first stage metal temperature from 1.2-TR-47-1 (Impulse Chamber Metal Temp., pt. 1).

Examinee the records first stage meta temperature from 1 -TR-47-1 Standard:

(Impulse Chamber Metal Temp., pt. #1) of 150 deg F from the initial conditions.

Comment

1201 NRC RO ADMIN A.1.a Page 5 of 7 5.3 Turbine Roll (continued)

STEP

[31]

DETERMINE the maximum ACCELERATION RATE for turbine roll using SAT recorder 1.2-TR-47-1 (Impulse Chamber Metal Temp., pt. #1) and Tl-28 Figure 6 Recommended Start-up and Loading Times.

by performing the UNSAT following:

Critical Step

[312]

DETERMINE the minimum allowable period for turbine roll-up from turning gear speed to synchronous speed USING TI-28, Figure 6.

C Examinee determines the minimum allowable period for turbine rolk1p Standard:

from turning gear speed to synchronous speed of 25(22.5 to 27.5) minutes.

Comment 5.3 Turbine Roll (continued)

STEP 4

[31]

DETERMINE the maximum ACCELERATION RATE for turbine roll using SAT recorder 1,2-TR-47-1 (Impulse Chamber Metal Temp,, pt. #1) and Tl-28 Figure 6 Recommended Start-up and Loading Times.

by performing the U NSAT following:

[313]

RECORD the time determined from Figure Ai5 in the equation below, Standard:

Examinee records the time obtained in JPM step 3.

Examiner If the examinee notes the difference in the step 31.2 and 31.3 between TI Note 28 figure 6 and A.1 5 reference state management has recognized that TI 28 figure 6 and A.15 are the same and has initiated a procedure change.

Comment

1201 NRC RO ADMIN A.1.a Page 6 of 7 5.3 TurbIne Roil (continued)

STEP 5

[31]

DETERMINE the maximum ACCELERATION RATE for turbine roll using SAT recorder 1.2-TR-47-1 (Impulse Chamber Metal Temp., pt. #1) and TI-28 Figure 6 Recommended Start-up and Loading Times. by performing the UNSAT following Critical Step

[31.4]

CALCULATE the Maximum Acceleration Rate from the equation below.

1800 RPM divided by

=

Sync Speed Minimum allowable Maximum Acceleration period for turbine rollup Rate Standard:

mu calculates the Maximum Acceleration Rate forturbine roll of 72 (65.5 to 80) rpm/miri Comment 5.4 Placing Main Generator in Service (continued)

STEP 6 :

SAT

[38]

DETERMiNE time to hold at 5% turbine load on TI-28 figure 6 and record minutes.

UNSAT minutes 1

Examinee determines the time to hold at 5% turbine Load from TI-2a Critical Step Standard:

figure 6 of 30(25 to 35) minutes.

Comment Terminating The JPM is complete when the Examinee returns the JPM Cue:

briefing sheet to the Examiner.

STOP Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

INITIAL CONDITIONS:

1.

A Plant Startup is in progress in accordance with 0-GO-4, Power Ascension from Less than 5%

Reactor Power to 30% Reactor Power.

2.

The crew is preparing to roll the turbine.

3.

Turbine Impulse Metal Temperature on 1-TR-47-1 Pt. 1 indicates 150 degrees F.

INITIATING CUES:

1.

0-GO-4, section 5.3, Turbine Roll, step 31 you have been directed to determine the following:

a)

Determine the minimum allowable period for turbine roll-up from turning gear speed to synchronous speed.

b)

Determine the maximum ACCELERATION RATE for turbine roll.

2.

Then from 0-GO-4, section 5.4, Placing Main Generator In Service, step 38, you have been directed to time to hold at 5% turbine load.

3.

Record your answers on the JPM briefing sheet.

4.

Notify the Examiner when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

Figure 6 (Page 1 of 1)

RECOMMENDED START-UP AND LOADING TIMES NUCLEAR STEAM SYSTEM UNITS INITIAL H.R flJNE F1RST STAGE METAL TDAPRTURE 4

qc.e(erd.j 4

f rL rLoi 1J1Lc-SQN CURVE BOOK TI-28 Rev. 0244 Page 21 of 96 Do

I7oi N1L P SQN CURVE BOOK TI-28 Rev. 0244 Page 21 of 96 Figure 6 (Pagelofl) 7 RECOMMENDED START-UP AND LOADING TIMES C,

w NUCLEAR STEAM SYSTEM UNITS INITIAL H.P TU1NE F1RSTSTAGE MTL TD.4PEPATIJRE

-rI 40 Li 0

JA

.A S, 9

1_________

iifr_____

t a

L3_ll i

I 4L r

0 Oo 400

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE SEQUOYAH NUCLEAR PLANT RO ADMIN A.1.b

1201 NRC RO ADMIN A.1.b Page 2 of 10 SRO JOB PERFORMANCE MEASURE Task:

Calculate Manual Makeup to the Volume Control Tank Task#:

0040250101 Task Standard:

Examinee calculates a manual make up to the VCT to raise level 10% of 173.03 (1 72.99-173.27) gallons water and 26.97 (26.73-27.01) gallons of boric acid.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.1.37 (4.6)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

I Name! Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

22 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC RO ADMIN A.1.b Page 3 of 10 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/Equipment/Procedures Needed:

1.

Calculator 2.

0-SO-62-7, Boron Concentration Control.

3.

TI-44, Boron Tables.

References:

Reference Title Rev No.

1.

0-SO-62-7 Boron Concentration Control 60 2.

TI-44 Boron Tables 12 Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1.

Unit2isinMODEl, 100%.

2.

Reactor coolant system boron concentration is 967 ppm.

3.

BAT boron concentration is 6820 ppm.

4.

B-b depletion value is 40 ppm 5.

REACTF software is not available for boron calculations.

INITIATING CUES:

1.

You are a Unit 2 AC and are to perform a calculation for a manual blended makeup to the Volume Control Tank level from 41% to 51% using 0-SO-62-7, Boron Concentration Control appendix C.

2.

Notify the Examiner when the calculation is completed.

1201 NRC RO ADMIN A.1.b Page 4 of 10 Start Time Obtain a copy of 0-SO-62-7 BORON CONCENTRATION CONTROL and STEP 1

TI-44 BORON TABLES SAT UNSAT Examinee obtains a copy of 0-SO-62-7 BORON CONCENTRATION Standard:

CONTROL and TI-44 BORON TABLES.

Cue Provide a copy of 0-SO-62-7 BORON CONCENTRATION CONTROL

Appendix C and TI-U BORON TABLES.

Comment STEP 2

[1]

OBTAIN Current RCS Boric Acid Concentration.

SAT UNSAT Examnee obtains the RCS Boric Acid Concentration of 967 ppm from the Standard:

initial conditions Comment STEP 3

[2]

OBTAIN Current BAT Boric Acid Concentration SAT UNSAT Standard:

Examinee obtains the BAT Boric Acid Concentration of 6820 ppm from the initial conditions Comment

1201 NRC ROADMINA.1.b Page 5 of 10

[4]

CALCULATE BAT Boric Acid Concentration Ratio (BACR): 6820 ppm ÷ Step [2] ppm =

Examinee calculates BAT Boric Acid Concentration Ratio to be 1.0.

STEP 4

[3]

OBTAIN B-b depletion value from Rx Eng Information page SAT UNSAT Standard Examinee obtains the B-b depletion value of 40 ppm from the initial conditions Corn rnent Procedure Result in Step 4] should be rounded to second decimal place.

STEP 5

SAT UNSAT Critical Step Standard Comment Examiner Step 2 BAT Boric Acid Concentration of 6820 ppm was from the initial Note:

conditions.

1201 NRC RO ADMIN A.i.b Page 6 of 10 STEP 6

[5]

CALCULATE 8-10 correoted boron concentration SAT UNSAT STEP til STEP [3]

B-b corrected boron Critical Step Standard:

Examinee calculates 8-10 corrected boron concentration to be 927 Corn ment Examiner Step 1 RCS Boric Acid Concentration of 967 ppm is from the initial Note:

conditions Ex 1

m,er Step 2 B-i 0 depletion value of 40 ppm from the initial conditions STEP 7

[6]

DETERMINE Corr?cted BoricAcid Flow Rate and Controller SAT Setting using appropriate Table from T1-44 Appendix C..

[a] RECORD Corrected Boric Acid Flow Rate from T144 UNSAT Appendix C Table i Critical Step S

d d*

Exarninee calculates the Corrected Boric Acki Flow Rate to be 1091 tan ar (10.8 tG1O.9S Comment Examiner The Corrected Boric Acid Flow Rate of was determined by calculating the Note:

difference of two points Tl-44 BORON TABLES Appendix C Page 3 of 10.

1201 NRC RO ADMIN A.1.b Page 7 of 10 STEP 8

[6]

DETERMINE Corrected Boric Acid Flow Rate and Controller SAT Setting using appropriate Table from Tl-44 Appendix C.

UNSAT

[b] RECORD Corrected Boric Acid Controller Setting from TI-44 Appendix C Table 2.

Critical Step Examinee calculates the Corrected Boric Acid Controller Setting to be Standard:

21.81 (21.6 to 21,9)

Comment Examiner The Corrected Boric Acid Flow Rate of was determined by calculating the Note:

difference of two points Tl-44 BORON TABLES Appendix C Page 8 of 10.

STEP 9

[

X (BACR from step 4) =

(Boric Acid Controller Setting)

SAT Corrected Boric Acid Controller Setting (Step [6][b)) J UNSAT Critical Step Examinee calculates the Corrected Boric Acid ContoIler Setting to be Standard:

21.81 (21.6 to 21.9).

Comment Examiner The BACR from step 4 was determined to be 1.0 Note:

1201 NRC RO ADMIN A.1.b Page 8 of 10

)

STEP 10

[8]

[

X 20 GAL/percent

=

GALS SAT (Desired VCT (Actual VCT 3

Total Volume level) level)

UNSAT Critical Step Examinee calculates the volume added to be added to the VCT to be 200 Standard:

gal.

Comment Examiner The desired value of 51% and the actual value of 41% were given in the Note:

initial conditions.

STEP 11 f9]

I GPM X((BACR from at 70 GPM GPM SAT ep4l,)+

=

I Corrected8oricAcidFlow (Primary II20)

Total Flow Fate I

I Rate (Step [6][a])

)

UNSAT Critical Step Standard:

Examinee. calculates the total flow rate to be 80..91 (80.8 to 80.93)

Comment Examiner The Corrected Boric Acid Flow Rate calculated in procedure step 6a was Note:

10.91 (10.8-10.93). The BACR calculated in procedure step 4 was 1.0.

GPM GALS SAT STEP 12 I

GPM X (BACR from step 4) ÷ Total Flow Rate X

Total Volume 1101 I

I Corrected8oricAcidFlow (Obtained from step [9]

(Obtained from step [81 U NSAT Rate (Step 161[al)

GALS Critical Step BORIC ACID INTEGRATOR

/

SETTING Exarninee calculates the Boric Acid Integrator setting to be 267 t26.73-Standard:

2701)

Comment

1201 NRC RO ADMIN A.1.b Page 9 of 10 STEP 13 GPM GALS SAT 70 6PM GALS (Pbnmy H 2

0]

TotalFlowRate X

TOt& Volume

=

PRIMARYWATERINTEGRATOR Obtained from step [9J Obtained from step [81 SETTING UNSAT Critical Step Examinee calculates the Primary Water Integrator setting to be 173.03 Standard:

(172.99-173.27)

Comment Examiner The total flow rate calculated in procedure step 9 was 80.89. The total Note:

volume calculated in procedure step 8 was 200 gallons.

Calculation check: Step [10] results + Step [111 results should

= Step [8]

STEP 14 results SAT UNSAT Standard:

Examinee verifies the calculation result Comment Terminating The JPM is complete when the Examinee returns the JPM Cue:

briefing sheet to the examiner.

STOP Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

INITIAL CONDITIONS:

1.

Unit 2 is in MODE 1,100%.

2.

Reactor coolant system boron concentration is 967 ppm.

3.

BAT boron concentration is 6820 ppm.

4.

B-i 0 depletion value is 40 ppm 5.

REACTF software is not available for boron calculations.

INITIATING CUES:

1.

You are a Unit 2 RO and are to perform a calculation for a manual blended makeup to the Volume Control Tank level from 41% to 51% using 0-SO-62-7, Boron Concentration Control appendix C.

2.

Notify the Examiner when the calculation is complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

CALCULATION OF BORIC ACID AND PRIMARY WATER INTEGRATOR SETTING FOR MANUAL MAKEUP TO VCT (RCS)

[1]

OBTAIN Current RCS Boric Acid Concentration

[2]

OBTAIN Current BAT Boric Acid Concentration

[3]

OBTAIN B-b depletion value from Rx Eng Information page NOTE Result in Step [4] should be rounded to second decimal place.

CALCULATE BAT Boric Acid Concentration Ratio (BACR) 6820 ppm Step [2] ppm

=

CALCULATE B-b corrected boron concentration

L/

0 I 1 2.7 STEP [1]

STEP [3]

d21fd boron DETERMINE Corrected Boric Acid Flow Rate and Controller Setting using appropriate Table from Tl-44 Appendix C

[a]

RECORD Corrected Boric Acid Flow Rate from TI-44 Appendix C Table 1.

[b]

RECORD Corrected Boric Acid Controller Selling from Tl-44 Appendix C Table 2.

ii I I

4 4

72 I, 0/

I I

I 4b

° X BACR from step 4)

I (ric Acid Controller Setting)

Corrected Boric Acid Controller Setting (Step [6][b})

l 4 2 1 (7. r.

21 c)

CONTINUED ON N EXT PAetW APPENDIX C Page 1 of 2

[4]

[5]

[6]

[7]

[

I

SQ.

1,2 ll t1ZL BORON CONCENTRATION CONTROL 12o Aib 0-SO-62-7 Rev. 60 Page 162 of 201 APPENDIX C Page 2 of 2 (Actual VCT level)

(io. io)

GPM Corrected Boric Acid Flow Rate (Step [6][a])

70 GPM (Primary H 2

0)

-.3o.q3)

Sc.?!

GPM Total Flow Rate X

(Obtained from step [9J

[5!

I (Desired VCT level)

[8]

[9]

[

[10]

X 20 GAL/percent

]

X (BACR from step 4) +

=

2oo GALS Total Volume

(.

GPM

Total Flow Rate I

focl GPM X 1

BACRfromstep4 ÷ Corrected Boric Acid Flow Rate (Step [6][a])

(o$

,A3) too 70 GPM

[11]

(Primary H 2

0)

OI?I GPM Total Flow Rate Obtained from step [9]

x GALS ZO GALS Total Volume

=

(Obtained fro1 step 2

GALS BORIC ACID INTEGRATOR (riz.n-J 73

ØSFJTING

=

113.o3 GALS PRIMARY WATER INTEGRATOR SETIING Total Volume Obtained from step [8]

Calculation check:

Step [101 results

+ Step [11] results should = Step [8] results

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC RO ADMIN A.2

1201 NRC RO Admin JPM A.2 Page 2 of 15 SRO JOB PERFORMANCE MEASURE Task:

Perform RCS Leakage Calculation (0-Sl-OPS-068-1 37.0)

Task#:

0020010201 Task Standard:

The examinee performs a manual RCS Leakrate calculation using 0-Sl-OPS-068-1 37.0 Appendix D obtains a result of 1.07 gpm, and discovers the acceptance criteria are not met.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name I Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

20 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC RO Admin JPM A.2 Page 3 of 15 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/U NSAT column by bold print Critical Step.

Tools/Equipment/Procedures Needed:

1.

0-Sl-OPS.068-137.0 Reactor Coolant System Water Inventory 2.

Chronological Test Log 3.

Calculator

References:

Reference Title Rev No.

1.

0-Sl-OPS-068-1 37.0 Reactor Coolant System Water Inventory 26 2.

Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1201 NRC RO Admin JPM A.2 Page 4 of 15 1.

Unit-2 is in MODE 1.

2.

The performance of 0-SI-OPS-068-1 37.0, Reactor Coolant System Water Inventory is due.

3.

ICS is not available.

4.

Data for the 0-SI-OPS-068-1 37.0, Reactor Coolant System Water Inventory Appendix C is as follows:

Parameter Final Initial VCT Level 42.9 50.1 Pressurizer level 58 58 RCS Temperature 578 578 RCS Pressure 2235 2235 PRT Level 12358 12349 RCDT Level 172.2 171.8 CLA#1 Level 7692 7692 CLA #2 Level 7677 7677 CLA#3 Level 7672 7672 CLA #4 Level 7680 7680 Appendix D is complete through section 1.4 SG leakages SG#1 SG#2 SG#3 SG#4 3.2 gpd 2.2 gpd 2.3 gpd 4.8 gpd 6.

0-SI-CEM-000-050.2 and 2-SI-CEM-068-137.5 were completed last night at 2300.

7.

Chemistry supervisor reported no additions or samples taken from the RCS during the selected performance period.

8.

CCPIT/HUT leakage and other leakage is 0 gpm.

9.

Data collected for performance of this leak rate is documented in Appendix C and was collected over 120 minutes INITIATING CUES:

1.

You are the Unit 2 CR0 and have been directed to complete a manual RCS Leakage calculation using 0-SI-OPS-068-137.0, Reactor Coolant System Water Inventory Appendix D 2.

Record your answers on the Appendix D.

3.

Identify all deviations (if any) on the Chronological Test Log 4.

Notify the Examiner when you are complete.

5.

1201 NRC RO Admin JPM A.2 Page 5 of 15 Start Time 10 TOTAL RCS LEAKAGE 1.1 Volume Control Tank (VCT)

%)(

%)JxlS.27galI%

VCT Leakage

=

Initial Level Final Level mirL )

Time VCT Leakage -

The Examinee determines VCT Leakage is 1.16 gpm.

Obtain a copy of 0-SI-OPS-068-1 37.0 Reactor Coolant System Water STEP 1

Inventory.

SAT UNSAT Standard Copy of 0-Sl-OPS-068-1 37.0 Reactor Coolant System Water Inventory, is obtained.

C Provide a marked up copy of opy of 0-Sl-OPS-068-1 37.0 Reactor

Coolant System Water Inventory.

Comment RCS LEAKAGE CALCULATIONS NOTE 1 All calculations must be performed manually. Verification may be performed using the computer.

NOTE 2 All calculations should be carried out to two places after the decimal.

Signs must be carried through all calculations.

NOTE 3 Calculations in this Appendix may need to be performed more than once.

If multiple calculations are required, copies of this Appendix may be used.

STEP 2

SAT UNSAT Critical Step Standard:

Comment

1201 NRC RO Admin JPM A.2 Page 6of 15 NOTE I Vf (specific volume of a fluid) and V9 (specific volume of a gas)values for current plant conditions can be found using a copy of the steam tables.

NOTE 2 In the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

STEP 3

1.0 TOTAL RCS LEAKAGE SAT 1.2 Pressurizer (Pzr) UNIT 2 only UNSAT

[1]

CALCULATE initial PZR volume by using the following equation:

tintial PZR Vol=I PZRLeveI ÷ l00%PZRLev1 lb

olevel 100 il*ft /

Initial PZR Vol =

+ 202 lb

%level Ab J

1 7

iii Itmial PZR Vol =

za1 Performed By Reviewed By Standard:

The Examinee determines Initial Pressurizer Volume is 4857.55 gallons Corn ment

1201 NRC RO Admin JPM A.2 Page 7 of 15 NOTE I Vf (specific volume of a fluid) and Vg (specific volume of a gas)values for current plant conditions can be found using a copy of the steam tables.

NOTE 2 in the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

STEP 4

1.0 TOTAL RCS LEAKAGE SAT 1.2 Pressurizer (Pzr) UNIT 2 only UNSAT t2] CALCULATE Final PZR volume by using the following equation:

fiualPZRVol PZRLeve[1OO%PZRLevel 202 gal*tt/

T 3

/

(lel Fm PZR Vol=

+**-1 2.O2/

i-,,

ft7

// lb olevel (lb (Ii,

/

UI

/

UI Final PZR Vol a1 Perfomied By Reviewed By Standard:

The Examinee determines Final Pressurizer Volume is 4857.55 gallons Comment STEP 5

1.0 TOTAL RCS LEAKAGE SAT 1.2 Pressurizer (Pzr) UNIT 2 only UNSAT

[3] CALCULATE PZR leakage rate by using the following equation:

LniiaI PZR \\oliune Final PZR Vhune PZR Leakape ItUUC PZR Leakage =

0770 PZR Leakage =

apui Performed By Reviewed By Standard:

The Examinee determines Pressurizer Leakage is 0 gpm.

Comment

1201 NRC RO Admin JPM A.2 Page 8 of 15 NOTE The multiplier used in the following equation is the RCS Temperature Correction Volumetric constant as determined by engineering.

STEP 6

1.0 TOTALRCS LEAKAGE SAT 1.3 Pressurizer (Pzr) UNIT I only U NSAT Standard:

The Examinee addresses the step as not applicable.

Comment NOTE The multiplier used in the following equation is the RCS Temperature Correction Volumetric constant as determined by engineering.

STEP 7

1.0 TOTAL RCS LEAKAGE SAT 1.4 RCS Temperature Correctlon[C.1]

UNSAT A.

Initial Conditions:

RCS pressure =

÷ 14.7 =

psia RCS temperature =

Specific volume (v) =

ft 3

/lbm Density = 1/v =

Ibm/ft 3

B.

Final Conditions:

RCS pressure =

+ 14.7 =

psia RCS temperature =

Specific volume (V%) =

ft 3

/lbm Density = 1/v =

Ibm/ft 3

C. Temperature Correction;

[ (

Ibmfft 3

)

(

bmift 3 )]

1298.19 gal*ft 3

/ ibm Temp Corr, =

nitaI Density Ftnal Density time Temperature Correction =

Standard:

The Examinee determines RCS Temperature Correction is 0 gpm.

Comment

1201 NRC RO Admin JPM A.2 Page 9 of 15 STEP 8

1.0 TOTAL RCS LEAKAGE SAT 1.5 Total RCS Leakage UNSAT Critical Step Total RCS Leakage = (

) ÷

+/-

VCT PZR Temp Corr Total RCS Leakage =

Performed By Reviewed By Standard:

The Examinee determines Tota[RCS leakage is 1.16 gpm.

Comment NOTE Leakage into the PRT or CLA tanks will always be a positive value (or zero) when RCS pressure is greater than the specific tanks pressure.

If negative leakage is calculated under these conditions, that tanks leakage must be set equal to zero.

STEP 9

2.0 IDENTIFIED LEAKAGE SAT 2.1 Pressurizer Relief Tank (PRT)

UNSAT 2.1 Pressurizer Relief Tank (PRT)

Critical Step al)(

Qal)1 PRT Leakage =

Final Volume initial Volume

(

mm)

Time PRTLeakage=

Standard The Examinee determines Pressurizer Relief Tank (PRT) leakage is between 0.075 and 0.08 gpm Comment

NOTE 1201 NRC RO Admin JPM A.2 Page 10 of 15 RCDT Level (2.2) may be marked N/A if conditions do NOT permit obtaining RCDT level.

2.0 IDENTIFIED LEAKAGE 2.2 Reactor Coolant Drain Tank (RCDT)

RCDT Leakage =

Final Volume Initial Volume SAT UNSAT miii)

Time 2 0 IDENTIFIED LEAKAGE 1

SAT 2.3 Steam Generator (SG)

UNSAT A. SG1 Leakage GPD Critical Step B. SG2 Leakage =

GPO C. SG3 Leakage =

GPD D. SG4 Leakage =

GPD Total SG leakage =

GPD

=

GPM 1440 mu/day The examinee determines the Steam Generator Leakage is between 0.009 and 0.01 gpm.

STEP 10 RCDT Leakage Performed By Reviewed By The Examinee determines Reactor Coolant Drain Tank (RCDT) is Standard:

between zero and 0.03 gpm.

Comment NOTE I SG leak rates are obtained by 1, 2Sl--CEM0681 37.5.

0SICEM000050.2 is performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and looks at activity levels which will determine if those leak rates from the last performance of 1, 2Sl--CEM---0681 37.5 have risen.

NOTE 2 Individual steam generator leakage may be NIAd if activity is too low to be separately identified.

STEP 11 Standard:

Comment

1201 NRC RO Admin JPM A.2 Page 11 of 15 STEP 12 2.0 IDENTIFIED LEAKAGE SAT 2.3 Steam Generator (SG)

UNSAT Date/Time OSI-CEM000--050.2 performed:

/

Date Time Date/Time 1. 2SICEM068137.5 performed:

/

Date Time Recorded by Standard The examinee obtains the date and time of the steam generator leakage surveillance from the initiating cue.

Comment

1 Tme CLA2Leakage=

C. CLA 3 Leakage = [(Fnaolume)(lnitialVume)]

-min k.

Time

[(Final Volume gal)

- (Initial Volume gal)]

mm Time GPM CLA4Leakage=

Total CLA leakage =

GPM 1201 NRC RO Admin JPM A.2 Page 12 of 15 STEP 13 NOTE CLA leakage calculations may be marked N/A if RCS pressure is greater than 683 psig and quantification of CLA leakage isj required.

2.0 IDENTIFIED LEAKAGE 2.4 Cold Leg Accumulators (CLA)

[(Final Volume gal)

- (Initial Volume gal)]

A. CLA I Leakage

-- f Tme CLA 1 Leakage

[(Final Volume gal)

- (Initial Volume gal)]

B. CLA 2 Leakage =

SAT UNSAT CLA 3 Leakage =

DCLA4 Leak Performed By Reviewed By Standard:

The Examinee determines the Cold Leg Accumulators (CLA) is zero.

Comment

1201 NRC RO Admin JPM A.2 Page 13 of 15 NOTE The following value may be marked NIA if no additional identified leakage sources (i.e., Appendix A) are to be included.

STEP 14 2.0 IDENTIFIED LEAKAGE SAT 2.5 Other Identified Leakage UNSAT Other identified leakage (Appendix A> =

GPM Standard:

The Examinee determines the Other Identified Leakage is zero.

Comment NOTE if NO CCPIT de-pressurization / pressurization activity has occurred during the stabilization period. COPIT/HUT leakage may be marked NIA.

STEP 15 2.0 IDENTIFIED LEAKAGE SAT 2.6 Total Identified Leakage UNSAT Total identified leakage

=

+

Critical Step PRT RCDT SG

+

Total CLA CCPIT/HUT Other Total identified leakage

GPM St d

d The Examinee determines Total Identified Leakage is between 0.084 and an ar 0.09 gpm.

Comment

1201 NRC RO Admin JPM A.2 Page l4of 15 STEP 16 3.0 UNIDENTIFIED LEAKAGE SAT Unidentified leakage

=

UNSAT Total leakage Identified leakage Critical Step Unidentified leakage =

GPM Performed By Reviewed By Standard:

The Examinee determines Unidentified Leakage is 1.07 gpm.

Comment

¶ 5.0 ACCEPTANCE CRITERIA STEP 17 SAT A. The Technical Specification acceptance criteria is as follows:

UNSAT

1. The primarytosecondary leakage must be less than or equal to 150 GPD Critical Step through any one steam generator.

2. The total identified RCS leakage must be less than or equal to 10 GPM.

3. The total unidentified leakage must be less than or equal to 1 GPM.

B.

If any of the Technical Specification criteria are NOT satisfied, the SM must be notified and action requirement (b) of LCQ3.4.6.2 satisfied, The Examinee discoversthe total unidentified leakage greaterthan I Standard.

GPM.

Comment Terminating The JPM is complete when the Examinee returns the JPM briefing STOP Cue:

sheet to the Examiner.

Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

1.

Unit2isinMODEl.

2.

The performance of 0-Sl-OPS-068-1 37.0, Reactor Coolant System Water Inventory is due.

3.

ICS is not available.

4.

Data for the 0-SI-OPS-068-1 37.0, Reactor Coolant System Water Inventory Appendix C is as follows:

Parameter Final Initial VCT Level 42.9 50.1 Pressurizer level 58 58 RCS Temperature 578 578 RCS Pressure 2235 2235 PRT Level 12358 12349 RCDT Level 172.2 171.8 CLA#1 Level 7692 7692 CLA #2 Level 7677 7677 CLA #3 Level 7672 7672 CLA #4 Level 7680 7680 10.

Appendix D is complete through section 1.4 SG leakages SG#1 SG#2 SG#3 SG#4 3.2 gpd 2.2 gpd 2.3 gpd 4.8 gpd 11.

0-SI-CEM-000-050.2 and 2-SI-CEM-068-1 37.5 were completed last night at 2300.

12.

Chemistry supervisor reported no additions or samples taken from the RCS during the selected performance period.

13.

CCPIT/HUT leakage and other leakage is 0 gpm.

14.

Data collected for performance of this leak rate is documented in Appendix C and was collected over 120 minutes INITIATING CUES:

1.

You are the Unit 2 CR0 and have been directed to complete a manual RCS Leakage calculation using 0-SI-OPS-068-1 37.0, Reactor Coolant System Water Inventory Appendix D 2.

Record your answers on the Appendix D.

3.

Identify all deviations (if any) on the Chronological Test Log 4.

Notify the Examiner when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

ts4jLo 1JVLRA A.7..

--.v-

SQN REACTOR COOLANT 0-SI-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page5Oof6l NOTE I NOTE 2 NOTE 3 kE(

All calculations must be performed manually. Verification may be performed using the computer.

All calculations should be carried out to two places after the decimal.

Signs must be carried through all calculations.

Calculations in this Appendix may need to be performed more than once.

If multiple calculations are required, copies of this Appendix may be used.

1.0 TOTAL RCS LEAKAGE 1.1 Volume Control Tank (VCT)

VCT Leakage

=

So, I %)-.(

I12A

%)}x19.27gal/%

Initial Level Final Level A Time mm.

VCT Leakage =

1.2 Pressurizer (Pzr) UNIT 2 only

[I]

NOTE I NOTE 2 GPM CALCULATE initial PZR volume by using the following equation:

Vf (specific volume of a fluid) and Vg (specific volume of a gas)values for current plant conditions can be found using a copy of the steam tables.

In the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

L gal*ft3/

2.02

/lb

%level

}

m RCS LEAKA13E cALcuLATIdNS APPENDIX D Page 1 of 9 PZRLeve1 100% PZRLeveI

+

Initial PZR Vol =[

fL flA A.1

rLQ,

)Rc SQN REACTOR COOLANT O-SI-OPS.O68-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page5lof6l APPENDIX D Page 2 of 9 Initial PZR Vo1[

100-ft3/

ft 3/

/lb

/

m

/

m ft3/

/lb

%level m

Initial PZR Vol =

1q67. gal

[2] CALCULATE Final PZR volume by using the following equation:

Performed By Reviewed By NOTE I NOTE 2 Vf (specific volume of a fluid) and Vg (specific volume of a gas)values for current plant conditions can be found using a copy of the steam tables.

In the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

/

I*2.021

/

/lb

%level

)

/

m Final PZR Vol Final PZR Vol =

ft3/

/lb

%level m

Performed By

+

Final PZR Vol [

PZRLeve1 100%

PZRLevel Vf Vg Reviewed By

iLoi MILL kD A.4 1

,b 1

n,%

a12

[3] CALCULATE PZR leakage rate by using the following equation:

APPENDIX D Page 3 of 9 PZR Leakage =

PZR Leakage =

Initial PZR Volume Final PZR Volume Ltime q

1

çç gal -

al tO mm PZR Leakage =

0 gpm 1.3 Pressurizer (Pzr) UNIT I only t%1

[1] CALCULATE initial PZR volume using the following equation:

Performed By Reviewed By NOTE I Vf (specific volume of a fluid) and Vg (specific volume of a gas)values for current plant conditions can be found using a copy of the steam NOTE 2 In the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

+

Initial PZR Vol

=(

PZRLeve1 100% PZRLeveI 2.01 gal

ft3 Vt Vg

ft3/

/lbm * %level

SQN REACTOR COOLANT O-Sl-OPS-068-I 37O SYSTEM WATER INVENTORY Rev: 26 1&2 Page52of6l tables.

%level Initial PZR Vol [

/lb 1

Initial PZR Vol =

100

+

ft3/

/lbm Performed By Reviewed By

lloi ND.L Ro& AL SQN REACTOR COOLANT 0-SI-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page53of6l

[2] NCALCULATE Final PZR volume using the following equation:

APPENDIX D Page 4 of 9 NOTE I V (specific volume of a fluid) and Vg (specific volume of a gas)values for current plant conditions can be found using a copy of the steam tables.

NOTE 2 In the following equation the multiplier is the Pressurizer Volumetric Constant determined by engineering.

Final PZR Vol =1 PZRLeve1 + 100%PZRLeve1 \\,

201 gal*ft 3

Vt Vg

)

Final PZR Vol =[

100 ft 3/

/lbm ft/

/lbm * %level Final PZR Vol =

[3] CALCULATE PZR leakage rate using the following equation:

Performed By Reviewed By PZR Leakage =

PZR Leakage =

Initial PZR Volume Final PZR Volume ztime gal

gal m

Performed By

%level PZR Leakage =

Reviewed By

riot sLC At SQN REACTOR COOLANT 0-SI-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page54of6l APPENDIX D Page 5 of 9 1.4 RCS Temperature Correction[C.1]

NOTE The multiplier used in the following equation is the RCS Temperature Correction Volumetric constant as determined by engineering.

A.

Initial Conditions:

RCS pressure = 2.3 psig + 14.7 22,1 psia RCS temperature =

Specific volume (v1) =

O12I ft 3

/lbm Density = 1/v1 =

(14(.Ii Ibm/ft 3

B. Final Conditions:

RCS pressure =

22 psig + 14.7 = 22,7 psia RCStemperature= S7

°F Specific volume (v1) =

f ft 3

/lbm Density = 1/v 1 =

6jL, r7 Ibm/ft 3

C. Temperature Correction:

[(

14tL17 Ibm/ft 3 )

( L!1 Ibm/ft 3 )j

1298.19 gal*ft 3

/ Ibm Temp Corr. =

Initial Density Fina Density time Temperature Correction =

9 GPM

I Loi tPc (Lo A.t A 414 SQN REACTOR COOLANT 0-Sl-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page55of6l APPENDIX 0 Page 6 of 9 1.5 Total RCS Leakage Total RCS Leakage = (

1iI(

) + (

)

VCT PZR Temp Corr

jotal RCS Leakage =

I Performed By Reviewed By 2.0 IDENTIFIED LEAKAGE NOTE Leakage into the PRT or CLA tanks will always be a positive value (or zero) when RCS pressure is greater than the specific tanks pressure.

If negative leakage is calculated under these conditions, that tanks leakage must be set equal to zero.

2.1 Pressurizer Relief Tank (PRT) cjal)(

iz.9 gal)]

PRT Leakage Final Volume Initial Volume J

mm)

... 4 Time LjeakPMj NOTE RCDT Level (2.2) may be marked N/A if conditions do NOT permit obtaining RCDT level.

2.2 Reactor Coolant Drain Tank (RCDT)

[(

Qal)(

I71i gal)]

RCDT Leakage =

Final Volume Initial Volume

(

IL mm) zTime RCDT Leakage =

GPM Performed By Reviewed By

rtoi IhJRL RD L

SQN REACTOR COOLANT 0-SI-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page56of6l APPENDIX D Page 7 of 9 2.3 Steam Generator (SG)

NOTE I SG leak rates are obtained by 1, 2SlCEM0681 37.5.

OSl-CEM000050.2 is performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and looks at activity levels which will determine if those leak rates from the last performance of 1, 2S[---CEM.-0681 37.5 have risen.

NOTE 2 Individual steam generator leakage may be NIAd if activity is too low to be separately identified.

A. SGI Leakage =

GPD B. SG2Leakage=

2.2.

GPD C. SG3 Leakage =

GPD D. SG4 Leakage =

GPD Total SG leakage =

i2..5 GPD O.j GPM 1440 mm/day Date/Time 0SICEM000050.2 performed: L* NiA /

Date Time Date/Time 1, 2SICEM0681 37.5 performed: -44 Nf Date Time Recorded by

1ei.ikL D A).t A.2 SQN REACTOR COOLANT O-SI-OPS-O68-137O SYSTEM WATER INVENTORY Rev: 26 1&2 Page57of6l 2A Cold Leg Accumulators (CLA)

APPENDIX D Page 8 of 9 NOTE CLA leakage calculations may be marked NIA if RCS pressure is greater than 683 psig and quantification of CLA leakage is j required.

CLA1 Leakage=_

B. CLA 2 Leakage =

CLA2 Leakage=_

11 7 gal

- I gal

[kFinal Volume

)

kinitialVolume

I mm ATime

/7 GPM I_I gal

- I 7C77 gal

[IFinal Volume

)

IInitial Volume (ii I

mm Tmme C

GPM C. CLA 3 Leakage =

CLA 3 Leakage =

D. CLA 4 Leakage =

CLA 4 Leakage =

II 12-gai - I

%72 gal

[Final Volume

)

jnitial Volume

(

2M mm zTmme O

GPM

[1

-)w&

al-1

[LFinal Volume g ijnitial Volume g mm ATime 0

GPM Total CLA leakage =

0 GPM Performed By A. CLA I Leakage Reviewed By

iio NP.c lb A.1 SQN REACTOR COOLANT 0-SI-OPS-068-1 37.0 SYSTEM WATER INVENTORY Rev: 26 1&2 Page58of6l 2.5 Other Identified Leakage APPENDIX D Page 9 of 9 NOTE The following value may be marked NIA ifj additional identified leakage sources (i.e., Appendix A) are to be included.

Other identified leakage (Appendix A) =

GPM NOTE If NO CCPIT de-pressurization / pressurization activity has occurred during the stabilization period, CCPIT/HUT leakage may be marked 2.6 Total Identified Leakage Total identified leakage

=

Otg

+

D. 0

+

PRT RCDT SG Total identified leakage =

cr), O?

GPM I

3.0 UNIDENTIFIED LEAKAGE Unidentified leakage =

Total leakage Identified leakage Unidentified leakage =

1.07 GPM Performed By N/A.

+

Total CLA CCPIT/HUT Other Reviewed By

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC RO ADMIN A3

1201 NRC RO ADMIN A.3 Page 2 of 6 SRO JOB PERFORMANCE MEASURE Task:

Pre Job Brief for Emergent Work in the RCA Task#:

3430290302 Task Standard:

In preparation for a Pre-job brief with an AUO, the Examinee chooses correct survey map and calculates a total dose of 480 mrem for the job.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.3.13 (3.4/3.8)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name! Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

14 mm Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC RO ADMIN A.3 Page 3 of 6 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/Equ ipmentlProcedures Needed:

1.

Calculator 2.

Survey map A216 U-i Pipe Chase 3.

Survey map A217 U-2 Pipe Chase 4.

Survey map 408 u-i Mixed Bed Valve Gallery 5.

Survey map 417 U-2 Mixed Bed Valve Gallery

References:

Reference Title Rev No.

2.

RCI-3 PERSONNEL MONITORING 48 Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1.

Unit2 is in MODEl 2.

You are the CRC and are preparing for a pre-job brief for an AUC to make an emergent entry in to the RCA.

3.

The work is 1 meter from valve 2-63-750 Test Conn on RHR Suction from RWST located in the Auxiliary Bldg 653 elevation pipe chase.

4.

The total time to perform the job is 36 mm INITIATING CUES:

1.

Calculate the total dose that the worker will accumulate performing the job 2.

Inform the evaluator when you are complete.

1201 NRC RO ADMIN A.3 Page 4 of 6 Start Time STEP 1

Obtain the correct copy of a Survey map.

SAT UNSAT Critical step Standard Examinee discriminates and chooses Survey map A217 U-2 Pipe Chase of the four given.

Provide a copy of the following survey maps:

Survey map A216 U-i Pipe Chase L

Survey map A217 U-2 Pipe Chase Survey map 408 U-i Mixed Bed Valve Gallery Survey map 417 U-2 Mixed Bed Valve Gallery Comment Examiner Both the Unit 1 and Unit 2 Pipe Chase and Mixed Bed Valve Gallery Note survey maps are provided.

1201 NRC RO ADMIN A.3 Page 5 of 6 STEP 2

Calculate the dose SAT UNSAT Critical step d

Examinee calculates a total dose of 480 mrem will be accumulated an ard.

perfoing the task Comment The work will be performed near valve 2-63-750 Test Conn on RHR Suction from RWST. General area dose rate from survey map A21 7 Examiner from a hot spot in the area is 800 mr/hr. The job as indicated from the Note initiating cue will take 36 minutes.

800 mrem/hr ÷ 60 mm/i hr X 36 mm

= 480 mrem Terminating The task is complete when the Examinee returns the cue sheet to ue:

the examiner.

Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

INITIAL CONDITIONS:

1.

Unit2isinMODEl 2.

You are the CRC and are preparing for a pre-job brief for an AUO to make an emergent entry in to the RCA.

3.

The work is 1 meter from valve 2-63-750 Test Conn on RHR Suction from RWST located in the Auxiliary Bldg 653 elevation pipe chase.

4.

The total time to perform the job is 36 mm INITIATING CUES:

1.

Calculate the total dose that the worker will accumulate performing the job 2.

Inform the evaluator when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

Rad Con MaDs:

A21 6.WMF A21 7.WMF A408.WMF A41 7.WMF

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE SEQUOYAH NUCLEAR PLANT SRO ADMIN A.1.a

1201 NRC RO ADMIN A.1.a Page 2 of 8 SRO JOB PERFORMANCE MEASURE Task:

Review Completed Refueling Checklist and Implement Technical Specification Requirements.

Task#:

0001430302 Task Standard:

Exam flee reviews the completed Refueling Checklist Log Appendices and discovers two embedded errors, than applies LCO 3.9.10 to stop fuel movement and LCO 3.9.8.2 action ato immediately commence corrective action to return the required RHR loops to OPERABLE status as soon as possible.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.1.40 (3.9)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name I Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

12 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC RO ADMIN A.1.a Page 3 of 8 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/EquipmentiProcedures Needed:

1.

In progress copy of 0-SI-OPS-000-065.S, Refueling Surveillance Log 2.

Technical Specifications, Unit 1 3.

Chronological Test Log

References:

Reference Title Rev No.

1.

0-Sl-OPS-000-065.S Refueling Surveillance Log Rev 17 2.

Unit 1 Technical Specifications Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM. I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1.

Unit 1 is in MODE 6 on day shift.

2.

Fuel movement is in progress using FHI-3 MOVEMENT OF FUEL, 35 fuel bundles have been loaded into the core.

3.

The loading of the 36th bundle was in progress when data was obtained.

4.

0-SI-OPS-000-065.S, Refueling Surveillance Log has been completed INITIATING CUES:

1.

You have been directed to review the completed copy of 0-SI-OPS-000-065.S, Refueling Surveillance Log.

2.

When you have completed your review, identify all deviations (if any) on the Chronological Test Log and take all required action.

3.

Inform the evaluator when you are complete.

1201 NRC RO ADMIN A.1.a Page 4 of 8 Start Time Obtain a copy of 0-S-OPS-00O-065.S, Refueling Surveillance Log STEP 1

SAT UNSAT Standard:

Copy of 0-SI-OPS-000-065.S, Refueling Surveillance Log is obtained Cue Provide an in progress copy of 0-SI-OPS-000-065.S, Refueling Surveillance Log Comment

1201 NRC RO ADMIN A.1.a Page 5 of 8 I

Refueling Canal Ladder in I

STEP 2

ReactorVessel SR49.IO 6,7 6

23feetabove I equipmentpitinsideupper rungson SAT Level reactor vessel I

containment vIe 734 ladder flange UNSAT Examinee reviews 0-SI-OPS-000-065.S, Refueling Surveillance Log

Appendix A for Reactor Vessel Level.

Critical Step Examinee determines Reactor Vessel Level of 9 rungs on the ladder (724 Standard:

10.5 is insufficient.

REACTOR CAVITY LEVEL INDICATOR (Refueling Canal Ladder)

RUNG NO.

- EL733-9 1(2 (TOP of CURB) zz EL 725-lO 112 NORMAL TECH SPEC WATER MIN. LEVEL LEVEL 725-l 112 726-l 112 j._

-EL7I7-lO 1I2 z2Qz

-EL713-10112 Examiner Note:

- EL 701-lO 112 REACTOR VESSEL FLANGE 702-l 1/2

42 Fl R1-1fl 112 REFUELING CANAL FLOOR 690-lO 112 REFUELING CANAL (EQUIPMENT PIT) LADDER 42 RUNGS SPACED AT 1-O Notes

7) Minimum TECH SPEC level is 725 ft and 1 1/2, If water level is at 8 rungs from the top of the Refueling Canal ladder, the level will be at elevation 725 ft and 10 1/2.

Comment

1201 NRC RO ADMIN A.1.a Page 6 of 8 RHR loop LCO 19.82 8

8 operable A-A RHR Pump STEP 3

operab8ty SAT operable B-B RHR Pump Examinee reviews 0-Sl-OPS-000-O6SS, Refueling Surveillance Log UNSAT Appendix A for RHR Loop operability.

Critical Step Examinee determines two loops are required operable if Reactor Vessel d

Cl*

Level is less than 23 ft above the Reactor Vessel flange and immediately an ar initiates correctiie action to return the required RHR loops to OPERABLE status as soon as possible.

Comment Notes Examiner Note:

8) Two loops are required operable if Reactor Vessel level is less than 23 ft above the Reactor Vessel flange.

STEP 4

UMLTINS CONDITIONS FOROPERTIONS SAT j 39.10 At east 23 feet of water shah be maintained over the top of the reactor pressure vessel flange.

APPUCABILIT(:

UNSAT Critical Step Owing movement of Irrathated fuel assembires wlthm conlalnmenL ACTION:

With the requirements of the above specification not satisfied, immediately suspend operations involving movement of irradiated fuel assembles within containment.

Standard Eiiiinee enters the LCO 31O action and suspends of movenientof irradiated fuel assemblies withiftcontainment Comment

1201 NRC RO ADMIN A.1.a Page 7 of 8 UMO1NG GOND111ON FOR OPERATION STEP 5 :

SAT 39.&2 Two independent Residual Heat Removal (RHR) bops shaH be OPERABLE*

APPUCABILITY: MODE $ wten the water level above the top of the reactor pressure vessel flange s U NSAT less than 23 feet,.

Critical Step ACTION:

a Wfh less than the required RHR loops OPERABLE, mmed[ately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b.

The provisions of SpecifIcation 3.0.3 are not applicable.

Standard:

xamjnee enters LCO 3.9.82 action &

Comment Terminating The JPM is complete when the Examinee returns the JPM STOP Cue:

briefing sheet to the Examiner.

Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

INITIAL CONDITIONS:

1.

Unit 1 is in MODE 6 on day shift.

2.

Fuel movement is in progress, 35 fuel bundles have been loaded into the core.

3.

The loading of the 36th bundle was in progress when data was obtained.

4.

O-SI-OPS-000-065.S, Refueling Surveillance Log has been completed INITIATING CUES:

1.

You have been directed to review the completed copy of O-SI-OPS-000-065.S, Refueling Surveillance Log.

2.

When you have completed your review, identify all deviations (if any) on the Chronological Test Log and take all required action.

3.

Inform the evaluator when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

(tot 1L I hI1 Sequoyah Nuclear Plant Unit I & 2 Surveillance Instruction O-Sl-OPS-000-065.S REFUELING SURVEILLANCE LOG Revision 0017 Quality Related Level of Use: Continuous Use Effective Date:

10-03-2007 Responsible Organization:

OPS, Operations Prepared By:

Patty Burchard / Joe Postell Approved By:

W. T. Leary Current Revision Description This procedure was converted from Word 95 to Word 2002 (XP) using Rev 16. No technical changes were made.

SQN REFUELING SURVEILLANCE LOG O-Sl-OPS-000-06&S Unitl&2 Rev.0017 Page 6 of 15 Date roo-1 PREREQUISITE ACTIONS Throughout this Instruction where an IF/THEN statement exists, the step should be NIA if condition does NOT exist.

,7/7 PRELIMINARY ACTIONS ENSURE Precautions and Limitations, Section 3.0, has been Y

reviewed.

ENSURE Instruction to be used is a copy of effective version.

4.2 Measuring and Test Equipment, Parts and Supplies None 4.3 Field Preparations None

r SQN REFUELING SURVEILLANCE LOG O-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Page 7 of 15 5.0 ACCEPTANCE CRITERIA

,/Specific quantitative and/or qualitative requirements that are intended to be Lf verified by this Instruction are noted in the instruction steps where the verifying actions are performed and documented.

,7%) If specific acceptance criteria stated in the instruction steps are NOT met, notify the SM as soon as practical after observation of the noncompliance.

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Page 8 of 15 Date I 4

j PERFORMANCE COMPLETE Appendix A for applicable shift p.4<

[2]

IF Xl-92-5 (backup source range indication) is being used, THEN RECORD WA M&TE Instrument ID and instrument calibration due date below.

M&TEIDNo.

Cal Due Date__________

NØS 1)

If core alterations have been suspended for greater than or equal to one hour, then step 6.0[3] must be performed.

2)

The following step may be left open until the end of the 1830-0630 shift and then marked NIA if core alterations are in progress.

[3]

RECORD the time and date that Communications between the Control Room and Containment is verified within one (1) hourprior to the start of core alterations.

TIME i-1r DATE______

SQN REFUELING SURVEILLANCE LOG 0-Sl-OPS-000-065.S Unitl&2 Rev.0017 Page 9 of 15 Date t

6.0 PERFORMANCE (continued)

If movement of irradiated fuel assemblies (F/As) within containment have been suspended for greater than or equal to one hour, then step 6.0[4] must be performed. The following step may be left open until the end of the 1830-0630 shift and then marked NIA if movement of irradiated F/As within containment is in progress.

RECORD the time and date that the reactor vessel level above the flange is verified to be greater than or equal to 23 ft. within two (2) hoursprior to the start of movement of irradiated fuel assemblies within containment.

CN Reactor cavity vessel seal must be installed before increasing reactor vessel water level above reactor vessel flange (el 702).

(j?

IF inflatable cavity seal is installed, THEN COMPLETE Appendix D to verify Reactor Cavity Seal air supply.

DELIVER SI Package to the US/SRO for review and approval.

0630-1830 1830-0630 7.0 POST PERFORMANCE ACTIVITIES None

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Page 10 of 15 Appendix A (Page 1 of 2)

REFUELING SURVEILLANCE LOG SURVEILLANCE REF Notes Mode TS Limits Instrument No./Location Units 0630-1830 1830-0630 REMARKS Source Range SR 4.9.2 1,2,3 6

Operable Xl-92-5001/M-4 CPS Channel indication 4

Operable XI-92-50021M-4 CPS Operable Xl-92-5 CPS

/ 4 Communications established from the control room to the refueling TRM 4 9 5 5

6 startiori and from the refueling station to the control room.

Operators Initials Notes 1)

At least 2 SRMs operable with continuous visual indication in the Control Room.

2)

At least 1 SR Channel having audible indication in Containment and Control Room.

3)

Xl-92-5 (Backup Source Range Monitor) can be used as one of the source range channels if visual indication is provided in the control room.

IF X-92-5 is NOT providing visual indication in the control room, THAN MARK N/A.

4)

Any channel Not in service should be marked N!A.

5)

Direct communications between the control room and refueling station with clear audible contact in both directions must be established during core alterations, is core alteration Not inprogress, NIA communicadstions established.

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Page 11 ofl5 Appendix A (Page 2 of 2)

REFUELING SURVEILLANCE LOG SURVEILLANCE REF Notes Mode TS Limits Instrument No./Location Units 0630-1830 1830-0630 REMARKS Refueling Canal Ladder in Reactor Vessel SR 4.9.10 6,7 6

23 feet above equipment pit/inside upper rungs on Level reactor vessel containment ele 734 ladder flange RHR loop LCO 3.9.8.2 8

6 operable A-A RHR Pump operability Naf operable B-B RHR Pump V

RHR flow SR 4.9. 8.1 9,10 6

2000 gpm FI-63-91A1M6 gpm p

Fl-63-91B/M6 gpm 0

Fl-63-92A/M6 gpm FI-63-92B/M6 gpm Z..9 o Operators Initials Notes 6)

Reactor Vessel Level requirement are applicable during movement of irradiated fuel assemblies within containment.

7)

Minimum TECH SPEC level is 725 ft and 1 1/2, If water level is at 8 rungs from the top of the Refueling Canal ladder, the level will be at elevation 725 ft and 10 1/2.

8)

Two loops are required operable if Reactor Vessel level is less than 23 ft above the Reactor Vessel flange.

9)

At least one RHR pump in operation and recirculating flow greater than or equal to 2000 gpm.

10)

Record flow for each flow indicator. Fl-63-91A + Fl-63-92A or +FI-63-91B +FI-63-92B must total greater than or equal to 2000 gpm.

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Pagel2ofI5 Appendix B (Page 1 of 2)

REFUELING CAVITYIREFUELING CANAL DIAGRAM 733 7 1 !2 Floor Elev.

TopotCurb-73391/2 cs Vent Duct Opening

726 7 7261 1/2 Reactor Cavity Rx Vessel Flange Refueling Canal 7021 1/2 Upper and Lower Internals Storage Area 69010 1/2 Total gallons required to fill Upper and Lower Internals 56,782 gals.

Storage Area from el. 690 10 112 to Rx. Vessel Flange el. 702 1 3/4.

Total gallons required to fill Rx. Cavity from el. 702 1 238,580 gals.

3/4 to 726 1 1/2.

Total gallons required to fill Rx. Cavity (initial fill).

295,362 gals.

Total gallons required to fill Fuel Transfer Canal (initial 98,288 gals.

fill).

SQN REFUELING SURVEILLANCE LOG I O-Sl-OPS-000-065.S Unitl&2 Rev.0017 Page 13 of 15 Appendix B (Page 2 of 2)

REACTOR CAVITY LEVEL INDICATOR (Refueling Canal Ladder)

RUNG NO.

- EL 733-9 112 (TOP of CURB) 8

- EL725-lO 112

NORMAL

TECH SPEC WATER MIN. LEVEL LEVEL 725-l 112 726-l 112 16

- EL717-lO 112 20

- EL 713-lO 112 32

- EL 701-lO 1I2 REACTOR VESSEL FLANGE 702-l 112 42

- Fl R1-1fl 112 REFUELING CANAL FLOOR 690-lO 112 REFUELING CANAL (EQUIPMENT PIT) LADDER 42 RUNGS SPACED AT 1-O

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Pagel4ofl5 Appendix C (Page 1 ofl)

DEFINITIONS Fuel Handling Supervisor (FHS) - Fuel Handling Supervisor duties and responsibilities are referenced in OPDP-1.

Refueling Coordinators

- Individuals responsible for the supervision and coordination of all refueling activities except the actual fuel movement and core alterations.

Core Alterations

- Core Alterations shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the head removed and fuel in the vessel. Suspension of CORE ALTERNATIONS shall not preclude completion of movement of a component to a safe position.

Refueling shutdown

- the scheduled shutdown to replace or reposition core fuel assemblies.

Manipulator Crane Testing - shall be performed 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations (during movement of drive rods or fuel assemblies within the reactor vessel, TRM 3/4.9.6). Subsequent retesting may be performed at the discretion of the Fuel Handling Supervisor between resumption of core alterations or if a modification or maintenance to the crane is performed.

Appendix D (Page 1 ofl)

INFLATABLE CAVITY SEAL AIR SUPPLY NOTES 11)

Normal poerating pressure to the reactor vessel cavity seal is greater than or equal to 33 psig and less than or equal to 37 psig.

12>

Backup air supply bottle pressure greater than 500 psig.

13)

Not applicable if a non-inflatable cavity seal is used.

14)

Verify proper alignment as follows:

A.

Service air isolation valve open.

B Backup air supply valve on bottle open.

C Manfold shutoff valve open.

D Record ID. number of service air isolation valve the manifold is connected to:

SQN REFUELING SURVEILLANCE LOG 0-SI-OPS-000-065.S Unit I & 2 Rev. 0017 Pagel5ofl5 Surveillance Notes Mode TS Umits Instrument No./Location Units 0630-1830 1830-0630 REI Ref Normal Operating Inflatable Reactor 11,13 6

Pressure Regulator located PSIG Vessel Cavity Seal in upper containment, ele IA

Backup Air Supply Bottle 12,13 6

Pressure located in upper PSIG containment, ele 734

Cavity Seal Air Supply 13,14 6

Manifold located in silver q

box in upper containment, ele_734 Operators Initials

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC SRO ADMIN SRO A.1.b

1201 SRO ADMIN JPM A.1.b Page 2 of 12 SRO JOB PERFORMANCE MEASURE Task:

Perform Technical Review of Shutdown Margin Calculation.

Task#:

0010040101 Task Standard:

Examinee completes a review of a completed 0-SI-NUC-000-038.0 Shutdown Margin, finds two embedded errors and ultimately enters the LCO 3.1.1 ACTION to immediately initiate a boration of 35 gpm.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.1.37 (4.3/4.6)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name! Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

25 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 SRO ADMIN JPM A.1.b Page 3 of 12 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/Equipment/Procedures Needed:

1.

Unit2Cyclel8NDR 2.

Unit 2 Tech Specs 3.

0-Sl-NUC-000-038.0 Shutdown Margin In Progress copy 4.

0-Sl-NUC-000-038.0 Att 2 Unit 2 Cycle Specific SDM Data 5.

0-Sl-NUC-000-038.0 Att 3, Correction To Boron Concentration (M-P) 6.

TI-33 Xenon Worth Calculation 7.

Chronological Test Log 8.

Calculator

References:

Reference Title Rev No.

1.

Unit2Cyclel8NDR a

Unit 2 Tech Specs O-Sl-NUC-000-038.O Shutdown Margin 59 4.

O-SI-NUC-000-038.O Att 2 Unit 2 Cycle Specific SDM Data 5.

O-Sl-NUC-000-038.O Att 3 Correction To Boron Concentration (M-P) 6.

Tl-33 Xenon Worth Calculation 27 Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM. I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

INITIAL CONDITIONS:

1.

Unit 2 is in MODE 3, 350 deg F.

2.

Core Average Burnup is 6000 MWD/MTU.

3.

The present RCS Boron concentration is 1299 ppm.

4.

Unit 2 tripped from full power hours 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months ago.

5.

Unit 2 had been at 100% power for 2 weeks prior to the trip.

6.

0-Sl-NUC-000-038.0, Shutdown Margin, Data Sheet 1 has been completed.

7.

REACTF is unavailable 8.

ICS is unavailable INITIATING CUES:

1.

You have been directed to review the completed copy of 0-SI-NUC-000-038.0, Shutdown Margin data Sheet 1.

2.

When you have completed your review, identify all deviations (if any) on the Chronological Test Log and take all required action.

3.

Inform the evaluator when you are complete.

1201 SRO ADMIN JPM A.1.b Page 4 of 12 Start Time Obtain a copy of the following:

STEP 1

SAT Unit2Cyclel8NDR UNSAT 0-SI-NUC-000-038.O Shutdown Margin In Progress copy 0-SI-NUC-000-038.0 All 2 Unit 2 Cycle Specific SDM Data 0-SI-NUC-000-038.0 All 3 Correction To Boron Concentration (M-P)

Unit2Tech Specs Tl-33 Xenon Worth Calculation.

Copies of the following are obtained.

Unit2Cyclel8NDR 0-Sl-NUC-000-038.0 Shutdown Margin In Progress copy Standard:

0-Sl-NUC-000-038.0 Att 2 Unit 2 Cycle Specific SDM Data 0-Sl-NUC-000-038.0 Aft 3 Correction To Boron Concentration (M-P)

Unit 2 Tech Specs TI-33 Xenon Worth Calculation.

Provide copies of:

Unit2Cyclel8NDR 0-SI-NUC-000038.0 Shutdown Margin In Progress copy 0-Sl-NUC-.000-038.0 All 2 Unit 2 Cycle Specific SDM Data 0-Sl-NUC-000-038.0 Aft 3 Correction To Boron Concentration (M-P)

Unit 2 Tech Specs TI-33 Xenon Worth Calculation.

Comment

1201 SRO ADMIN JPM A.1.b Page 5 of 12 Data Sheet I STEP 2

(Page 1 of 4)

SAT SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 1.0 MInimum Boron Concentration (C 8)

UNSAT

[1]

A.

Dateltirne unit was shutdown Date Time B.

Current date/time Date Time C.

Mode SDM calculation is performed for:

MODE_____________

D.

Minimum expected core average temperature for surveillance interval E,

Time since plant shutdown F.

Steady-state core power level before plant shutdown G.

Core average burnup H.

Currentlprojected RCS C8 and associated date/time (N/A date and time for projected RCS boron concentration)

Date Time I,

Surveillance interval hours The examinee reviews Data Sheet 1 section 1.0 step 1 and notes no Standard:

errors.

Comment

1201 SROADMINJPMA.1.b Page 6 of 12 Data Sheet I STEP 3

(Pagelof4>

SAT SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 1.0 MInimum Boron Concentration (CB)

UNSAT Critical Step i

[2]

Minimum 08 I

The examinee reviews Data Sheet 1 section 1.0 step 2 and discovers the Standard:

error for minimum boron concentration for SDM from Table 6-4 of Nuclear Design Report and identifiesthe correctvalue of 1320 ppm.

Examiner The number entered on the Data Sheet 1 section 1.0 step 2 was 1231 Note ppm which applies to 450 deg F, whereas the correct answer is 1320 ppm which applies to 350 deg F.

Comment Data Sheet I STEP 4

(Page 1 of 4)

SAT SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 1.0 Minimum Boron Concentration (Ca)

UNSAT Critical Step

[3]

Correction to Minimum 08 (M-P)

The examinee reviws Data, Sheet t section tO step 3 and discovers for Standard the correction to boron concentration from 0Sl-NUC-00Q-Q38.0 Att 3 Correction To Boron Concentration (M-P) and identifies the correct value of 40 ppm.

Examiner The number entered on the Data Sheet 1 section 1.0 step 3 was zero Note which applies to Unit 1 whereas the correct answer is 40 which applies to Unit 2.

Comment

1201 SRO ADMIN JPM A.1.b Page 7 of 12 STEP 5

SAT SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 UNSAT

1.0 Minimum Boron Concentration (C 9)

Critical Step

[4]

Adj. Mm. Boron Conc. = Step 1.O[2] + Step tO[3]

Adjusted Minimum C 6

=

+

The examinee reviews Data Sheet 1 SectiOn tO step 4 and discovers the Standard:

error carried forward from steps 2 and 3. The examinee identifies the correct answer for the adjusted Minimjm GB is 1360 ppm.

Examiner The number entered on the Data Sheet 1 section 1.0 step 4 was 1231 Note:

ppm whereas the correct answer is 1360 ppm.

Comment STEP 6

SAT SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 1.0 Minimum Boron Concentration (C 9)

[5]

Was the minimum boron concentration provided by the Reactor Engineering memorandum used in Step 1.0(2]?

Yes D

No C

If yes, proceed to Section 6.0 and NIA (or discard) intermediate steps.

Standard The examinee reviews Data Sheet 1 section 1.0 step 5 identifies the answer to step 5 was No and proceeds to review step 6.

Comment

1201 SRO ADMIN JPM A.1.b Page 8 of 12 STEP 7

Data Sheet 1 SAT (Page 2 of 4)

UNSAT 1.0 Minimum Boron Concentration (C

8) (continued)

[6]

Are corrections to the adjusted minimum C6 recorded in 1.0[4] desired?

Yes D

No U

If No, proceed to Section 6.0 and NIA (or discard> intermediate steps.

Standard The examinee reviews Data Sheet 1 section 1.0 step 6 identifies the answer to step 6 was Yes and proceeds to review section 2.

Comment NOTE Xenon worths shall be entered as negative values.

STEP 8

Data Sheet SAT (Page 2 of 4)

UNSAT 2.0 Xe Worth

[1]

Xe Worth Standard:

The examinee reviews Data Sheet 1 section 2.0 step 1 and verifies the value of Xe Worth of -99 pcm from Table 6-40 of the Unit 2 NDR.

Comment

1201 SRO ADMIN JPM A.1.b Page 9 of 12 STEP 9

Data Sheet 1 SAT (Page 2 of 4) 30 Correction for Reduced SDM Requirement in Mode 5 UNSAT

[1]

Mode 5 SDM correction (negative value if other than zero) pcm Standard The examinee reviews Data Sheet 1 section 3.0 step 1 and verifies the Correction for Reduced SDM Requirement in Mode 5 is zero pcm.

Comment STEP 10 Data Sheet I SAT (Page 2 of 4) 4.0 Correction for Immovable or Untrippable Rods UNSAT

[1]

No. immovable/untrippable rods

[2]

A.) Maximum double stuck rod worth B.> Maximum stuck rod worth

[3]

Stuck rod worth = Step 4.O[2]A

- Step 4.O[2]B

=

[4]

Correction for immovable/untrippable rods Step 4.O[1] : Step 4.O[3]

x Standard The examinee reviews Data Sheet 1 section 4.0 step 1 and verifies the Correction for Immovable or Untrippable Rods is zero pcm.

Comment

1201 SROADMINJPMA.1.b Page 10 of 12 Data Sheet I STEP 11 (Page3of4)

SAT 5.0 Minimum Boron Concentration Calculation UNSAT (1)

Correction to Mm. Boron Conc.

2.0[1] + 3.O[1] + 4.0(4]

Critical Step Correction to Mm. Boron Conc. =

+

[2)

Inverse Boron Worth

[3]

Correction to Mm. Boron Conc. = Step 5.0[1]

Istep 5.0(2)1 I

I ppm

[4)

Correction to Mmn. SDM Boron Conc. = Step 1.0(4) + Step 5.0[3]

The examinee reviews Data Sheet I section 5.0 and d)scovers the error Standard rned forward from Section tO steps 2 and 3. The examihee identifies the correct answer for the correction to the Minimum SOM Boron Concentration is 1348 ppm.

Comment 1.

Data Sheet 1 STEP 12 (Page 3 of 4)

SAT 6.0 Acceptance Criteria Verification UNSAT

[1]

Required Boron Concentration for Shutdown Y4argin Critical Step

[2)

CHECK appropriate box and sign to indicate whether the following acceptance criteria was satisfied.

Acceptance Criteria:

The RCS boron concentration 1.0[1]H ms greater than the required boron concentration of 6.0(1].

Yes U

No U

Test Director Signature Date The examinee revIews Data Sheet 1 section 6.0 and discovers the Standard acceptance criteria is not met since current Boron Concentration given in the initial conditions is less than the Required Boron Concentration for Shutdown Margin specified in step 1.

Comment

1201 SRO ADMIN JPM A.1.b Page 11 of 12 STEP 13 oata:ti SAT 6.0 Acceptance Criteria Verification (continued)

UNSAT

[31 IF acceptance criteria from Step 6.0[2J was NOT satisfied, THEN IMMEDIATELY NOTIFY the US that acceptance criteria has NOT been satisfied and borate per action requirement of LCO 3.1.1.1 or LCO 3.1.1.2.

Test Director Signature Date

[41 Was boron concentration provided by Reactor Engineering memorandum used in Step 1.0[21?

Yes 0

No C

IF yes, proceed to Section 8.0 and NIA (or discard) Section 7.0.

Standard The examinee reviews Data Sheet 1 section 6.0 step 3 and transitions to Unit2LCO3.1.1.

Comment LMrnNc: CDNITiCN FGR OPERAT.ON STEP 14 SAT 3.1 1

1 The SHUTQN tRGIN shai be grea:er than or equa :o 1 % delta li for4 loop operaton.

UNSAT

.PLlOBlLITY: MOSES 1. 2. 3. arc 4.

Critical Step cTlCN:

A th the ShUTDC.VN MRGIN less than 1 6% dea Mc. nin-edarely n bate ird ccinue bora.on at reaterran or equal to 35 qpm ala sou on corta n.ng gre..ter than c equal to 6120 ppm boron or eqi1a:ent util the required SHUTDGAN MARGIN is restared Examinee enters the LCO 3.1.1.1 action which calls for immediate d

d initiation of boration at greater than or equal to 35 gpm of a solution an ar containing greater than or equal to 6120 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

Comment Terminating The JPM is complete when the Examinee returns the JPM STOP Cue:

briefing sheet to the Examiner.

Stop Time

1201 SRO ADMIN JPM A.1.b Page 12 of 12 JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

1.

Unit 2 is in MODE 3, 350 deg F.

2.

Core Average Burnup is 6000 MWD/MTU.

3.

The present RCS Boron concentration is 1299 ppm.

4.

Unit 2 tripped from full power hours 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months ago.

5.

Unit 2 had been at 100% power for 2 weeks prior to the trip.

6.

0-SI-NUC-000-038.0, Shutdown Margin, Data Sheet 1 has been completed.

7.

REACTF is unavailable 8.

ICS is unavailable INITIATING CUES:

1.

You have been directed to review the completed copy of 0-SI-.NUC-000-038.0, Shutdown Margin data Sheet 1.

2.

When you have completed your review, identify all deviations (if any) on the Chronological Test Log and take all required action.

3.

Inform the evaluator when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

17DI tRc-Mn AJ A.Lh

SQN SHUTDOWN MARGIN 0-Sl-NUC-000-038.0 Unit 0 Rev. 0059 Page 44 of 72 Appendix A Aotc*

(Page 1 of 13)

SHUTDOWN MARGIN CALCULATION FOR MODES 3, 4, AND 5 1.0 Maximum Boron Concentration (C 8)

NOTES 1)

The output of the computer code REACTF can be substituted and attached to satisfy the calculation of Appendix A, or Appendix A can be used if calculations are performed by hand.

If computer code REACTF output is attached, then Appendix A can be discarded. When running REACTF interactively, from the WINDOWS icon, a STAND ALONE RUN is being performed. Therefore, enter zero for the CURRENT RUN #

(refer to REACTF Release Documentation Package for additional information).

2)

If the computer code REACTF is being used to calculate SDM, then refer to the REACTF Users Guide. The Xe and I concentrations are entered at time of trip or when the unit goes subcritical from a controlled shutdown. The above concentrations can be obtained from the plant computer post-trip review, if the unit is tripped, or from the plant computer at the time the unit went subcritical from a controlled shutdown from the plant computer addresses U2107 and U2106. The values for Xe and I must be retained for the specific outage so that future SDM calculations can be performed.

3)

Appendix A can be performed for current RCS boron concentrations or projected boron concentrations. Projected boron concentration for this appendix and an official calculation is a boron concentration less than the current boron concentration while for a scoping calculation the projected boron concentration may be less or greater than the current boron concentration.

4)

All tables and figures referenced in this Instruction are contained in the Nuclear Design Report for the applicable fuel cycle unless specified otherwise. In the event the Nuclear Design Report has not been formally received, other applicable vendor data may be used provided it is properly documented.

5)

Absolute value sign is designated in this appendix by RECORD the following information on Data Sheet 1:

Date/time unit was shutdown. (Following a refueling outage write After Refueling Outage)

Current date/time.

Mode calculation is performed for.

Minimum expected core average temperature for surveillance interval.

SQN SHUTDOWN MARGIN 0-Sl-NUC-000-038.0 Unit 0 Rev. 0059 Page 45 of 72 Appendix A (Page 2 of 13) 1.0 Maximum Boron Concentration (CB) (continued)

Time since plant shutdown. (Time at start of surveillaçce interval)

(Following a refueling outage use zero).

Reactor power prior to shutdown (steady state core power level from daily Nuclear Log/Reactor Conditions Report, Operations Log or averaged over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from other data). (Following a refueling outage the Reactor Power Prior to Shutdown is zero.)

Core Average Burnup.

Current/projected RCS boron concentration, date and time. (N/A date and time for projected RCS boron concentration).

()

Surveillance interval.

RECORD on Data Sheet 1 the minimum boron concentration for SDM from Table 6-4 of Nuclear Design Report using minimum expected core average temperature,1.O[1]D, and core average burnup, 1.O[1jG, OR USE the boron concentration provided by Reactor Engineering memorandum.

RECORD on Data Sheet I correction to boron concentration (M-P) to account for differences between measured and predicted core conditions (see Attachment 3).

NOTE The boron concentration in the following step assumes shutdown banks are fully inserted, no xenon is present, HZP equilibrium samarium and plutonium concentrations, 1600 pcm of shutdown margin is provided, and a safety factor of at least 55 ppm has been added.

CALCULATE on Data Sheet 1 the adjusted minimum boron concentration using the following equation:

Adj. Mm. Boron Conc. = Step I.0[2j + Step 1.0[3]

SQN SHUTDOWN MARGIN 0-SI-NUC-000-0380 Unit 0 Rev. 0059 Page 46 of 72 Appendix A (Page 3 of 13) 1.0 Maximum Boron Concentration (CB) (continued)

IF the boron concentration provided by Reactor Engineering memorandum was used in Step 1.0[2], THEN MARK Data Sheet 1 YES, AND PROCEED to Section 6.0, AND N/A (or discard) intermediate steps, OTHERWISE MARK Data Sheet 1 NO, AND PROCEED with procedure.

IF corrections to the adjusted minimum boron concentration recorded in Step 1.0[4] are desired for any of the following:

Xenon concentration, OR B.

Reduced shutdown margin requirement in Mode 5, OR C.

Untrippable or immovable rods, THEN MARK Data Sheet 1 YES, AND PROCEED to Section 2.0, OTHERWISE MARK Data Sheet 1 NO, AND PROCEED to Section 6.0, AND N/A (or discard) intermediate steps.

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 47 of 72 Appendix A (Page4of 13) 2.0 Xenon Worth 1)

Ensure correct section of TI-33 is performed with regard to steady state or transient plant conditions. Steady state is defined as operation at a given power level (plus or minus 5%) for the last 3 days before plant shutdown.

2)

Xenon worth should be calculated for both the beginning and end of the surveillance nterval. The least negative of the xenon worths should be used in the following step.

[Ci]

3)

Xenon worth must be negative.

DETERMINE Xenon worth by performing TI-33, AND RECORD on Data Sheet 1, AND ATTACH applicable supporting documentation to the Data Package. [Ci]

3.0 cJ/iorrection for Reduced SDM Requirement in Mode 5

[1]

IF the SDM calculation is for Mode 5, THEN RECORD Mode 5 SDM correction of -600 pcm on, Data Sheet 1, OTHERWISE ENTER zero (0) pcm.

Ai 4.0 jCorrection for Immovable or Untrippable Rods

[1]

IF the number of immovable/untrippable rods is zero, THEN RECORD zero for number of rods, AND N!A Step 4.0[2], AND RECORD zero (0) pcm for the reactivity correction in Step 4.0[3], OTHERWISE RECORD the number of immovable/untrippable rods on Data Sheet I

/bJ 1

4 Q

r SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 I

Unit 0 Rev. 0059 L

Page 48 of 72 Appendix A (Page 5 of 13>

4.0 Correction for immovable or Untrippable Rods (continued)

NOTE The worth of the most reactive rod from NDR Table 635 used in Step 4,O[2j shall be consistent with plant conditions (i.e., core burnup, temperature).

[2]

RECORD on Data Sheet 1, the maximum double stuck rod worth and maximum stuck rod worth from Table 635 of Nuclear Design Report at the burnup specified in Step 1.O[1]G

[3]

CALCULATE and RECORD on Data Sheet 1, the stuck rod worth using the following equation:

Stuck Rod Worth = Maximum Double Stuck Rod Worth Maximum Stuck Rod Worth

[4]

CALCULATE and RECORD on Data Sheet 1, the reactivity correction for immovable/untrippable rods using the following equation.

Reactivity correction = Step 4.O[lj x Step 4.O[3]

Minimum Boron Concentration Calculation

(

CALCULATE on Data Sheet 1, the correction for Xe, Mode 5 and stuck rods as

appropriate using the following equation:

Correction to Mm Boron Concentration in pcm =

Step 2.O[1] + Step 3.O[1] + Step 4.O[4j RECORD on Data Sheet 1, the inverse boron worth from Tables 6-11 and 6-12 of Nuclear Design Report based on the core average burnup using Step 1.O[1]G, minimum expected core average temperature from Step 1.O[1]D, and boron concentration from Step 1.O[4].

CALCULATE and RECORD on Data Sheet 1, the correction to the minimum boron concentration using the following formula:

Correction to minimum boron concentration = Step 5.O[1]

Step 5.O[2]I

T SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 49 of 72 Appendix A (Page 6 of 13) 50 Minimum Boron Concentration Calculation (continued)

CALCULATE and RECORD on Data Sheet 1, the correction to the minimum SDM boron concentration (BC) using the following formula:

Correction to minimum SDM BC = Step 1.O[4] + Step 5.O[3]

9 Acceptance Criteria Verification NOTE The boron concentration recorded in the following step ensures that the shutdown margin required by Technical Specifications is available when all rods (including shutdown banks) are fully inserted.

DETERMINE required boron concentration for shutdown margin from 1 O[4]

or 5.O[4j, AND RECORD on Data Sheet 1.

CHECK appropriate box on Data Sheet 1, to indicate whether the following Acceptance Criteria was satisfied.

Acceptance Criteria:

The RCS boron concentration I.O[1JH is greater than the required boron concentration of 6.O[1}.

[3]

IF Acceptance Criteria from Step 6.O[2] was NOT satisfied, THEN IMMEDIATELY NOTIFY the Unit Supervisor (US) that acceptance criteria has NOT been satisfied and borate per action requirement of LCO 3.1.1.1 or LCO 3.1.1.2.

D

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 UnitO Rev. 0059 Page 50 of 72 Appendix A (Page 7 of 13) 6.0 Acceptance Criteria Verification (continued)

[4]

IF the boron concentration provided by Reactor Engineering memorandum was used in Step 1.0[2], THEN MARK Data Sheet 1 YES, AND PROCEED to Section 8.0, AND NIA OR (DISCARD) Section 7.0, OTHERWISE MARK Data Sheet 1 NO, AND PROCEED with procedure

T SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 UnitO Rev. 0059 Page 51 of 72 Appendix A (Page 8 of 13) 7.0 Boron Dilution Protection Evaluation NOTES 1)

This section determines if the current/projected boron concentration provides adequate protection against an inadvertent boron dilution event.

If any of the following conditions exist, this section may be marked NIA (or omitted from the data package).

Residual Heat Removal (RHR) system is NOT aligned to RCS.

Two out of four, primary water makeup pumps (IA, 18, 2A and 28) are tagged out with the pump suction or discharge valve closed and tagged; AND valve

[0-81-5191 is CLOSED and TAGGED.

If 0-81-519 is not closed and tagged then all four primary water pumps (1A, 18, 2A, and 28) need to be tagged along with their pump suction or discharge valves.

Shutdown banks are fully withdrawn.

Dilution flow paths are isolated.

2)

The flow of primary water is determined by the number of primary water pumps running (IA, 18, 2A, or 28) and the position of the cross over tie valve 0-81-519. Each primary water pump is equivalent to 150 gpm. Maximum primary water flow rate of concern is 300 gpm for protection against inadvertent dilution. Therefore, without 0-81-519 closed and tagged, two out of the four primary water pumps (1A, 18, 2A, or 2B) must be tagged off and the pump suction or discharge valves closed.

[1]

RECORD on Data Sheet 2, the current/projected RCS boron concentration from I.0[1]H.

D

[2]

RECORD on Data Sheet 2, the maximum predicted dilution flow rate for the surveillance interval.

D

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 52 of 72 Appendix A (Page 9 of 13) 7.0 Boron Dilution Protection Evaluation (continued)

NOTE The RHR flow rate in the following step must be greater than or equal to 1000 gpm. The reduction in the RHR flow due to swapping RHR trains does not need to be considered in the following step.

[3]

RECORD on Data Sheet 2, the minimum predicted RHR flow rate for the surveillance interval. (This flow rate must be greater than or equal to 1000 gpm.)

[4]

RECORD on Data Sheet 2, the all-rods-in-i (ARI-i) critical boron concentration from Table 63 of Nuclear Design Report (which includes at least 55 ppm conservatism) using minimum expected core average temperature 1.0[i]D and core average burnup 1.0[ijG.

D

[5]

RECORD on Data Sheet 2, correction to boron concentration total (M-P) to account for differences between measured and predicted core conditions (see Attachment 3).

D

[6)

CALCULATE on Data Sheet 2, the adjusted ARI-1 critical boron concentration (with at least 55 ppm conservatism) using the following equation.

Adj. ARI-1 Critical Boron Conc. = Step 7.0[4] + Step 7.0[5]

EJ

[7]

CHECK the appropriate box on Data Sheet 2, to indicate the status of the RCS as either filled or less than full.

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 53 of 72 Appendix A (Page 10 of 13) 7.0 Boron Dilution Protection Evaluation (continued)

NOTE Limiting Boron Concentration Ratio is defined as ARt-i Critical Boron Concentration divided by the boron concentration required to provide protection against an inadvertent boron dilution event.

[8]

DETERMINE the limiting boron concentration ratio by finding the intersection of Steps 7.0[2] and 7.0[3] on the appropriate figure/table as determined in Step 7.0[7], AND RECORD on Data Sheet 2.

D

[9]

CALCULATE on Data Sheet 2, the boron concentration required to provide protection against an inadvertent boron dilution event using the following equation:

C8 (dilution protection) = Step 7.0[6] ÷ Step 7.0[8]

[10]

CHECK appropriate box on Data Sheet 2, to indicate whether the following acceptance criteria is satisfied.

D Acceptance Criteria:

Current/Projected Boron Concentration in Step 7.0[1] is greater than or equal to the Boron Concentration in Step 7.O[9].

[ii]

IF acceptance criteria from Step 7.0[10] was NOT satisfied, THEN IMMEDtATELY NOTIFY the US that current RCS boron concentration does not provide adequate protection against an inadvertent boron dilution event, and borate per action requirement of LCO 3.1.1.1 or LCO 3.1.1.2.

D

SQN SHUTDOWN MARGIN 0-Sl-NUC-000-038.0 Unit 0 Rev. 0059 Page 54 of 72 Appendix A (Page 11 of 13) 8.0 Corrected Subcritical Boron Concentration with SID Banks Withdrawn NOTES 1)

An RCS boron concentration equal to or greater than the refueling boron concentration ensures that the reactor will remain in mode 3 (k-eff < 0.99) following shutdown bank withdrawal, If the current boron concentration is equal to or greater than the refueling boron concentration, or the boron concentration specified in memorandum directive by Reactor Engineering addresses k-eff being less than 0.99 with shutdown banks withdrawn, this section can be marked N/A (or omitted from the data package).

2)

If S/D banks are not to be withdrawn, then this section can be N/A (or omitted from the data package).

3)

This section determines if the current boron concentration is adequate to ensure k-eff will remain below 0.99 (i.e., mode 3) following shutdown bank withdrawal.

It is required to be performed if EITHER of the following conditions exist.

If neither of these conditions exist, this section may be marked NIA ( or omitted from the data package).

0-SI-OPS-085-01 1.0, 0-Sl-SXX-085-043.O, I,2-Pl-ICC-085-050.0, 1,2-Pl-ICC-085-051.0,or other testing that requires control rod movement in Modes 3, 4, or 5, will be performed, or Shutdown banks are to be withdrawn and an Estimated Critical Position (ECP) calculation has not been performed or verified within the last 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 0-SI-NUC-000-001.0.

4)

If a set of control banks is to be withdrawn for testing, then this section may be used to ensure that k-eff of the reactor will remain <0.99 following control bank withdrawal provided that the worth of the set of control banks is less then the total worth of the shutdown banks per the core nuclear design.

If shutdown banks are being used to take credit for the withdrawal of a set of control banks, then it must be documented on the SDM package.

[1]

RECORD on Data Sheet 3, the critical boron concentration with S/D banks withdrawn from Table 62 of Nuclear Design Report based on minimum expected core average temperature 1.0[1jD and core average burnup 1.0[1JG.

D

[2]

RECORD on Data Sheet 3, the correction to maximum boron concentration total (M-P) to account for differences between measured and predicted core conditions (see Attachment 3).

D

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 55 of 72 Appendix A (Page 12 of 13) 8M Corrected Subcritical Boron Concentration with SID Banks Withdrawn (continued)

[3]

CALCULATE on Data Sheet 3, the adjusted critical boron concentration with S!D banks withdrawn using the following equation:

Adjusted Critical OB with SID banks withdrawn = Step 8.0[1] +

Step 8.0[2]

0

[4]

RECORD on Data Sheet 3, the Inverse Boron Worth (IBW) from Tables 6I 1 and 6-12 of Nuclear Design Report based on core average burnup 1 0[1]G using minimum expected core average temperature tO[1]D and boron concentration from Step 8.0[3j 0

NOTE The Xenon worth is a negative value

[5]

RECORD on Data Sheet 3, the Xenon worth from Step 2.0[1].

0

[61 CALCULATE and RECORD on Data Sheet 3, the boron equivalent worth of Xenon using the following formula:

Boron Equivalent Worth of Xenon = Step 8.0[5]

Step 80[411 0

[7]

CONVERT 1000 pcm to ppm using the following formula AND RECORD on Data Sheet 3:

1000 pcm = 1000 pcm

Istep 8.0[4]I

[8]

RECORD on Data Sheet 3, the corrected subcritical boron concentration using the following equation:

Corrected Subcritical BC = Step 80[3] + Step 8.0[6] +

Step 8.0[7]

0

[9]

CHECK appropriate box on Data Sheet 3, to indicate whether the following criteria for shutdown bank withdrawal was satisfied.

0

r SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 56 of 72 Appendix A (Page 13 of 13) 8.0 Corrected Subcritical Boron Concentration with SID Banks Withdrawn (continued)

Acceptance Criteria:

The RCS boron concentration in 1.0 [1]H is greater than the corrected subcritical boron concentration recorded in Step 8.0[8].

[101 IF criteria in Step 8.0[9j was NOT satisfied, THEN NOTIFY the US that shutdown banks may NOT be withdrawn with the current boron concentration.

SQN Unit 0 Data Sheet I i-4 4AØ SHUTDOWN MARGIN CALCULATiON IN MODES 3,4, OR 5 Minimum Boron Concentration (C 8)

Date/time unit was shutdown Date 7o Time Current date/time Date Time Mode SDM calculation is performed for:

MODE S

,Øj Minimum expected core average temperature for surveillance interval Time since plant shutdown 7 0 hours0 days 0 hours 0 weeks 0 months Steady-state core power level before plant shutdown to 9 Core average burnup 00 0 MWD/MTU Current/projected RCS C8 and associated date/time (N/A date and time

for projected RCS boron concentration) 1 2 ppm Date Time krc Surveillance interval

/ 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Minimum C8 11 3 ppm Correction to Minimum C8 (M-P)

(2 ppm Adj. Mm. Boron Conc. = Step I.O[2] + Step 1.O[3]

Adjusted Minimum C8 =

) I

+

0

- Yf ppm

,$j Was the minimum boron concentration provided by the Reactor Engineering K

memorandum used in Step I.0[2]?

Yes D

No If yes, proceed to Section 6.0 and NIA (or discard) intermediate steps.

lIo( NLD SHUTDOWN MARGIN 0-SI-NUC-000-038.O Rev. 0059 Page 57 of 72

SQN SHUTDOWN MARGIN O-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 58 of 72 Data Sheet I (Page 2 of 4)

Minimum Boron Concentration (CB) (continued)

Are corrections to the adjusted minimum CB recorded in 1.0[4j desired?

Yes No D

If No, proceed to Section 6.0 and N/A (or discard) intermediate steps.

O Xe Worth TE Xenon worths shall be entered as negative values.

Xe Worth Correction for Reduced SDM Requirement in Mode 5 Mode 5 SDM correction (negative value if other than zero) pcm Correction for Immovable or Untrippable Rods No. immovable/untrippable rods

[2]

A.) Maximum double stuck rod worth

! fA pcm B.) Maximum stuck rod worth Stuck rod worth = Step 4.0[2]A - Step 4.0[2]B 0

pcm

I SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 L

Page 59 of 72 Data Sheet I (Page 3 of 4)

Correction for Immovable or Untrippable Rods (continued)

Correction for immovable/untrippable rods = Step 4.O[1] x Step 4.O[3J

=

x Minimum Boron Concentration Calculation Correction to Mm. Boron Conc. = 2.O[1] ÷ 3.O[1] ÷ 4.O[4J Correction to Mm. Boron Conc. =

C)

+

()

+

Inverse Boron Worth

0. 1 3

ppm/pcm Correction to Mm. Boron Cone. = Step 5.O[1]

Istep 5.O[2}I

fr(j3 I 2_

ppm Correction to Mm. SDM Boron Cone. = Step 1.O[4] + Step 5.O[3j

/2.31

+_____

Acceptance Criteria Verification Required Boron Concentration for Shutdown Margin ppm CHECK appropriate box and sign to indicate whether the following acceptance criteria was satisfied.

Acceptance Criteria:

The RCS boron concentration 1.O[1}H is greater than the required boron concentration of 6.0th.

Yes No C

Test Director Signature Datd

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 60 of 72 Data Sheet I (Page 4of4) 6.0 Acceptance Criteria Verification (continued)

[3]

IF acceptance criteria from Step 6.O[2] was NOT satisfied, THEN IMMEDIATELY NOTIFY the US that acceptance criteria has NOT been satisfied and borate per action requirement of LCO 3.1.1.1 or LCO 3.1.1.2.

Test Director Signature Date

[4]

Was boron concentration provided by Reactor Engineering memorandum used in Step 1.O[2]?

Yes D

No D

IF yes, proceed to Section 8.0 and NIA (or discard) Section 7.0.

Data Sheet I (Page 1 of4)

SHUTDOWN MARGIN CALCULATION IN MODES 3,4, OR 5 1.0 Minimum Boron Concentration (CB)

[1]

A.

Date/time unit was shutdown Date Time B.

Current date/time Date Time C.

Mode SDM calculation is performed for:

MODE_____________

D.

Minimum expected core average temperature for surveillance interval

°F E.

Time since plant shutdown F.

Steady-state core power level before plant shutdown G.

Core average burnup H.

Current/projected RCS C8 and associated date/time (N/A date and time for projected RCS boron concentration)

_____pp Date Time I.

Surveillance interval

[2]

Minimum C8 i320 ppm

[3]

Correction to Minimum CB (M-P)

I 0 ppm

[4]

Adj. Mm. Boron Conc. = Step I.O[2] + Step 1.O[3]

Adjusted Minimum CB =

137--p

+

_____pp

[5]

Was the minimum boron concentration provided by the Reactor Engineering memorandum used in Step I.O[2]?

Yes D

No D

If yes, proceed to Section 6.0 and NIA (or discard) intermediate steps.

SHUTDOWN MARGIN

SQN SHUTDOWN MARGIN 0-Sl-NUC-000-038.0 Unit 0 Rev. 0059 Page 58 of 72 Data Sheet 1 (Page 2of4) 1.0 Minimum Boron Concentration (CB) (continued)

[6]

Are corrections to the adjusted minimum CB recorded in 1.0[4] desired?

Yes D

No 0

If No, proceed to Section 6.0 and NIA (or discard) intermediate steps.

2.0 Xe Worth NOTE Xenon worths shall be entered as negative values.

[1]

XeWorth

_____p T3qp%d

-33 3.0 Correction for Reduced SDM Requirement in Mode 5

[1]

Mode 5 SDM correction (negative value if other than zero) pcm 4.0 Correction for Immovable or Untrippable Rods

[1]

No. immovable/untrippable rods

[2]

A.) Maximum double stuck rod worth B.) Maximum stuck rod worth

[3]

Stuck rod worth = Step 4.0[21A

- Step 4.O[2]B

_____p

______p

SQN SHUTDOWN MARGIN 0-SI-NUC-000-0380 Unit 0 Rev. 0059 Page 59 of 72 Data Sheet I (Page 3 of 4) 4.0 Correction for Immovable or Untrippable Rods (continued)

[4]

Correction for immovable/untrippable rods = Step 4.O[1] x Step 4.O[3]

=

x p

5.0 Minimum Boron Concentration Calculation

[1]

Correction to Mm. Boron Conc. = 2.O[1] + 3.O[1] + 4.O[4]

Correction to Mm. Boron Conc. =

14

+

p

[2]

Inverse Boron Worth

- 3 ?

pp p

M

+

[3]

Correctton to Mm. Boron Conc. = Step 5.O[1]

IStep 5.O[2]I o

1 i

ç ppm

[4]

Correction to Mm. SDM Boron Conc.

Step 1.O[4] + Step 5.O[3]

i3O +

pp

I 3 i-I3S 6.0 Acceptance Criteria Verification

[1]

Required Boron Concentration for Shutdown Margin 3S ppm

[2]

CHECK appropriate box and sign to indicate whether the following l3St4I36(o acceptance criteria was satisfied.

Acceptance Criteria:

The RCS boron concentration 1.O[1]H is greater than the required boron concentration of 6.O[1].

Yes D

No Test Director Signature Date

SQN SHUTDOWN MARGIN 0-SI-NUC-000-038.0 Unit 0 Rev. 0059 Page 60 of 72 Data Sheet I (Page 4 of 4) 60 Acceptance Criteria Verification (continued)

[3]

IF acceptance criteria from Step 6.O[2] was NOT satisfied, THEN IMMEDIATELY NOTIFY the US that acceptance criteria has NOT been satisfied and borate per action requirement of LCO 3.1.1.1 or LCO 3.1.1.2.

Test Director Signature Date

[4]

Was boron concentration provided by Reactor Engineering memorandum used in Step I.O[2]?

Yes D

No D

IF yes, proceed to Section 8.0 and N/A (or discard) Section 7.0.

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC SRO ADMIN A.3

1201 NRC SRO ADMIN A.3 Page 2 of 10 SRO JOB PERFORMANCE MEASURE Task:

Screen Workers for Emergency Exposure.

Task#:

3430290302 Task Standard:

Given five different workers during a life saving emergency, the Examinee will choose two (Roberts and Crabtree) of the five for the authorization to exceed the TVA Administrative Dose Levels, determines authorization from the Site Emergency Director is required and the maximum TEDE exposure limit for the conditions given is 25 Rem.

Time Critical Task:

YES:

NO:

X K/A Reference/Ratings:

2.3.13 (3.8)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

I Name I Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

14 mm Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC SRO ADMIN A.3 Page 3 of 10 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

Tools/Equipment/Procedures Needed:

2.

EPIP-15, Emergency Exposure Guidelines 3.

Calculator

References:

Reference Title Rev No.

1.

EPIP-15 EmergencyExposureGuidelines 9

Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM. I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

1201 NRC SRO ADMIN A.3 Page 4 of 10 INITIAL CONDITIONS:

1.

Unit 1 is in MODE 3 2.

An event is in progress involving due to failed fuel.

3.

The TSC is not operational.

4.

An RP Technician reports that an operator working with him in the Unit 1 690 Penetration Room has fallen and is severely injured. He has moved the injured person to an area that is somewhat shielded. Due to rapidly increasing dose rates, the RP Technician leaves to get help. He believes the injuries are life threatening. He also stated that the individual could be retrieved but it would take two people to do so.

5.

RP has been contacted and estimates it will take at least 30 minutes to retrieve the victim.

Auxiliary Building Area Radiation Monitors indicate extremely high radiation levels.

6.

Dose rates at the area needing access are greater than 9.8 Rem/Hr.

7.

There are five people that are available to perform the rescue operation.

Worker Name Age Sex Shearer 23 years old Male Roberts 26 years old Male Volunteer YTD TEDE Dose Yes 137 mrem Yes 132 mrem Yes 121 mrem Crabtree 57 years old Female Yes 125 mrem Collins 55 years old Male No 25 mrem 8.

None of the five people that are available to perform the rescue operation has received a planned special exposure.

INITIATING CUES:

1.

You are the Shift Manager in the Control Room.

2.

Perform the following:

a)

Calculate to two decimal places the total dose each worker would receive performing the rescue operation.

b)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, assess which two of the five workers are to be used to perform the rescue operation.

c)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, determine the required approval prior to the rescue operation.

d)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, determine the maximum TEDE exposure limit for the conditions given.

3.

Inform the evaluator when you are complete.

Ferry 27 years old Female not declared pregnant

1201 NRC SRO ADMIN A.3 Page 5 of 10 Start Time STEP 1

Calculate the dose.

SAT UNSAT Critical step St nd rd Examinee calculates a total dose of 4.9 Rem will be accumulated a

a performing the task.

Comment Examiner Note 9.8 Rem/hr ÷ 60 mm/i hr X 30 mm

= 4.9 Rem Obtain a copy of EPIP-15, Emergency Exposure Guidelines.

STEP 2

SAT UNSAT Standard:

Copy of is obtained EPIP-15, Emergency Exposure Guidelines.

Q Provide a copy of EPIP-15, Emergency Exposure Guidelines.

Comment

1201 NRC SRO ADMIN A.3 Page 6 of 10 4.0 REQUIREMENTS STEP 3

1 SAT 4.1 Guidance for Emergency Dose Limits UNSAT D.

Receipt of emergency dose limits shall be on a voluntary basis.

Critical Step Standard:

Ecamlnee determines Collins is excluded due to not volunteering,.

Comment

4.0 REQUIREMENTS STEP 4

i SAT 4.1 Guidance for Emergency Dose Limits E.

Other factors being equal, older volunteers should be selected first.

UNSAT Critical Step Standard:

Examinee determines Shearer is excluded duebeing, youngest worker Comment 4.0 REQUIREMENTS STEP 5

4.1 Guidance for Emergency Dose Limits SAT F.

Other factors being equal, selection of female volunteers capable UNSAT of reproduction should be avoided.

Critical Step St d

d Examinee determines Frry is excluded dues being capable of an ar reproduction..

Comment

1201 NRC SRO ADMIN A.3 Page 7 of 10 4.0 REQUIREMENTS STEP 6 :

SAT 4.1 Guidance for Emergency Dose Limits G.

During declared emergencies TVA Administrative Dose Levels UNSAT (ADLs) shall be amended as shown in Appendix D, Emergency Exposure Reference Guide, however, efforts shall be taken to hold Critical Step doses to the lowest practicable level that the emergency permits.

An individuals remaining allowable dose (RAD) is determined by subtracting the year-to-date dose from the applicable 10CFR2O exposure limit. Emergency responders that require additional dose in excess of 100FR2O limits may obtain consent through the completion of Appendix A.

APPENDIX A ACKNOWLEDGMENT AND AUTHORIZATION TO EXCEEL OCCUPATIONAL DOSE LIMITS READ THE FOLLOWING STATEMENT BEFORE SIGNING THIS FORM:

Authorized by:

Site Emergency Director

Date I Time I

Consult with the most senior Radiation Protection person prior to authorization The Examinee determines the Site Emergency Director approval is Standard:

raquirecL Examiner Both Crabtree and Roberts are expected to exceed the 1 OCFR limit of 5 Note:

rem/year during the rescue.

Examiner The Examinee may include Site Rad Protection Mgr, RSO, Plant Note:

Manager, and Site VP, however using the guidance of EPIP-15 these approvals are not required.

Comment

1201 NRC SRO ADMIN A.3 Page 8 of 10 UNSAT Critical Step Examiriee determihesthe maximumTEDEexoure limitfor the Standard:

conditions givenis 25 Bern.

Comment Terminating The task is complete when the Examinee returns the cue sheet to STOP Cue:

the examiner.

STEP 7

4.2 J

Lifesaving or Protection of Large Populations A.

Appendix 0, Emergency Exposure Reference Guide, may be used as a reference for exposure limits APPENDIX 0 EMERGENCY EXPOSURE REFERENCE GUIDE Emergency responders are vounteers who are qualihed radmtron workers at WA and have not decarcd pregnancy.

SAT 000 2

Continue accessing and modify exposure actions as necessary UPON DETERMINATION OF EXPOSURE LIMIT EPA PAGS are: I rem TEDE or REFER TO INSTRUCTIONS PRIOR TO 5 rem Thyroid CDE AUTHORIZING EXPOSURE Stop Time

JPM BRIEFING SHEET DIRECTIONS TO TRAINEE:

The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

INITIAL CONDITIONS:

1.

Unit 1 is in MODE3 2.

An event is in progress involving due to failed fuel.

3.

The TSC is not operational.

4.

An RP Technician reports that an operator working with him in the Unit 1 690 Penetration Room has fallen and is severely injured. He has moved the injured person to an area that is somewhat shielded. Due to rapidly increasing dose rates, the RP Technician leaves to get help. He believes the injuries are life threatening. He also stated that the individual could be retrieved but it would take two people to do so.

5.

RP has been contacted and estimates it will take at least 30 minutes to retrieve the victim.

Auxiliary Building Area Radiation Monitors indicate extremely high radiation levels.

6.

Dose rates at the area needing access are greater than 9.8 Rem/Hr.

7.

There are five people that are available to perform the rescue operation.

Worker Name Age Sex Volunteer YTD TEDE Dose Shearer 23 years old Male Yes 137 mrem Roberts 26 years old Male Yes 132 mrem Ferry 27 years old Female not declared pregnant Yes 121 mrem Crabtree 57 years old Female Yes 125 mrem Collins 55 years old Male No 25 mrem 8.

None of the five people that are available to perform the rescue operation has received a planned special exposure.

1201 NRC SRO ADMIN A.3 Page 10 of 10 INITIATING CUES:

1.

You are the Shift Manager in the Control Room.

2.

Perform the following:

a)

Calculate to two decimal places the total dose each worker would receive performing the rescue operation.

b)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, assess which two of the five workers are to be used to perform the rescue operation.

c)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, determine the required approval prior to the rescue operation.

d)

Using the guidance provided in EPIP-15 Emergency Exposure Guidelines, determine the maximum TEDE exposure limit for the conditions given.

3.

Inform the evaluator when you are complete.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNED TASK.

SEQUOYAH NUCLEAR PLANT JOB PERFORMANCE MEASURE 1201 NRC SRO ADMIN A.4

1201 NRC ADMIN SRO A.4 Page 2 of 10 SRO JOB PERFORMANCE MEASURE Task:

Classify the Event per the REP (Degraded Core With Possible Loss of Coolable Geometry and Likely Cntmt Failure)

Task#:

3440030302 Task Standard:

The Exam inee classifies the event as a GENERAL EMERGENCY based on LOSS of any two barriers and Potential LOSS of third barrier (1.1.1.L AND 1.2.2.L or 1.2.L P AND 1.3.1.P or 1.3.2.P) and the Examinee completes a TVA Initial Notification for General Emergency form with no errors on items noted with an.

Time Critical Task:

YES:

X NO:

K/A Reference/Ratings:

2.4.41. (2.9/4.6)

Method of Testing:

Simulated Performance:

Actual Performance:

X Evaluation Method:

Simulator In-Plant Classroom X

Main Control Room Mock-up Performer:

Trainee Name Evaluator:

/

Name! Signature DATE Performance Rating:

SAT:

UNSAT:

Validation Time:

19 minutes Total Time:

Performance Time:

Start Time:

Finish Time:

COMMENTS

1201 NRC ADMIN SRO A.4 Page 3 of 10 SPECIAL INSTRUCTIONS TO EVALUATOR:

1.

Critical steps are identified in step SAT/UNSAT column by bold print Critical Step.

2.

Any UNSAT requires comments.

Tools/EquipmentlProcedures Needed:

1.

EPIP-1, Emergency Plan Classification Matrix 2.

EPIP-5, General Emergency 3.

A clock must be available in classroom that all exam inees and evaluator can see

References:

Reference Title Rev No.

1.

EPIP-1 Emergency Plan Initiating Conditions Matrix 46 2.

EPIP-5 General Emergency 39

1201 NRC ADMIN SRO A.4 Page 4 of 10 Read to the examinee:

DIRECTIONS TO TRAINEE:

I will explain the initial conditions, and state the task to be performed. All control room steps shall be performed for this JPM.

I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. 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.

HAND JPM BRIEFING SHEET TO EXAMINEE AT THIS TIME!

1.

Unit 1 is in MODE 3 following a plant event 2.

Both Containment Spray pumps have failed.

3.

All other required equipment started and is operating as designed 4.

The TSC has not been manned.

5.

The following data was observed Parameter Indication Source Range Counts cps and decreasing RCS Pressure 1000 psig Pressurizer Level Off scale low Core exit thermocouples All 710° F and increasing slowly RCS Subcooling 10°F superheat RVLIS Lower range 38% and decreasing Highest Steam Generator 22%

Level RCS Cold Leg Temp 245° Fand decreasing slowly Containment Pressure 14 psig and decreasing slowly Wind speed at 46 meters 5 mph Wind direction at 46 meters 235 degrees INITIATING CUES:

1.

Using the data provided and the applicable procedure (s) classify the event.

2.

Raise your hand when classification is complete.

3.

When classification is complete, request from the examiner the correct TVA Initial Classification form, then complete the required form.

4.

Determine what, if any, Protective Action Recommendations are required.

5.

Raise your hand when you have completed the notification form.

6.

The is (are) an element (s) of this task that is (are) time critical.

1201 NRC ADMIN SRO A.4 Page 5 of 10 Start Time Obtain a copy of EPIP-1, EMERGENCY PLAN CLASSIFICATION STEP 1

MATRIX.

SAT UNSAT Stand rd Examinee obtains a copy of EPIP-1, EMERGENCY PLAN a

CLASSIFICATION MATRIX.

Cue Provide a copy of EPIP-1, EMERGENCY PLAN CLASSIFICATION

MATRIX.

Comment Examiner Annotate start time when the examinee acknowledges the task is Note understood. Start time_____________

INSTRUCTIONS Note:

A condition is considered to be MET if, in the judgment of the SED, the Procedure condition will be MET IMMINENTLY (i.e.:

with two hours). The classification shall be made as soon as this determination is made.

STEP 2

1.

In the matrix to the left, REVIEW the SAT initiating conditions in all three barrier columns and circle the conditions that UNSAT are MET.

2.

In each of the three barrier columns, IDENTIFY If any Loss or Potential Loss INITIATING CONDITIONS have been MET.

Examinee reviews the EMERGENCY PLAN CLASSIFICATION MATRIX Standard.

and identifies initiating conditions provided in the initial conditions.

Comment

1201 NRC ADMIN SRO A.4 Page 6 of 10 3.

COMPARE the number of barrier STEP 3

Losses and Potential losses to the SAT criteria below and make the appropriate declaration.

UNSAT Emergency Class Criteria Critical Step General Emeraencv LOSS of any two barriers and Potential LOSS of third barrier The examihee compares barrier losses and classifies the event as a General Emergency within IS minutes of starting thetasL LOSS of the Fuel Clad Barrier ttl.Ldueto Core Coling Red (FR-C.I entry ANJ Standard:

LOSS Of the ROS Barrier 1.22,L (ROS teak results in subcoolin<4Q *F)

OR LOSS of the RCS Barrier 124 L (VAlJD RVUS level<42% or U-Ge-368 or Li-68371 with no RCP wnnlng).

AND PotentiaLOSS of the Containmentt3.t.P (Containment Red Path FR-ZI) OR Potential LOSS of the Containment I.322 (Pressure>Z& PSIG (Phase B with one fUlltrain of containment spray Comment EXAMINER This is a critical step because of the requirement to arrive at the NOTE:

correct classification within 15 minutes.

EXAMINER NOTE:

Annotate the stop time for the event classification here.

EXAMINER Examinee continues with the task to complete the State Notification NOTE:

Form using EPIP-5, GENERAL EMERGENCY. The critical time element continues.

EXAMINER NOTE:

Annotate the start time for the State Notification here.

EXAMINER NOTE:

The start data is provided to the examinee on the JPM briefing sheet.

EXAMINER NOTE:

Examinee transitions to EPIP-5, GENERAL EMERGENCY

1201 NRC ADMIN SRO A.4 Page 7 of 10 Obtain a copy of EPIP-5, GENERAL EMERGENCY STEP 4

SAT UNSAT Standard:

Examinee obtains a copy of EPIP-5, GENERAL EMERGENCY.

Cue Provide a copy of EPIP-5, GENERAL EMERGENCY Comment NOTE:

IF there are personnel injuries, THEN IMPLEMENT EPIP-1O, Medical Emergency Response.

Procedure Note rNoTE; IF there are immediate hazards to plant personnel, THEN consider immediately implementing EPIP-8 Personnel Accountability and Evacuation in parallel with this procedure 3.1 GENERAL EMERGENCY DECLARATION BY THE MAIN CONTROL ROOM STEP 5

SAT Upon classifying events as a GENERAL EMERGENCY, the SM/SED shall:

UNSAT

[1]

IF TSC is OPERATIONAL. (SED transferred to TSC), THEN GO TO Section 3.2.

[2]

RECORD time of Declaration Time

[3]

ACTIVATE Emergency Paging System (EPS) as follows.

[a]

IF EPS has already been activated, THEN GO TO Step 4.

[b]

IF ongoing onsite Security events may present risk to the emergency responders, THEN CONSULT with Security to determine if site access is dangerous to the life and health of emergency responders.

[C]

IF ongoing events makes site access dangerous to the life and health of emergency responders, THEN SELECT STAGING AREA button on the EPS terminal INSTEAD of the EMERGENCY button.

[d]

ACTIVATE EPS using touch screen terminal. IF EPS fails to activate, THEN continue with step 4.

El Standard:

The examinee addresses steps 3.1.1 through 3.1.3 Comment

1201 NRC ADMIN SRO A.4 Page 8 of 10

[4]

EVALUATE Protective Action Recommendations (PARS) using Appendix B.

U STEP 6

SAT UNSAT Examinee determines Protective Action Recqmmendations (PARs)

Critical Step Standard:

RecommendatIon 2 using Appendix B.

Appendix B PROTECTIVE ACTION RECOMMENDATIONS GENERAL EMERGENCY DECLARED Zerea

</

ort term controlled N release whereas near N

NO At Nccuated before

.7 4

plume arrival?

7/

Lond (see Note 2)/V plant areas cannot be Dose greater,,>

Meades, is the N Projected OR N.

NO than Table 1 C?

RECOMMENDATION 3 TABLE I Protective Action Guides (PAG)

TYPE LIMIT Measured 3.9 E-6 micro Ci/cc of Iodine 131 or 1 REM per hour External Dose Projected 1 REM TEDE or 5 REM Thyroid CDE Note: Unknown conditions are assumed less than listed conditions.

Note 1:

If conditions are unknown utilizing the flowchart, then answer is NO.

Note 2: A short term release is defined as a release that does not exceed a 15 minute duration.

CONTINUE ASSESSMENT Modify protective actions based on available plant and field monitoring information.

Locate and evaluate localized hot spots.

Examiner Note:

N RECOMMENDATION I EVACUATE 2 miles radius and 10 miles downwind AND SHELTER remainder of 10 mile EPZ I

RECOMMENDATION 2 EVACUATE 2 miles radius and 5 miles downwind AND SHELTER remainder of 10 mile EPZ Comment

1201 NRC ADMIN SRO A.4 Page 9 of 10 3.1 GENERAL EMERGENCY DECLARATION BY THE MAIN CONTROL ROOM STEP 7

SAT 15]

COMPLETE Appendix C (TVA Initial NotificaUon for General Emergency).

L

UNSAT The examinee completes an AppendbcB TVA Initial Notification for Critical Step Standard:

General Emergency With no errors on items noted with an k on the answer keywithin 14 minutes of event declaration.

EXAMINER This is a critical step because of the requirement to provide notification NOTE:

of an event to the state within 15 minutes.

Examiner Note:

Annotate the stop time for the initiation of state notification._______

Comment Terminating The task is complete when the Examinee completes the WA STOP Cue:

Initial Notification for General Emergency.

Stop Time

DIRECTIONS TO TRAINEE:

JPM BRIEFING SHEET The examiner will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

1.

Unit 1 is in MODE 3 following a plant event 2.

Both Containment Spray pumps have failed.

3.

All other required equipment started and is operating as designed 4.

The TSC has not been manned.

5.

The following data was observed Parameter Indication Source Range Counts cps and decreasing RCS Pressure 1000 psig Pressurizer Level Off scale low Core exit thermocouples All 710° F and increasing slowly RCS Subcooling 10°Fsuperheat RVLIS Lower range 38% and decreasing Highest Steam Generator 22%

Level RCS Cold Leg Temp 245° F and decreasing slowly Containment Pressure 14 psig and decreasing slowly Wind speed at 46 meters 5 mph Wind direction at 46 meters 235 degrees INITIATING CUES:

1.

Using the data provided and the applicable procedure (s) classify the event.

2.

Raise your hand when classification is complete.

3.

When classification is complete, request from the examiner the correct TVA Initial Classification form, then complete the required form.

4.

Determine what, if any, Protective Action Recommendations are required.

5.

Raise your hand when you have completed the notification form.

6.

The is an (are) element (s) of this task that is (are) time critical.

Acknowledge to the examiner when you are ready to begin.

HAND THIS PAPER BACK TO YOUR EVALUATOR WHEN YOU HAVE SATISFACTORILY COMPLETED THE ASSIGNEb TASK.

iiou iVC A.L(

2. This Sequoyah has declared a GENERAL EMERGENCY affect4t Unit I EJ Unit 2 3

EAL De atojjij A I

4. Brief Description of the Event:

Minor releases within federally approved limits*I Minor releases within federally approved limits 1

Releases above federally approved limits 1 b a! AEI Releases above federally approved limits 1

/

X Release information not known

.J Release information not known c

7 The Meteorological Conditions are:

(Use 46 meter data from the Met Tower)

WdSpeec

8. Provide Protective Action Recommendation:

(Check eithd[1 or 2 or 3.)

5 m.p.h Recommendation I Recommendation 2 4 I EVACUATE LISTED SECTORS (2 mile Radius WIND FROM and 10 miles downwind)

DEGREES Radius and 5 mile downwind)

SHELTER remainder of 10 mile EPZ DIRECTION SHELTER remainder of 10 mile EPZ CONSIDER issuance of POTASSIUM IODIDE in CONSIDER issuance of POTASSIUM accordance with the State Plan.

a IODIDE in accordance with the State Plan.

A-1,B-i,C-1,D-1, C-2,-6,-7,-8,D-2,-3,-5,-6 12-49 A-1,B-1,C-1,D-i, C-2,D-2 A-i,B-i,C-1,D-1, D-2,-3,-4,-5,-6 50-70 A-i,B-1,C-i,D-i, D-2 A-1,B-i,C-1,D-i, A-3,-4, D-2,-3,-4,-5 71-112 A-lB-iC-iD-i, A-3,D-2 A-i,B-i,C-1,D-i, A-2,-3,-4,-5,-6, D-4 113-146

A-1,B-1,C-1D-1, A-2,A-3, A-i, B-i, C-i, D-i, A-2, -3, -4, -5,-6, B-2

147-173 A-JiC-1,D-i,A-2.A3,B-2 A-i,B-1,C-i,D-i, A-2,-5,-6, 8-2,-3,-4 174-21f A-i,B-i,C-1,D-i, A2B-2 A-i,B-1,C-1,D-1, B-2,-3,-4,-5,-6,-7,-8

215-25 A-1,B-i,C-i,D-i, B-2,B-5, A-i,B-i,C-1,D-i, B-2,-3,-5,-6,-7,-8, C-2,-3,-4-5,-6

259-33k A-i B-i,C-i,D-i, B-5, C-2-3-4,-5-6,-7-8 332-11

=

A-1,B-i,C-1,D-i, B-5,C-2 El Recommendation 3 SHELTER all sectors.

CONSIDER issuance of Potassium Iodide in accordance with the State Plan.

9. Please repeat back the information you have received to ensure accuracy.

El

10. When completed, FAX this information to the ODS or TEMA as required by Sections 3.1 or 3.2. El PAGE 15 of 16 REVISION 39

[

SEQUOYAH GENERAL EMERGENCY EPIP5 4

Appendix C WA INITIAL NOTIFICATION OF GENERAL 1

El This is an Actual Event

- Repeat - This IS an Actual Event

6. Event Declared:

Time: ti4r 4

(osTI*

Date: To FORWARD COMPLETED PROCEDURE TO EMERGENCY PREPAREDNESS MANAGER

SEQUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP1 I

INSTRUCTIONS Containment Red (FR-Z.1)

Actions of FR-C.1 (Red Path) are INEFFECTIVE (i.e.: core TCs trending up)

-OR ment Pressure I F Potential LOSS I-<apo unexplained pressure Containment Hydrogen decrease following initial increases to >4% by volume on increase on PDI-30-44 or 45 H21-43-200 or 210 Containment pressure or sump Pressure >2.8 PSIG (Phase B) level not increasing on Ll with < one full train of 178 and 179 with a LOCA in containment spray progress

-OR-

3. Containment Isolation Status

I*

Póteritia[ LOSS Containment isolation, when Not Applicable required is incomplete and a release path to the environment exists.

-OR-

4. Containment Bypass Potential LOSS RUPTURED SIG that is also Unexpected VALID increase in faulted outside containment (E2 area or ventilation RAD and E3) monitors adjacent to OR containment (with LOCA in

>4 hour secondary side release progress).

outside containment from a SIG with a S/C tube leak >TIS limits (AOP R.01 App A)

-OR-

5. Significant Radiation in Containment

(.*

Potential LOSS Not Applicable VALID reading of greater than:

5.8E÷03 RIhr on RM-90-271A and RM-90-272A OR 3.4E÷03 RIhr on RM-90-273A and 4.9E+03 RJhr on RM-90-274A (see instruction note 4)

-OR-

6. SED Judgment Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the Containment Barrier comparable to the conditions listed above.

1.

In the matrix to the left, REVIEW the initiating conditions in all three barrier columns and circle the conditions that are MET.

2.

In each of the three barrier columns, IDENTIFY if any Loss or Potential Loss INITIATING CONDITIONS have been MET.

3.

COMPARE the number of barrier Losses and Potential losses to the criteria below and make the appropriate declaration.

4.

Containment Radiation Monitors are temperature sensitive and can be affected by temperature-induced currents. These monitors should be used for trending only until containment temperature has been stable for approximately 5 minutes after a Steam Line Break or LOCA.

Note:

MONITOR the respective status tree criteria if a CSF is listed as an INITIATING CONDITION.

1. Critical Safety Function Status Not Applicable Note:

A condition is considered to be MET if, in the judgment of the SED, the condition will be MET IMMINENTLY (i.e.:

with two hours). The classification shall be made as soon as this determination is made.

LOSS Emeraencv Class Criteria General Emergency LOSS of any two barriers Potential LOSS of third barrier Site Area Emergency LOSS or Potential LOSS of any two barriers Alert Any LOSS or Potential LOSS of Fuel Clad barrier OR Any LOSS or Potential LOSS of RCS barrier Unusual Event LOSS or Potential LOSS of Containment barrier Page 10 of 47 Revision 46

[QUOYAH EMERGENCY PLAN CLASSIFICATION MATRIX EPIP-1 11 Fuel Clad Barrier

1. Critical Safety Function Status Potential LOSS Core Cooling Red[

(FR-C.1 Heat Sink RED (FR-H.1) and RHR Shutdown Cooling not in service

- OR -

2. Primary Coolant Activity Level I.i*

Potential LOSS RCS sample activity is Not Applicable greater than 300 pCi/gm dose equivalent 1131

-OR-

3. Incore Thermocouple HI Quad Average s1*

Potential LOSS Greater than 1200 °F on Greater than or equal to Xl-94-101 or 102 700 °F on Xl-94-101 or (EXOSENSOR) 102 (EXOSENSOR)

-OR-

4. Reactor Vessel Water Level Potential LOSS Not Applicable VALID RVLIS level

<42% on Ll-68-368 or Ll-68-371 with no RCP running

-OR-

5. Containment Radiation Monitor

.I*

Potential LOSS VALID reading of Not Applicable greater than:

2.5E÷02 RJhr on RM-90-271A and -272A OR I.5E+02 R/hr on RM-90-273A and 2.IE+02 Rlhron RM-90-274A (see instruction note 4)

-OR-

6. SED Judgment Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the Fuel Clad Barrier comparable to the conditions listed above.

PoentLa[LOSS Pressurized Thermal Shock Red (FR-P.1)

OR Heat Sink RED (FR-Hi) and RHR Shutdown Cooling not in service

-OR

2. RCS Leakae I LOCA eM Potential LOSS RCS leak results in !

Non Isolatable RCS leak subcooling <40 °F as f exceeding the capacity indicated on Xl-94-101 of one charging pump in or 102 (EXOSENSOR)I the normal charging 4

alignment OR RCS leakage results in entry into E-1

-OR-

3. Steam Generator Tube Rupture

I*

PontIaU*

SGTR that results in a Not Applicable Safety Injection actuation OR Entry into E-3

-OR-4.

e1)t1aL LØ.

VALID RVLISir Not Applicable

<42% on Ll-68-368 or Ll-68-371 with no RCP jinjing 1-OR-

5. SED Judgment Any condition that, in the judgment of the SM or SED, indicates loss or potential loss of the RCS Barrier comparable to the conditions listed above.

Core Cooling Orange (FR-C.2)

OR

1. Critical Safety Function Status Not Applicable Page 9 of 47 Revision 46

SEQUOYAH GENERAL EMERGENCY EPIP-5 Appendix B PROTECTIVE ACTION RECOMMENDATIONS Note 1:

If conditions are unknown utilizing the flowchart, then answer is NO.

Note 2: A short term release is defined as a release that does not exceed a 15 minute duration.

/

Istherea

/ short term controll

/

release whereas near

(

plant areas cannot be evacuated before

/

N plume arrival? /

(see Note 2)

YES 1

TABLE I Protective Action Guides (PAG)

TYPE LIMIT Measured 3.9 E-6 micro Cl/cc of Iodine 131 or 1 REM per hour External Dose Projected 1 REM TEDE or 5 REM Thyroid CDE PAGE 14 of 16 Note: Unknown conditions are assumed less than listed conditions.

REVISION 39 GENERAL EMERGENCY Modify protective actions based on available plant and field monitoring information.

Locate and evaluate localized hot spots.

RECOMMENDATION 3 SHELTER 10 mile EPZ YES F-RECOMMENDATION I I

EVACUATE 2 miles radius and 10 miles downwind AND SHELTER remainder of 10 mile EPZ EVACUATE 2 miles radius and 5 miles downwind AND SHELTER remainder of 10 mile EPZ FORWARD COMPLETED PROCEDURE TO EMERGENCY PREPAREDNESS MANAGER

ML120720356

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