Text

10 Draft Ops Test
	ML12297A450
Person / Time
Site:	Cooper
Issue date:	10/30/2012
From:	Sean Hedger; Operations Branch IV
To:	; Nebraska Public Power District (NPPD)
	Larua Hurley
References
	Download: ML12297A450 (264)
	v • d • e

Nebraska Public Power District SKL034-50-04(6804)

Cooper Nuclear Station Page 1 of 4 Job Performance Measure for Operations Revision 01 Task No.: 205036G0403

Title:

PERFORM A TIME TO BOIL DETERMINATION Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 10 minutes

5. NRC K/A 2.1.23 (4.3/4.4); 2.1.25 (3.9/4.2)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to perform a Time To Boil Determination.

2. Observe the trainee during performance of the JPM for proper use of self-checking methods.

3. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

4. Give the trainee his copy of the Directions to the Trainee (Attachment 1) when ready to start the JPM.

5. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Directions to Trainee:

When I tell you to begin, you are to perform a Time To Boil Determination. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

Nebraska Public Power District SKL034-50-04(6804)

Cooper Nuclear Station Page 2 of 4 Job Performance Measure for Operations Revision 01 Task No.: 205036G0403

Title:

PERFORM A TIME TO BOIL DETERMINATION General Conditions:

1. The plant is shutdown for refueling.

2. The reactor has been shutdown for 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months .

3. The steam separator is being removed and flood up of the reactor cavity has begun.

4. RHR HX inlet temperature indicates 110°F as read on RHR-TI-131.

5. RPV water level just reaches the reactor vessel flange when RHR isolates on high pressure (instrument failure).

6. The isolation signal cannot be reset.

7. No other method of decay heat removal can be established.

General

References:

1. Abnormal Procedure 2.4SDC, SHUTDOWN COOLING ABNORMAL.

General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standards:

1. The operator correctly determines how many hours until boiling will occur in the reactor cavity.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

You are to perform a Time To Boil Determination. Inform the SM of the amount of time until reactor coolant temperature reaches 212°F.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-50-04(6804)

Cooper Nuclear Station Page 3 of 4 Job Performance Measure for Operations Revision 01 Task No.: 205036G0403

Title:

PERFORM A TIME TO BOIL DETERMINATION Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Refer to 2.4SDC. Referred to 2.4SDC, Attachment 5, Figure 2.

Applied hours after shutdown to diagram on

2. Determine correct graph. Figure 2.

3. Applies 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months after shutdown and Plotted intersection of 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months and the 110°F 110°F RHR HX curve.

temperature.

Related the 45 hour5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months - 110°F intersect to the hours to boiling axis and determined

4. Determines hours to approximately 2.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

boiling.

NOTE: (2.3-2.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months is acceptable).

Informed SM the time to boil is approximately 2.6

5. Informs SM. hours.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-04 (6804)

Cooper Nuclear Station Page 4 of 4 Job Performance Measure for Operations Revision 01 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin, you are to perform a Time To Boil Determination. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

General Conditions:

1. The plant is shutdown for refueling.

2. The reactor has been shutdown for 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months .

3. The steam separator is being removed and flood up of the reactor cavity has begun.

4. RHR HX inlet temperature indicates 110°F as read on RHR-TI-131.

5. RPV water level just reaches the reactor vessel flange when RHR isolates on high pressure (instrument failure).

6. The isolation signal cannot be reset.

7. No other method of decay heat removal can be established.

Initiating Cues:

You are to perform a Time To Boil Determination. Inform the SM of the amount of time until reactor coolant temperature reaches 212°F.

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Perform DW Unidentified Leak Rate Checks Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A 2.1.18 (3.6)

Directions to Examiner:

1. This JPM determines if the RO can correctly determine DW leakage rates given previous data and the current Totalizer reading.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Perform DW Unidentified Leak Rate Checks Directions to Trainee:

When I tell you to begin, you are to determine the current DW unidentified leakage rate and the change in leak rate. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The DW flood drain sump has just stopped after manual start.

General

References:

1. Procedure 6.LOG.60, DAILY SURVEILLANCE LOG - MODES 1, 2, AND 3 General Tools and Equipment:

1. Calculator Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

2. Simulator cues denoted by "#".

Task Standards:

1. Correctly calculate unidentified leak rate and change in leak rate.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

You are the on watch Reactor Operator. You have just manually pumped the floor drain sump pump. It is currently 0800 and you are to perform the unidentified leak rate checks. The Totalizer is now indicating 855600. Turn in your unidentified leak rate checks when completed.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Perform DW Unidentified Leak Rate Checks Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Form completed. Trainee completed 0800 reading column.

2. Calculates leak Leak rate determined and matched Answer rate. Key +/- 0.01

3. Calculate leak rate Change in leak rate determined and matched change. Answer Key +/- 0.01

4. Completed Attachment 2 Completed Attachment compared to Answer Key handed in.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Perform DW Unidentified Leak Rate Checks ATTACHMENT 1 ANSWER KEY Time Last Pumped (Previous Day): 1600 (F-2) 0000 0800 1600 OPERABILITY APPLICABLE ATT. 22 RW-FQ-527 (a) READING READING READING LIMIT MODES NOTES Pump Used N/A F-1 (F-1/F-2)

Present Grand 855210 855600 Total

(-) Previous Grand N/A N/A N/A 855210 855210 Total

(=) Total Gallons 0 390

(÷) Time Interval 480 960 (minutes)

(=) Leak Rate (gpm) (b) 0 0.41 5 gpm 1, 2, 3 51

(-) Previous Day 0.1 0.092 0.094 N/A N/A N/A Leak Rate (gpm)

(=) Change In Leak 0.092 0.31 +2 gpm 1 51 Rate (gpm)

(a)

If Sump F Totalizer has failed, determine Total Gallons per failed Sump F Totalizer table and ensure the Conditions and Required Actions of LCO 3.4.5 are entered.

(b)

If leak rate > 0.25 gpm, refer to Procedure 0-CNS-OP-109, Drywell Leakage Investigation.

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to determine the current DW unidentified leakage rate and the change in leak rate. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The DW flood drain sump has just stopped after manual start.

Initiating Cues:

You are the on watch Reactor Operator. You have just manually pumped the floor drain sump pump. It is currently 0800 and you are to perform the unidentified leak rate checks. The Totalizer is now indicating 855600. Turn in your unidentified leak rate checks when completed.

Nebraska Public Power District SKL034-50-74(49062)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Time Last Pumped (Previous Day): 1600 (F-2) 0000 0800 1600 OPERABILITY APPLICABLE ATT. 22 RW-FQ-527 (a) READING READING READING LIMIT MODES NOTES Pump Used N/A (F-1/F-2)

Present Grand 855210 Total

(-) Previous Grand 855210 N/A N/A N/A Total

(=) Total Gallons 0

(÷) Time Interval 480 (minutes)

(=) Leak Rate (gpm) (b) 0 5 gpm 1, 2, 3 51

(-) Previous Day 0.092 0.1 0.094 N/A N/A N/A Leak Rate (gpm)

(=) Change In Leak 0.092 +2 gpm 1 51 Rate (gpm)

(a)

If Sump F Totalizer has failed, determine Total Gallons per failed Sump F Totalizer table and ensure the Conditions and Required Actions of LCO 3.4.5 are entered.

(b)

If leak rate > 0.25 gpm, refer to Procedure 0-CNS-OP-109, Drywell Leakage Investigation.

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 1 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 15 minutes

5. NRC K/As 2.1.24 2.8/3.1 Directions to Examiner:

1. This JPM evaluates the trainees ability to determine appropriate mechanical isolation boundaries.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 2 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries Directions to Trainee:

When I tell you to begin, you are to use station documents to determine the minimum mechanical/piping boundaries required to isolate the RHR Pump A for removal. RHR Motor A has already been tagged and removed. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. N/A General

References:

1. Procedure 0.49 General Tools and Equipment:

1. Set of Mechanical Prints

2. Copy of 2040 Sheet 1

3. Copy of 2031 Sheet 2 Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

Task Standards:

1. Correctly identified all the valves that require controlling for isolating RHR Pump A.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The Control Room Supervisor directs you to determine the minimum mechanical/piping isolation boundaries required to isolate the RHR 1A pump for removal. RHR 1A Motor has already been tagged and removed. Find and highlight all valves required to be positioned to support this task.

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 3 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries NOTE: Tell the trainee to begin.

Start Time: ____________

Performance Checklist Standards Sat Unsat

1. May reference Referenced Procedure 0.49 procedure 0.49 Note to examiner: Steps 2 through 12 can be completed in any order.

Found B&R 2040 Sheet 1

2. References BR CUE: When the candidate successfully 2040 Sheet 1 locates 2040 Sheet 1, provide a copy for the candidate to mark up.

3. Highlights RHR- RHR-MO-15A was highlighted.

MO-15A

4. Highlights RHR- RHR-MO-13A was highlighted.

MO-13A

5. Highlights RHR-98 RHR-98 was highlighted.

6. Highlights RHR-58 RHR-58 was highlighted.

7. Highlights RHR-11 RHR-11 was highlighted.

8. Highlights RHR RHR-35, RHR 36 and RHR 34 were highlighted.

drain path Found B&R 2031 Sheet 2.

9. References BR 2031 Sheet 2 to CUE: When the candidate successfully find REC Isolation locates 2031 Sheet 2 Provide a copy for valves the candidate to mark up.

10. Highlights REC-75 REC-75 was highlighted.

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 4 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries Performance Checklist Standards Sat Unsat

11. Highlights REC-76 REC-76 was highlighted.

12. Highlights REC REC-87, 397 and 83 were highlighted Drain Path.

Gave highlighted drawings to the examiner.

13. Informs the CRS that the boundaries

CUE: Acknowledges receipt of the highlighted are identified.

prints. This JPM is complete.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 5 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries ATTACHMENT 1 ANSWER KEY B&R 2040 Sheet 1

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 6 of 7 Job Performance Measure for Operations Revision 01 Task No.: 299012O0301 Determine Isolation Boundaries ATTACHMENT 1 ANSWER KEY B&R 2031 Sheet 2

Nebraska Public Power District SKL034-50-57 (36402)

Cooper Nuclear Station Page 7 of 7 Job Performance Measure for Operations Revision 01 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to use station documents to determine the minimum mechanical/piping boundaries required to isolate the RHR 1A pump for removal. RHR 1A Motor has already been tagged and removed. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. The plant is shutdown in a refueling outage.

2. RHR Loop A is not required to be operable.

Initiating Cue(s):

The Control Room Supervisor directs you to determine the minimum mechanical/piping isolation boundaries required to isolate the RHR 1A pump for removal. RHR 1A Motor has already been tagged and removed. Find and highlight all valves required to be positioned to support this task.

Nebraska Public Power District SKL034-50-75(49063)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 0 Task No.: 200148P0501

Title:

Calculate Liquid Release Curie Content Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 10 minutes

5. NRC K/A 2.3.11 (3.8/4.3)

Directions to Examiner:

1. This JPM addresses the trainees ability to compute the curie content of a liquid release.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-75(49063)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 0 Task No.: 200148P0501

Title:

Calculate Liquid Release Curie Content Directions to Trainee:

When I tell you to begin, you are calculate the Curie content of a release from the South Condensate Storage Tank per Procedure 5.7.16, Release Rate Determination. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The South Condensate Storage Tank has developed a leak in the side of the tank.

2. The plant was operating normally prior to the leak developing.

3. Tank level was at 24 feet prior to the leak.

4. Chemistry personnel are not available to sample the tank.

General

References:

1. 5.7.16, Release Rate Determination General Tools and Equipment:

1. Calculator Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

2. Simulator cues denoted by "#".

Task Standards:

1. The trainee will correctly calculate the total Curies to be released from the South Condensate Storage Tank.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The CRS has been informed the leak on the side of the tank will result in the tank level lowering to 13 feet. The CRS has directed you to determine the total Curies that will be released using Procedure 5.7.16.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-50-75(49063)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 0 Task No.: 200148P0501

Title:

Calculate Liquid Release Curie Content Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Obtain Procedure Obtained Procedure 5.7.16.

2. Calculated release Calculated release rate and recorded results in rate Step 12 of Attachment 2.

3. .Hand in work Signed and handed in completed Att.2 NOTE to Examiner: Acceptable Curies released range is 4.24 to 5.4 Ci.

4. Release rate The calculated total Curies released was agrees with 4.91 Ci +/-0.49 Ci.

Answer Key Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-75(49063)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 0 Task No.: 200148P0501

Title:

Calculate Liquid Release Curie Content ATTACHMENT 1 ANSWER KEY (3) (4) (5)

(1) GROSS BEGINNING ENDING (6)

GROSS (2) ACTIVITY TANK VOLUME* TANK VOLUME* VOLUME ACTIVITY CONVERSION OF LIQUID (%) (%) RELEASED*

OF LIQUID FACTOR (µCi/gal) (if not known, (if not known, (%)

(µCi/ml) (ml to gal) (1) x (2) use 100%) use 0%) (4) - (5) 1.0E-2 3785 37.85 N/A N/A N/A (9) (10)

(7) (8) GALLONS TOTAL µCi (11) (12)

CONVERSION TANK CAPACITY RELEASED RELEASED CONVERSION TOTAL CURIES FACTOR* (gal)* (gal) (µCi) FACTOR RELEASED (Ci)

(% to decimal) (see Table 1) (6) x (7) x (8) (3) x (9) (µCi to Ci) (10) x (11) 0.01 N/A 129,800 4,912,930 1E-6 4.91 Completed By: Date:

TABLE 1 - CAPACITY OF TANKS (GALLONS)

CONTAINING POTENTIALLY CONTAMINATED LIQUIDS South Condensate Storage Tank** 450,000 Condensate Backwash Tank 12,500 North Condensate Storage Tank*** 700,000 Condensate Phase Separators (each) 12,500 Floor Drain Sample Tank 20,000 Waste Sludge Tank 17,000 Waste Sample Tank (each) 22,000 RWCU Phase Separators (each) 4,500 Waste Surge Tank 65,000 Spent Resin Tank 2,000 Waste Collector Tank 22,000 Torus/Suppression Pool 700,000 Floor Drain Collector Tank 20,000 Lab Drain Tank (each) 500 Chemical Waste Tank 5,200

For conversion purposes the SCST and NCST display in feet, perform Steps 1 through 3 above, then substitute the following calculation for Steps 4 through 8, with the result being entered into Step 9.

[Beginning Tank Level in Feet - Ending Tank Level in Feet] (Conversion Factor) = Gallons Released (Step 9)

- SCST = 11,800 gal/ft

- - NCST = 19,090 gal/ft

Nebraska Public Power District SKL034-50-75 (49063)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are calculate the Curie content of a release from the South Condensate Storage Tank per Procedure 5.7.16, Release Rate Determination. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The South Condensate Storage Tank has developed a leak in the side of the tank.

2. The plant was operating normally prior to the leak developing.

3. Tank level was at 24 feet prior to the leak.\

4. Chemistry personnel are not available to sample the tank.

Initiating Cue(s):

The CRS has been informed the leak on the side of the tank will result in the tank level lowering to 13 feet. The CRS has directed you to determine the total Curies that will be released using Procedure 5.7.16.

Nebraska Public Power District SKL034-50-75 (49063)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 (3) (4) (5)

(1) GROSS BEGINNING ENDING (6)

GROSS (2) ACTIVITY TANK VOLUME* TANK VOLUME* VOLUME ACTIVITY CONVERSION OF LIQUID (%) (%) RELEASED*

OF LIQUID FACTOR (µCi/gal) (if not known, (if not known, (%)

(µCi/ml) (ml to gal) (1) x (2) use 100%) use 0%) (4) - (5) 3785 (9) (10)

(7) (8) GALLONS TOTAL µCi (11) (12)

CONVERSION TANK CAPACITY RELEASED RELEASED CONVERSION TOTAL CURIES FACTOR* (gal)* (gal) (µCi) FACTOR RELEASED (Ci)

(% to decimal) (see Table 1) (6) x (7) x (8) (3) x (9) (µCi to Ci) (10) x (11) 0.01 1E-6 Completed By: Date:

TABLE 1 - CAPACITY OF TANKS (GALLONS)

CONTAINING POTENTIALLY CONTAMINATED LIQUIDS South Condensate Storage Tank** 450,000 Condensate Backwash Tank 12,500 North Condensate Storage Tank*** 700,000 Condensate Phase Separators (each) 12,500 Floor Drain Sample Tank 20,000 Waste Sludge Tank 17,000 Waste Sample Tank (each) 22,000 RWCU Phase Separators (each) 4,500 Waste Surge Tank 65,000 Spent Resin Tank 2,000 Waste Collector Tank 22,000 Torus/Suppression Pool 700,000 Floor Drain Collector Tank 20,000 Lab Drain Tank (each) 500 Chemical Waste Tank 5,200

For conversion purposes the SCST and NCST display in feet, perform Steps 1 through 3 above, then substitute the following calculation for Steps 4 through 8, with the result being entered into Step 9.

[Beginning Tank Level in Feet - Ending Tank Level in Feet] (Conversion Factor) = Gallons Released (Step 9)

- SCST = 11,800 gal/ft

- - NCST = 19,090 gal/ft

Nebraska Public Power District SKL034-50-76(49064)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 01 Task No.: 345022O0103

Title:

Determine Actions for a Control Rod Mispositioned > 2 Hours Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/As 2.1.19 (3.8), 2.1.37 (4.6)

Directions to Examiner:

1. This JPM evaluates the trainees ability to identify a mispositioned control rod and direct actions to recover from the situation.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 1 and the prepared 6.CRD.301 and PMIS RPIS printouts when ready to start the JPM.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-76(49064)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 01 Task No.: 345022O0103

Title:

Determine Actions for a Control Rod Mispositioned > 2 Hours Directions to Trainee:

When I tell you to begin, you are to perform a review of the weekly Withdrawn Control Rod Operability IST Test and perform actions as required. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. The plant is operating at rated conditions.

2. Surveillance procedure 6.CRD.301 has been completed by the At the Controls Operator earlier in the shift and needs to be reviewed.

General

References:

1. Surveillance Procedure 6.CRD.301, Withdrawn Control Rod Operability Test.

2. 10.13, Control Rod Sequence and Movement Control.

General Tools and Equipment:

1. PMIS RPIS printout for control rods BEFORE positions.

2. PMIS RPIS printout for control rods AFTER positions.

3. Completed 6.CRD.301 Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Task Standards:

1. The examinee determines the control rod has been mispositioned for two hours and requests guidance from Reactor Engineering.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

Surveillance 6.CRD.301 was completed 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago and needs to be reviewed. Perform a Shift Managers assessment of 6.CRD.301, Withdrawn Control Rod Operability Test and determine applicable actions based upon your review. Inform the evaluator when you have completed your assessment.

Nebraska Public Power District SKL034-50-76(49064)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 01 Task No.: 345022O0103

Title:

Determine Actions for a Control Rod Mispositioned > 2 Hours Start Time: ____________

Performance Checklist Standards Sat Unsat Reviewed the procedure and RPIS printout to

1. Review 6.CRD.301 verify all required steps performed and acceptance criteria was satisfied.

2. Identify Identified control rod 34-43 inserted 1 notch discrepancy from where it was prior to performance of between BEFORE procedure.

and AFTER PMIS RPIS printouts.

3. Verify control rod Reviewed 10.13 and Control Rod Sequence mis-positioned per Package to determine proper control rod position Procedure 10.13. and determined control rod was left at a position other than the intended position.

4. Notify SM of mis- Shift Manager notified of mis-positioned positioned control control rod rod.

5. Determine the number of notches Reviewed Control Rod Sequence Package and the rod is mis- PMIS RPIS printouts and determined rod 34-43 positioned. is one notch from its intended position.

6. Determine how Reviewed the RPIS printout time/date stamp long the control and determined the rod had been mis-rod has been mis- positioned for > 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

positioned.

Contact Reactor Engineer for recovery

7. Contact Rx Eng instructions.

Nebraska Public Power District SKL034-50-76(49064)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 01 Task No.: 345022O0103

Title:

Determine Actions for a Control Rod Mispositioned > 2 Hours Performance Checklist Standards Sat Unsat Notifies the Fuels and Reactor Engineering

8. Make notifications Manager, Operations Manager, and the Plant per 10.13. Manager of the mis-positioned control rod and actions taken.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-76(49064)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 01 Task No.: 345022O0103

Title:

Determine Actions for a Control Rod Mispositioned > 2 Hours ATTACHMENT 1 ANSWER KEY Verify control rod 34-43 mispositioned as follows:

A control rod "left" at a position other than the intended position per the Control Rod Sequence Package Notify SM control rod.34-43 is mispositioned.

Determine control rod has been mispositioned for > 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Contact Reactor Engineer for recovery instructions.

The Nuclear Engineering Manager, Operations Manager, and General Manager of Plant Operations shall be notified of the mispositioned control rod and recovery actions taken.

Nebraska Public Power District SKL034-50-76 (49064)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to perform a review of the 31 day test of the Withdrawn Control Rod Operability Test and perform actions as required. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. The plant is operating at rated conditions.

2. Surveillance procedure 6.CRD.301 has been completed by the At the Controls Operator and needs to be reviewed.

Initiating Cue(s):

Surveillance 6.CRD.301 was completed 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago and needs to be reviewed. Perform a Shift Managers assessment of 6.CRD.301, Withdrawn Control Rod Operability Test and determine applicable actions based upon your review. Inform the evaluator when you have completed your assessment.

Record your assessment below:

Signature

USE: CONTINUOUS CNS OPERATIONS MANUAL QUALITY: QAPD RELATED SURVEILLANCE PROCEDURE 6.CRD.301 EFFECTIVE: 12/3/09 WITHDRAWN CONTROL ROD APPROVAL: ITR-RDM OPERABILITY IST TEST OWNER: OSG SUPV DEPARTMENT: OPS P&I: 1

1. PURPOSE ......................................................................................................................... 1

2. PRECAUTIONS AND LIMITATIONS ................................................................................. 1

3. REQUIREMENTS .............................................................................................................. 2

4. 31 DAY TEST..................................................................................................................... 4

5. 92 DAY IST TEST ................................................................................................................

6. RESTORATION ................................................................................................................. 7

7. ACCEPTANCE CRITERIA ................................................................................................. 7 ATTACHMENT 1 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #11 .................................................................................... 8 ATTACHMENT 2 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #21 ......................................................................................

ATTACHMENT 3 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #31 ......................................................................................

ATTACHMENT 4 SIGN-OFF AND REVIEW SHEET................................................... 10 ATTACHMENT 5 INFORMATION SHEET .................................................................. 12 REVISION VERIFICATION:

(initial use + every 7 days) LO Todays Date REV. DATE CHANGES 27 11/13/09 Clarified and removed limitations and steps for failed fuel performance.

Changed to align with TS Amendment 235, Control Rod Notch Testing 28 12/3/09 Frequency, per TSTF-475.

1. PURPOSE 1.1 This procedure provides instructions for Operations personnel to perform an OPERABILITY test of the withdrawn control rods.

2. PRECAUTIONS AND LIMITATIONS 2.1 Do not attempt to operate a control rod that has an inoperable drive.

2.2 Do not withdraw or insert more than one control rod at a time. Testing of a control rod shall be completed prior to testing another control rod.

2.3 Closely monitor nuclear instrumentation and CRD System instrumentation while moving control rods.

2.4 Do not withdraw control rod without first ensuring it is only rod selected. A malfunction of Reactor Manual Control System could cause more than one rod to be selected.2

2.5 All control rod movements performed when REACTOR MODE switch is in START &

HOT STBY or RUN shall be Concurrently Verified by a second Licensed Operator or an individual certified as an STE.

2.6 Data sheets used to record OPERABILITY test shall be checked with rod position printout or hard copy of PMIS RPIS display.

2.7 Whenever possible, for a long term lay-up (> 28 days), continue normal cooling water flow for a minimum of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months once per week plus cycle drives a minimum of one notch in and out once each week.

2.8 To provide an acceptable IST closure test of HCU cooling water check valve 138, control rod shall be inserted with normal drive water differential pressure ( 265 psid).

2.9 IST test frequency is once every 92 days.

2.10 If a fuel leaker is suspected or confirmed, the following additional limitations apply:

2.10.1 With a fuel leak, testing partially inserted control rods or control rods in flux suppression locations requires reactor power to be reduced at least 5% below current steady state power level or as indicated by Reactor Engineering, prior to testing. If leak location is not known, power will be reduced by 10% for performance of this procedure.

2.10.2 If a control rod double notches during testing, do not recover the control rod without Reactor Engineering concurrence. Further power reduction may be required to recover without causing fuel damage. Reactor Engineering must evaluate and provide a recovery plan.

2.10.3 If required power reduction extends 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , restoration of power to current steady state power level is limited to 1% per hour or as directed by Reactor Engineering.

2.11 Proper rod selection consists of the Operator ensuring that all indications of rod selection (i.e., Rod Select Matrix, Full Core Display, and Four Rod Display) respond normally or as expected and that LPRM signals on the 16 LPRM readout windows are normal or as expected for the power level and rod selected.4

3. REQUIREMENTS 3.1 Check section to be performed and discard section not to be performed:

[ X ] 4, 31 DAY TEST

[ ] 5, 92 DAY IST TEST

3.2 Check Attachment to be performed this shift and discard attachments not to be performed:7

[ X ] ATTACHMENT 1

[ ] ATTACHMENT 2

[ ] ATTACHMENT 3 3.3 Obtain Shift Manager's permission to perform test.

SM Signature/Date: Shift Manager/ Todays date 3.4 Inform and obtain signature from Control Room Operator.

CR Operator Signature/Date: BOP/ Todays date 3.5 Ensure no other tests affecting CRD System are in progress.

3.6 Obtain rod position printout or hard copy of PMIS RPIS display and attach it to procedure.

N/A 3.7 (Checked By) For each control rod at 00, mark its row N/A on Attachment 1, 2, or 3, as applicable.

Performed By:

Checked By:

4. 31 DAY TEST 4.1 Mark DRIVE WATER DP columns N/A on Attachment 1, 2, or 3, as applicable.

NOTE - Person performing Checked By shall be a Licensed Operator or Shift Technical Engineer.

4.2 (Checked By) For each control rod at 02 to 44, mark its COUPLING CHECK and WITHDRAW STALL FLOW blocks N/A, as required, on Attachment 1, 2, or 3.6 Performed By: LO Checked By: STE NOTE - Withdraw timer malfunction test needs only to be performed once for 31 DAY TEST, normally performed with Attachment 1, and may be N/A for other attachments.

4.3 Check withdraw timer malfunction by performing following for any control rod:

4.3.1 Select desired control rod and ensure following:2,4 4.3.1.1 Desired control rod is only button that backlights on ROD SELECT MATRIX.

4.3.1.2 Desired control rod is only control rod that backlights on FULL CORE DISPLAY.

4.3.1.3 LPRM signals on the 16 LPRM readout windows are normal or as expected for the power level and rod selected.

4.3.2 Place and hold TIMER TEST SWITCH to TEST and check following:

4.3.2.1 White TEST light turns on while switch is in TEST.

4.3.2.2 Red timer MALFUNCTION SELECT light turns on after ~ 3 to 5 seconds.

4.3.2.3 Selected control rod light on full core display turns off.

4.3.3 Release TIMER TEST SWITCH.

4.3.4 AC Verify a control rod cannot be selected.

4.3.5 Place TIMER TEST SWITCH to RESET and check following:

4.3.5.1 Red timer MALFUNCTION SELECT light turns off.

4.3.5.2 Control rod can be selected.

NOTE 1 - If control rod to be tested is at 02 to 46, Step 4.4 provides guidance for testing control rod.

NOTE 2 - If control rod to be tested is at 48, Step 4.5 provides guidance for testing control rod.

NOTE 3 - Step 4.4 or 4.5 may be performed in any order as directed by SM or CRS.

4.4 For each control rod at 02 to 46, perform following:

4.4.1 Using concurrent verification, select desired control rod and ensure following:3,4 4.4.1.1 Desired control rod is only button that backlights on ROD SELECT MATRIX.

4.4.1.2 Desired control rod is only control rod that backlights on FULL CORE DISPLAY.

4.4.1.3 LPRM signals on the 16 LPRM readout windows are normal or as expected for the power level and rod selected.

4.4.1.4 Record in DESIRED ROD SELECTED block for the selected rod.

4.4.2 If control rod at Position 46, perform following (Steps 4.4.3 and 4.4.4 are N/A):

4.4.2.1 Withdraw selected control rod by performing following:

a. Apply continuous withdraw signal.

b. Check red DRIFT light on full core display for selected rod does not turn on.

c. Check Annunciator 9-5-1/B-4, ROD OVERTRAVEL, does not alarm.

d. Record WITHDRAW STALL FLOW from CRD-FI-305 for selected control rod.

e. Release continuous withdraw signal.

f. If Steps 4.4.2.1b and 4.4.2.1c are satisfied, record in COUPLING CHECK block for selected control rod.5 4.4.2.2 Insert selected control rod one notch.

4.4.3 Insert selected control rod one notch.

4.4.4 Withdraw control rod one notch and perform following:

4.4.4.1 Check control rod does not continue to slowly withdraw additional notches.

4.4.4.2 If continued withdrawal is observed, insert control rod using normal or emergency insert and take action per Procedure 2.4CRD.

4.4.5 If control rod inserted and withdrew one notch, record in CONTROL ROD EXERCISED block for selected control rod.

4.5 For each control rod at 48, perform following:

4.5.1 Using concurrent verification, select desired control rod and ensure following:3,4 4.5.1.1 Desired control rod is only button that backlights on ROD SELECT MATRIX.

4.5.1.2 Desired control rod is only control rod that backlights on FULL CORE DISPLAY.

4.5.1.3 LPRM signals on the 16 LPRM readout windows are normal or as expected for the power level and rod selected.

4.5.1.4 Record in DESIRED ROD SELECTED block for the selected rod.

4.5.2 Insert selected control rod one notch.

4.5.3 Withdraw selected control rod by performing following:

4.5.3.1 Apply continuous withdraw signal.

4.5.3.2 Check red DRIFT light on full core display for selected rod does not turn on.

4.5.3.3 Check Annunciator 9-5-1/B-4, ROD OVERTRAVEL, does not alarm.

4.5.3.4 Record WITHDRAW STALL FLOW from CRD-FI-305 for selected control rod.

4.5.3.5 Release continuous withdraw signal.

4.5.3.6 If Steps 4.5.3.2 and 4.5.3.3 are satisfied, record in COUPLING CHECK block for selected control rod.5 4.5.4 If control rod inserted and withdrew one notch, record in CONTROL ROD EXERCISED block for selected control rod.

5. RESTORATION 5.1 Obtain rod position printout or hard copy of PMIS RPIS display from PMIS and attach it to procedure.

NOTE - Person performing Checked By shall be a Licensed Operator or Reactor Engineer.

5.2 (Checked By) Ensure control rods have been restored to pattern prior to start of test or if control rods were adjusted concurrently with this procedure, to a pattern specified by a Reactor Engineer.

Performed By: RO Checked By: STE 5.3 (Checked By) Deselect control rod by turning ROD SELECT power switch to OFF and back to ON.

Performed By: RO Checked By: STE

6. ACCEPTANCE CRITERIA 6.1 Non-TS A control rod could not be selected when withdraw timer malfunction was tested.

6.2 [SR 3.1.3.3, SR 3.1.3.5] Check marks recorded in all applicable shaded blocks on Attachment 1, 2, or 3 indicate checks were performed satisfactorily.

6.3 If Step 7.1 is not satisfied, ensure no control rod is withdrawn until repairs are made.

N/A

ATTACHMENT 1 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #11 ATTACHMENT 1 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #11

1. If any abnormal CRDM conditions are noted, such as high drive water pressure to move control rod, double notching, slow movement, or failure to move or complete notch on first attempt, initiate a Condition Report/Notification and evaluate OPERABILITY per Procedure 0.5.OPS.

CONTROL DESIRED DRIVE WITHDRAW CONTROL ROD ROD WATER COUPLING STALL ROD PERFORMED CONCURRENT NUMBER SELECTED DP CHECK FLOW EXERCISED BY VERIFICATION 18-51 N/A 0.9 LO STE 22-51 N/A 1.3 LO STE 26-51 N/A 1.3 LO STE 30-51 N/A 2.1 LO STE 34-51 N/A 1.5 LO STE 10-47 N/A 0.9 LO STE 14-47 N/A 1.3 LO STE 18-47 N/A 1.3 LO STE 22-47 N/A 2.1 LO STE 26-47 N/A 1.5 LO STE 30-47 N/A 1.5 LO STE 34-47 N/A 1.8 LO STE 38-47 N/A 2.2 LO STE 42-47 N/A 2.5 LO STE 06-43 N/A 1.2 LO STE 10-43 N/A 1.9 LO STE 14-43 N/A 1.9 LO STE 18-43 N/A 2.2 LO STE 22-43 N/A 2.5 LO STE 26-43 N/A 1.8 LO STE 30-43 N/A 0.9 LO STE

ATTACHMENT 1 WITHDRAWN CONTROL ROD OPERABILITY TEST DATA SHEET #11 CONTROL DESIRED DRIVE WITHDRAW CONTROL ROD ROD WATER COUPLING STALL ROD PERFORMED CONCURRENT NUMBER SELECTED DP CHECK FLOW EXERCISED BY VERIFICATION 34-43 N/A 1.3 LO STE 38-43 N/A 2.2 LO STE 42-43 N/A 2.5 LO STE 46-43 N/A 1.8 LO STE 06-39 N/A 1.3 LO STE 10-39 N/A 2.1 LO STE 14-39 N/A 1.5 LO STE 18-39 N/A 1.5 LO STE 22-39 N/A 1.8 LO STE 26-39 N/A 0.9 LO STE 30-39 N/A 1.3 LO STE 34-39 N/A 1.3 LO STE 38-39 N/A 2.1 LO STE 42-39 N/A 1.5 LO STE 46-39 N/A 2.8 LO STE

ATTACHMENT 4 SIGN-OFF AND REVIEW SHEET ATTACHMENT 4 SIGN-OFF AND REVIEW SHEET 31 DAY 92 DAY Initials Printed Name Initials Printed Name LO / Licensed Operator /

STE / Shift Technical Engineer /

/ /

Acceptance Criteria Satisfied: [ X ] YES; [ ] NO Initials/Date: LO/Todays date Shift Manager Review: Date:

IST Engineer Review (required): Date:

RECORDS Entire procedure is sent to CNS Records (quality record upon final review signature).

ATTACHMENT 4 SIGN-OFF AND REVIEW SHEET Initial/date by each discrepancy or resolution listed.

DISCREPANCIES # RESOLUTIONS

ATTACHMENT 5 INFORMATION SHEET ATTACHMENT 5 INFORMATION SHEET

1. SURVEILLANCE REQUIREMENTS - TECHNICAL SPECIFICATIONS 1.1 This procedure satisfies the requirements of:

1.1.1 SR 3.1.3.3.

1.1.2 SR 3.1.3.5.

1.1.3 SR 3.9.5.1.

1.1.4 Part of Section 5.5.6.

2. DISCUSSION 2.1 This procedure provides steps to test the withdrawn OPERABLE control rods by moving the control rods and checking for proper response. Also, withdraw stall flows are recorded 31 day for control rods that are withdrawn to 48 and a check for control rod coupling is performed on these control rods. Stall flows are periodically reviewed by System Engineer for determining maintenance on CRDMs.

2.2 The control rod inserting with normal drive water pressure ( 265 psid) satisfies the CNS IST Program closure exercise test for the HCU cooling water check valve CRD-CV-138 associated with the tested control rod.

2.3 Satisfactory control rod insertion and withdrawal checks proper operation of CRD insert, withdrawal, and exhaust (two) solenoid valves CRD-SOV-SO123, CRD-SOV-SO122, CRD- SOV-SO120, and CRD-SOV-SO121. This satisfies CNS IST Program requirements for closure test and fail safe test of these valves associated with the tested control rod.

3. REFERENCES 3.1 TECHNICAL SPECIFICATIONS 3.1.1 LCO 3.1.3, Control Rod OPERABILITY.

3.1.2 LCO 3.9.5, Control Rod OPERABILITY - Refueling.

3.1.3 Section 5.5.6, Inservice Testing Program.

3.2 UPDATED SAFETY ANALYSIS REPORT 3.2.1 Section III-5.0, Control Rod Drive Mechanical System.

3.2.2 Section VII-7.4.2, Reactor Manual Control System Operation.

ATTACHMENT 5 INFORMATION SHEET 3.3 CODES AND STANDARDS 3.3.1 ASME Code for Operation and Maintenance of Nuclear Power Plants.

3.3.2 CNS IST Program.

3.4 DRAWINGS 3.4.1 FLOW DIAGRAMS 3.4.1.1 B&R Drawing 2039, CRD Hydraulic System.

3.4.1.2 GE Drawing 945E617, CRD System ARI Modification.

3.4.2 ELECTRICAL DIAGRAMS 3.4.2.1 B&R Drawing 3002, Auxiliary One Line Diagram.

3.4.2.2 B&R Drawing 3007, Auxiliary One Line Diagram.

3.4.2.3 B&R Drawing 3010, Vital One Line Diagram.

3.4.2.4 B&R Drawing 3028, Elementary Diagram.

3.4.2.5 GE Drawing 791E254, CRD Hydraulic Instruction.

3.5 VENDOR MANUALS 3.5.1 CNS Number 0023, CRD HCU.

3.5.2 CNS Number 0025, Worthington Pumps.

3.6 PROCEDURES 3.6.1 Administrative Procedure 0.5.OPS, Operations Review of Notifications/Operability Determination.

3.6.2 General Operating Procedure 2.1.10, Station Power Changes.

3.6.3 Abnormal Procedure 2.4CRD, CRD Trouble.

3.6.4 Engineering Procedure 3.9, ASME Code Testing of Pumps and Valves.

3.7 MISCELLANEOUS 3.7.1 1 CR 2007-06350, CA-00001, Concurrent Verification. Affects Attachments 1, 2, and 3.

3.7.2 2 GE SIL-174, Improved Rod Select Pushbuttons. Affects Steps 2.4, 4.3.1, and 5.2.1.

ATTACHMENT 5 INFORMATION SHEET 3.7.3 3 GE SIL-292, Inadvertent Control Rod Withdrawal. Affects Steps 4.4.1, 4.5.1, 5.3.1, and 5.4.1.

3.7.4 4 GE SIL-625, Rod Block Monitor Rod Select Failure, Issued 26APR00.

Affects Steps 2.11, 4.3.1, 4.4.1, 4.5.1, 5.2.1, 5.3.1, and 5.4.1.

3.7.5 5 NEDO-33091, Improved OBPWS Control Rod Insertion Process, Revision 2, April 2003. Affects Steps 4.4.2.1f, 4.5.3.6, 5.3.3.1f, and 5.4.4.6.

3.7.6 6 SER 14-91, Inadvertent Withdrawal of a Control Rod During Testing.

Affects Steps 4.2 and 5.1.

3.7.7 7 TIP Plan 5.2.1.1, Revision 2, Action Step 18h, Reactivity Management and RCR 2002-1085, Action #4. Affects Step 3.2.

Post-surveillance RPIS printout.

Pre-surveillance RPIS printout.

Nebraska Public Power District SKL034-50-66 (44052)

Cooper Nuclear Station Page 1 of 4 Job Performance Measure for Operations Revision 01 Task No.: 299015O0301 Task

Title:

Assess Non-Scheduled Call-Out Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR, SIM, Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A 2.1.23 (3.9/4.0)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to perform a Shift Managers assessment of Non-Scheduled Work Call-Outs.

2. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

3. Give the trainee his copy of the Directions to the Trainee (Attachment 1) when ready to start the JPM.

4. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes:_____________________________________________________________________

Directions to Trainee:

When I tell you to begin, you are to perform a Shift Managers assessment of Non-Scheduled Work Call-Outs. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

Nebraska Public Power District SKL034-50-66 (44052)

Cooper Nuclear Station Page 2 of 4 Job Performance Measure for Operations Revision 01 Task No.: 299015O0301 Task

Title:

Assess Non-Scheduled Call-Out General Conditions:

1. The plant is operating at 100% power.

2. Shift Staffing has not been met by two (2) Station Operators.

3. 2 Hour LCO is in effect for minimum staffing not met.

4. Call-outs have been initiated, but both responders have indicated that they have consumed alcohol within the past 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

5. John reported consuming 2 beers with dinner 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago, and he feels no way impaired now.

6. Bill reported consuming 1 beer 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago and feels no way impaired now.

General

References:

1. Procedure 0-FFD-01

2. Tech Specs General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Task Standards:

1. The student determines a breathalyzer test is required and BAC limits apply.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

You are to perform a Shift Managers assessment of 0-FFD-01 Non-Scheduled Work Call-Outs, and determine applicable actions based upon your review. Inform the evaluator when you have completed your assessment.

Nebraska Public Power District SKL034-50-66 (44052)

Cooper Nuclear Station Page 3 of 4 Job Performance Measure for Operations Revision 01 Task No.: 299015O0301 Task

Title:

Assess Non-Scheduled Call-Out Start Time: __________

Performance Checklist Standards Sat Unsat Referred to 0-FFD-01 section 16 for call outs

1. Refer to 0-FFD-01.

for unscheduled work.

2. Determine that the Determined the two individuals that were two individuals have called in have consumed alcohol in the past 5 consumed alcohol. hours.

3. Determine that the Determined the exception allowed in step exception can be 16.2.1 can be applied to both people.

applied.

Determined that once the individuals had

4. Determine that a arrived on site, that a Breathalyzer test breathalyzer test was mandatory, and that the individuals is required. had to pass it with a BAC of < 0.040% to be allowed on site.

5. Notify the evaluator. Notified the evaluator that both individuals can report to site, and turned in the procedure.

Stop Time: __________ Total Time: __________

Nebraska Public Power District SKL034-50-66 (44052)

Cooper Nuclear Station Page 4 of 4 Job Performance Measure for Operations Revision 01 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin, you are to perform a Shift Managers assessment of Non-Scheduled Work Call-Outs. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

General Conditions:

1. The plant is operating at 100% power.

2. Shift Staffing has not been met by two (2) Station Operators.

3. 2 Hour LCO is in effect for minimum staffing not met.

4. Call-outs have been initiated, but both responders have indicated that they have consumed alcohol within the past 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

5. John reported consuming 2 beers with dinner 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ago, and he feels no way impaired now.

6. Bill reported consuming 1 beer 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ago and feels no way impaired now.

You are to perform a Shift Managers assessment of 0-FFD-01 Non-Scheduled Work Call-Outs, and determine applicable actions based upon your review. Inform the evaluator when you have completed your assessment.

Record your assessment below:

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 1 of 8 Job Performance Measure for Operations Revision 03 Task No.: 202012O0201

Title:

PERFORM JET PUMP OPERABILITY CHECK (SRO)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: Classroom / SIM

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 18 minutes

5. K/A 202001 K1.06 3.6/3.6 Directions to Examiner:

1. This JPM evaluates the trainee's ability to perform the daily Jet Pump and Recirc Pump Flow Check of the Daily Tech Specs Surveillance Log.

2. Observe the trainee during performance of the JPM for proper use of self-checking methods.

3. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

4. Provide the Reactor Recirc and Jet Pump values to the candidate (Attachment 1);

do not allow them to take the readings from the simulator.

5. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 2 of 8 Job Performance Measure for Operations Revision 03 Task No.: 202012O0201

Title:

PERFORM JET PUMP OPERABILITY CHECK (SRO)

Directions to Trainee:

When I tell you to begin, you are to perform the activities associated with daily Jet Pump AND Recirc Pump Flow Check. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is operating at 100% power with DEH in Mode 4.

2. Both Reactor Recirculation pumps are operating with pump flows balanced.

3. Data is provided for the readings to be taken.

General

References:

1. Procedure 6.LOG.601 General Tools and Equipment:

1. Calculator.

2. Jet pump operability curves.

Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Task Standards:

1. Determines Jet Pump Operability incorrectly checked SAT for Check 3 and Jet Pump 13 d/p is out of specification low.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

Perform the Control Room Supervisors 0630-1830 review of the daily Jet Pump (6.LOG.601 2) and Recirc Pump Flow Check (6.LOG.601 Attachment 13) as part of the routine shift activities to ensure they have been performed correctly. Record your conclusions on Page 4 of Attachment 1.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 3 of 8 Job Performance Measure for Operations Revision 03 Task No.: 202012O0201

Title:

PERFORM JET PUMP OPERABILITY CHECK (SRO)

Start Time: ____________

Performance Checklist Standards Sat Unsat Reviewed Attachment 13 Jet Pump Operability Items A, B, C and D. Noted all have been filled

1. Reviews Attachment out correctly.

13 Jet Pump Operability Items.

CUE: IF asked to verify data provide, report the data has been verified.

2. Reviews Reviewed Attachment 13 Jet Pump Attachment 13 Jet Operability Items 1, 2, and 3. Noted Check 3 Pump Operability should be checked Unsat for loop A and Checks 1, 2 and 3. denoted it.

3. Reviews Reviewed Attachment 14 RECIRC Pump Flow Attachment 14 Checks, NBI-DPR/FR-95 was outside the RECIRC Pump Exclusion Region. Checked that it was Flow Checks marked SAT correctly.

Checks

4. Reviews Attachment Reviewed Attachment 14 RECIRC Pump Flow 14 RECIRC Pump Checks, NBI-FI-92A/B Mismatch should be N/A Flow Checks for this flow.

Checks

5. Reviews Attachment 14 Reviewed Attachment 14 RECIRC Pump Flow RECIRC Pump Checks, NBI-FI-92A/B Mismatch and checked Flow Checks it was marked SAT correctly.

Checks

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 4 of 8 Job Performance Measure for Operations Revision 03 Task No.: 202012O0201

Title:

PERFORM JET PUMP OPERABILITY CHECK (SRO)

6. Reviews Attachment Reviewed Attachment 14 RECIRC Pump Flow 14 RECIRC Pump Checks, Recirc Pump operating or RHR Pump Flow Checks operating in SDC is marked N/A.

Checks NOTE to Examiner: Rx Engineer may be notified after conclusions given to SM.

Notified the Reactor Engineer that Jet Pump 13 d/p is low out of spec.

7. Notifies Reactor Engineer CUE: As reactor engineer respond to the report and tell the CRS that it was not low last time the surveillance was run.

8. Notifies the Shift Denoted Jet Pump 13 was low out of speculation Manager and Attachment 13 information was correct.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 5 of 8 Job Performance Measure for Operations Revision 03 ATTACHMENT 1 (Page 1 of 4)

Directions to Trainee:

When I tell you to begin, you are to perform the activities associated with daily Jet Pump AND Recirc Pump Flow Check. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is operating at 100% power with DEH in Mode 4.

2. Both Reactor Recirculation pumps are operating with pump flows balanced.

3. Data is provided for the readings to be taken.

Initiating Cue(s):

Perform the Control Room Supervisors 0630-1830 review of the daily Jet Pump (6.LOG.601 2) and Recirc Pump Flow Check (6.LOG.601 Attachment 13) as part of the routine shift activities to ensure they have been performed correctly. Record your conclusions on Page 4 of Attachment 1.

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 6 of 8 Job Performance Measure for Operations Revision 03 ATTACHMENT 1 (Page 2 of 4) 3 JET PUMP OPERABILITY ATTACHMENT 13 JET PUMP OPERABILITY NOTE - If in single loop operation, mark idle loop N/A.

JET PUMP P (%)

ITEMS LOOP A LOOP B JP # LOOP B JP # LOOP A 6

A Core Flow (10 lb/hr) NBI-DPR/FR-95 62 1 31 11 32 3

B RR Pump Flow (10 gpm) RR-FR-163 38.5 38.5 2 32 12 30 C RRMG Set Speed (%) RRFC-SI-1A/B 79 80 3 31 13 20 6

D JP Flow (10 lb/hr) NBI-FI-92A/B 31 31 4 31 14 32 5 36 15 34 6 34 16 35 7 33 17 32 8 32 18 31 9 33 19 33 10 33 20 32 LOOP B LOOP A Avg 32.6 Avg 31.1 OPERABILITY APPLICABLE ATT. 21 CHECKS SAT UNSAT LIMIT MODE NOTE Item B and C values within Loop A:

1 (a) curve limits Loop B:

Item C and D values within Loop A:

2 (a) SAT curve limits Loop B:

(b) (b) 1 ,2 50 Jet Pump P differs by 20% Loop A:

3 (a, c) from established patterns Loop B:

Loop A:

Checks 1 and 2 SAT, or SAT Check 3 SAT Loop B:

(a) Refer to Jet Pump Operability Curves maintained in Control Room.

(b) Required to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after > 25% RTP and required to be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after associated recirculation loop is in service.

(c) If any Jet Pump P vs. established pattern is not within curve limits, immediately notify Reactor Engineering.

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 7 of 8 Job Performance Measure for Operations Revision 03 ATTACHMENT 1 (Page 3 of 4)

Attachment 14 RECIRC PUMP FLOW CHECKS ATTACHMENT 14 RECIRC PUMP FLOW CHECKS 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 21 CHECKS SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES Core Flow, NBI-DPR/FR-95, value is not in Stability SAT MCO SAT 1, 2 49 Exclusion Region of Power (a) to Flow Map JP Flow, NBI-FI-92A/B, values for Loop A and Loop B flow mis-match is (c) (c) 6 N/A MCO SAT 1 ,2 48, 127 7.35x10 lbs/hr at 6

< 51.45x10 lbs/hr Rated (b) (d)

Core Flow JP Flow, NBI-FI-92A/B, values for Loop A and Loop B flow mis-match is (c) (c) 6 SAT MCO SAT 1 ,2 48, 127 3.67x10 lbs/hr at 6

51.45x10 lbs/hr Rated (b) (d)

Core Flow (a) Refer to power to flow map in Procedure 2.1.10.

(b) Per Technical Specification and TRM Bases, a recirculation loop is considered not in operation when the mis-match between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation.

(c) Required to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in service.

(d) If a flow mismatch results in one loop considered not in operation, contact FRED to manually insert single loop operation limits into GARDEL.

0700-1000 1900-2200 READING READING OPERABILITY APPLICABLE ATT. 21 (e) (e)

CHECK () () LIMIT MODES NOTES 3 (when Reactor Verify Recirc Pump operating Pressure is less RR Pump or or RHR Pump operating in N/A than SDC 66 SDC in Operation SDC Pressure (f)

Permissive)

(e) indicates at least one RR pump or SDC Subsystem is in service.

(f) Required to be met within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure is less than the shutdown cooling permissive pressure.

Nebraska Public Power District SKL034-20-113(29979)

Cooper Nuclear Station Page 8 of 8 Job Performance Measure for Operations Revision 03 ATTACHMENT 1 (Page 4 of 4)

==

Conclusion:==

Signature xxxxxxxxx

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 01 Task No.: 200150P0501

Title:

Release Rate Determination Based On Drywell Curie Content And Vent Flow Rate Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 16 minutes

5. NRC K/A Rating 295038 EA2.01 (3.3/4.3) 2.3.11 (2.7/3.2)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to perform a release rate determination based on curie content and vent flow rate.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 01 Task No.: 200150P0501

Title:

Release Rate Determination Based On Drywell Curie Content And Vent Flow Rate Directions to Trainee:

When I tell you to begin, you are to perform a release rate determination based on Drywell curie content and vent flow rate. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The plant had been in a normal line up with the reactor at 100% power for last 10 days.

2. The Reactor is shutdown.

3. A LOCA has occurred.

4. Torus-to-drywell vacuum breakers have stuck open.

5. Drywell spray is unavailable.

6. Kaman ERP effluent monitors are out of service.

7. The reactor was scrammed at 15:05.

8. Level fell to -50" on fuel zone instruments prior to recovery.

9. The PMIS system is out of service.

10. RMA-RR-40, high range containment radiation recorder, is out of service.

11. RMA-RR-40A indicates 70 R/hr.

12. RMA-RR-40B indicates 80 R/hr.

General

References:

1. Procedure 5.7.16 General Tools and Equipment:

1. Scientific calculator.

Special Conditions, References, Tools, Equipment:

1. .Critical checks denoted in bold.

2. Simulator cues denoted by "#".

Task Standards:

1. Correctly calculates Noble Gas Release Rate within tolerance.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 01 Task No.: 200150P0501

Title:

Release Rate Determination Based On Drywell Curie Content And Vent Flow Rate Initiating Cue(s):

It is now 22:35 hours. The decision has been made to vent the Drywell through SBGT using the 1" line IAW 5.8.18. The Shift Manager has directed you to perform a release rate determination based on Drywell Curie content. Inform the Shift Manager when you have completed the task.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 01 Task No.: 200150P0501

Title:

Release Rate Determination Based On Drywell Curie Content And Vent Flow Rate Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Obtain Procedure Obtained Procedure 5.7.16.

2. Calculate release Calculated release rate and recorded results in rate Step 10 of Attachment 2.

3. Hand in work Signed and handed in completed Att.2 Note to Examiner: Acceptable release rate range is 4.122 x 105 to 5.038 x 105 µCi/sec

4. Release rate The calculated Curie release rate was agrees with 4.58 x 105 µCi/sec (+/- 4.58 x 104).

Answer Key Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 01 Task No.: 200150P0501

Title:

Release Rate Determination Based On Drywell Curie Content And Vent Flow Rate ATTACHMENT 1 ANSWER KEY (5)

(3) ESTIMATED PROJECTED (4) DRYWELL DBA-LOCA PROJECTED NOBLE GAS (2) EXPOSURE DBA-LOCA CURIE (8)

ACTUAL RATE AT NOBLE GAS CONTENT (7) DRYWELL (9) (10)

(1) CONT. EFFECTIVE CURIE (Ci) (6) NOBLE VENTING CONV. NOBLE GAS EFFECTIVE EXPOSURE AGE CONTENT D/W GAS CONC. FLOW FACTOR RELEASE RATE (2) 3 (Ci/min to AGE RATE (rem/hr) (Ci) x (4) VOLUME (Ci/ft ) RATE (µCi/sec)

(hours) (rem/hr) (from Att. 4) (from Att. 6) (3) 3 (ft ) (5) ÷ (6) (cfm) µCi/sec) (7) x (8) x (9) 7.5 80R/hr 9.0 x 105 1.5 x 108 1.24 x 104 1.45E5 8.6 x 10-2 319 cfm 1.67E4 4.58 x 105 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 1.45E5 1.67E4 Completed By: Date:_____________

Nebraska Public Power District SKL034-50-25(8904)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 01 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to perform a release rate determination based on Drywell curie content and vent flow rate. Before you start, I will state the general plant conditions, the initiating cues and answer any questions you may have.

General Conditions:

1. The plant had been in a normal line up with the reactor at 100% power for last 10 days.

2. The Reactor is shutdown.

3. A LOCA has occurred.

4. Torus-to-drywell vacuum breakers have stuck open.

5. Drywell spray is unavailable.

6. Kaman ERP effluent monitors are out of service.

7. The reactor was scrammed at 15:05.

8. Level fell to -50" on fuel zone instruments prior to recovery.

9. The PMIS system is out of service.

10. RMA-RR-40, high range containment radiation recorder, is out of service.

11. RMA-RR-40A indicates 70 R/hr.

12. RMA-RR-40B indicates 80 R/hr.

Initiating Cues:

It is now 22:35 hours. The decision has been made to vent the Drywell through SBGT using the 1" line IAW 5.8.18. The Shift Manager has directed you to perform a release rate determination based on Drywell Curie content. Inform the Shift Manager when you have completed the task.

Nebraska Public Power District SKL034-50-77(49065)

Cooper Nuclear Station Page 1 of 5 Job Performance Measure for Operations Revision 0 Task No.: 344018O0303

Title:

EAL Table Top 9 Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: STE/SRO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A 2.4.41 (4.6)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to make an EAL classification, given a set of plant conditions.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachments 1 and 2.

6. Brief the trainee, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-50-77(49065)

Cooper Nuclear Station Page 2 of 5 Job Performance Measure for Operations Revision 0 Task No.: 344018O0303

Title:

EAL Table Top 9 Directions to Trainee:

When I tell you to begin, you are to use the provided attachment to classify the event at the highest EAL attained. Also identify the basis for your classification including the procedural section. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. The plant has experienced the events listed on Attachment 2.

General

References:

1. Procedure 5.7.1, EAL Matrix.

2. Procedure 5.7.6, Attachment 1.

General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Task Standards:

1. The correct EAL is determined.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The plant has experienced the events listed on Attachment 2 of this JPM. The parameters listed have been present for the past 16 minutes. You are to classify the event at the highest EAL attained. Return the completed Attachment 2 to the examiner when you have completed this task.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-50-77 (49065)

Cooper Nuclear Station Page 3 of 5 Job Performance Measure for Operations Revision 0 Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Refer to Procedure Referred to Procedure 5.7.1, EAL Classification 5.7.1, EAL Matrix Matrix, Attachment 4.

Classified the event as an ALERT based on EPIP 5.7.1, SA2.1 (Automatic scram failed and

2. Classify the event. manual actions successfully shut down reactor as indicated by power <3%)

3. Return completed Attachment 2 to Returned the completed Attachment 2 to the examiner. examiner.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-50-77 (49065)

Cooper Nuclear Station Page 4 of 5 Job Performance Measure for Operations Revision 0 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin, you are to use the provided attachment to classify the event at the highest EAL attained. Also identify the basis for your classification including the procedural section. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

General Conditions:

1. The plant has experienced the events listed on Attachment 2.

Initiating Cues:

The plant has experienced the events listed on Attachment 2 of this JPM. The parameters listed have been present for the past 16 minutes. You are to classify the event at the highest EAL attained. Return the completed Attachment 2 to the examiner when you have completed this task.

Nebraska Public Power District SKL034-50-77 (49065)

Cooper Nuclear Station Page 5 of 5 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Parameter/System/Component Status The plant is operating at 100% power and the following occur:

Both Reactor Feed Pumps trip.

When reactor water level passes through 0" Narrow Range all the RPS Group white lights stay illuminated.

The Reactor Operator depresses the Manual Scram Push Buttons and the white lights extinguish.

10 control rods fail to fully insert.

All APRMs indicate downscale.

4160V bus 1F de-energized on the scram and remains de-energized.

Highest EAL: _________________________________________________________________

Based on (Section and Description):_______________________________________________

Signature of Operator/STE:_____________________________________________

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 1 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 10 minutes

5. NRC K/A 239001 A4.02 (3.7 / 4.7)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to conduct alternate pressure control using the Main Steam Line drains.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 2 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Directions to Trainee:

When I tell you to begin, you are to conduct alternate pressure control using the Main Steam Line Drains as directed in procedure EOP 5.8.1. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The Plant had been operating at 100% power since startup from a refueling outage 3 months ago.

2. A turbine trip was experienced, followed by the expected Reactor Scram.

3. Due to lowering vessel level and various malfunctions of systems, the CRS has entered the EOPs.

4. The MSIVs are closed.

5. Radiation conditions through the plant are normal for plant conditions.

General

References:

1. EOP 5.8.1 General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

Task Standards:

1. Accurately locate, identify, operate and/or manipulate all component controls required to be utilized to conduct alternate pressure control using the MSL Drains.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 3 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Initiating Cue(s):

The CRS has directed you to conduct Alternate Pressure Control per EOP 5.8.1 using Main Steam Line Drains. Inform the CRS when a cooldown rate has been established.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 4 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Start Time: ____________

Performance Checklist Standards Sat Unsat Checked Main Condenser Vacuum on MS-PI-72A/B, A/B VACUUM on Panel B.

1. Check MC vacuum CUE: Main Condenser Vacuum is 7 Hg.

Ensured Reactor Mode Switch NOT in RUN.

2. Rx Mode Switch not in RUN. CUE: RX Mode Switch pointing to Shutdown.

Ensured all MSIV control switches in CLOSE.

___VLV AO 80A ___VLV AO 86A

___VLV AO 80B ___VLV AO 86B

___VLV AO 80C ___VLV AO 86C

3. MSIV C/S In close. ___VLV AO 80D ___VLV AO 86D CUE: VLV AO 80A-D pointing to CLOSE.

VLV AO 86A-D pointing to CLOSE.

Placed CONDENSER LOW VACUUM LOGIC TEST keylock switches to bypass.

___16A-S17A

___16A-S17B

4. Keylock switches ___16A-S17C to close. ___16A-S17D CUE: Keylock switches 16A-S17A-D have keys installed in switches and switches pointing to CLOSE.

Ensured Turbine Stop Valves Closed on DEH HMI.

5. Stop valves closed.

CUE: MT Stop Valve indications all GREEN.

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 5 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Performance Checklist Standards Sat Unsat At Panel 9-5, placed Channel A and B GROUP ISOL RESET switches to Gp 1 RESET (LEFT) and released to NOR.

CUE: CHANNEL A and B GROUP ISOL

6. Reset Group 1 RESET switches pointing to left RESET position and position to NOR when released.

GROUP 1 Channel A and B indicating lights all illuminated on Panel 9-5.

On Panel 9-4, opened MS-MO-74, INBD ISOL VLV.

7. Open MO-74 CUE: VLV MO 74 control switch pointing to OPEN. RED light on GREEN light off.

On Panel 9-4, opened MS-MO-77, OUTBD ISOL VLV.

8. Open MO-77 CUE: VLV MO 77 control switch pointing to OPEN. RED light on GREEN light off.

Note to Examiner: Opening MO-78 or MO-79 may be performed prior to opening MO-203, 204 and 205 if opening MO-78 or 79 is not expected to change Heater Bay dose rates.

Directed NLO to open following valves in Heater Bay

___MS-MO-205

___MS-MO-203

9. Open BV strainer ___MS-MO-204 blowdown valves

CUE: Five minutes have elapsed and the NLO reports MO-205, 203 and 204 are open.

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 6 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains Performance Checklist Standards Sat Unsat Throttled open either of following valves to control cooldown rate:

10. Begin pressure ___MS-MO-78 control. ___MS-MO79 CUE: Reactor pressure is lowering.

Reported to CRS the MSL drain header is being utilized and a cooldown rate has been

11. Inform CRS established.

CUE: CRS acknowledges report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-134(49057)

Cooper Nuclear Station Page 7 of 8 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Establish RPV Cooldown Using MSL Drains ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC IC-15 C. Run Batch File None D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial

3. Remotes Number Title Trigger TD Value Ramp 0

4. Overrides Instrument Tag Trigger TD Value Ramp None

a. Scram the reactor and place mode switch in SHUTDOWN.

b. Trip the turbine.

c. Open AR-MO-150 Vacuum Breaker until condenser vacuum < 8 Hg, and

5. Panel Setup then close MO-150.

d. Leave MSIV switches in OPEN.

e. Ensure MSIVs are closed.

f. Allow LLS to control RPV pressure.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-134 (49057)

Cooper Nuclear Station Page 8 of 8 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to conduct alternate pressure control using the Main Steam Line Drains as directed in procedure EOP 5.8.1. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The Plant had been operating at 100% power since startup from a refueling outage 3 months ago.

2. A turbine trip was experienced, followed by the expected Reactor Scram.

3. Due to lowering vessel level and various malfunctions of systems, the CRS has entered the EOPs.

4. The MSIVs are closed.

5. Radiation conditions through the plant are normal for plant conditions.

Initiating Cue(s):

The CRS has directed you to conduct Alternate Pressure Control per EOP 5.8.1 using Main Steam Line Drains. Inform the CRS when a cooldown rate has been established.

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 1 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A 264000 K4.07(3.3/3.4), A4.04(3.7/3.7)

Directions to Examiner:

NOTE: This JPM may be performed in conjunction with SKL034-21-67.

1. This JPM evaluates the trainee's ability to shut down Diesel Generator 2 from the Main Control Room.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 2 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

Directions to Trainee:

When I tell you to begin, you are to shut down diesel generator #2. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. DG2 was started per section 7 of 2.2.20.1.

2. All data has been taken and the load has been reduced to ~1000kW.

3. DG2 load has been steady at 1000 kW for the past 17 minutes.

4. DG2 is no longer required.

General

References:

1. Procedure 2.2.20.1 General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

Task Standards:

1. DG2 output breaker is opened and the engine is shutdown.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 3 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

Initiating Cue(s):

The Control Room Supervisor directs you to shut down DG 2 from the control room per Procedure 2.2.20.1 and inform the CRS when the task is complete.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 4 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

Start Time: ____________

Performance Checklist Standards Sat Unsat Verified DG2 was synchronized with the grid with

1. Verify conditions load <1000kW but > 400 kW.

Lowered DG2 load to 400 kW with the Governor

2. Lower KW to ~400 switch.

Reduced DG2 VARs as low as possible with the

3. Lower VARs Voltage Regulator switch.

4. Open Breaker Opened DIESEL GEN 2 BKR EG2 Note to Examiner: The booth operator must perform the action of placing the Droop Parallel switch to ISOC if this JPM is NOT being performed in conjunction with JPM SKL034-21-67.

Contacted NLO in DG2 room to place Droop Parallel switch to ISOCH.

5. Place Droop Parallel to ISOCH #CUE: If NOT being performed with SKL034-21-67, report Droop Parallel switch is in ISOCH.

Adjusted DG2 speed to ~60 Hz with the Governor

6. Verify speed switch.

Adjusted voltage to ~4200 VAC as necessary with

7. Verify voltage Voltage Regulator switch.

Note to Examiner: If asked, inform trainee this procedure is not being performed in conjunction with any maintenance procedures.

Placed and held DIESEL GEN 2 STOP/START

8. DG2 Engine switch to STOP for 1 to 2 seconds, then shutdown released.

Note to Examiner: Do NOT provide following cue if this JPM is being performed with JPM SKL034-21-67.

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 5 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

Performance Checklist Standards Sat Unsat Contacted the NLO and informed him that DG2 has been shut down and is ready for the following actions:

Check Lube Oil Press at Turbocharger is

< 5 psig..

Check that the Pre-Post Lube Oil Pump starts

Check butterfly valves on both banks open; locking device is installed and latch pawls engaged

After 5 min. place the Pre-Post Lube Oil Pump

9. Inform the Diesel in AUTO Room Operator of

Ensure the Start Mode Selector Switch is in actions. FAST

Fill the Day Tank.

SYNCH relay target is reset.

Voltage Permissive relay target is reset.

CUE: The NLO acknowledges the report and informs the operator that the above tasks are complete.

CUE: Inform trainee DG Data Acquisition Cabinet and DL750 recorders were NOT installed.

Directed NLO to fill DG2 Day Tank per Section 10.

10. Fill Day Tank.

CUE: Acknowledge request and report you are filling the Day Tank per Section 10.

Notified CRS that DG2 was shutdown.

11. Inform CRS.

CUE: CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-14(2173)

Cooper Nuclear Station Page 6 of 7 Job Performance Measure for Operations Revision 03 Task No.: 264023O0101

Title:

DG2 Shutdown From Control Room (2.2.20.1)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC Any IC C. Run Batch File None D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial

3. Remotes Number Title Trigger TD Value Ramp 0

4. Overrides Instrument Tag Trigger TD Value Ramp None

a. Start DG2 per Procedure 2.2.20.1, Section 7.

5. Panel Setup b. Load DG2 to ~ 1000 kW.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-14 (2173)

Cooper Nuclear Station Page 7 of 7 Job Performance Measure for Operations Revision 03 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to shut down diesel generator #2. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. DG2 was started per section 7 of 2.2.20.1.

2. All data has been taken and the load has been reduced to ~1000kW.

3. DG2 load has been steady at 1000 kW for the past 17 minutes.

4. DG2 is no longer required.

Initiating Cues:

The Control Room Supervisor directs you to shut down DG 2 from the control room per Procedure 2.2.20.1 and inform the CRS when the task is complete.

Nebraska Public Power District SKL034-21-135(49058)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Operate Reactor Water Sample Valves with Group 7 Present Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 5 minutes

5. NRC K/A 223002 K4.08 (3.3/3.7)

Directions to Examiner:

1. This JPM tests the ability of the trainee to bypass the Group 7 isolation and open the reactor water sample valves.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-135(49058)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Operate Reactor Water Sample Valves with Group 7 Present Directions to Trainee:

When I tell you to begin you are to open the reactor water sample valves with a PCIS Group 7 isolation present per Procedure 2.2.68.1.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in a post accident condition

2. RPV level had lowered to -158 inches and then recovered..

3. A PCIS Group 7 isolation is present..

4. Chemistry has requested the reactor water sample valve be opened per Procedure 8.PASS.1.

General

References:

1. Procedure 2.2.68.1

2. Procedure 8.PASS.1 General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

Task Standards:

1. Manually bypass PCIS Group 7 isolation signal and open the reactor water sample valves.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-21-135(49058)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Operate Reactor Water Sample Valves with Group 7 Present Initiating Cue(s):

Chemistry has contacted the control room and requested reactor water sample valves be opened so they can sample per their post-accident sampling system procedure 8.PASS.1.

Inform the CRS when the sample valves have been opened.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-135(49058)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Operate Reactor Water Sample Valves with Group 7 Present Start Time: ____________

Performance Checklist Standards Sat Unsat Requested permission to bypass the Group 7 isolation and open the reactor water sample valves.

1. Obtain SM approval

CUE: As SM give permission to perform the task.

2. Obtain two switch Obtained two switch keys form the Control keys. Room Key Depository.

At Panel 9-4, inserted key and placed PASS

3. Bypass 741 VLV 741 ISOL SIG BYPASS S-53 keylock isolation signal. switch to BYPASS.

At Panel 9-4, inserted key and placed PASS

4. Bypass 740 VLV 740 ISOL SIG BYPASS S-54 keylock isolation signal. switch to BYPASS.

Placed switch for RR-740-AV, OUTBD ISOL

5. Open RR-740AV. VLV, to OPEN.

Placed switch for RR-741-AV, INBD ISOL VLV,

6. Open RR-741AV. to OPEN.

Informed CRS reactor water sample valves RR-740AV and RR-741AV were opened.

7. Inform CRS.

CUE: CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-135(49058)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 0 Task No.:

Title:

Operate Reactor Water Sample Valves with Group 7 Present ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC IC-20 C. Run Batch File D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial

3. Remotes Number Title Trigger TD Value Ramp 0

4. Overrides Instrument Tag Trigger TD Value Ramp None

a. Scram the reactor and trip RFPs.

b. Put in malfunction RR 20a at 50% until RPV level lowers to -25 inches FZ then delete malfujction..

5. Panel Setup

c. Transfer RPS A and B to alternate power (causes Group 7)

d. Secure all RPV injection except for CRD and RCIC.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-135 (49058)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin you are to open the reactor water sample valves with a PCIS Group 7 isolation present per Procedure 2.2.68.1.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in a post accident condition

2. RPV level had lowered to -158 inches and then recovered..

3. A PCIS Group 7 isolation is present..

4. Chemistry has requested the reactor water sample valve be opened per Procedure 8.PASS.1.

Initiating Cues:

Chemistry has contacted the control room and requested reactor water sample valves be opened so they can sample per their post-accident sampling system procedure 8.PASS.1.

Inform the CRS when the sample valves have been opened.

Nebraska Public Power District SKL034-20-107(22658)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 04 Task No.: 249005P0401 Lowering DEH pressure setpoint (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

THIS IS AN ALTERNATE PATH JPM Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A 241000 A4.06 (3.9/3.9)

Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. During pressure reduction, the DEH Throttle Pressure signals will fail causing Pressure Set Point to fail to respond.

1. This JPM evaluates the trainee's ability to manipulate the DEH controls in order to lower reactor pressure during a Cooldown.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee his copy of the Directions to the Trainee (Attachment 2) when ready to start the JPM.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-20-107(22658)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 04 Task No.: 249005P0401 Lowering DEH pressure setpoint (Alternate Path)

Directions to Trainee:

When I tell you to begin, you are to lower reactor pressure to 800 psig at a rate of 10 psig /min using the hard card. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant has scrammed and pressure is approximately 930 psig.

General

References:

1. Procedure 2.2.77.1 (hard card)

General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Alternate path denoted by.

Task Standards:

1. The student enters the correct depressurization target setpoint and rate; recognizes DEH failure causes BPVs to close and takes manual control to open BPVs.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The CRS Orders you to lower Reactor Pressure to 800 Psig at a rate of 10 psig/min using the hard card.

Inform the CRS when the Reactor is at 800 psig.

Nebraska Public Power District SKL034-20-107(22658)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 04 Task No.: 249005P0401 Lowering DEH pressure setpoint (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Determine Determined depressurization rate does not depressurization does exceed a cooldown rate by looking in steam not exceed cooldown tables.

rate limits.

2. Press TARGET SP on Pressed TARGET SP to display keypad.

HMI.

3. Enter required set Cleared existing value and entered 800 then point. selected OK.

4. Verify HOLD. Verified the HOLD button backlit yellow.

5. Press RATE button. Pressed the RATE to display keypad.

Cleared the existing value and entered 10

6. Enter ramp rate.

then selected OK.

Pressed the GO button and verified it backlit

7. Press GO.

yellow.

8. When Pressure has lowered for Reported that Throttle Pressure was INVALID approximately one and all Bypass Valves have closed.

minute, insert Trg 1 Malfunction TC14A-C.

Reported to the CRS that annunciators B-1/C-1.

DEH CONTROL TROUBLE and B-1/F-1 DEH

9. Informs CRS. SYSTEM TROUBLE were alarming.

CUE: CRS acknowledged the report.

Nebraska Public Power District SKL034-20-107(22658)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 04 Task No.: 249005P0401 Lowering DEH pressure setpoint (Alternate Path)

Performance Checklist Standards Sat Unsat Requested guidance for continuing, and took the controller to manual and lowered pressure to 800 psig.

10. CRS orders BPV control in Manual. #CUE: As CRS, Inform the Operator to take manual control and lower Reactor Pressure to 800 psig.

11. Operator takes On the BPV screen pressed MANUAL and manual control of acknowledged the request.

BPVs.

Pressed the JOG UP and FAST/SLOW buttons as necessary to start lowering

12. Open BPVs. pressure.

Reported to CRS that pressure is being

13. Inform CRS. manually controlled and pressure is 800 psig.

Stop Time: ____________ Total Time: ___________

Nebraska Public Power District SKL034-20-107(22658)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 04 Task No.: 249005P0401 Lowering DEH pressure setpoint (Alternate Path)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC IC-20 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None DEH Throttle

2. Malfunctions TC14A Press Signal 1 N/A -1 N/A N/A failure DEH Throttle TC14B Press Signal 1 N/A -1 N/A N/A failure DEH Throttle TC14C Press Signal 1 N/A -1 N/A N/A failure N/A N/A

3. Remotes None N/A N/A

4. Overrides None N/A N/A

a. Depress Manual Scram PBs and place Mode Switch in Refuel.

5. Panel Setup Allow plant to stabilize at 926 pressure set point and BPVs open to control pressure Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-20-107 (22658)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 04 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to lower reactor pressure to 800 psig at a rate of 10 psig /min using the hard card. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant has scrammed and pressure is approximately 930 psig.

Initiating Cue(s):

The CRS Orders you to lower Reactor Pressure to 800 Psig at a rate of 10 psig/min using the hard card.

Inform the CRS when the Reactor is at 800 psig.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 1 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

ALTERNATE PATH Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 20 minutes

5. NRC K/A 203000 A4.05 (4.3/4.1)

Directions to Examiner:

Note: This JPM is an Alternate Path JPM. RCIC flow controller fails in auto and manual control requiring test mode operation.

1. This JPM tests the trainees ability to place RCIC in operation; recognize auto and manual control fail and utilize the test mode.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 2 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Directions to Trainee:

When I tell you to begin, you are to perform a manual startup of RCIC per Procedure 2.2.67.1, so maintenance personnel can observe the turbine coupling.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. RCIC is in a standby lineup.

2. Suppression Pool temperature is 80°F.

General

References:

1. Procedure 2.2.67.1 General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical steps denoted in bold.

3. Simulator cues denoted by "#".

4. Alternate path denoted by.

Task Standards:

1. Recognize auto and manual controller failure and utilize the test mode to operate RCIC in the test flow path.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The CRS has directed you to manually start RCIC per Procedure 2.2.67.1 and place RCIC in the test flow path. Maintenance is present in the room to observe the turbine coupling. The NLO is in the area to assist as needed. Inform the CRS when RCIC is operating in the test flow path.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 3 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat Monitored SP temperatures on SPDS and PC-TR-24 and PC-TR-25.

1. Monitor SP temperature

CUE: Another operator is monitoring Suppression Pool temperatures.

At VBD-M ensured REC-MO-711 or REC-714 open.

2. REC Critical Loop supply CUE: REC-MO-711 control switch is RED Flagged.

RED light ON. GREEN light OFF.

Notified Shift Manager RCIC is inoperable and to enter TS LCO 3.5.3.

3. Notify SM

CUE: Shift Manager acknowledges.

At Panel 9-4, started Gland Seal Vacuum Pump.

4. Start GS Pump CUE: Gland Seal Vacuum Pump control switch pointing to START.

RED light ON. Green light OFF.

At Panel 9-4, opened RCIC-MO-132, TURB OIL COOLING WTR SUPPLY.

5. Open MO-132.

CUE: RCIC-MO-132 control switch pointing to OPEN.

RED light ON. GREEN light OFF.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 4 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Performance Checklist Standards Sat Unsat At Panel 9-4, ensured RCIC-FIC-91 in BAL.

6. FIC in BAL CUE: RCIC-FIC-91 knob pointing to BAL At Panel 9-4, opened RCIC-MO-33, ECST TEST LINE SHUTOFF VLV.

7. Open MO-33 CUE: RCIC-MO-33 control switch pointing to OPEN.

RED light ON. GREEN light OFF.

At Panel 9-4, opened RCIC-MO-131, STM SUPP TO TURB VLV and observed turbine started.

8. Open MO-131 CUE: RCIC-MO-131 pointing to OPEN.

RED light ON. GREEN light OFF.

RCIC-SI-3067 indicating idle speed.

Note to Examiner: Turbine speed rises to idle due to FIC-91 AUTO control failing low.

At Panel 9-4, verified turbine started and came up

9. Verify turbine comes to speed by observing RCIC-SI-3067.

up to speed.

CUE: RCIC-SI-3067 indicating idle speed.

At Panel 9-4, placed RCIC-FIC-91 to MAN.

10. Place FIC-91 in MAN.

CUE: RCIC-FIC-91 knob pointing to MAN.

At Panel 9-4, turned RCIC-FIC-91 MAN knob clockwise to raise turbine speed.

11. Observe turbine speed changes.

CUE: RCIC-FIC-91 MAN knob turning clockwise.

RCIC-SI-3067 indicating idle speed.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 5 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Performance Checklist Standards Sat Unsat Note to Examiner: If trainee questions continuing, as CRS direct trainee to continue and bring RCIC up to operating speed.

Reported FIC-91 controller failure to CRS.

12. Reports failure to CRS.

CUE: CRS acknowledges report.

At Panel 9-4, rotated TURB TEST SPEED CONTROL potentiometer fully counter-clockwise.

13. Adjust Test Pot CUE: .TURB TEST SPEED CONTROL potentiometer turns fully counter-clockwise.

At Panel 9-4, placed RCIC TURB SPEED TEST switch in ON.

14. Turb Speed Test in Test CUE: TURB SPEED TEST switch pointing to ON.

At Panel 9-4, placed TURB TEST POWER switch to ON.

15. Test Power on.

CUE: TURB TEST POWER switch pointing to ON.

At Panel 9-4, verified TURB TEST POWER light

16. Test Power light on.

on.

CUE: TURB TEST POWER light illuminated.

Note to Examiner: RCIC-MO-33 and RCIC-MO-131 may already be open if not reclosed above. If it is open, the following two steps will not be critical steps.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 6 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Performance Checklist Standards Sat Unsat At Panel 9-4, opened RCIC-MO-33, ECST TEST LINE SHUTOFF VLV.

17. Open MO-33 CUE: RCIC-MO-33 control switch pointing to OPEN.

RED light ON. GREEN light OFF.

At Panel 9-4, opened RCIC-MO-131, STM SUPP TO TURB VLV and observed turbine started.

18. Open MO-131 CUE: RCIC-MO-131 pointing to OPEN.

RED light ON. GREEN light OFF.

RCIC-SI-3067 indicating 1000 RPM.

At Panel 9-4, turned TURB TEST SPEED CONTROL potentiometer clock-wise and

19. Raise speed raised speed above 2200 RPM.

CUE: RCIC-SI-3067 indicating 2500 RPM.

At Panel 9-4, opened RCIC-MO-30, TEST BYP TO ECST VLV.

20. Open MO-30 CUE: RCIC-MO-30 control switch pointing to OPEN.

RED light ON. GREEN light OFF.

At Panel 9-4, verified RCIC-MO-27, MIN FLOW BYP VLV closes.

21. Verify MO-27 closed.

CUE: RCIC-MO-27 GREEN light ON. RED light OFF.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 7 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Performance Checklist Standards Sat Unsat At Panel 9-4, rotate potentiometer clock-wise in small increment to maintain turbine speed

22. Raise turbine

> 2000 RPM and < 5000 RPM.

speed.

CUE: RCIC-SI-3067 indicating 4200 RPM.

At Panel 9-4, adjusted RCIC-MO-30 until pump discharge pressure within ~100 psig of RPV

23. Adjust discharge pressure.

pressure CUE: RCIC-PI-93 indicating 1050 psig.

At Panel 9-4, observed RCIC-AO-34 and RCIC-AO-35, STM LINE DR TO COND ISOL VLVs

24. AO-34 and AO-35 closed.

observed closed.

CUE: RCIC-AO-34 and 35 GREEN lights OFF and RED lights ON.

At Panel 0-4, observed RCIC-PI-96 indicating

25. Pump suction > 15 Hg Vac.

pressure observed CUE: RCIC-PI-96 indicating 11 psig.

At Panel 9-4, observed RCIC-PI-94 approximately

26. Turbine inlet steam equal to RPV pressure.

pressure observed CUE: RCIC-PI-94 indicating 980 psig.

At Panel 9-4, observed RCIC-PI-95 < 25 psig.

27. Turbine exhaust observed CUE: RCIC-PI-95 indicating 4.5 psig.

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 8 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

Performance Checklist Standards Sat Unsat At Panel 9-4, observed RCIC-SI-3067 > 2200

28. Turbine speed RPM.

observed CUE: RCIC-SI-3067 indicating 4200 RPM.

Contacted NLO to obtain RCIC-PI-3064, RCIC TURBINE GOVERNOR END GLAND SEAL

29. Gov end gland seal EXHAUSE PRESSURE GAUGE pressure exhaust pressure reading.

checked locally

CUE: NLO reports RCIC-PI-3064 indicating 1.5 Hgv.

Contacted NLO to obtain RCIC-PI-3065, RCIC TURBINE COUPLING END GLAND SEAL

30. Turbine coupling EXHAUST PRESSURE GAUGE pressure end gland seal reading.

exhaust pressure checked locally

CUE: NLO reports RCIC-PI-3065 indicating 1.5 Hgv.

Reported to CRS that RCIC was operating in a test flow utilizing the Test POT.

31. Inform CRS

CUE: CRS Acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-136(49059)

Cooper Nuclear Station Page 9 of 10 Job Performance Measure for Operations Revision 0 Task No.:

Title:

RCIC Operation in Test Mode (Alternate Path)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC Any IC above 35% power.

C. Run Batch File D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial RC04 RCIC Flow Controller failure A N/A 0 N/A

3. Remotes Number Title Trigger TD Value Ramp

4. Overrides Instrument Tag Trigger TD Value Ramp FIC-91 Manual Pot ZAIRCICFIC91(1) A N/A 0 N/A FIC-91 Dev Meter ZAORCIC91(2) A N/A -0.000399 N/A

a. Ensure RCIC in standby status.

5. Panel Setup Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-136 (49059)

Cooper Nuclear Station Page 10 of 10 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to perform a manual startup of RCIC per Procedure 2.2.67.1, so maintenance personnel can observe the turbine coupling.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. RCIC is in a standby lineup.

2. Suppression Pool temperature is 80°F.

Initiating Cue(s):

The CRS has directed you to manually start RCIC per Procedure 2.2.67.1 and place RCIC in the test flow path. Maintenance is present in the room to observe the turbine coupling. The NLO is in the area to assist as needed. Inform the CRS when RCIC is operating in the test flow path.

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 1 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

ALTERNATE PATH Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 10 minutes

5. NRC K/A 261000 A4.06 (3.3/3.6)

Directions to Examiner:

Note: This JPM is an Alternate Path JPM. The RX BLDG/SGT DP Controller fails to operate In automatic and must be placed in manual.

1. This JPM evaluates the trainees ability to restore Reactor Building differential pressure following a failure of RB HVAC. The trainee must manually start a SGT train per the guidance of procedure 2.2.73.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 2 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

Directions to Trainee:

When I tell you to begin, you are to restore Reactor Building differential pressure using SGT in accordance with 2.2.73. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

If being simulated In Control Room:

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in cold shutdown.

2. Reactor Building HVAC fans have been secured due to being unable to maintain Reactor Building differential pressure.

3. Abnormal Procedure 2.4HVAC Attachment 1 has been entered and actions up to Step 3 have been marked N/A.

General

References:

1. Procedure 2.4HVAC, Building Ventilation Abnormal

2. Procedure 2.2.73, Standby Gas Treatment System.

General Tools and Equipment:

1. None Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical steps denoted in bold.

3. Simulator cues denoted by "#".

4. Alternate path denoted by.

Task Standards:

1. The trainee places the controller in manual and establishes manual pressure control.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 3 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

Initiating Cue(s):

The Control Room Supervisor directs you to start SGT System per Procedure 2.2.73 to maintain Reactor Building pressure negative. Another operator is assigned to shutdown RB HVAC.

Inform the CRS when Reactor Building negative pressure is restored.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 4 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Place Placed HV-DPIC-546, RX BLDG/SGT DP HV-DPIC-546 in controller in AUTO; and adjusted set tape to AUTO with set tape 0.25"wg.

adjusted to

- 0.25"wg CUE: Controller in AUTO; set tape indicates

-25"wg.

Placed the control switch for EF-R-1E, SGT A

2. Start EF-R-1E, SGT EXHAUST FAN in RUN.

A EXHAUST FAN.

CUE: RUN red ON, Green light OFF.

Verified SGT-AO-249, SGT A INLET red indicating light is on and green closed

3. Check SGT-AO-light is off.

249, SGT A INLET opens.

CUE: SGT-AO-249 Red light ON, Green light OFF.

Verified SGT-AO-251, SGT A DISCHARGE red indicating light is on and green closed light is off.

4. Check SGT-AO-251, CUE: SGT-AO-251 Red light ON, Green light SGT A OFF.

DISCHARGE, opens.

CUE: If trainee asks if RBHVAC is to be isolated, direct him to continue in procedure that another operator is isolating RBHVAC.

Responded per Annunciator alarm procedure

5. Recognize and SGT A LOW FLOW K-1/D-2 and checked HV-respond to SGT A DPIC-546 for proper operation.

Low Flow Condition.

CUE: HV-DPIC-546, output signal full scale.

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 5 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

Performance Checklist Standards Sat Unsat Verified annunciator SGT HI D/P K-1/D-1 is not in

6. Verify filter DP not alarm.

HIGH.

CUE: SGT HI D/P is clear.

Observed indicating lights for SGT-DV-PC546A

7. Check SGT-DV-for red light on.

PC546A is not closed.

CUE: Green light ON, Red light OFF.

8. Check pressure on Observed HV-DPR-835B RX BLDG / SGT DP HV-DPR-835, RX failed to maintain -0.25" wg in AUTO.

BLDG /ATMOS DP is being maintained CUE: HV-DPR-835 RX BLDG / ATMOS DP, at - 0.25"wg. Indicates 0.00" wg.

Placed HV-DPIC-546, RX BLDG /SGT DP in

9. Establish manual MANUAL and established negative pressure pressure control using the manual control knob.

with HV-DPIC-546, RX BLDG /SGT DP CUE: When controller is in manual and controller. adjusted, indicates mid scale and RX BLDG DP lowering.

10. Report to CRS that Reactor Building CRS informed that negative RB pressure is Pressure has been restored.

restored negative with SGT with #CUE: CRS acknowledges report.

manual control.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 6 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC Any IC with Rx Bldg HVAC in service.

C. Run Batch File D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial

3. Remotes Number Title Trigger TD Value Ramp

4. Overrides Instrument Tag Trigger TD Value Ramp HV-DPIC-546 RX BLDG/SGT DP CNTLR - ZAIHV546SVAL A 0 0.5 0 SETPOINT

Nebraska Public Power District SKL034-21-59(8782)

Cooper Nuclear Station Page 7 of 8 Job Performance Measure for Operations Revision 02 Task No.: 288024C0401

Title:

Respond To Loss Reactor Building Ventilation Equipment (Alternate Path)

a. Trip RB HVAC fans.

b. Close RB HVAC valves.

c. Place Preferred Train tag on SGT A.

5. Panel Setup

d. Place Simulator in FREEZE.

e. Ensure HV-DPIC-546 set to V and 100.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-59 (8782)

Cooper Nuclear Station Page 8 of 8 Job Performance Measure for Operations Revision 02 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to restore Reactor Building differential pressure using SGT in accordance with 2.2.73. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

If being simulated In Control Room:

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in cold shutdown.

2. Reactor Building HVAC fans have been secured due to being unable to maintain Reactor Building differential pressure.

3. Abnormal Procedure 2.4HVAC Attachment 1 has been entered and actions up to Step 3 have been marked N/A.

Initiating Cues:

The Control Room Supervisor directs you to start SGT System per Procedure 2.2.73 to maintain Reactor Building pressure negative. Another operator is assigned to shutdown RB HVAC.

Inform the CRS when Reactor Building negative pressure is restored.

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 1 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 10 minutes

5. NRC K/A 215005 A1.07(3.0/3.4)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to perform APRM calibration with a valid Official Case available with two RR loops operating.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 2 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Directions to Trainee:

When I tell you to begin, you are to perform any corrective actions for the data given. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in normal operation with two Reactor Recirculation Pumps in service.

2. No main turbine bypass valves are open.

3. DEH is operating in sequential valve control.

General

References:

1. Procedure 10.1 General Tools and Equipment:

1. Small screwdriver

2. NPP 10.1, Attachment 1 and 3 Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

Task Standards:

1. Adjust APRM gain adjustments so APRM readings are within -2.0 APRM - %CTP 2.0.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 3 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Initiating Cue(s):

The Control Room Supervisor directs you to review the provided Official Case and complete any required actions. Inform the CRS when the task is complete and provide him with any completed paperwork.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 4 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Obtain copy of NPP Obtained a copy of NPP 10.1.

10.1 At Panel 9-5, ensured APRM recorders were energized, and APRM/IRM switches were in APRM,

2. Verify APRM Status CUE: All APRM/IRM switches pointing to APRM Record CTP and %CTP from Periodic Case.

3. Record CTP on Attachment 1 CUE: CTP and %CTP recorded.

Obtained main turbine 1st stage pressure from

4. Obtain MT 1st DEH HMI.

Stage Pressure CUE: 1st Stage Pressure = 680 psig.

Checked 1st stage pressure versus %CTP is

5. Compare %CTP to within UPPER and LOWER LIMIT lines on 10.1 1 st stage pressure Attachment 3.

6. Mark SAT Recorded SAT on Step 3 of Attachment 1.

Calculated and recorded APRMdesired on

7. %CTP recorded Attachment 1.

8. Record APRMdesired Determined and recorded APRMdesired reading reading on Att. 1.

At Panel 9-5, bypassed APRM B with the Manual

9. Bypass APRM Bypass Joystick.

Channel B CUE: APRM B bypassed.

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 5 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Performance Checklist Standards Sat Unsat At Panel 9-14, verified or placeed APRM B meter function switch in AVERAGE.

10. APRM B in AVERAGE CUE: Switch pointing to AVERAGE.

At Panel 9-14, adjusted the Gain Adjustment Potentiometer to obtain APRMdesired reading

11. Adjust gain to from Attachment 1.

obtain APRMdesired reading CUE: APRM B indicates APRMdesired reading.

At Panel 9-5, un-bypassed APRM B by placing the Manual Bypass Joystick in NEUTRAL

12. Unbypass APRM position.

Channel CUE: Switch is in NEUTRAL.

13. Repeat if necessary Repeated Gain adjustment as necessary.

Demanded Periodic Case or Heat Balance.

14. Generate Periodic Case.

CUE: Periodic Case obtained from typer.

Recorded (APRM - %CTP) from Periodic Case on Attachment 1.

15. Record (APRM -

%CTP)

CUE: APRM Final recorded.

Ensured APRM-%CTP within -2.0 APRM -

16. Ensure APRM- %CTP 2.0.

%CTP within limits.

CUE: Attachment 1 completed.

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 6 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

Performance Checklist Standards Sat Unsat

17. Attach Periodic Case to Attachment Attached Periodic Case to Attachment 1.

1.

Signed Performed By and entered Time/Date on

18. Sign form Attachment 1.

Forward Periodic Case and Attachment 1 to CRS.

19. Provide paperwork to CRS.

CUE: The CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-20-18(2120)

Cooper Nuclear Station Page 7 of 9 Job Performance Measure for Operations Revision 11 Task No.: 215038P0101

Title:

Perform APRM Gain Adjustment (With A Valid CTP Available With Two RR Loops Operating)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC Any IC at full power.

C. Run Batch File D. Change the simulator conditions as follows:

1. Triggers Number File Name Description

2. Malfunctions Number Title Trigger TD Severity Ramp Initial

3. Remotes Number Title Trigger TD Value Ramp 0

4. Overrides Instrument Tag Trigger TD Value Ramp None

a. Place Simulator in RUN.

b. Adjust APRM B gain to lower 3-5%.

5. Panel Setup c. Ensure 2419 MWTh.

d. Ensure Official Case indicates only B APRM requires adjustment.

e. Place Simulator in FREEZE.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-20-18 (2120)

Cooper Nuclear Station Page 8 of 9 Job Performance Measure for Operations Revision 11 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to perform any corrective actions for the data given. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. The plant is in normal operation with two Reactor Recirculation Pumps in service.

2. No main turbine bypass valves are open.

3. DEH is operating in sequential valve control.

Initiating Cue(s):

The Control Room Supervisor directs you to review the provided Official Case and complete any required actions. Inform the CRS when the task is complete and provide him with any completed paperwork.

Nebraska Public Power District SKL034-20-18 (2120)

Cooper Nuclear Station Page 9 of 9 Job Performance Measure for Operations Revision 11 ATTACHMENT 2

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 1 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

ALTERNATE PATH Additional Program Information:

1. Appropriate Performance Locations: CR / SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate Perform

4. Performance Time: 15 minutes

5. NRC K/A 201003 A2.01 (3.4/3.6)

Directions to Examiner:

Note: This JPM is an Alternate Path JPM. Control Rod 26-11 does not withdraw from Position 00 until alternate methods are used.

1. This JPM evaluates the trainees ability to withdraw a control rod that is temporarily stuck at position 00.

2. If this JPM is performed on the Simulator, only the cues preceded by "#" should be given.

3. Observe the trainee during performance of the JPM for proper use of self-checking methods.

4. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

5. Give the trainee Attachment 2.

6. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 2 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

Directions to Trainee:

When I tell you to begin, you are to withdraw control rod 26-11 to position 02 per Procedure 2.2.8.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. A plant startup is in progress.

2. The normal attempts to withdraw control rod 26-11 from position 00 have failed.

General

References:

1. Procedure 2.2.8 General Tools and Equipment:

1. Key 19 Special Conditions, References, Tools, Equipment:

1. Simulator Setup: See Attachment 1.

2. Critical checks denoted in bold.

3. Simulator cues denoted by "#".

4. Alternate path denoted by.

Task Standards:

1. The control rod is withdrawn after using the INDIVIDUAL ROD SCRAM switch.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The CRS directs you to withdraw control rod 26-11 to position 02 per procedure 2.2.8. Inform the CRS when the control rod has been withdrawn to position 02.

NOTE: Ensure the Simulator is in RUN and tell the trainee to begin.

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 3 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat Placed ROD MOVEMENT CONTROL switch to OUT NOTCH and HOLD.

1. Hold Rod Movement control switch CUE: ROD MOVEMENT CONTROL switch pointing to OUT NOTCH position.

Placed EMERGENCY NOTCH OVERRIDE switch to EMER ROD IN and released.

2. Emergency Notch CUE: EMERGENCY NOTCH OVERRIDE Override to Emer switch pointing to EMER ROD IN and Rod In then OFF when released.

Control rod position indicating 00.

Note to Examiner: The previous step is repeated several times.

Released ROD MOVEMENT CONTROL.

3. Release Rod Movement switch. CUE: ROD MOVEMENT CONTROL switch pointing to OFF position Placed EMERGENCY NOTCH OVERRIDE switch

4. Emergency Notch to EMER ROD IN and held for several seconds.

Override to Emer Rod In CUE: EMERGENCY NOTCH OVERRIDE switch pointing to EMER ROD IN.

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 4 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

Performance Checklist Standards Sat Unsat Simultaneously released EMERGENCY NOTCH OVERRIDE switch and placed ROD MOVEMENT CONTROL switch to NOTCH OUT.

5. Release Emergency Notch Override CUE: EMERGENCY NOTCH OVERRIDE switch switch and place pointing to OFF and ROD MOVEMENT Rod Movement CONTROL switch POINTING to NOTCH switch to Out Notch OUT Control rod position indicating 00.

Note to Examiner: The previous TWO steps are repeated several times.

Inquired if double-notching allowed.

6. Double Notching not allowed. #CUE: Inform trainee double-notching of the control rod cannot be tolerated.

Obtained Key 19 and opened INDIVIDUAL ROD

7. Open Individual SCRAM switches panel.

Rod Scram Switch panel CUE: Key inserted in lock and panel door open.

Note to Examiner: The individual rod scram unsticks the control rod.

At Panel 9-16, placed INDIVIDUAL ROD SCRAM switch DOWN for 2 to 3 minutes.

8. Scram control rod CUE: INDIVIDUAL ROD SCRAM switch pointing DOWN.

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 5 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

Performance Checklist Standards Sat Unsat At Panel 9-16, placed INDIVIDUAL ROD SCRAM switch up.

9. Reset individual rod scram CUE: INDIVIDUAL ROD SCRAM switch pointing UP.

Placed ROD MOVEMENT CONTROL switch to OUT NOTCH and released.

10. Withdraw control rod using normal CUE: ROD MOVEMENT CONTROL switch to method OUT NOTCH and returned to OFF.

Control rod position indicating 02.

Informed CRS control rod 26-11 had been withdrawn to position 02.

11. Report to CRS

CUE: CRS acknowledges.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-21-138(49061)

Cooper Nuclear Station Page 6 of 7 Job Performance Measure for Operations Revision 0 Task No.: 200025O0401

Title:

Withdrawal Of Control Rod From Position 00 (Alternate Path)

ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC IC-16 C. Run Batch File JPM/3421XX D. Change the simulator conditions as follows:

1. Triggers Number File Name Description E1 None Trgset 1

2. Malfunctions Number Title Trigger TD Severity Ramp Initial RD12 Control Rod 26-11 stuck A 0 N/A N/A N/A

3. Remotes Number Title Trigger TD Value Ramp 0

4. Overrides Instrument Tag Trigger TD Value Ramp None

a. Ensure correct Control Rod Sequence book at 9-5.

b. Ensure Control Rod Sequence book marked indicating current

5. Panel Setup control rod movement attempt.

c. Select control rod 26-11.

Note: If this JPM is to be performed more than once, take a SNAPSHOT after the panel setup is complete.

Nebraska Public Power District SKL034-21-138 (49061)

Cooper Nuclear Station Page 7 of 7 Job Performance Measure for Operations Revision 0 ATTACHMENT 2 Directions to Trainee:

When I tell you to begin, you are to withdraw control rod 26-11 to position 02 per Procedure 2.2.8.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

General Conditions:

1. A plant startup is in progress.

2. The normal attempts to withdraw control rod 26-11 from position 00 have failed.

Initiating Cue(s):

The CRS directs you to withdraw control rod 26-11 to position 02 per procedure 2.2.8. Inform the CRS when the control rod has been withdrawn to position 02.

Nebraska Public Power District SKL034-10-95(7456)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 06 Task No.: 200134I0504

Title:

Respond to No Break Power Panel Failure (Control Bldg. Actions) (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

ALTERNATE PATH Additional Program Information:

1. Appropriate Performance Locations: Plant

2. Appropriate Trainee level: SO / RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 25 minutes

5. K/A: 262002 K4.01(3.4/3.4)

Note: The alternate source supplying load will not work and the manual bypass switch must be used.

Directions to Examiner:

1. This JPM evaluates the trainee's ability to respond to a no-break power panel failure.

2. Observe the trainee during performance of the JPM for proper use of self-checking methods.

3. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

4. Brief the trainee and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-10-95(7456)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 06 Task No.: 200134I0504

Title:

Respond to No Break Power Panel Failure (Control Bldg. Actions) (Alternate Path)

Directions to Trainee:

When I tell you to begin you are to perform the assigned Control Building Operator=s actions to respond to a no break power panel failure. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

General Conditions:

1. The plant has experienced a no break power panel failure.

2. The Rx has scrammed.

3. An attempt to transfer NBPP to MCC-R by placing NBPP PWR TRANSFER switch to MCC-R has failed.

4. The electricians have determined that the fault is not in the NBPP itself.

5. Both switches have been placed to ALT at the 120v AC supplies for the Gaitronics.

6. Control Room Operators have performed actions 4.1 through 4.4.11 of 5.3NBPP.

General

References:

1. 5.3NBPP (NO BREAK POWER FAILURE)

General Tools and Equipment:

1. Key for access to Control Building 903' doors.

Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

2. Alternate path steps are denoted by Task Standards:

1. Re-energize NBPP using the Manual Bypass Switch.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Nebraska Public Power District SKL034-10-95(7456)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 06 Task No.: 200134I0504

Title:

Respond to No Break Power Panel Failure (Control Bldg. Actions) (Alternate Path)

Initiating Cue(s):

You have been assigned to carry out actions for a no break power panel failure in accordance with 5.3NBPP. The Control Room Supervisor (CRS) directs you to perform all the necessary Control Building Operator=s actions for restoring power from the Alternate Supply (MCC-R).

Notify the CRS when the NBPP is energized.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-10-95(7456)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 06 Task No.: 200134I0504

Title:

Respond to No Break Power Panel Failure (Control Bldg. Actions) (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat

1. Obtain copy of Obtained a copy of 5.3NBPP, NO BREAK procedure. POWER FAILURE Ensured EE-DSC-NBPP(AC), MCC-R FEED TO

2. Ensure MCC-R NBPP TRANSFMR (Cable Spreading near FEED TO NBPP is NBPP), is in ON.

in ON.

CUE: The handle is in the UP position.

Contacted the Reactor Building Operator to ensure that MCC-R, Breaker 2B, NO BREAK AC POWER SUPPLY, is closed and reset.

3. Contact the Rx Building Operator. CUE: Respond as the Rx Bldg Operator, that MCC-R, Breaker 2B, NO BREAK AC POWER SUPPLY, is closed and reset.

At Inverter A, opened INVERTER OUTPUT breaker (125/250 A Switchgear Room).

4. Open breaker.

CUE: The handle is in the DOWN position.

At Inverter A, ensured SUPPLY TO NBPP

5. Ensure breaker is breaker is closed.

closed.

CUE: The handle is in the UP position.

At Inverter A, ensured ALTERNATE AC INPUT

6. Ensure breaker is TO STATIC SWITCH breaker is closed.

closed.

CUE: The handle is in the UP position.

Nebraska Public Power District SKL034-10-95(7456)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 06 Task No.: 200134I0504

Title:

Respond to No Break Power Panel Failure (Control Bldg. Actions) (Alternate Path)

Performance Checklist Standards Sat Unsat At Inverter A, depressed ALTERNATE SOURCE SUPPLYING LOAD button.

7. Depress button. CUE: The button is depressed and the red ALT SOURCE SUPPLYING LOAD light is OFF.

(NBPP is still de-energized)

Placed MANUAL BYPASS SWITCH to

8. Place MAN BP ALTERNATE SOURCE TO LOAD.

SW to ALT SOURCE TO CUE: Point to ALTERNATE SOURCE TO LOAD. LOAD position.

Depressed ALTERNATE SOURCE SUPPLYING

9. Depress ALT LOAD button.

SOURCE SUPPLYING LOAD CUE: The red ALT SOURCE SUPPLYING button. LOAD light is ON. (NBPP is energized)

Note to Examiner: The Gaitronics do not need to be transferred Notified CRS that NBPP was energized.

10. Notifies CRS.

CUE: The CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-10-95 (7456)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 06 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin you are to perform the assigned Control Building Operator=s actions to respond to a no break power panel failure. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

General Conditions:

1. The plant has experienced a no break power panel failure.

2. The Rx has scrammed.

3. An attempt to transfer NBPP to MCC-R by placing NBPP PWR TRANSFER switch to MCC-R has failed.

4. The electricians have determined that the fault is not in the NBPP itself.

5. Both switches have been placed to ALT at the 120v AC supplies for the Gaitronics.

6. Control Room Operators have performed actions 4.1 through 4.4.11 of 5.3NBPP.

Initiating Cues:

You have been assigned to carry out actions for a no break power panel failure in accordance with 5.3NBPP. The Control Room Supervisor (CRS) directs you to perform all the necessary Control Building Operator=s actions for restoring power from the Alternate Supply (MCC-R).

Notify the CRS when the NBPP is energized.

Nebraska Public Power District SKL034-40-84(18851)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 04 Task No.: 200390A0501

Title:

5.3ALT-STRATEGY, Inject Fire Protection Water to RHR Trainee: Examiner:

Pass Fail Examiner Signature: Date:

Additional Program Information:

1. Appropriate Performance Locations: PLANT

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 15 minutes

5. NRC K/A 295003.AA1.03 (4.4 / 4.4)

Directions to Examiner:

1. This JPM evaluates the trainee's ability to inject fire protection water into the RHR System during a complete loss of AC and DC power.

2. Observe the trainee during performance of the JPM for proper use of self-checking methods.

3. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

4. Give the trainee Attachment 1.

5. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-40-84(18851)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 04 Task No.: 200390A0501

Title:

5.3ALT-STRATEGY, Inject Fire Protection Water to RHR Directions to Trainee:

When I tell you to begin, you are to perform the actions to align Fire Protection to RHR Subsystem A, using SWBP A in accordance with Procedure 5.3ALT-STRATEGY. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

ALL PLANT ANNOUNCEMENTS WILL BE SIMULATED!

General Conditions:

1. The Plant has experience a total loss of A/C and D/C power.

2. Reactor Pressure is 600 psig.

3. Procedure 5.3ALT-STRATEGY, Alternate Core Cooling Mitigating Strategies, has been entered.

4. Another NLO has completed the breaker alignments per step 1.1.1 Attachment 1 of 5.3ALT-STRATEGY.

5. Fire Protection header is intact.

General

References:

1. Procedure 5.3ALT-STRATEGY General Tools and Equipment:

1. 3 Fire Hose

2. Fire hose Wrench Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Nebraska Public Power District SKL034-40-84(18851)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 04 Task No.: 200390A0501

Title:

5.3ALT-STRATEGY, Inject Fire Protection Water to RHR Task Standards:

1. Inject Fire Protection water into RHR Subsystem A utilizing SWBP A.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

Initiating Cue(s):

The CRS has directed you to align Fire Protection Water to RHR Subsystem A using SWBP A in accordance with 5.3ALT-STRATEGY, Attachment 1. You are to inform the CRS when Fire Protection Water is aligned to RHR.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-40-84(18851)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 04 Task No.: 200390A0501

Title:

5.3ALT-STRATEGY, Inject Fire Protection Water to RHR Start Time: ____________

Performance Checklist Standards Sat Unsat Obtained a copy of the current revision of

1. Obtain procedure procedure 5.3ALT-STRATEGY Note to Examiner - Equipment is stored C-882-N by MCC-T.

Note to Examiner - The Operator must show that they are capable of obtaining the required equipment, but does not have to take them to the specified locations, to get credit for the task element.

Obtained the 3 hose.

2. C-882 N, gets hose.

CUE: Hose is obtained.

Removed the 21/2 hose at Hose Station No. 16

3. Remove existing 21/2 hose CUE: The hose is removed.

Attached the 3 hose to FP-315, Hose Station

4. C-882 N, attaches No. 16.

hose.

CUE: The hose is attached.

Removed the pipe cap from SW-641, Gland Water System Supply from SWB Pump A.

5. C-882 N, removes (Outlet of SWBP A Pump Inlet Strainer) pipe cap.

CUE: The pipe cap is removed.

Attached the 3 hose to SW-641 connection.

6. C-882 N, attaches hose.

CUE: The hose is attached.

Ensured SW-77, SWBP A Suction valve is closed.

7. C-882 N, ensures SW-77 is CLOSED. CUE: The hand wheel has stopped moving in the clockwise direction.

Nebraska Public Power District SKL034-40-84(18851)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 04 Task No.: 200390A0501

Title:

5.3ALT-STRATEGY, Inject Fire Protection Water to RHR Performance Checklist Standards Sat Unsat Ensured SW-121, Emergency Core Flooding Tell-Tale Drain valve is CLOSED. (SW-121 is in a contaminated area. He can call for RP

8. C-882 N, ensures support or obtain PCs, or since this is an the SW-121 is Emergency Procedure, he can complete the CLOSED. task and then ask for RP support.)

CUE: The hand wheel has stopped moving in the clockwise direction.

Ensured SW-120, Emergency Core Flooding Supply Shutoff valve is OPEN.

9. C-882 N, ensures valve is OPEN. CUE: The hand wheel has stopped moving in the counter-clockwise direction.

Ensured SW-118 Emergency Core Flooding Supply Root valve is OPEN.

10. C-882 N, ensures valve is OPEN. CUE: The hand wheel has stopped moving in the counter-clockwise direction.

Ensured SW-641, Gland Water System Supply from SWBP A is OPEN.

11. C-882 N, ensures CUE: The valve Operator has been turned valve is OPEN.

Counter Clockwise now is pointing straight up.

Slowly opened FP-315.

12. Slowly opens FP-CUE: The hand wheel has stopped moving in 315.

the counter-clockwise direction and the hose is charged.

Informed the CRS that Fire Water is aligned to the

13. Notifies CRS that RHR System via SWBP A.

task is completed.

CUE: The CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-40-84 (18851)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 0 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin, you are to perform the actions to align Fire Protection to RHR Subsystem A, using SWBP A in accordance with Procedure 5.3ALT-STRATEGY. Before you start, I will state the general plant conditions, the Initiating Cues and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

During task performance, state the actions you are taking, e.g.: repositioning controls and observing instrumentation.

ALL PLANT ANNOUNCEMENTS WILL BE SIMULATED!

General Conditions:

1. The Plant has experience a total loss of A/C and D/C power.

2. Reactor Pressure is 600 psig.

3. Procedure 5.3ALT-STRATEGY, Alternate Core Cooling Mitigating Strategies, has been entered.

4. Another NLO has completed the breaker alignments per step 1.1.1 Attachment 1 of 5.3ALT-STRATEGY.

5. Fire Protection header is intact.

Initiating Cue(s):

The CRS has directed you to align Fire Protection Water to RHR Subsystem A using SWBP A in accordance with 5.3ALT-STRATEGY, Attachment 1. You are to inform the CRS when Fire Protection Water is aligned to RHR.

Nebraska Public Power District SKL034-11-03(8792)

Cooper Nuclear Station Page 1 of 6 Job Performance Measure for Operations Revision 01 Task No.: 20100200104

Title:

Placing Cold Reference Leg Continuous Backfill System in Service (Alternate Path)

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

ALTERNATE PATH Additional Program Information:

1. Appropriate Performance Locations: PLANT

2. Appropriate Trainee level: NLO / RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 15 minutes

5. NRC K/As 2.1.29 (3.4/3.3), 2.1.30 (3.9/3.4)

Directions to Examiner:

Note: This JPM is an Alternate Path JPM. The flow cannot be adjusted within required values without coarse adjustment.

1. This JPM evaluates the trainees ability to ability to place the cold reference leg continuous backfill system in service per Procedure 4.6.1.

2. Observe the trainee during performance of the JPM for proper use of self-checking methods.

3. Check off either satisfactory or unsatisfactory performance. If Unsat state why in the notes section below.

4. Give the trainee Attachment 1.

5. Brief the trainee, place the simulator in run, and tell the trainee to begin.

Notes: ______________________________________________________________________

Nebraska Public Power District SKL034-11-03(8792)

Cooper Nuclear Station Page 2 of 6 Job Performance Measure for Operations Revision 01 Task No.: 20100200104

Title:

Placing Cold Reference Leg Continuous Backfill System in Service (Alternate Path)

Directions to Trainee:

When I tell you to begin, you are to place the cold reference leg continuous backfill system in service. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

General Conditions:

1. The plant has entered Procedure 2.1.1, Startup Procedure.

2. You have been provided a drain hose and wrench.

General

References:

1. Instrumentation Operating Procedure 4.6.1 General Tools and Equipment:

1. 1/32" hex key

2. Drain hose

3. Wrench Special Conditions, References, Tools, Equipment:

1. Critical checks denoted in bold.

Task Standards:

1. The flow rate is adjusted utilizing the course adjust.

2. 100% of critical elements successfully completed without error.

3. 100% of safety and radiological work practices.

4. Alternate path denoted by.

Initiating Cue(s):

The Control Room Supervisor directs you to place reference leg 3A continuous backfill system in service per Procedure 4.6.1, Reactor Vessel Water Level Indication. Notify the CRS when the task is complete.

NOTE: Tell the trainee to begin.

Nebraska Public Power District SKL034-11-03(8792)

Cooper Nuclear Station Page 3 of 6 Job Performance Measure for Operations Revision 01 Task No.: 20100200104

Title:

Placing Cold Reference Leg Continuous Backfill System in Service (Alternate Path)

Start Time: ____________

Performance Checklist Standards Sat Unsat At IR 25-51, used wrench and removed the pipe cap from the drain on NBI-636. (R-903-

1. Remove pipe cap NW) from drain on NBI-636 CUE: The cap rotates counter-clockwise (when viewed from below). The cap is removed.

At IR 25-51, attached tubing to NBI-636, CV-55CV OUTBOARD TEST VALVE, and directed

2. Attach tubing to tubing to nearest equipment drain. (R-903-NW).

drain on NBI-636 CUE: Tubing is attached.

At IR 25-51, slowly opened NBI-636 while ensuring tube did not displace from equipment drain.

3. Open NBI-636 CUE: Air and water are coming out of the end of the tube and tube is not displaced from the drain.

On Rack 119, slowly throttled vernier handle on NBI-FCV10, REF LEG 3A MAN FLOW CONT

4. Throttle vernier VLV (R-903-SE).

handle on NBI-FCV10 CUE: Air free water flow is coming out of the end of the tube.

At IR 25-51, closed NBI-636 and removed tubing.

5. Close NBI-636 and remove tubing. CUE: The valve handle rotated clockwise and will not turn any more. The tubing has been removed.

Nebraska Public Power District SKL034-11-03(8792)

Cooper Nuclear Station Page 4 of 6 Job Performance Measure for Operations Revision 01 Task No.: 20100200104

Title:

Placing Cold Reference Leg Continuous Backfill System in Service (Alternate Path)

Performance Checklist Standards Sat Unsat Slowly throttled close vernier handle on NBI-

6. Throttle close FCV10 until flow on NBI-FI-10 is zero.

vernier handle on NBI-FCV10 CUE: NBI-FI-10 is indicating zero.

Ensured NBI-FI-10 display indicates RATE mode.

7. Ensure NBI-FI-10 Depress 3" key as necessary.

Display indicates rate mode.

CUE: NBI-FI-10 display indicates RATE mode.

At IR25-51, slowly opened NBI-620, CONTINUOUS REF LEG FILL ROOT (R-903-NW).

8. Open NBI-620 CUE: The valve handle rotates counterclockwise. The valve handle will not move any more.

Note to Examiner: Irrespective of the operators attempts to adjust flow in the next step, flow will not rise above the value provided and the valve will reach its stop and not move Slowly throttled vernier handle on NBI-FCV10

9. Throttle vernier until flow on NBI-FI-10 is 0.0095 to 0.0105 gpm.

handle on NBI-FCV10 CUE: Flow on NBI-FI-10 is 0.0080 gpm Established communications with Control Room

10. Establish to provide monitoring of RPV level indication while communication to adjusting NBI-FCV10.

monitor RPV level indication CUE: Communications has been established with Control Room.

Adjusted vernier handle on NBI-FCV10 to middle

11. Adjust vernier of operating range.

handle on NBI-FCV10 CUE: The middle of the operating range is indicated.

Nebraska Public Power District SKL034-11-03(8792)

Cooper Nuclear Station Page 5 of 6 Job Performance Measure for Operations Revision 01 Task No.: 20100200104

Title:

Placing Cold Reference Leg Continuous Backfill System in Service (Alternate Path)

Performance Checklist Standards Sat Unsat Obtained a 1/32" hex key.

12. Obtain 1/32" kex key.

CUE: You have a 1/32" kex key.

Note to Examiner: In the next step, the set screw is on the side of the knurled knob on the bottom of the valve.

Loosened top setscrew on coarse adjust knob

13. Loosen top using 1/32" hex key.

setscrew on coarse adjust knob CUE: Top setscrew rotates counter-clockwise and is loose.

Adjusted coarse adjust in 0.0010 gpm flow increments to establish 0.0100 gpm flow on

14. Adjust flow to NBI-FI-10.

0.0100 gpm CUE: 0.0100 gpm flow is indicated.

Tightened the setscrew.

15. Tighten the setscrew CUE: Setscrew is tight.

Adjusted vernier handle to establish middle of operating range or 0.0095 to 0.0105 gpm.

16. Adjust vernier handle CUE: 0.0100 gpm is indicated and vernier handle is in the middle of its operating band.

Notified CRS Reference Leg 3A Continuous

17. Inform the CRS of Backfill System had been placed in service.

completion.

CUE: The CRS acknowledges the report.

Stop Time: __________ Total Time: ___________

Nebraska Public Power District SKL034-11-03 (8792)

Cooper Nuclear Station Page 6 of 6 Job Performance Measure for Operations Revision 01 ATTACHMENT 1 Directions to Trainee:

When I tell you to begin, you are to place the cold reference leg continuous backfill system in service. Before you start, I will state the general plant conditions, the Initiating Cues, and answer any questions you may have.

When simulating, physically point to any meters, gauges, recorders and controls you would be using. State the position of controls as you would have manipulated them in order to complete the assigned task.

General Conditions:

1. The plant has entered Procedure 2.1.1, Startup Procedure.

2. You have been provided a drain hose and wrench.

Initiating Cue(s):

The Control Room Supervisor directs you to place reference leg 3A continuous backfill system in service per Procedure 4.6.1, Reactor Vessel Water Level Indication. Notify the CRS when the task is complete.

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: ___1____

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 70% Power (EOL), DEH Pump B tagged out for maintenance on motor.

Turnover: Raise reactor power 30 MWe net with RR and then perform 6.RCIC.201, RCIC POWER OPERATED VALVE OPERABILITY TEST (IST)

Event Malf. No. Event Event No. Type* Description 1 N/A R (ATC) Raise reactor power with RR.

2 N/A N (BOP) 6.RCIC.201, RCIC POWER OPERATED VALVE OPERABILITY TEST (IST). RCIC is inoperable during testing (LCO 3.5.3)

T (CRS) 3 Override T (CRS) RCIC-MO-16 fails to close during stroke timing.

C/S CRS declares valve inoperable per TS LCO 3.6.1.3.

4 Override C (ATC) CRD-FC-301, CRD flow controller automatic setpoint fails downscale. Manual control required. 2.4CRD entry.

ZAICRDF C301(2) 5 MS02a C (BOP Small steam leak in PC-Abnormal 2.4PC (within vent capability).

T (CRS) CRS enters TS LCO 3.6.1.4 (0.75 psig) 6 MS02a M(ALL) Steam leak propagates > vent requiring manual scram per 2.4PC.

Overrides C, © DW pressure > 1.5 psig.

ZRORPS RELK4B/ ATC 4D 7 RD15 C (ATC) Control rod 14-23 fails to insert on the scram.

8 MC01 C (ALL) Main condenser loss of vacuum on air in leakage. (2.4VAC).

9 N/A (ALL) EOP 1A and 3A entry on high DW pressure.

10 HP12 C (BOP) HPCI oil line fails and HPCI will not start or inject.

11 RD08b C (ATC) CRD Pump trips.

12 N/A © Emergency Depressurize when drywell temperature cannot be (BOP) maintained below 280°F.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, M)ajor, (T)echnical Specifications, (©) Critical Task Rev 1

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: The reactor is operating at 70% power at the end of the current fuel cycle.

One of the DEH Pumps is tagged out for maintenance.

Turnover: Raise reactor power with the reactor recirculation pumps and then perform a RCIC valve operability surveillance test.

The ATC will raise reactor power with reactor recirculation to receive credit for a reactivity manipulation.

A normal activity of performing a valve operability surveillance on RCIC is performed. CRS declares RCIC inoperable for testing (TS LCO 3.5.3). During valve stroking of a RCIC steam supply isolation valve, the valve will not close requiring the CRS to declare the valve inoperable.

The valve is a primary containment isolation valve. The penetration must be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . (TS LCO 3.6.1.3)

CRD flow controller CRD-FC-301 automatic controller output fails downscale. CRD cooling water lowers and charging water pressure rises. CRS enters 2.4CRD. ATC takes manual control of the flow controller and returns system flows and pressures to normal.

A Main Steam Line leak inside primary containment begins requiring entry into Abnormal Procedure 2.4PC. The primary containment is vented with the Standby Gas Treatment System.

The venting can keep up with the leak at this severity. The CRS enters TS LCO 3.6.1.4 when drywell pressure is above 0.75 psig. After venting is ongoing, the leak will propagate requiring a manual reactor scram (critical task) in accordance with 2.4PC (before drywell pressure reaches 1.5 psig). Note: The high drywell pressure RPS relays that feed B1 and B2 RPS scram logic are frozen in an energized state and will not de-energize and trip RPS B logic should drywell pressure rise to the level of the RPS trip (1.84 psig).

One control rod will fail to insert on the scram but the reactor is considered shutdown because the shutdown margin is met. The ATC may insert the control rod per Procedure 2.1.5 if directed by the CRS.

Drywell pressure rises to the ECCS initiation level of 1.84 psig requiring entry into EOP 1A for RPV pressure and level control and EOP 3A of primary containment control.

The HPCI oil piping fails and HPCI will not start so HPCI is not available for injection. BOP utilizes Feedwater or RCIC as alternate level control system.

A main condenser air in-leakage causes condenser vacuum to be lost and the Reactor Feed Pump Turbines and Main Steam bypass valves close. The CRS enters abnormal procedure 2.4VAC in response to the loss of vacuum.

RCIC is used for RPV level control if its steam line was not isolated earlier. If RCIC was isolated earlier, the CRD system can be set up in maximum RPV injection lineup.

Rev 1

Appendix D Scenario Outline Form ES-D-1 RPV pressure control is transferred to SRVs when the Main Steam Lines close on loss of condenser vacuum or the operators manually close the MSIVs. The BOP controls RPV pressure in the correct band and initiates suppression pool cooling with RHR to cool the suppression pool. RHR is used to spray the torus air space and drywell due to high drywell temperature.

The operating CRD pump trips and the applicable alarm procedure directs starting the standby pump.

The steam leak rate reaches the point where drywell temperature cannot be restored and maintained below 280°F and EOP 3A requires Emergency Depressurization.

The scenario ends when RPV level is being controlled in a band set by the CRS.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 2 Event

Description:

Surveillance 6.RCIC.201, RCIC POWER OPERATED VALVE OPERABILITY TEST (IST).

Time Position Applicants Action or Behavior BOP Inform Shift Manager that RCIC System is inoperable for testing.

CRS reviews TS LCO 3.5.3 and declares RCIC inoperable:

COMPLETION CONDITION REQUIRED ACTION TIME A. RCIC System inoperable A.1 Verify by administrative 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months means High Pressure Coolant CRS Injection System is OPERABLE.

AND A.2 Restore RCIC System to OPERABLE status.

14 days END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 3 Event

Description:

RCIC-MO-16 fails to close during stroke timing.

Time Position Applicants Action or Behavior NOTE to Examiner: RCIC-MO-16 failure is already active.

Press AUTO ISOL SIGNAL 13A-S16A button.

Press AUTO ISOL SIGNAL 13A-S16B button.

AC (Time) Close RCIC-MO-16, OUTBD STM SUPP ISOL VLV.

Check Annunciator 9-4-1/G-2, RCIC-MO-15/16 NOT FULL OPEN, BOP alarms.

Check Alarm CRT displays 1691 RCIC-MO-16 NOT FULL OPEN ALARM.

Check PMIS indicates N505 RCIC CONTAINMENT ISO VA -

OUTBD CLSD.

BOP Report to CRS failure of RCIC-MO-16 to close.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 3 Event

Description:

RCIC-MO-16 fails to close during stroke timing.

Time Position Applicants Action or Behavior Reviews TS LCO 3.6.1.3 (Primary Containment Isolation Valves).

COMPLETION CONDITION REQUIRED ACTION TIME A. ------NOTE------- A.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except penetration flow path by for main steam Only applicable to use of at least one line penetration flow closed and de-activated paths with two AND automatic valve, closed PCIVs.

manual valve, blind 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for

flange, or check valve main steam line with flow through the One or more valve secured.

penetration flow paths with one AND PCIV inoperable A.2 ---------NOTES------------

except for MSIV leakage not within 1. Isolation devices in limit. high radiation areas may be verified by use of CRS administrative means.

2. Isolation devices that Once peer 31 are locked, sealed, or days for otherwise secured may isolated be verified by use of devices outside administrative means. primary containment AND Verify the affected penetration flow path is Prior to isolated. entering MODE 2 or 3 from MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 3 Event

Description:

RCIC-MO-16 fails to close during stroke timing.

Time Position Applicants Action or Behavior Booth As WCC SRO, when CRS calls on RCIC-MO-16 inform him you Operator will put together a repair team to investigate the valve.

CRS Direct halting 6.RCIC.201.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 4 Event

Description:

CRD-FC-301, CRD Flow Controller automatic signal fails low Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Override ZAICRDFC301(2)=0 to fail CRD-FC-301 output to zero flow (Trigger 4).

Operator Report CRD system flows and pressures abnormal. Respond to alarm 9-ATC 5-2/E-6, Charging Header High Pressure.

Enter Abnormal Procedure 2.4CRD, Direct ATC perform subsequent CRS operator actions of 2.4CRD.

Take responsibility for 2.4CRD scram action:

ATC If more than one rod is drifting, SCRAM and concurrently enter Procedure 2.1.5.

Enter 2.4CRD Attachment 5 for abnormal cooling water flows:

ATC Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 4 Event

Description:

CRD-FC-301, CRD Flow Controller automatic signal fails low Time Position Applicants Action or Behavior ATC Take manual control of FC-301 and adjust drive water flow to 45-50 gpm.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 5 Event

Description:

Small steam leak in PC (Within SGT venting capacity)

Time Position Applicants Action or Behavior Booth When directed by Lead Examiner Insert Trigger 3, Malfunction Operator MS02a at 0.0075% with a ramp time of 5 minutes (Trigger 5).

CREW Report drywell temperature and pressure rising.

Enter Abnormal Procedure 2.4PC. Assign critical parameters and direct CRS BOP perform 2.4PC actions.

NOTE to Examiners: CRS may select a critical parameter for manual scram at less than the procedure required 1.5 psig.

Responsible for Scram action per 2.4PC or as assigned by CRS.

RO If while performing this procedure, drywell pressure cannot be maintained 1.5 psig, SCRAM and enter Procedure 2.1.5.

Assigned 2.4PC Subsequent Actions:

Maintain drywell pressure between 0.25 psig and 0.45 psig by venting containment through SGT System per Procedure 2.2.60 HARD CARD.

1.1 Ensure PC-AD-R-1B is open and PC-AD-R-1A is closed.

1.2 Start preferred SGT fan.

1.3 Open SGT-DPCV-546A(B) valve.

1.4 Vent Torus by performing following:

BOP 1.4.1 Ensure PC-MO-1308 is closed.

1.4.2 Open PC-AO-245AV.

1.4.3 Open PC-MO-305MV.

1.4.4 When Torus pressure ~ 0.25 psig, close PC-MO-305MV.

1.4.5 Close PC-AO-245AV.

1.4.6 Place switch for PC-AO-245AV to AUTO.

1.5 Vent Drywell by performing following:

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 5 Event

Description:

Small steam leak in PC (Within SGT venting capacity)

Time Position Applicants Action or Behavior 1.5.1 Open PC-AO-246AV.

1.5.2 While ensuring Torus pressure does not exceed Drywell pressure by > 0.1 psig, open PC-MO-306.

1.5.3 When Drywell pressure ~ 0.25 psig, close PC-MO-306.

1.5.4 Close PC-AO-246AV.

1.5.5 Place switch for PC-AO-246AV to AUTO.

1.6 Place switch for running SGT fan to AUTO.

1.7 Place switch for SGT-DPCV-546A(B) to AUTO.

NOTE to Evaluators: The drywell pressure will level out below the assigned scram setpoint.

When drywell pressure cannot be maintained below 0.75 psig, review TS LCO 3.6.1.4.

CRS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not A.1 Restore pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within limit. within limit.

If drywell pressure cannot be restored and maintained below 0.75 psig, CRS direct RO perform rapid power reduction per Procedure 2.1.10.

If directed perform Rapid Power Reduction per Procedure 2.1.10:

RO While monitoring rod line and feedwater flow, reduce core flow to 40x106 lbs/hr using Reactor Recirculation.

Assign Procedure 2.1.5 Attachments 1, 2, and 3 to the RO and CRS Attachments 4 and 5 to the BOP.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 5 Event

Description:

Small steam leak in PC (Within SGT venting capacity)

Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior Booth When directed by Lead Examiner raise leak by changing Malfunction Operator MS02a to 0.015% and ramp over 3 minutes.

NOTE to Examiner: Overrides ZROSPRRELK4B and ZRORPSRELK4D are already active (An RPS trip on high drywell pressure will not occur)

Ensure all available drywell FCU control switches in RUN.

Ensure following valves are open:

REC-MO-702, DRYWELL SUPPLY ISOLATION.

BOP REC-MO-709, DRYWELL RETURN ISOLATION.

If plant conditions permit, lower REC temperature as much as possible per Procedure 2.2.65.1.

Per Abnormal Procedure 2.4PC:

1. If leak in drywell is indicated, perform following:

NOTE - Next Step shall be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

1.1 If OWC Injection System in service, place OWC INJECTION SYS ENABLE SWITCH (Panel A) to SHUTDOWN.

1.1.1 When time permits, secure OWC Injection System BOP per Procedure 2.2.98.

1.2 Have Chemistry evaluate isotopic makeup of drywell atmosphere/sump discharge to assist in determining source of leak.

1.3 Monitor REC SURGE TANK for possible source of leakage.

1.4 If RCS Operational Leakage occurs, enter Condition and Required Actions of TS LCO 3.4.4.

BOP Report drywell pressure rise.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior Critical Before drywell pressure reaches 1.84 psig, scram the reactor.

Task When drywell pressure cannot be maintained below critical parameter directed by the CRS, manually scram the reactor per Procedure 2.1.5, Attachment 1, Mitigating Task Scram Actions:

1. MITIGATING TASK SCRAM ACTIONS 1.1 Press both RX SCRAM buttons.

RO 1.2 Place REACTOR MODE switch to REFUEL.

Provides scram report:

Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

Performs Procedure 2.1.5, Attachment 2, Reactor Power Control actions:

1. REACTOR POWER CONTROL 1.1 Ensure REACTOR MODE switch is in SHUTDOWN.

1.2 Verify all SDV vent and drain valves are closed.

NOTE - RR pump(s) will be tripped if on Normal Transformer or if ARI/RPT has automatically initiated.

RO 1.3 Ensure operating RR pumps have run back to 22% speed.

NOTE - Steps 1.4 and1.5 may be performed concurrently.

1.4 Verify all control rods are fully inserted.

1.41 If necessary, insert control rods as directed by CRS.

1.5 Observe nuclear instrumentation and perform following:

1.5.1 Insert SRM detectors.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior 1.5.2 Insert IRM detectors.

1.5.3 Change APRM recorders to IRMs.

1.5.4 Range IRMs on scale.

1.5.5 Check reactor power is lowering.

When RPV water level lowers below +3 inches, enter EOP 1A CRS Direct RO maintain RPV water level between +3 and +54 inches.

Performs Procedure 2.1.5, Attachment 3 Reactor Water Level Control, actions:

1.1 After FW Sequence has reached Mode 2 or level has stabilized, place RFC-SW-S1, SETPOINT SETDOWN, switch to DISABLE/RESET.

1.2 Maintain RPV level in prescribed band using following systems, as required, based on plant conditions:

RO 1.2.1 Verify preferred RFP is controlling level in FW Sequence Mode 2 with controlling RFP in RX PRESS FOLLOW Mode.

1.2.2 Note to Examiner, Step is N/A.

1.2.3 If EMER CLOSE button is yellow, press EMER CLOSE button on either FCV-11AA or FCV-11BB.

1.2.4 Ensure following controllers are in AUTO:

1.2.4.1 FCVs 11AA and 11BB.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior 1.2.4.2 STARTUP MASTER CONTROL.

1.2.5 Note to Examiner, Step is N/A.

1.2.6 Adjust STARTUP MASTER controller using UP/DOWN arrows or RAMP FUNCTION to adjust LEVEL SETPOINT as desired.

1.2.6.1 During plant cooldown, for further guidance, adjust Reactor Feedwater System/Condensate System per Procedure 2.2.28.1/2.2.6.

1.4 HPCI per Procedure 2.2.33.1.

1.5 RCIC per Procedure 2.2.67.1.

1.6 Trip non-preferred RFP, if not needed, or minimum flow is isolated.

1.7 Trip all but one condensate booster pump.

1.8 Trip all but one condensate pump.

Direct BOP to control RPV pressure between 1050 and 800 psig.

CRS Performs Procedure 2.1.5, Attachment 4 Reactor Pressure Control, actions:

NOTE - Steps may be performed concurrently.

BOP 1.1 If necessary to stabilize or reduce reactor pressure, BPVs can be operated in manual by performing following:

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior 1.1.1 Transfer bypass valve control from AUTO to MANUAL by pressing BPV MANUAL button and check it backlights.

1.1.1.1 Press BPV RAISE or LOWER buttons to adjust impulse pressure or reactor pressure.

1.2 Maintain RPV pressure in the prescribed band by using the following systems based on plant conditions:

1.2.1 DEH per Procedure 2.2.77.1.

1.2.2 SRVs per Procedure 2.2.1.

1.2.3 HPCI per Procedure 2.2.33.1.

1.2.4 RCIC per Procedure 2.2.67.1.

Performs Procedure 2.1.5, Attachment 5 Balance of Plant, actions:

1.1 Verify main turbine automatically tripped or perform following when main generator output 80 MWe:

1.1.1 At Panel B, simultaneously press TURB TRIP 1 and TURB TRIP 2 buttons, and verify turbine trips.

1.2 Note to Examiner, Step is N/A.

BOP 1.3 When main turbine trips, observe following valves close:

1.3.1 Both stop valves.

1.3.2 All governor valves.

1.3.3 All reheat stop valves.

1.3.4 All interceptor valves.

1.4 Verify station service is transferred to Startup Transformer.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 6 Event

Description:

Steam leak propagates > vent capacity requiring manual scram per 2.4PC.

Time Position Applicants Action or Behavior 1.5 Ensure PCB-3310 open (Panel C).

1.6 Ensure PCB-3312 open (Panel C).

1.7 Ensure GEN EXCITER FIELD BKR is open (Panel C).

When drywell temperature reaches 150°F, enter EOP 3A. Direct operating all available drywell cooling. (Note all cooling is already in service).

CRS RO Control RPV water level +3 inches to +54 inches.

BOP Stabilize RPV pressure below 1050 psig.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

Control rod 14-23 fails to insert on the scram.

Time Position Applicants Action or Behavior NOTE to Examiner: The one control rod failure to insert is already active.

RO Report failure of control rod 14-23 to fully insert on the scram.

CRS Direct RO to manually insert control rod 14-23.

Insert control rod 14-23 by manually inserting it with the reactor manual RO control system.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 8 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Malfunction MC01 at 100%

with a ramp time of 3 minutes (Trigger 8).

Operator BOP Report main condenser vacuum going away.

Enter Abnormal Procedure 2.4VAC and direct BOP to perform subsequent CRS operator actions.

CRS Provide BOP critical parameter for closing MSIVs.

Per 2.4VAC, if vacuum cannot be maintained 12Hg, close the MSIVs at Panel 9-3:

1.1 VLV AO 80A, INBOARD STEAM ISOLATION.

1.2 VLV AO 86A, OUTBOARD STEAM ISOLATION.

1.3 VLV AO 80B, INBOARD STEAM ISOLATION.

1.4 VLV AO 86B, OUTBOARD STEAM ISOLATION.

BOP 1.5 VLV AO 80C, INBOARD STEAM ISOLATION.

1.6 VLV AO 86C, OUTBOARD STEAM ISOLATION.

1.7 VLV AO 80D, INBOARD STEAM ISOLATION.

1.8 VLV AO 86D, OUTBOARD STEAM ISOLATION.

Close the MSL drain valve at Panel 9-4:

1.1 MS-MO-74, INBD ISOL VLV.

When MSIVs are closed, direct BOP to control pressure with SRVs/RCIC CRS operation (Alternate Pressure control systems).

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 8 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 8 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior Procedure 2.2.1 provides guidance for SRV manual operation:

NOTE: Relief valves should be opened in sequence listed on Panel 9-3, with at least 3 seconds between openings.

1.1 Open desired relief valve and verify following:

1.1.1 At switch, red and amber lights - on, and green light - off.

1.1.2 Annunciator 9-3-1/A-2, RELIEF VALVE OPEN, alarms.

1.1.3 Annunciator 9-3-1/C-1, SAFETY/RELIEF VALVE LEAKING, alarms if no other relief valve tailpipe temperature is high.

1.2 When RPV pressure has been lowered 100 to 150 psig, place BOP switch for open valve to AUTO and verify following:

1.2.1 At switch, red and amber lights - off and green light - on.

1.2.2 Annunciator 9-3-1/A-2, RELIEF VALVE OPEN, clears.

NOTE - PMIS points specified in Attachment 1 may also be used to verify position and temperature.

1.3 MS-TR-166, SAFETY AND RELIEF VALVE LEAKAGE TEMPS (Panel 9-21), or PMIS points indicate valve is closed by lowering tailpipe temperature.

1.4 Place RHR System in Suppression Pool Cooling Mode per Procedure 2.2.69.3, as required, to maintain suppression pool temperature < 95°F.

Per Procedure 2.2.67.1, Attachment 1 RCIC (HARD CARD)

RPV PRESSURE CONTROL WITH RCIC 2.1 Ensure RCIC auto initiation signal is clear.

2.2 Open RCIC-MO-33.

RO 2.3 Start GLAND SEAL VACUUM PUMP.

2.4 Open RCIC-MO-132.

2.5 Open RCIC-MO-30.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 8 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior 2.6 Open RCIC-MO-131.

2.7 Adjust RCIC-FIC-91, RCIC flow controller, as required, to maintain desired RPV pressure.

2.7.1 If operating < 300 gpm and flow oscillations are occurring, place RCIC-FIC-91 to MANUAL.

2.8 If available, at VBD-M, ensure REC-MO-711 or REC-MO-714 (associated with an in service REC HX) is open.

2.9 If RCIC is needed for RPV Injection:

2.9.1 Open RCIC-MO-21.

2.9.2 Adjust RCIC-FIC-91, RCIC flow controller until discharge pressure is greater than RPV pressure.

2.9.3 Throttle RCIC-MO-30, as required, to maintain desired RPV level.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 9 Event

Description:

Drywell pressure rises requiring EOP 1A and EOP 3A entry.

Time Position Applicants Action or Behavior NOTE to Examiner: Drywell steam leak is already active.

CRS When drywell pressure reaches 1.84 psig, enter EOP 1A and re-enter EOP 3A.

Before torus pressure reaches 10 psig direct the BOP to place Torus Sprays in service.

CRS Booth When Drywell pressure reaches 3 psig, change malfunction MS02 to 10% with 2 Operator minute ramp.

Place torus sprays in service.

Per Procedure 2.2.69.3, Attachment 1 Containment Sprays (HARD CARD) 2.1 If required, place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

2.2 Ensure RHR-MO-39A(B) open.

2.3 If reactor pressure 300 psig and injection not desired, close RHR-MO-27A(B), OUTBD INJECTION VLV.

BOP 2.4 Ensure RHR PUMP(s) running.

NOTE - RHR pump operation at minimum flow should be limited to < 15 minutes or pump damage may result.

2.5 Throttle RHR-MO-38A(B) to maintain desired containment pressure.

2.6 Throttle RHR-MO-66A(B) to obtain desired cooling rate.

2.7 If PCIS Group 6 lights lit on Panel 9-5, ensure one of following open:

2.7.1 REC-MO-711; or Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 9 Event

Description:

Drywell pressure rises requiring EOP 1A and EOP 3A entry.

Time Position Applicants Action or Behavior 2.7.2 REC-MO-714.

2.8 Place RHR SW System in service:

2.8.1 Start SWBP(s).

2.8.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

BOP Keep CRS informed of drywell pressure trends.

When torus pressure exceeds 10 psig, or before drywell temperature reaches 280°F, direct drywell sprays placed in service.

CRS Direct RR Pumps and Drywell FCUs secured.

CRS RO At Panel 9-4, place running RR pump control switch(es) to STOP.

RO At VBd H, place 4 drywell FCU control switches to OFF.

CRS Provide drywell pressure band to control.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 9 Event

Description:

Drywell pressure rises requiring EOP 1A and EOP 3A entry.

Time Position Applicants Action or Behavior Place drywell sprays in service.

Per Procedure 2.2.69.3, Attachment 1 Containment Sprays (HARD CARD)

BOP 1.1 If Drywell Spray required:

1.1.1 Open RHR-MO-31A(B).

1.1.2 Throttle RHR-MO-26A(B) to maintain desired containment pressure.

BOP Control drywell pressure in band provided by CRS (Generally 2 psig to 10 psig.)

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 10 Event

Description:

HPCI fails to start (Oil system leak)

Time Position Applicants Action or Behavior NOTE to Examiners: HPCI failure is already active.

BOP Report failure of HPCI to start on high drywell pressure.

BOP Send NLO to investigate HPCI failure.

Booth When directed to investigate HPCI, wait 3 minutes and report there Operator is an oil leak on the AOP discharge piping.

BOP Report HPCI oil leak to CRS and place HPCI AOP in PTL.

BOP Update crew HPCI Aux Oil Pump is in PTL.

CRS Direct WCC to repair HPCI oil leak.

Booth As WCC SRO inform CRS a team is being put together to repair the Operator HPCI oil leak.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 11 Event

Description:

Operating CRD pump trips Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction Rd08b, trip of Operator CRD Pump B (Trigger 11).

RO Report trip of CRD Pump B.

CRS Direct starting standby CRD Pump.

Start CRD Pump A per alarm 9-5-2/C-6:

1.1.1 Place CRD-FC-301 in MAN.

1.1.2 Adjust CRD-FC-301 to minimum.

1.1.3 When FCV indicates closed on CRD-FC-301, start CRD Pump A.

RO 1.1.3.1 Note to Examiner, Step is N/A.

1.1.4 Slowly adjust CRD-FC-301 to obtain flow of 50 gpm.

1.1.5 Balance CRD-FC-301.

1.1.6 Place CRD-FC-301 to BAL.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 12 Event

Description:

Steam leak raises drywell temperature to the point Emergency Depressurization is required.

Time Position Applicants Action or Behavior Booth At direction of lead examiner, adjust MS02b in small increments until Operator the leak causes drywell temperature to remain above 280°F.

CREW Report change in drywell temperature and pressure trend.

CRS Direct other RHR loop placed into drywell spray mode.

Place torus sprays in service.

Per Procedure 2.2.69.3, Attachment 1 Containment Sprays (HARD CARD) 2.1 If required, place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

2.2 Ensure RHR-MO-39A(B) open.

2.3 If reactor pressure 300 psig and injection not desired, close RHR-MO-27A(B), OUTBD INJECTION VLV.2 2.4 Ensure RHR PUMP(s) running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

2.5 Throttle RHR-MO-38A(B) to maintain desired containment pressure.

BOP 2.6 Throttle RHR-MO-66A(B) to obtain desired cooling rate.

2.7 If PCIS Group 6 lights lit on Panel 9-5, ensure one of following open:

2.7.1 REC-MO-711; or 2.7.2 REC-MO-714.

2.8 Place RHR SW System in service:

2.8.1 Start SWBP(s).

2.8.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 12 Event

Description:

Steam leak raises drywell temperature to the point Emergency Depressurization is required.

Time Position Applicants Action or Behavior Place drywell sprays in service.

Per Procedure 2.2.69.3, Attachment 1 Containment Sprays (HARD CARD) 1.1 If Drywell Spray required:

1.1.1 Open RHR-MO-31A(B).

1.1.2 Throttle RHR-MO-26A(B) to maintain desired containment pressure.

Critical When drywell temperature cannot be restored and maintained < 280°F, Task emergency depressurize the RPV.

When emergency depressurization is required enter EOP 2A.

CRS CRS Direct stop and prevent of Core Spray and RHR injection per Hard Card.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 12 Event

Description:

Steam leak raises drywell temperature to the point Emergency Depressurization is required.

Time Position Applicants Action or Behavior Perform prevent injection per EOP 5.8, Attachment 4 (HARD CARD):

1. If non-ATWS conditions, ensure one of following:

1.1 RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3 and system pressure maintained

< RPV pressure.

1.2 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

BOP 1.3 Ensure both RHR Systems secured with pumps in PULL-TO-LOCK.

2. Prevent CS by performing following:

2.1 Ensure CS-MO-12A is closed.

2.2 Ensure CS Pump A control switch in PULL-TO-LOCK.

2.3 Ensure CS-MO-12B is closed.

2.4 Ensure CS Pump B control switch in PULL-TO-LOCK.

CRS Direct BOP to open 6 SRVs.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 12 Event

Description:

Steam leak raises drywell temperature to the point Emergency Depressurization is required.

Time Position Applicants Action or Behavior When directed to ED, verify PC water level is above 6 ft. and open 6 SRVS BOP by taking their control switches to OPEN.

Direct RO maintain RPV level band of -110 inches to -60 inches FZ during CRS ED.

Utilize low pressure systems to control an injection flow rate of a minimum of RO 3000 gpm to 4000 gpm (1.5 to 2 Mlbs/hr) and maintain RPV level in band.

When RPV pressure is < 50 psig above drywell pressure report ED is BOP complete.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 1 Event No.: 12 Event

Description:

Steam leak raises drywell temperature to the point Emergency Depressurization is required.

Time Position Applicants Action or Behavior Notes When ED is complete (RPV pressure < 50 psig above Torus pressure), RPV level is controlled in desired band and lead examiner has observed enough of the scenario, then terminate the scenario.

END OF THE SCENARIO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Simulator Setup Initialize the simulator in IC120 (EOC)

Triggers and Malfunctions E1 - None E2 - None E3 - None E4 - ZAICRDFC301(2)=0 (Override on CRD-FC-301 automatic setpoint)

E5 - MS02a set to 0,0075% with a 5 minute ramp.

E6 - None E7 - None E8 - MC01 set at 100% with a 3 minute ramp.

E9 - RD08b, CRD Pump B trip.

- HP12, HPCI Oil Piping break

- RD15 14-23, Control Rod 14-23 fails to scram (Active)

Overrides

ZDIRCICSWS2(2)=OPEN, RCIC-MO-16 control switch to OPEN.(Active)

ZRORPSRELK4B=ON, RPS Relay 5A-K4B does not de-energize on high DW pressure.

(Active)

ZRORPSRELK4D=ON, RPS Relay 5A-K4D does not de-energize on high DW pressure.

(Active)

Panel Set-up

Ensure PMIS IDTs are blank.

Place the Shutdown EOC Rod Sequence Book on Panel 9-5

Reduce Reactor Power with Rx Recirculation to 70%

Ensure Recirculation Controllers are selected to AP@

Procedures Needed 6.RCIC.201, RCIC POWER OPERATED VALVE OPERABILITY TEST (IST)

Tags Hung DEH Pump B (C/S in PTL)

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Turnover Sheet:

Plant Status: The plant is operating at approximately 70% at the end of the current fuel cycle.

Risk: Green Activities in Progress: None LCOs in effect: None Equipment out of service: DEH Pump B tagged out for coupling adjustment.

Activities for the Shift: Raise reactor power 30 MWe net with RR and then perform 6.RCIC.201, RCIC POWER OPERATED VALVE OPERABILITY TEST (IST)

Rev 1

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: __1____

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: 100% power (BOL) Air Compressor 1B tagged out for maintenance. 6.1SGT.301, SGT Operability Test/Off Gas Flow Monitor Channel Functional Test IST (DIV 1) ongoing, SGT fan has been running for 10.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Turnover: Maintain steady state. Mechanics performing motor vibration PMs on EF-T-1A, Turbine Building Exhaust Fan. Complete 6.1SGT.301. Currently at Step 4.10. (Section 5 already complete).

Event Malf. Event Event No. No. Type* Description 1 N/A N Secure SGT from 6.1SGT.301.

(BOP) 2 FW12a I (ATC) FW Flow transmitter input to RVLCS fails high.

3 RD11 T (CRS) Accumulator nitrogen pressure low.

TS LCO 3.1.5 Associated control rod inoperable.

(BOP)

TS LCO 3.5.1, HPCI inoperable.

T (CRS) 5 HP06 M HPCI Steam Leak in Secondary Containment (EOP 5A).

(ALL) 6 HP09 N/A HPCI steam lines fail to isolate and cannot be isolated from the main control room or ASD room.

MO- (BOP) 15/16 override open 7 RP01a,b I, © EOP-1A (from EOP-5A) and scram reactor (manual scram fails and and c ARI is used to insert control rods.

(ATC)

RD26 EOP-1A to EOP-6A entered. Manual ARI initiation inserts control rods and the CRS transitions back to EOP-1A.

RD27 8 RR17 I, (ATC) One RR pump fails to runback post scram. (ATC trips pump).

(BOP)

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: The plant is at 100% power at the beginning of the cycle. Plant Air Compressor 1B is tagged out for maintenance. The monthly SGT 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months run surveillance 6.1SGT.301 is in progress.

Turnover: Maintain power level. Complete 6.1SGT.301.

BOP operator secures SGT and places system in standby status.

While BOP is securing SGT one feedwater flow transmitter input to reactor vessel level control oscillates and causes small RPV level oscillations. Abnormal Procedure 2.4RXLVL is entered and the RVLCS is placed in 1 ELEMENT CONTROL.

One control rod drive mechanism accumulator alarms due to low nitrogen pressure. The CRS declares the associated control rod inoperable per TS LCO 3.1.5.

HPCI inadvertently starts on a false initiation signal and the BOP operator secures HPCI (critical task) per immediate operator actions of Abnormal Procedure 2.4CSCS. The CRS declares HPCI inoperable per TS LCO 3.5.1.

HPCI steam line begins leaking in secondary containment (Reactor building). The CRS enters EOP-5A on high secondary containment temperatures.

The crew determines the leak is on the HPCI steam line and tries to isolate the steam line. The steam line isolation valves (HPCI-MO-15 & 16) will not close from the main control room. If isolation is tried from the Alternate Shutdown Room the valves also fail to close. Any attempt to manually close an isolation valve also fails.

As temperatures rise in secondary containment, the CRS enters EOP-1A and directs the reactor scrammed.

When a manual reactor scram is attempted, the manual scram buttons or the mode switch in shutdown fail to insert the control rods. The CRS exits EOP-1A and Enters EOP-6A and 7A and directs ARI to be initiated. Automatic ARI initiation does not function and manual ARI is successful (critical task) and the CRS exits EOP-6A and 7A and re-enters EOP-1A.

One reactor recirculation pump fails to automatically run back to minimum post-scram requiring the ATC to trip the pump.

Secondary containment temperatures continue to rise requiring the CRS to direct RPV emergency depressurization. The BOP EDs with SRVs. (critical task)

The scenario ends when RPV level is stabilized post-ED.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 1 Event

Description:

Secure from SGT surveillance 6.1SGT.301.

Time Position Applicants Action or Behavior 4.10 AC After SGT Subsystem A has run for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , check both red lights above SGT-HTR-SGHA switch are on.

NOTE - SGT-AO-249 is timed, when required, as follows: (1) closing time is measured with a stopwatch to nearest 1/10 of a second from time SGT fan switch is placed to AUTO until red light turns off; (2) time is recorded on Attachment 2.

4.11 (Independent Verification) Stop EF-R-1E by placing its switch to AUTO.

BOP 4.12 AC (Time) Check SGT-AO-249 closes.

4.13 AC Check SGT-AO-251 closes.

4.14 Place SGT-DPCV-546A switch to AUTO.

4.15 Record EF-R-1E RUN TIME INTEGRATOR value in FINAL INTEGRATOR (If) VALUE block on Attachment 1, Table 1, and calculate RUN TIME.

4.16 Record ERP FLOW MEASURING DEVICE POST-TEST data on Attachment 1, Table 2; N/A if not a 92 day test or already performed.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 2 Event

Description:

FW Flow transmitter input to RVLCS fails high.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction FW12a at 50%

(Trigger 2).

Operation RO Report RPV level, A FW flow oscillations.

CRS Enter Abnormal Procedure 2.4RXLVL.

Announces ownership of following scram action:

lf either of following occur at any time, SCRAM and concurrently enter Procedure 2.1.5:

RO RPV level cannot be maintained above +12" on narrow range instruments.

RPV level cannot be maintained below +50" on narrow range instruments.

RO Place LEVEL CONTROL SELECT switch to 1 ELEMENT CONT.

If level control is still erratic place following in MAN, MDEM, or MDVP:

RO MASTER LEVEL controller.

RFPT-1A/RFPT-1B controller.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 3 Event

Description:

Control Rod 06-39 accumulator nitrogen pressure low.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, put in malfunction RD11 06-39, Operator accumulator low pressure (Trigger 3).

RO Respond to Alarm 9-5-2/G-6.directs entering LCO 3.1.5.

BOP Direct NLO to investigate control rod 06-39 accumulator.

Role Play-Booth When called as NLO to investigate 06-39 accumulator. Wait 2 minutes Operator and report the alarm is low accumulator pressure and the pressure is 930 psig.

Reviews TS LCO 3.1.5.

ACTIONS

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

Separate Condition entry is allowed for each control rod scram accumulator.

CONDITION REQUIRED ACTION COMPLETION TIME A. One control rod A.1 --------NOTE---------------

scram Only applicable if the accumulator associated control rod inoperable with scram time was within reactor steam the limits of Table 3.1.4-CRS dome pressure 1 during the last scram 900 psig. time Surveillance.

Declare the associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod scram time slow.

OR A.2 Declare the associated 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months control rod inoperable.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 3 Event

Description:

Control Rod 06-39 accumulator nitrogen pressure low.

Time Position Applicants Action or Behavior Role Play-Booth Operator If contacted as Reactor Engineering for control rod 06-39, respond that you will check your records.

Role Play-Booth One minute later call the CRS and report control rod 06-39 was NOT Operator within the limits of Table 3.1.4-1 during the performance of the last scram time surveillance.

CRS CRS control rod 06-39 inoperable per TS LCO 3.1.5.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 4 Event

Description:

Inadvertent HPCI initiation.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction HP05, inadvertent Operator HPCI initiation Trigger 4).

BOP Report HPCI has initiated.

Check drywell pressure and RPV level and determine HPCI does NOT have RO/BOP a valid initiation signal.

NOTE to Examiners: Immediate Operator Actions are performed from memory.

Perform Immediate Operator Action of Abnormal Procedure 2.4CSCS:

1.1.1 Ensure AUXILIARY OIL PUMP control switch in START.

1.1.2 Press and hold TURBINE TRIP button.

BOP 1.1.3 After turbine stops, place AUXILIARY OIL PUMP in PULL-TO-LOCK.

1.1.4 Release TURBINE TRIP button.

Critical HPCI AOP placed into PTL before 3 minutes of sustained injection.

Task Time:____________

RO Report any power rises if HPCI allowed to inject into the RPV.

Enter Abnormal Procedure 2.4CSCS.

CRS BOP Provide crew an update that HPCI Auxiliary Oil Pump is in PTL.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 4 Event

Description:

Inadvertent HPCI initiation.

Time Position Applicants Action or Behavior Review TS LCO 3.5.1:

ACTIONS

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

LCO 3.0.4.b is not applicable to HPCI CONDITION REQUIRED ACTION COMPLETION TIME CRS C. HPCI System C.1 Verify by administrative 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable means RCIC System is OPERABLE.

AND C.2 Restore HPCI System 14 days to OPERABLE status.

CRS Declare HPCI inoperable.

Role Play-Booth Operator As WCC SRO receive report about HPCI being inoperable. Report you will put a repair team together to investigate HPCI.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 5 Event

Description:

HPCI Steam Leak in Secondary Containment Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction HP06 at 3%

Operator (Trigger 5).

BOP responds to high building area temperature alarm (9-3-1/E-10).

BOP Reports high area temperature in the SW Torus Area.

Crew determines the leak is on the HPCI steam line.

BOP Direct NLO to determine cause of alarm.

Role Play-Booth When contacted to investigate for steam leaks in Reactor Building, wait Operator 5 minutes and report you hear loud hissing noise in torus area. Report the torus area humidity level has risen above normal levels.

CRS enters EOP 5A to control Secondary Containment temperatures.

Monitors area temperature for any temperature rising above 195°F.

CRS Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 5 Event

Description:

HPCI Steam Leak in Secondary Containment Time Position Applicants Action or Behavior Places area coolers and available Reactor Building HVAC into service.

At VBd R: FC-R-1E SE Quad, Control Switch to RUN FC-R-1F NE Quad, Control Switch to RUN BOP FC-R-1H SW Quad, Control Switch to RUN FC-R-1J NW Quad, Control Switch to RUN FC-R-1G HPCI ROOM, Control Switch to RUN END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 6 Event

Description:

HPCI steam lines fail to isolate and cannot be isolated from the main control room or ASD room.

Time Position Applicants Action or Behavior NOTE to Examiners: HPCI will not isolate on any isolation signal and cannot be isolated with control switches form the control room or alternate shutdown room.

CRS Direct HPCI steam line isolated.

At Panel 9-4, attempt to close HPCI-MO-15 and HPCI-MO-16 with their BOP control switches.

Report HPCI will not isolate from control room.

BOP Direct NLO to isolate HPCI from the ASD panel in the Reactor Building.

Role Play-Wait 3 minutes when directed to isolate HPCI from the ASD panel:

Override alarm 9-3-3/G-5, Inner Door to ON.

Booth Operator Place HPCI-MO-15 and HPCI-MO-16 ASD ISOLATION SWITCHES TO ISOLATE.

Then report you cannot get HPCI-MO-15 or 16 to close.

Delete override on alarm 9-3-3/G-5 indicating door is closed.

BOP Direct HPCI-MO-16 manually closed.

Role Play-Booth Operator If directed to manually isolate HPCI-MO-16, wait 5 minutes and report you cannot get the valve to move.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 6 Event

Description:

HPCI steam lines fail to isolate and cannot be isolated from the main control room or ASD room.

Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 7 Event

Description:

Reactor scram (RPS fails to insert control rods) and ARI is used to insert control rods.

Time Position Applicants Action or Behavior NOTE to Examiners: Malfunctions to RPS Rod Groups 1, 2 and 3 prevents these control rods from inserting from an RPS signal. RPS Group 4 control rods insert. Automatic ARI initiation fails.

When Secondary Containment area temperatures rise above 195°F, in one area, enter EOP 1A and direct the reactor scrammed.

CRS Critical ARI is initiated to insert all control rods.

Task When drywell pressure cannot be maintained below critical parameter directed by the CRS, manually scram the reactor per Procedure 2.1.5, Attachment 1, Mitigating Task Scram Actions:

2. 1. MITIGATING TASK SCRAM ACTIONS 1.1 Press both RX SCRAM buttons.

RO 1.2 Place REACTOR MODE switch to REFUEL.

1.3 lf reactor power > 3% perform following:

1.3.1 Place REACTOR MODE switch to SHUTDOWN 1.3.2 lnitiate ARl.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 7 Event

Description:

Reactor scram (RPS fails to insert control rods) and ARI is used to insert control rods.

Time Position Applicants Action or Behavior Provides scram report:

Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

Performs Procedure 2.1.5, Attachment 2, Reactor Power Control actions:

1. REACTOR POWER CONTROL 1.1 Ensure REACTOR MODE switch is in SHUTDOWN.

1.2 Verify all SDV vent and drain valves are closed.

NOTE - RR pump(s) will be tripped if on Normal Transformer or if ARI/RPT has automatically initiated.

1.3 Ensure operating RR pumps have run back to 22% speed.

NOTE - Steps 1.4 and 1.5 may be performed concurrently.

1.4 Verify all control rods are fully inserted.

RO 1.41 If necessary, insert control rods as directed by CRS.

1.5 Observe nuclear instrumentation and perform following:

1.5.1 Insert SRM detectors.

1.5.2 Insert IRM detectors.

1.5.3 Change APRM recorders to IRMs.

1.5.4 Range IRMs on scale.

1.5.5 Check reactor power is lowering.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 7 Event

Description:

Reactor scram (RPS fails to insert control rods) and ARI is used to insert control rods.

Time Position Applicants Action or Behavior Performs Procedure 2.1.5, Attachment 3 Reactor Water Level Control, actions:

1.1 After FW Sequence has reached Mode 2 or level has stabilized, place RFC-SW-S1, SETPOINT SETDOWN, switch to DISABLE/RESET.

1.2 Maintain RPV level in prescribed band using following systems, as required, based on plant conditions:

1.2.1 Verify preferred RFP is controlling level in FW Sequence Mode 2 with controlling RFP in RX PRESS FOLLOW Mode.

1.2.2 Note to Examiner, Step is N/A.

1.2.3 If EMER CLOSE button is yellow, press EMER CLOSE button on either FCV-11AA or FCV-11BB.

1.2.4 Ensure following controllers are in AUTO:

1.2.4.1 FCVs 11AA and 11BB.

RO 1.2.4.2 STARTUP MASTER CONTROL.

1.2.5 Note to Examiner, Step is N/A.

1.2.6 Adjust STARTUP MASTER controller using UP/DOWN arrows or RAMP FUNCTION to adjust LEVEL SETPOINT as desired.

1.2.6.1 During plant cooldown, for further guidance, adjust Reactor Feedwater System/Condensate System per Procedure 2.2.28.1/2.2.6.

1.4 HPCI per Procedure 2.2.33.1.

1.5 RCIC per Procedure 2.2.67.1.

1.6 Trip non-preferred RFP, if not needed, or minimum flow is isolated.

1.7 Trip all but one condensate booster pump.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 7 Event

Description:

Reactor scram (RPS fails to insert control rods) and ARI is used to insert control rods.

Time Position Applicants Action or Behavior 1.8 Trip all but one condensate pump.

3. Performs Procedure 2.1.5, Attachment 4 Reactor Pressure Control, actions:

NOTE - Steps may be performed concurrently.

1.1 If necessary to stabilize or reduce reactor pressure, BPVs can be operated in manual by performing following:

1.1.1 Transfer bypass valve control from AUTO to MANUAL by pressing BPV MANUAL button and check it backlights.

1.1.1.1 Press BPV RAISE or LOWER buttons to adjust BOP impulse pressure or reactor pressure.

1.2 Maintain RPV pressure in the prescribed band by using the following systems based on plant conditions:

1.2.1 DEH per Procedure 2.2.77.1.

1.2.2 SRVs per Procedure 2.2.1.

1.2.3 HPCI per Procedure 2.2.33.1.

1.2.4 RCIC per Procedure 2.2.67.1.

Performs Procedure 2.1.5, Attachment 5 Balance of Plant, actions:

1.1 Verify main turbine automatically tripped or perform following when main generator output 80 MWe:

1.1.1 At Panel B, simultaneously press TURB TRIP 1 and BOP TURB TRIP 2 buttons, and verify turbine trips.

1.2 Note to Examiner, Step is N/A.

1.3 When main turbine trips, observe following valves close:

1.3.1 Both stop valves.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 7 Event

Description:

Reactor scram (RPS fails to insert control rods) and ARI is used to insert control rods.

Time Position Applicants Action or Behavior 1.3.2 All governor valves.

1.3.3 All reheat stop valves.

1.3.4 All interceptor valves.

1.4 Verify station service is transferred to Startup Transformer.

1.5 Ensure PCB-3310 open (Panel C).

1.6 Ensure PCB-3312 open (Panel C).

1.7 Ensure GEN EXCITER FIELD BKR is open (Panel C).

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 8 Event

Description:

Reactor Recirculation pump fails to run back post-scram Time Position Applicants Action or Behavior NOTE to Examiners: The RR runback failure is already active.

RO Report failure of operating RR pump to runback.

NOTE to Examiners: Procedure 2.0.1.2 states: Upon recognition of a failure of an automatic safety feature, Operators shall manually perform those actions necessary to fulfill the safety function.

CRS Direct RR pump tripped.

RO At Panel 9-4 place control switch for RR pump to STOP.

RO Report pump tripped.

Enter Abnormal Procedure 2.4RR for both RR pumps tripped. Direct RO to CRS perform 2.4RR actions to address RPV stratification.

Perform 2.4RR, Attachment 1

1. lf both RR pumps trip, enter Attachment 5, 6, or 7, as required by RO plant conditions.

2. Align RRMG H&V System per Procedure 2.2.85.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 9 Event

Description:

Emergency Depressurize on SC temperatures.

Time Position Applicants Action or Behavior Booth After all control rods have been inserted by ARI change the HPCI Operator steam line break by changing malfunction HP06 to 4%.

When second Secondary Containment area temperature rises above 195°F, enter EOP 2A and direct emergency depressurization.

CRS Critical When second Secondary Containment area temperature rises above 195°F six SRVs are opened within 10 minutes.

Task Time:____________

When directed to ED, verify PC water level is above 6 ft. and open 6 SRVS BOP by taking their control switches to OPEN.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 9 Event

Description:

Emergency Depressurize on SC temperatures.

Time Position Applicants Action or Behavior Direct RO maintain RPV level band of -110 inches to -60 inches FZ during CRS ED.

Utilize low pressure systems to control an injection flow rate of a minimum RO of 3000 gpm to 4000 gpm (1.5 to 2 Mlbs/hr) and maintain RPV level in band.

When RPV pressure is < 50 psig above torus pressure report ED is BOP complete.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: 9 Event

Description:

Emergency Depressurize on SC temperatures.

Time Position Applicants Action or Behavior Notes When ED is complete (RPV pressure < 50 psig above Torus pressure), RPV level is controlled in desired band and lead examiner has observed enough of the scenario, then terminate the scenario.

END OF THE SCENARIO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Simulator Setup Initialize the simulator in IC120 (EOC)

Triggers and Malfunctions E1 - None E2 - FW12a at 50% severity.

E3 - RD11, accumulator nitrogen low pressure for control rod 06-39 E4 - HP05 E5 - HP06 at 3% severity.

- HP09, Failure of HPCI to isolate. (Active)

- RP01a, Failure of RPS Group 1 (Active)

- RP01b, Failure of RPS Group 2 (Active)

- RP01c, Failure of RPS Group 3 (Active)

- RD26, Failure of Backup Scram Valves (Active)

- RD27 Automatic ARI failure (Active)

- RR17b, RRMG B Jordan failure at 85 (Active)

Overrides

ZDIHPCISWS2(2)=OPEN. (Active)

ZDIHPCISWS1(2)=OPEN. (Active)

ZDIHPCISWMO15(2)=OPEN (Active)

ZDIHPCISWMO16(2)=OPEN (Active)

Panel Set-up

Ensure PMIS IDTs are blank.

Place the Shutdown BOL Rod Sequence Book on Panel 9-5

Ensure Recirc Controllers are selected to AP@

Procedures Needed 6.1SGT.301. Currently at Step 4.10. (Section 5 already complete).

Tags Hung Air Compressor B Rev 1

Appendix D Required Operator Actions Form ES-D-2 Turnover Sheet:

Plant Status: The plant is operating at approximately 100%.

Risk: Green Activities in Progress: None LCOs in effect: None Equipment out of service: Air Compressor B for maintenance.

Activities for the Shift: Mechanics are performing motor vibration PMs on EF-T-1A.

Complete surveillance 6.1SGT.301.

Rev 1

Appendix D Scenario Outline Form ES-D-1 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: ___1____

Examiners: ____________________________ Operators: _____________________________

Initial Conditions: Plant operating at 75% power following load schedule. TEC Pump C tagged out for bearing replacement, Turnover: Continue with plant startup. IAC will be performing 6.1APRM.305, APRM System (Flow Bias and Startup) Channel Calibration (DIV 1) later in the shift. Pump Vibration PMs being performed on RFP-1A A-2 oil pump.

Event Malf. Event Event No. No. Type* Description 1 RR17b I (ATC) RR runback. (2.4RR)

T (CRS) CRS declares RR flow mismatch per TS LCO 3.4.1.

2 FW01a N/A RFPT 1A trip (Mechanic bumped local trip PB) (2.4MC-RF)

(BOP) 3 N/A N Procedure 2.2.28.1 Quick restart RFPT 1A.

(BOP) 4 RP03a C RPS EPA 1A1/1A2 trip. Half Scram and Half Group 2 and 6 (ALL) isolations. BOP transfers RPS to Alternate Power. ATC resets reactor scram.

5 RP12 I,© PCIS Group 3 fails on RPS trip and RWCU Pump A seal leakage (BOP) raises pump room temperature to the isolation setpoint. BOP CU01a manually isolates RWCU. EOP 5A on high temperature alarm.

T (CRS)

CRS declares PCIS Function 5 (RWCU) inoperable per TS LCO 3.3.6.1.

6 ED9a C 480V Bus 1A trips (5.3AC.480 and 2.4TEC). ATC manually scrams (ATC) on loss of TEC.

8 N/A © (ATC) ATC utilizes multiple manual scrams and control rod insertion to insert all control rods. CRS exits EOP-6A and 7A and re-enters EOP-1A.

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: The plant is starting up at 75% power (EOL) and following the load schedule.

TEC Pump C is tagged out for bearing replacement.

Turnover: Continue with the startup. IAC will be performing an APRM surveillance later in the shift. Mechanics are currently performing vibration PMs on reactor feedwater pump oil pumps.

RR Pump B receives a runback signal and the ATC locks out the scoop tube. The CRS enters Abnormal Procedure 2.4RR and TS LCO 3.4.1 for RR flow mismatch > 10%.

The mechanics bump the local trip pushbutton causing reactor feedwater pump 1A to trip. The CRS enters abnormal procedure 2.4MC-RF. RFP-1A is restarted utilizing the quick restart operating procedure.

RPS EPA trip causes a loss of RPS bus 1A power and PCIS Half Group 2 and Full Group 6 isolation. Group 3 (RWCU) fails to isolate and RWCU Pump A seal leakage rises in the RWCU pump room. The BOP manually isolates RWCU to stop the leak (critical task). The CRS declares PCIS Function 5 inoperable for RWCU per TS LCO 3.3.6.1. CRS enter EOP 5A on RWCU room temperature rising above temperature switch setting.

480V Bus 1A trips on a fault causing the remaining TEC pumps to trip. The CRS enters 2.4TEC and 5.3AC480. Per 2.4TEC the ATC manually scrams the reactor on loss of TEC.

When the reactor is scrammed, a 95% ATWS exists and all control rods do not fully insert on the scram. The CRS enters EOP-1A and transitions to EOP-6A and 7A and directs lowering RPV water level for level/power control. The BOP inhibits ADS and performs Stop and Prevent on RPV injections systems to lower RPV level (critical task). The ATC injects boron with SLC.

(critical task) The ATC lowers power utilizing a combination of manual control rod insertions and manual scrams.

The ATC fully inserts control rods utilizing multiple manual scrams (critical task) and the CRS exits EOP-6A and 7A and re-enters EOP-1A.

The scenario ends when RPV level is being controlled in the required band and the control rods are fully inserted.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 1 Event

Description:

RR Pump B runback Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction RR17b at 22%

with a ramp of 2 minutes (Trigger 1).

Operator RO Report reactor power lowering.

Report RR Pump B lowering in speed and perform Immediate Operator RO Action of Emergency Procedure 2.4RR:

lf recirculation flow is lowering, press SCOOP TUBE LOCKOUT button.

Report RR Pump scoop tube locked out and the speed of both RR pumps RO is not equal.

Enter Emergency Procedure 2.4RR and assign subsequent operator CRS actions to RO.

Review TS LCO 3.4.1 If RR loop flows are >10% apart, the CRS enters TS LCO 3.4.1:

LCO 3.4.1 Two recirculation loops with matched flows shall be in operation outside of the Stability Exclusion Region of the power/flow map specified in the COLR.

CRS ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Requirements of B.1 Satisfy the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the LCO not met for requirements of reasons other than the LCO.

Condition A.

Performs 2.4RR, Attachment 4 actions for locked out scoop tube:

RO 1.1 Operate scoop tube locally per Procedure 2.2.68.1; or Directs WCC to find a licensed operator to take manual control of the RR RO pump.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 1 Event

Description:

RR Pump B runback Time Position Applicants Action or Behavior Role Play-Booth Operator When contacted as WCC, report you will find a licensed operator and have him report to the RRMG set area.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 2 Event

Description:

Inadvertent trip of RFPT A.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction FW01a, trip of RFPT A (Trigger 2).

Operator Booth As soon as RFPT A trips, delete malfunction FW01a.

Operator BOP Report trip of RFPT A.

RO Report RPV water level trend.

Role Play-Booth Contact control room as maintenance personnel in the RFP room and Operator report you bumped the trip button on RFP A.

Enter Abnormal Procedure 2.4MC-RF and assign subsequent operator CRS actions to BOP.

Take responsibility for scram action of 2.4MC-RF:

RO lf RPV level cannot be maintained above +12" on Narrow Range lnstruments, SCRAM and concurrently enter Procedure 2.1.5.

RO Ensure RFPT B controls RPV level in normal band.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 3 Event

Description:

Quick restart RFPT 1A.

Time Position Applicants Action or Behavior CRS Direct BOP to perform a Quick Restart of RFPT A.

Perform Quick Restart of RFPT A per the HARD CARD 1.1 Press RFPT-A(B) TURBINE TRIP pushbutton (Panel A).

1.2 Ensure all trips including high level trips are reset.

1.3 At a RFPT/RVLC HMI on MAIN CONTROL screen, ensure RFPT-A(B) control is in MDVP with OUTPUT at 0.

1.4 Press and hold RFPT-A(B) TRIP RESET button until RFPT-A(B) HP and LP STOP valves are open.

1.4.1 If RFPT does not reset, press and hold RFPT-A(B)

OVERSPEED TRIP BLOCK and RFPT-A(B) OVERSPEED TRIP RESET.

1.5 Ensure RF-FCV-11A(B), MINIMUM FLOW, is open.

1.6 At a HMI, select FEEDPUMP A(B) screen for desired RFP to be started.

1.7 Select QUICK RESTART start type.

BOP 1.8 If extraction steam is available, select LP START. Otherwise, select HP START.

1.9 Press green START button and confirm start in pop-up box.

1.10 After RFP A(B) reaches MINIMUM GOVERNOR, depress green CONTINUE button.

1.11 Ensure injection path is aligned to the reactor vessel, as dictated by plant conditions.

1.12 Use UP arrow to raise RFP speed to raise RFP discharge pressure.

1.13 Place RFP in desired mode (e.g., AUTO or REACTOR PRESSURE FOLLOW).

1.14 If required, adjust STARTUP MASTER controller using UP/DOWN arrows or RAMP FUNCTION to adjust LEVEL SETPOINT as desired.

1.15 When feedwater flow > 1.5 Mlbm/hr, ensure RF-FCV-11A(B) is closed.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 3 Event

Description:

Quick restart RFPT 1A.

Time Position Applicants Action or Behavior RO Ensure RFPTs are controlling RPV level in desired band.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 4 Event

Description:

RPS EPA 1A1/1A2 trip.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction RP03a, trip of RPS EPA 1A1/1A2 (Trigger 4).

Operator The BOP responds per alarm C-1/F-1.

2.1 If an RPS power supply is available, transfer RPS A to available BOP source.

2.2 Reset RPS Channel "A" half scram per Procedure 2 .1 .5.

2.3 Reset Group Isolations per Procedure 2 .1 .22.

Reports status of reactor and status of PCIS Group isolations:

Half Group 1 RO Half Group 2 Full Group 6 Half Group 7 BOP Reports failure of Group 3 isolation.

Directs BOP to transfer RPS A bus to alternate power per alarm card CRS guidance.

At Panel 9-16, verifies ALT SOURCE AVAIL white light illuminated and BOP transfers RPS A to alternate power. Verifies Red ALT SOURCE ON light illuminated.

Reset half scram per Procedure 2.1.5:

1.1 Place REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM.

1.2 Ensure eight SCRAM GROUP lights (Panels 9-15 and 9-17) or SCRAM INDICATIONS GROUP A and GROUP B lights are on.

RO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 4 Event

Description:

RPS EPA 1A1/1A2 trip.

Time Position Applicants Action or Behavior END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 5 Event

Description:

PCIS Group 3 failure and RWCU Pump A seal failure Time Position Applicants Action or Behavior NOTE to Examiner: When RPS Bus A de-energized, RP12 is already active and CU01a, RWCU Pump seal failure (100% with no ramp time),

automatically becomes active.

Critical RWCU-MO-15 or RWCU-MO-18 closed before RWCU room temperature Task exceeds 195°F (Max Safe temperature).

Recognize failure of RWCU to isolate and isolates system by closing BOP RWCU-MO-15 and RWCU-MO-18 with their control switches.

BOP Opens RWCU-MO-74 after suction piping isolated.

CRS reviews TS LCO 3.3.6.1 and declares PCIS Function 5 (RWCU) inoperable per TS LCO 3.3.6.1.

CONDITIONS REQUIRED ACTIONS COMPLETION TIME B. One or more B.1 Restore CRS Functions with isolation 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolation capability capability.

not maintained.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 6 Event

Description:

480V Bus 1A trips Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert malfunction ED09A, 480V Bus 1A trip (Trigger 6).

Operator BOP Respond to alarm C-2/G-4 and report 480V Bus 1A ground.

CRS Enter Emergency Procedure 5.3AC480.

BOP Report all TEC pumps tripped.

CRS Enter Abnormal Procedure 2.4TEC.

Take responsibility for scram action of 2.4TEC 1.1 lf at any time TEC PRESSURE cannot be restored and maintained above 55 psig and component temperatures rise and do not stabilize:

RO 1.1.1 SCRAM and enter Procedure 2.1.5.

1.1.2 Trip Main Turbine.

1.1.3 Rapidly reduce reactor pressure to 500 to 600 psig using main turbine BPVs per Procedure 2.2.77.1 .

CRS Direct reactor scrammed.

RO Depresses both manual scram pushbuttons.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior NOTE to Examiners: Malfunctions RD02a and RD02b, (40% power ATWS) are already active.

RO Report ATWS conditions and reactor power level.

Enter EOP 1A and transition to EOP 6A (Power/Pressure control) and EOP CRS 7A (RPV Level control).

EOP 6A Power CRS Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior EOP 7A RPV Level CRS Place Reactor Mode Switch in SHUTDOWN.

Initiate ARI.

RO Report reactor power level to CRS.

Run RR pumps back to minimum when directed by CRS.

Trip RR pumps when directed by CRS.

Critical Inhibit ADS before RPV automatic blowdown occurs.

Task Direct inhibiting ADS and installing PTMs for any open MSIV.

CRS Inhibit ADS when directed by CRS:

BOP At Panel 9-3 place ADS A and ADS B INHIBIT switches to INHIB.

Defeat MSIV low level interlocks, when directed by CRS, by installing EOP PTMs 57 through 60 per EOP 5.8.20 in Panels 9-15 and 9-17.

BOP 1.2.1 Install EOP PTM Number 57 jumper between Terminals DD-1 and DD-2 (BAY-1, PNL 9-15).

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior 1.2.2 Install EOP PTM Number 58 jumper between Terminals BB-1 and BB-2 (BAY-3, PNL 9-15).

1.2.3 Install EOP PTM Number 59 jumper between Terminals DD-1 and DD-2 (BAY-1, PNL 9-17).

1.2.4 Install EOP PTM Number 60 jumper between Terminals BB-1 and BB-2 (BAY-3, PNL 9-17).

Direct RPV level lowered below -60 inches by using stop and prevent.

CRS Perform Stop and Prevent per EOP 5.8 HARD CARD:

STOP INJECTION 1.3 Stop HPCI by performing one of following:

1.3.1 TRIP HPCI turbine:

1.3.1.1 If running, press and hold TURBINE TRIP button.

1.3.1.2 When the Turbine is at zero rpm, place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

1.3.1.3 If applicable, release TURBINE TRIP button.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior 1.3.2 Close HPCI-MO-16, STM SUPP OUTBD ISOL VLV.

1.3.3 Depress MANUAL ISOLATION PUSHBUTTON, if initiation signal present.

1.4 Stop Feedwater by performing following:

BOP 1.4.1 Ensure RFP A is tripped.

1.4.2 Ensure RFP B is tripped.

1.4.3 If Reactor pressure 600 psig, ensure all condensate booster pumps are tripped.

1.4.3.1 CBP A.

1.4.3.2 CBP B.

1.4.3.3 CBP C.

1.5 Stop Core Spray by ensuring following:

1.5.1 CS System A secured with pump in PULL-TO-LOCK.

1.5.2 CS System B secured with pump in PULL-TO-LOCK.

1.6 Stop RHR by ensuring one of following:

1.6.1 Both RHR Systems secured with pumps in PULL-TO-LOCK.

1.6.2 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

PREVENT INJECTION 1.7 Prevent RHR by performing Step 1.7.1 or 1.7.2:

1.7.1 If ATWS conditions, ensure one of following:

1.7.1.1 Both RHR Systems secured with pumps in PULL-TO-LOCK.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior 1.7.1.2 RHR outboard injection valves automatic open signal BOP bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

1.7.2 If non-ATWS conditions, ensure one of following:

1.7.2.1 RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3 and system pressure maintained < RPV pressure.

1.7.2.2 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

1.7.2.3 Ensure both RHR Systems secured with pumps in PULL-TO-LOCK.

1.8 Prevent Feedwater by performing following:

1.8.1 At a RVLC/RFPT HMI, select STARTUP VALVE screen, press EMER CLOSE button, and confirm "YES" in pop-up box.

1.8.2 Ensure RF-MO-29 is closed.

1.8.3 Ensure RF-MO-30 is closed.

1.8.4 Trip condensate and condensate booster pump(s), as required.

1.9 Prevent CS by performing following:

1.9.1 Ensure CS-MO-12A is closed.

1.9.2 Ensure CS Pump A control switch in PULL-TO-LOCK.

1.9.3 Ensure CS-MO-12B is closed.

Ensure CS Pump B control switch in PULL-TO-LOCK.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior Direct NLO to install EOP PTMs 97 through 100 for RHR injection valve BOP control.

Role Play-Booth When directed by BOP to install EOP PTMs97-100, wait 3 minutes then Operator put in the overrides for the PTMs. Report back to BOP when PTMs installed.

Critical Inject SLC before BIIT curve exceeded.

Task RO When directed by the CRS inject SLC.

1.10 Place both keys in SLC PUMP A and SLC PUMP B keylock switches on Panel 9-5 and place switches to START.

1.11 Check both SLC pumps start.

1.12 Check white SQUIB VALVE READY DS-3A (1106A) and SQUIB VALVE READY DS-3B (1106B) lights turn off (Panel 9-5).

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior 1.13 Check pressure on SLC-PI-65, PUMP PRESSURE (Panel 9-5), is greater than reactor pressure.

1.14 Check Annunciator 9-5-2/G-7, LOSS OF CONT TO SQUIB VLVS, alarms.

1.15 Ensure RWCU-MO-15, INBD ISOL VLV (Panel 9-4), is closed.

1.16 Ensure RWCU-MO-18, OUTBD ISOL VLV (Panel 9-4), is closed.

1.17 Ensure both RWCU pumps are off (Panel 9-4).

1.18 Ensure RWCU-MO-74, DEMIN SUCTION BYPASS VLV (Panel 9-4), is throttled open.

RO Provide CRS initial SLC tank level.

BOP Report RPV level trend as it lowers.

When RPV level lowers to less than -60 inches, maintain RPV level between

- 60 inches and -183 inches using EOP 5.8.13 systems.

BOP Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior Direct BOP stabilize RPV pressure below 1050 psig:

CRS Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior Open SRVs while maintaining Main Turbine Bypass valves open to control BOP RPV pressure.

Monitor and report torus level and temperature as SRVs are utilized for BOP pressure control. Announce EOP 3A entry conditions as necessary.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior Direct suppression pool cooling placed into service when torus temperature CRS rises above 95°F.

Use RHR and place into suppression pool cooling (after EOP PTMs97-100 are installed) per the HARD CARD.

1.19 Place RHR SW System in service:

1.19.1 Start SWBP(s).

1.19.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

1.20 If required, with CRS permission, place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD.

1.21 If required, place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

BOP 1.22 Open RHR-MO-39A(B).

1.23 If reactor pressure 300 psig and injection not desired, close RHR-MO-27A(B), OUTBD INJECTION VLV.

NOTE - If directed by EOP 3A, maximize cooling.

1.24 Ensure RHR PUMP running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

1.25 Throttle RHR-MO-34A(B), as required to obtain desired cooling flow.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 7 Event

Description:

40% ATWS Time Position Applicants Action or Behavior 1.26 Throttle RHR-MO-66A(B), as required to obtain desired cooling rate.

1.27 If PCIS Group 6 lights lit on Panel 9-5, ensure one of following open:

1.27.1 REC-MO-711; or 1.27.2 REC-MO-714.

If additional cooling required, initiate cooling in non-running RHR Loop and start additional pumps.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 8 Event

Description:

Multiple manual scrams and control rod insertion to insert all control rods.

Time Position Applicants Action or Behavior Insert control rods per EOP5.8.3 The RO will insert approximately 10-15 control rods manually prior to bypassing the RPS trips.

RO RO Place both CRD Pumps in service Ensure CRD-FC-301 is in Manual to maintain drive water d/p approximately RO 265 psid.

Selects the rods starting in the center and works out in a spiral pattern using RO the 5.8.3 Board depicted below.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 8 Event

Description:

Multiple manual scrams and control rod insertion to insert all control rods.

Time Position Applicants Action or Behavior RO Bypasses RPS trips by installing EOP PTMs 31 through 34 in Panel 9-15 and 9-16.:

1.28 Install EOP PTM Number 31 by installing a jumper between Terminals DD-41 and DD-42 (BAY-1, PNL 9-15).

1.29 Install EOP PTM Number 32 by installing a jumper between RO Terminals BB-41 and BB-42 (BAY-3, PNL 9-15).

1.30 Install EOP PTM Number 33 by installing a jumper between Terminals DD-41 and DD-42 (BAY-1, PNL 9-17).

1.31 Install EOP PTM Number 34 by installing a jumper between Terminals BB-41 and BB-42 (BAY-3, PNL 9-17).

RO Direct NLO to defeat ARI trips by installing EOP PTMs 60 and 61.

Role Play-Booth Operator When directed to install EOP PTMs 61 and 62, wait 2 minutes and then install them and report to RO they are installed.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 8 Event

Description:

Multiple manual scrams and control rod insertion to insert all control rods.

Time Position Applicants Action or Behavior When RO resets the scram lower malfunction RD02a and RD02b Booth severity to 25%. On the second scram reset Delete malfunctions to Operator allow all control rods to be fully inserted.

RO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 8 Event

Description:

Multiple manual scrams and control rod insertion to insert all control rods.

Time Position Applicants Action or Behavior RO Critical The RO resets the reactor scram and continues re-scrams to lower Task power.

RO Resets the scram and repeats scrams as rod motion is verified Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 3 Event No.: 8 Event

Description:

Multiple manual scrams and control rod insertion to insert all control rods.

Time Position Applicants Action or Behavior RO Report all control rods fully inserted.

CRS Order SLC injection stopped.

RO Place both SLC Pump control switches to OFF.

CRS Exits EOP 6A and 7A and enters EOP 1A.

END OF EVENT Notes When all the control rods have been inserted and RPV level is within band, stop the scenario as directed by the lead examiner.

END OF SCENARIO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Simulator Setup Initialize the simulator in IC20 (EOL)

Triggers and Malfunctions E1 - RR17b at 22% with a ramp of 2 minutes.

E2 - FW01a, local trip of RFPT A.

E3 - None E4 - RP03a, RPS EPA 1A1/1A2 trip.

E5 - CU01a, RWCU Pump A seal failure E6 - ED9a, 480V Bus 1A trips

- RP12, PCIS Group 3 Failure (Active)

- RD02a at 95% (Active)

- RD02bat 95% (Active)

Overrides

ZDITECSWTECPC(2)=STOP, TEC Pump C Control Switch (Active)

ZLOTECSWTECPC(2)=OFF, TEC Pump C Red Light off. (Active)

ZLOTECSWTECPC(1)=OFF, TEC Pump C Green Light off.(Active)

Panel Set-up

Ensure PMIS IDTs are blank.

Place the Shutdown EOL Rod Sequence Book on Panel 9-5.

Lower reactor to power to 75% with Reactor Recirculation pumps.

Ensure Recirculation Controllers are selected to AP@

Procedures Needed Tags Hung TEC Pump C c/s Rev 1

Appendix D Required Operator Actions Form ES-D-2 Turnover Sheet:

Plant Status: The plant is operating at approximately 75% at the end of the current fuel cycle.

Risk: Green Activities in Progress: None LCOs in effect: None Equipment out of service: TEC Pump C tagged out for bearing replacement.

Activities for the Shift: Continue with plant startup.

IAC will be performing 6.1APRM.305, APRM System (Flow Bias and Startup) Channel Calibration (DIV 1) later in the shift.

Pump Vibration PMs being performed on RFP-1A A-2 oil pump.

Rev 1

Appendix D Scenario Outline Form ES-D-1 Facility: _Cooper Nuclear Station_______ Scenario No.: 4 Op-Test No.: ___1____

Examiners: ____________________________ Operators: _____________________________

Initial Conditions:

Startup in the heating range (Mode 2) <5% power after the plant scrammed on a turbine trip after a 42 day operation following the latest refueling outage.

Turnover:

On Steps 4.18.2.5 and 5.35 of Procedure 2.1.1. Continue the startup. Condensate pump 1A breaker tagged out for lubrication.

Event Malf. Event Event No. No. Type* Description 1 N/A N, R Continue withdrawing control rods.

(ATC) 2 NM07a I, T IRM A full in signal lost causing control rod withdrawal block.

ATC bypasses IRM.

(ATC)

CRS declares IRM inoperable per TRM TLCO 3.3.1 3 HV02A N/A Minor earthquake. CRS enters 5.1QUAKE 4 SW11d C, T REC Pump D trips. BOP starts standby pump.

(BOP) CRS declares tripped REC pump inoperable per TS LCO 3.7.3 5 AD06c/ C SRV fails open and the tailpipe fails pressurizing torus air space. Abnormal Procedure 2.4SRV AD10c (BOP)

CRS Enters EOP-3A on high drywell temperature.

6 ED04 M Loss of all AC power.

(BOP) BOP manually starts and loads both DGs.

(BOP) CRS re-enters EOP-3A on low SP level. HPCI prevented from operating at 11 foot torus level.

Appendix D Scenario Outline Form ES-D-1 Initial Conditions: Plant is starting up from a scram at the beginning of core life. Reactor power

<5% and on IRM range 9. Reactor pressure is 650 psig with main turbine bypass valves controlling pressure.

Turnover: Continue with startup.

The ATC continues withdrawing control rods to raise reactor power. BOP is controlling heatup rate with the bypass valves.

IRM A full-in indication lost to the reactor manual control system resulting in a control rod withdrawal block. The ATC bypasses the IRM to clear the rod block. CRS declares IRM A inoperable per TRM TLCO 3.3.1.

A minor earthquake takes place and the CRS enters Emergency Procedure 5.1QUAKE.

REC Pump D trips and the BOP starts another REC pump to maintain system pressure. The CRS declares REC Pump D inoperable per TS LCO 3.7.3.

SRV C inadvertently opens and its tailpipe breaks in the torus air space. The CRS enters Abnormal Procedure 2.4SRV. Attempts to close the valve from the main control room fail. The CRS enters EOP-3A on high drywell temperature. When the fuses for SRV E are pulled, the valve closes.

A loss of all AC occurs. The CRS enters Abnormal Procedure 5.3SBO. Both Diesel Generators fail to automatically start. The BOP manually starts and loads both DGs (critical task). The CRS transitions to 5.3EMPWR, REC and Instrument Air recovery are prioritized.

A torus water leak develops requiring the CRS to re-enter EOP-3A on low SP level.

When torus water level reaches 11 feet, HPCI auxiliary oil pump must be placed in PTL (critical task).

When torus water level reaches 9.6 feet, emergency depressurization is required.

The BOP opens SRVs to ED (critical task). The ATC controls RPV water level during ED.

The scenario ends once ED is complete and RPV water level is being controlled in the desired band.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 1 Event

Description:

Continue withdrawing control rods Time Position Applicants Action or Behavior Take Reactivity Control Supervisor position per Procedure 2.0.3:

1. The CRS may fulfill this function in the following situations:

a. Control rod manipulations resulting in power changes < 25%

over a shift.

2. A SRO with no concurrent duties shall assume the role of Reactivity Manager during significant reactivity manipulations or power changes including plant startup or shutdown, or any power changes 25% with control rods.

a.The Reactivity Manager shall provide direct oversight of all reactivity manipulations performed by CRO-Reactor Operator (CRO-RO) and ensure all requirements and expectations associated with the standard for reactivity manipulation with control rods are carried out by CRO-RO and Reactor Engineering personnel.

b.For abnormal conditions, the CRS, when acting as the CRS Reactivity Manager, has to balance the priorities of new events against the importance of reactivity changes. It is recognized that the CRS has multiple duties and responsibilities during a transient or abnormal, emergency, or EOP condition. The CRS has at his discretion the level of the Reactivity Manager function he will perform during these emergent conditions.

3. The Reactivity Manager shall reinforce proper rod selection and movement by ensuring that the CRO-RO utilizes the standard protocol per Section 7.6 Standard Protocol for Non-EOP Control Rod Movement.

a.The Reactivity Manager shall ensure CRO-RO follow conservative operating practices and their actions reflect safety and core integrity taking precedence over power production.

b.The Reactivity Manager shall notify CRS/SM immediately of the occurrence of reactivity events as described in Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 1 Event

Description:

Continue withdrawing control rods Time Position Applicants Action or Behavior Attachment 1 of Procedure 0-CNS-61.

c.The Reactivity Manager shall lead the pre-activity briefing in a manner that is consistent with Management's expectations.

4. The Reactivity Manager shall ensure control rod manipulations or other reactivity change activities are not jeopardized by Control Room distractions.

a.If distractions occur, then control rod manipulations or other reactivity changes are stopped until the distraction is resolved.

RO ACTIONS FOR MOVING A CONTROL ROD

b. Acknowledge intended target rod position by circling the position in the "TO" column on the Rod Movement Sheet; N/A if not using Rod Movement Sheet.

c. Point to and announce the control rod intended to be selected. Wait for verification from Concurrent Verifier.

d. Select the desired control rod.

e. Verify the desired control rod is the only control rod selected.

f. Place hand(s) on the Rod Movement Control RO Switch/Emergency Override Switch in such a manner as to allow the Concurrent Verifier to verify the switch(es) will be properly positioned to cause the intended rod movement.

g. Announce to the Concurrent Verifier the control rod selected and the intended movement of the control rod including the final position and method of movement (continuous or notch, withdrawal or insertion).

h. After receiving concurrence from the Concurrent Verifier, position the Rod Movement Control Switch/Emergency Override Switch as specified to cause the intended rod motion.

i. Prior to deselecting the control rod or selecting the next Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 1 Event

Description:

Continue withdrawing control rods Time Position Applicants Action or Behavior control rod, verify that the final position of the control rod just positioned is correct per the controlling document.

j. Announce to the Concurrent Verifier the final position of the control rod.

k. Annotate completion of each step immediately after its performance.

Withdraw control rods per Procedure 10.13 package:

4.0.2 Proceed with consecutive step order on Attachment 5 in the Control Rod Sequence Package. Continuous rod withdrawal may be used unless otherwise specified by a Reactor Engineer.

4.0.3 The Operator or Concurrent Verifier shall place a check mark in the DESIRED ROD SELECTED block for the selected rod.11 RO 4.0.4 The Operator shall initial in the "PERFORMED BY" column on Attachment 5.

4.0.5 Note to Examiner: Step is N/A.

4.0.6 The Operator or Concurrent Verifier shall place a check mark in the "CC" column if a coupling check was performed.

4.0.7 Note to Examiner: Step is N/A.

4.0.8 Note to Examiner: Step is N/A.

Withdraws control rods controlling main turbine bypass valve position RO 5% to 50% open and controlling RPV heat up rate.

Monitor main turbine bypass valve position and coordinate with ATC BOP reactor heat up.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 1 Event

Description:

Continue withdrawing control rods Time Position Applicants Action or Behavior Notes After the RO has withdrawn control rods to change reactor power or as directed by lead examiner, proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 2 Event

Description:

IRM full in signal lost Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction NM07a, IRM A retract failure (Trigger 2).

Operator Respond to alarm 9-5-1/A-4, Rod Withdrawal Block and report alarm to RO CRS.

RO Report IRM A full in light lost on SRM/IRM Detector Position display.

CRS Direct RO to bypass IRM A.

NOTE to Examiners: The required number of IRM channels are met (6) but the IRM is inoperable.

Reviews TRM TLCO 3.3.1 and declares IRM A inoperable (Table T3.3.1-1 Function 2a).

CRS END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 3 Event

Description:

Minor earthquake Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction HV02 at 25%,

minor earthquake (Trigger 3).

Operator Respond to alarm B-3/B-1, Seismic Event.

BOP 2.1 Enter Procedure 5.1QUAKE.

Enter Emergency Procedure 5.1QUAKE. Assign procedure actions to CRS BOP.

Contact Ft. Calhoun Station Control Room and/or National Earthquake BOP Information Center.

Role Play-As Ft. Calhoun control room personnel, report you have received Booth reports of small ground movement at the station.

Operator As National Earthquake Information Center, report your instrumentation indicates the epicenter to be about 20 miles east of CNS.

Direct NLO to inspect Independent Spent Fuel Storage Installation (ISFSI)

BOP for signs of damage to Horizontal Storage Module (HSM) damage.

Role Play-Booth Operator As NLO, when requested to inspect ISFSI for damage to HSM, wait 2 minutes and report back no visible damage at the site.

Contact NLOs to perform detailed inspection of vital and major equipment BOP with emphasis on CSCS and engineered safeguard systems.

Role Play-Booth Operator As NLO, when requested to inspect equipment, wait 5 minutes and report back no equipment issues have been discovered.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 3 Event

Description:

Minor earthquake Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 4 Event

Description:

REC Pump D trips Time Position Applicants Action or Behavior NOTE to Examiner: Malfunction SW11d, REC Pump D trip activates automatically after a 5 minute time delay.

NOTE to Examiner: REC non-critical loads will isolate in approximately 40 seconds if system pressure remains below 62 psig.

Respond to alarm M-1/B-4:

1.1 Start another REC pump.

BOP 1 .2 Monitor REC pump discharge pressures and ensure valve line-up is correct.

NOTE to Examiner: If REC non-critical loads have isolated, the BOP will respond per REC Restoration HARD CARD.

If REC non-critical loads isolated on low pressure restore system per HARD CARD:

1.1 Ensure low pressure isolation not due to leakage or leak isolated.

1.2 Ensure two REC pumps are running.

1.3 Ensure one of following valves are OPEN:

1.3.1 REC.MO-711, NORTH CRITICAL LOOP SUPPLY.

1.3.2 REC-MO-714, SOUTH CRITICAL LOOP SUPPLY.

BOP 1.4 NOTE to Examiner, Step is N/A.

1.5 Throttle open REC HX outlet valve for a HX that was in service, as necessary, while maintaining REC CRIT LOOP SUPPLY PRESS in green band.

1.5.1 REC.MO.712, HX A OUTLET VLV.

1.5.2 REC.MO-713, HX B OUTLET VLV.

1.6 Start third REC pump.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 4 Event

Description:

REC Pump D trips Time Position Applicants Action or Behavior 1.7 Throttle open REC HX outlet valve, as necessary, to obtain following conditions:

1.7.1 REC CRIT LOOP SUPPLY PRESS >62 psg.

1.7.2 REC HEADER PRESSURE in t of green band.

1.8 Perform following simultaneously:

1.8.1 Open REC-MO-700, NON-CRITICAL HEADER SUPPLY.

1.8.2 Continue throttling open REC HX outlet valve, as necessary, to maintain REC HEADER PRESSURE in green band.

1.9 Ensure REC HX outlet valve full open.

1.10 lf REC-AO-710, RWCU NON-REGEN HX INLET, not closed for leak isolation, open REC-AO-710.

1.11 lf REC-MO-í329, AUGMENTED RADWASTE SUPPLY, not closed for leak isolation and cooling desired, open REC-MO-1329.

1.12 Place DRYWELL REC ISOL VALVE CONTROL switch to AUTO.

Reviews TS LCO 3.7.3 and declares REC Pump D inoperable.

CONDITION REQUIRED ACTION COMPLETION TIME B. One REC B.1 Restore the REC 30 days subsystem subsystem to inoperable for OPERABLE CRS reasons other status.

than Condition A.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 4 Event

Description:

REC Pump D trips Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 5 Event

Description:

SRV C sticks open and tailpipe in torus breaks.

Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunctions AD06c, SRV stuck open, at 100%, and AD10c, tailpipe failure in torus air space, at Operator 25% (Trigger 5).

Respond to alarms 9-3-1/A-2 and 9-3-1/C-1 2.1 Check amber lights on Panel 9-3 to determine which valve(s) has BOP opened.

2.2 Enter Procedure 2.4SRV if valve does not close.

RO Report reactor pressure trend.

Enter Abnormal Procedure 2.4SRV and direct BOP to perform subsequent CRS operator actions.

Take responsibility for scram action of 2.4SRV:

RO Before average suppression pool temperature reaches 110'F, SCRAM and concurrently enter Procedure 2.1.5.

When drywell temperature reaches 150°F, enter EOP 3A.

CRS Direct all drywell cooling to be operated.

BOP At Panel 9-3. place SRV C control switch to OPEN.

RO Monitor power while BOP is attempting to close the SRV.

BOP Note that no testing is taking on SRVs is in-progress.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 5 Event

Description:

SRV C sticks open and tailpipe in torus breaks.

Time Position Applicants Action or Behavior BOP At Panel 9-3, place ADS INHIBIT A and B to INHIB.

BOP At Panel 9-3, place SRV C control switch to AUTO.

RO Report no change in RPV pressure or power.

BOP Report SRV C still open.

BOP Direct NLO to Auxiliary Relay room to pull fuses.

RO Report torus pressure rise, and drywell pressure rise.

CRS Report torus to drywell vacuum breakers open.

Role Play-Booth Operator As NLO reporting to Auxiliary Relay room wait 2 minutes and then call the BOP to report you are in the room.

Direct NLO to pull following fuses:

BOP 2E-F3C 2E-F11C.

Role Play-Booth When directed to pull the fuses, wait one minute then insert trigger E5 Operator (SRV lights off) and delete malfunction AD06c and report to BOP the fuses are pulled.

RO Report RPV pressure rise and reactor power rise.

BOP Report SRV closed.

BOP At Panel 9-3, place ADS INHIBIT A and B to AUTO.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 5 Event

Description:

SRV C sticks open and tailpipe in torus breaks.

Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 6 Event

Description:

Loss of all AC power Time Position Applicants Action or Behavior Booth When directed by the lead examiner, insert malfunction ED04, loss of all AC power (Trigger 6).

Operator CRS Enters Emergency Procedure 5.3SBO.

BOP Report failure of both DGs to start.

Performs Procedure 2.1.5, Attachment 2, Reactor Power Control actions:

1. REACTOR POWER CONTROL 1.1 Ensure REACTOR MODE switch is in SHUTDOWN.

1.2 Verify all SDV vent and drain valves are closed.

1.3 Note to Examiner, Step is N/A.

NOTE - Steps 1.4 and 1.5 may be performed concurrently.

1.4 Verify all control rods are fully inserted.

1.41 If necessary, insert control rods as directed by CRS.

RO 1.5 Observe nuclear instrumentation and perform following:

1.5.1 Note to Examiner, Step is N/A.

1.5.2 Note to Examiner, Step is N/A.

1.5.3 Note to Examiner, Step is N/A.

1.5.4 Note to Examiner, Step is N/A.

1.5.5 Check reactor power is lowering.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 6 Event

Description:

Loss of all AC power Time Position Applicants Action or Behavior Respond to Reactor Scram by reporting following :

Provides scram report:

RO Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior NOTE to Examiner: DG failure to start is already active.

Critical BOP manually starts DGs before suppression pool temperature Task reaches 95°F.

Manually starts both DGs and they automatically load onto their respective BOP buses.

Exits Emergency Procedure 5.3SBO and enters 5.3EMPWR. Prioritizes CRS restoring REC and instrument air.

Role Play-Booth Operator If asked to inspect both DGs, wait 3 minutes and report back that both DGs are operating with all local parameters normal.

Performs Procedure 2.1.5, Attachment 3 Reactor Water Level Control, actions:

1.1 Note to Examiner, Step is N/A.

RO 1.2 Maintain RPV level in prescribed band using following systems, as required, based on plant conditions:

1.2.1 HPCI per Procedure 2.2.33.1.

1.2.2 RCIC per Procedure 2.2.67.1.

Reactor pressure controlled utilizing a combination of HPCI, RCIC and RO SRV control.

BOP Verify Service Water pumps operating to support DG and REC operation.

lf SAC(s) not running:

1.6.1 Place COMPRESSOR 1A control switch to OFF (PANEL A).

BOP 1.6.2 At 480V, SUBSTATION 1F, press TRIP button on Breaker 4C, SAC 1A (Critical Switchgear Room F).

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior 1.6.3 Place COMPRESSOR 1A control switch to AUTO (PANEL A).

When directed to trip air compressor 480V breaker, wait one minute and then insert remote functions Both ED31 Reset 480V Breaker Operator IA04 Air Compressor Reset Then inform the control room that it has been reset.

Restore REC system per 2.2.65.1 HARD CARD:

1.1 Ensure low pressure isolation not due to leakage or leak isolated.

1.2 Ensure two REC pumps are running.

1.3 Ensure one of following valves are OPEN:

1.3.1 REC.MO-711, NORTH CRITICAL LOOP SUPPLY.

1.3.2 REC-MO-714, SOUTH CRITICAL LOOP SUPPLY.

1.4 NOTE to Examiner, Step is N/A.

1.5 Throttle open REC HX outlet valve for a HX that was in service, as necessary, while maintaining REC CRIT LOOP SUPPLY PRESS in green band.

BOP 1.5.1 REC.MO.712, HX A OUTLET VLV.

1.5.2 REC.MO-713, HX B OUTLET VLV.

1.6 Start third REC pump.

1.7 Throttle open REC HX outlet valve, as necessary, to obtain following conditions:

1.7.1 REC CRIT LOOP SUPPLY PRESS >62 psg.

1.7.2 REC HEADER PRESSURE in t of green band.

1.8 Perform following simultaneously:

1.8.1 Open REC-MO-700, NON-CRITICAL HEADER SUPPLY.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior 1.8.2 Continue throttling open REC HX outlet valve, as necessary, to maintain REC HEADER PRESSURE in green band.

1.9 Ensure REC HX outlet valve full open.

1.10 lf REC-AO-710, RWCU NON-REGEN HX INLET, not closed for leak isolation, open REC-AO-710.

1.11 lf REC-MO-í329, AUGMENTED RADWASTE SUPPLY, not closed for leak isolation and cooling desired, open REC-MO-1329.

1.12 Place DRYWELL REC ISOL VALVE CONTROL switch to AUTO.

BOP Contact NLO to ensure SW Zurn strainers returned to service.

Role Play-Booth Operator As NLO contacted to return Zurn strainers to service, wait 2 minutes and report back that both strainers have been returned to service.

Ensure following DC lube oil pumps have started:

1.4.1 Hydrogen AIR SIDE SEAL OIL BACKUP PUMP (PANEL B).

1.4.2 Main Turbine EMERG BEARING OIL PUMP (PANEL B).

BOP 1.4.3 RFPT A and B EMERGENCY OIL PUMPS (PANEL A).

1.4.4 RRMG LUBE OIL PUMP C and LUBE OIL PUMP D (PNL 9-4),

Direct NLO to close following valves to prevent draining CST to hotwell:

1.7.1 MC-807, CST RECIRC THROTTLING VALVE (Rw-877-basement above Condensate Backwash Transfer Pump).

BOP 1.7.2 MC-38, FCV-17 OUTLET (T-882-behind TEC HXs).

1.7.3 MC-777, LCV-2D OUTLET ISOLATlON (T-882-behind TEC HXs).

1.7.4 CM-12, LCV-2C OUTLET (T-882-behind TEC HXs).

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior 1.7.5 CM-18, CONDENSATE PPS SEAL SUPP ROOT (T-882-behind TEC HXs).

1.7.6 DW-462, LT 1A and 1B REFERENCE SUPPLY ROOT (T-882-Condenser Area).

1.7 .7 CM-10, CONDENSATE STORAGE TK OUTLET TO TURB BLDG (near CST A, northeast area of tank).

Re-enter EOP 3A when torus water temperature reaches 95°F.

CRS Role Play-Booth Operator When contacted, use remote functions to close valves directed by BOP.

Place suppression pool cooling in service as necessary to maintain temperature below 95°F. (HARD CARD) 4.1 Place RHR SW System in service:

4.1.1 Start SWBP(s).

RO 4.1.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

4.2 If required, with CRS permission, place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior 4.3 If required, place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

4.4 Open RHR-MO-39A(B).

4.5 If reactor pressure 300 psig and injection not desired, close RHR-MO-27A(B), OUTBD INJECTION VLV.

NOTE - If directed by EOP 3A, maximize cooling.

4.6 Ensure RHR PUMP running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

4.7 Throttle RHR-MO-34A(B), as required to obtain desired cooling flow.

4.8 Throttle RHR-MO-66A(B), as required to obtain desired cooling rate.

4.9 If PCIS Group 6 lights lit on Panel 9-5, ensure one of following open:

4.9.1 REC-MO-711; or 4.9.2 REC-MO-714.

If additional cooling required, initiate cooling in non-running RHR Loop and start additional pumps.

END OF EVENT Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 7 Event

Description:

DGs manually started.

Time Position Applicants Action or Behavior Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 8 Event

Description:

Torus leak Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert malfunction PC08 at 33%

over 3 minutes (Trigger 8).

Operator BOP Report lowering Suppression Pool level.

When SP (torus) water level below -2 in, re-enter EOP 3A.

CRS Direct BOP maintain PC level above 11 ft with following systems, EOP 5.8.14:

RCIC

HPCI

RHR-A

RHR-B

CS-A

CS-B.

Role Play-Booth If sent to investigate damage in the Reactor Building, wait 4 Operator minutes and call back and report that the Suppression Pool has a leak at a weld below the water line and water is pouring out.

Contacts the Reactor Building Station Operator to perform section 10 of BOP Emergency Procedure 5.8.14.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 8 Event

Description:

Torus leak Time Position Applicants Action or Behavior Role Play-Booth Operator Respond as the RB Station Operator that you will perform Section 10 of Emergency Procedure 5.8.14.

Enter Procedure 5.8.14:

If using HPCI and/or RCIC, the system Minimum flow valve will be opened:

HPCI-MO-25 BOP

RCIC-MO-27 If using CS and/or RHR the test return path is used:

RHR-MO-39 and RHR-MO-34

CS-MO-26 When PC water level cannot be maintained above 11 ft, direct BOP to:

CRS

1. Stop and prevent HPCI.

2. Maintain PC water level above 9.6 ft.

BOP Report when PC water level is approaching 11 ft.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 8 Event

Description:

Torus leak Time Position Applicants Action or Behavior Critical Place HPCI Auxiliary Oil Pump in Pull-to-Lock, before PC water level Task reaches 11 ft.

CRS Direct stop and prevent with HPCI.

BOP Place HPCI Aux Oil Pump control switch in Pull-To-Lock.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 9 Event

Description:

Emergency Depressurize on low PC water level.

Time Position Applicants Action or Behavior CRS Direct BOP maintain PC water level above 9.6 ft.

BOP Report when PC water level approaches 9.6 ft.

When PC water level cannot be restored and maintained above 9.6 ft., enter EOP 2A and direct emergency depressurization.

CRS Critical When PC water level cannot be restored and maintained above 9.6 ft.,

Emergency Depressurize the RPV with the SRVs before PC water level Task is below 6 ft.

Direct ED per EOP 2A CRS Rev 1

Appendix D Required Operator Actions Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 9 Event

Description:

Emergency Depressurize on low PC water level.

Time Position Applicants Action or Behavior When directed to ED, verify PC water level is above 6 ft. and open 6 SRVS BOP by taking their control switches to OPEN.

Direct RO maintain RPV level band of -110 inches to -60 inches FZ during CRS ED.

Utilize low pressure systems to control an injection flow rate of a minimum of RO 3000 gpm to 4000 gpm (1.5 to 2 Mlbs/hr) and maintain RPV level in band.

When RPV pressure is < 50 psig above drywell pressure report ED is BOP complete.

END OF EVENT Notes When Emergency Depressurization is complete and RPV level is being controlled in desired band, stop scenario as directed by the lead examiner.

END OF SCENARIO Rev 1

Appendix D Required Operator Actions Form ES-D-2 Simulator Setup Initialize the simulator in IC09 (BOL)

Triggers and Malfunctions E1 - None E2 - NM07a, IRM A full-in position lost E3 - HV02b at 25%, Minor earthquake E3 - SW11d, REC Pump D trip after 5 minute time delay E4 -

E5 - OR ZLOMSSWS1C(1)=OFF, SRV C Green light off.

OR ZLOMSSWS1C(2)=OFF, SRV C Amber light off.

OR ZLOMSSWS1C(3)=OFF, SRV C Red light off.

E6 - ED04, loss of all AC power.

E7 -

E8 - PC08 at 33% ramp over 3 minutes, suppression pool leak Overrides

ZDIMCSWCPA(2)=STOP, CP A control switch to STOP.

ZLOMCSWCPA(3)=OFF, CP A red light OFF.

ZLOMCSWCPA(1)=OFF, CP A green light OFF.

Panel Set-up

Ensure PMIS IDTs are blank.

Place the Shutdown BOL Rod Sequence Book on Panel 9-5

Ensure Recirc Controllers are selected to AP@

Ensure REC Pump D operating.

Procedures Needed On Steps 4.18.2.5 and 5.35 of Procedure 2.1.1 (Mark previous steps complete as needed)

Tags Hung CP A c/s caution tag.

CP and CBP discharge valve tags.

Rev 1

Appendix D Required Operator Actions Form ES-D-2 Turnover Sheet:

Plant Status: The plant performing a startup from a refueling outage. The IRMs are on Range

9. Reactor pressure is 650 psig with main turbine bypass valves controlling pressure.

Risk: Green Activities in Progress: Raising power with control rods and controlling RPV heatup.

LCOs in effect: None Equipment out of service: Condensate pump 1A breaker tagged out for lubrication.

Expected completion time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Activities for the Shift: Continue with plant startup. Currently on Steps 4.18.2.5 and 5.35 of Procedure 2.1.1. Condensate pump 1A breaker tagged out for lubrication.

Rev 1

ML12297A450

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