Text

2017-03 Proposed Operating Test
	ML17122A080
Person / Time
Site:	Cooper
Issue date:	03/16/2017
From:	Vincent Gaddy; Operations Branch IV
To:	; Nebraska Public Power District (NPPD)
References
	Download: ML17122A080 (400)
	v • d • e

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 1 of 16 Cooper Nuclear Station Job Performance Measure JPM A1

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 2 of 16 (JPM A1)

Perform Jet Pump Operability Check (RO)

Revision Statement: This is a simplified format of bank JPM SKL-034-50-65 Rev 2.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: RO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A: 2.1.25 (3.9/4.2)

General

References:

1. Procedure 6.LOG.601, Daily Surveillance Log - Modes 1, 2, And 3 General Tools and Equipment:

1. Calculator

2. Jet Pump Operability Curves

3. Procedure 2.1.10, (Attachment 1) Power-To-Flow map.

Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant correctly recorded Jet Pump, RR, core flow parameters, averaged Jet Pump Ps and determined Jet Pump #13 falls outside the limit and is marked UNSAT for Check 3 in accordance with Procedure 6.LOG.601.

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 3 of 16 (JPM A1)

Directions to Examiner:

1. This JPM evaluates the applicant'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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2, Jet Pump curves and Procedure 2.1.10, Attachment 1, Power to Flow Map.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 4 of 16 Read the following to the JPM performer.

General Conditions:

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

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

Initiating Cue(s):

You have been directed to perform the daily Jet Pump (6.LOG.601 Attachment 12) and Recirc Pump Flow Check (6.LOG.601 Attachment 13) as part of the routine shift activities. Notify the CRS when the task is complete.

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 5 of 16 PERFORMANCE:

Start Time: ____________

1. Procedure Step: A. Record indicated core flow.

Standard Recorded Core flow from Recorder NBI-DPR/FR-95 (Blue Pen).

Cue Notes Results SAT UNSAT

2. Procedure Step: B. Record RR pump flow for both loops.

Standard Recorded RR pump flow, from RR-FR-163 for Pumps A & B.

Cue Notes Results SAT UNSAT

3. Procedure Step: C. Record RRMG Set Speed for both loops.

Standard Recorded RRMG Set speed from the following:

a. RRFC-SI-1A for RRMG A

b. RRFC-SI-1B for RRMG B Cue Notes Results SAT UNSAT

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 6 of 16

4. Procedure Step: D. Record JP Flow for both loops.

Standard Recorded Jet Pump Flow from the following:

a. NBI-FI-92A for LOOP A

b. NBI-FI-92B for LOOP B Cue Notes Results SAT UNSAT

5. Procedure Step: Record Jet Pump P (%) for all 20 Jet Pumps.

Standard Record differential pressures from individual jet pump instruments NBI-FI-78A through NBI-FI-78Z on Panel 9-38 in control room.

Cue Notes Results SAT UNSAT

6. Procedure Step: Averaged Loop B and Loop A flows.

Standard Added JP #1 through 10 and divided by 10 for LOOP B, then added JP #11 through 20 and divided by 10 for LOOP A (+/- 2 from number in Key).

Cue Notes Results SAT UNSAT

7. Procedure Step: Check 1 Verified B and C values within curve limits for both loops.

Standard Determined the values recorded in Items B and C are within the limits of the curve (SAT checked for both Loop A and B).

Cue Notes Results SAT UNSAT

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 7 of 16

8. Procedure Step: Check 2 Verified C and D values within curve limits.

Standard Determined the values recorded in Items C and D are within or outside the limits of the curve (SAT checked for Loop A and B).

Cue Notes Results SAT UNSAT

9. Procedure Step: Check 3 Verified Jet Pump P differ by 20% from curves.

Standard Determined Jet Pump p differs by 20% from established patterns for (UNSAT checked for Loop A, SAT checked for Loop B).

Cue Notes Results SAT UNSAT

10. Procedure Step: VERIFY Checks 1 and 2 SAT or Check 3 SAT Standard Verified check 1 and check 2 SAT or check 3 is SAT (Check 3 is SAT).

Cue Notes Results SAT UNSAT

11. Procedure Step: VERIFY Core Flow, NBI-DPR/FR-95, value is not in Stability Exclusion Region of Power to Flow Map.

Standard Determined operation is outside Stability Exclusion Region of Power to Flow Map using core flow value recorded in Item A on previous page and 2.1.10 Power to Flow Map Cue Notes Results SAT UNSAT

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 8 of 16

12. Procedure Step: Next check is N/A.

Standard Cue Notes Results SAT UNSAT

13. Procedure Step: VERIFY JP Flow, NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is 3.67x106 lbs/hr at 51.45x106 lbs/hr Rated Core Flow Standard Determined Item D values (previous page of 6.LOG.601) for Loop A and Loop B flow mismatch is 3.67x106 lbs/hr (SAT entered into block).

Cue Notes Results SAT UNSAT

14. Procedure Step: N/A Standard Informed Control Room Supervisor that the daily Jet Pump and Recirculation Pump Flow Check is Complete and Jet Pump 13 d/p is low out of specification.

Cue None Notes Results SAT UNSAT Stop Time: __________

DO NOT GIVE TO APPLICANTS ATTACHMENT 12 JET PUMP OPERABILITY ATTACHMENT 1

1. IF in single loop operation, THEN MARK idle loop N/A.

JET PUMP P (%)

PARAMETERS LOOP A LOOP B JP # LOOP B JP # LOOP A A Core Flow (106 lb/hr) NBI-FRDPR-95 62 1 31 11 32 B RR Pump Flow (103 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 D JP Flow (106 lb/hr) NBI-FI-92A/B 31 31 4 31 14 32 5 36 15 34 KEY 6 34 16 35 7 33 17 32 8 32 18 31 9 33 19 33 10 33 20 32 LOOP B LOOP 32.6 31.1 Avg A Avg OPERABILITY APPLICABLE ATT. 22 CHECKS SAT UNSAT LIMIT MODE NOTE VERIFY B and C values Loop A:

1 within curve limits (a) Loop B:

VERIFY C and D values Loop A:

2 SAT within curve limits (a) Loop B:

Loop A: 1 (b), 2 (b) 50 VERIFY Jet Pump P differ 3

by 20% from curves (a,c) Loop B:

VERIFY Checks 1 and 2 SAT or Loop A:

SAT Check 3 SAT Loop B:

(a)

REFER to Jet Pump Operability Curves maintained in Control Room.

(b)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after greater than 25% RTP and 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, THEN immediately NOTIFY Reactor Engineering.

DO NOT GIVE TO APPLICANTS ATTACHMENT 13 RECIRC PUMP FLOW ATTACHMENT 1 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 CHECKS SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES VERIFY Core Flow, MCO NBI-DPR/FR-95, value is not in Stability SAT SAT 1, 2 49 Exclusion Region of Power to Flow Map (a)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B (c) (c) flow mismatch is N/A SAT 1 ,2 48 7.35x106 lbs/hr at

< 51.45x106 lbs/hr Rated Core Flow (b,d)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is SAT SAT 1 (c)

,2 (c) 48 3.67x106 lbs/hr at 51.45x106 lbs/hr Rated Core Flow (b,d)

(a)

REFERENCE power to flow map in Procedure 2.1.10.

(b)

NOTE - Per Technical Specification, a recirculation loop is considered not in operation when total jet pump flows of the two loops is greater than mismatch limits.

1. IF greater than mismatch limit, THEN ENSURE loop with lower flow considered not in operation.

(c)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops in service.

(d)

IF flow mismatch results in one loop considered not in operation, THEN CONTACT FRED to manually insert single loop operation limits into GARDEL.

KEY

DO NOT GIVE TO APPLICANTS ATTACHMENT 1 0700-1000 1900-2200 READING READING OPERABILITY APPLICABLE ATT. 22 CHECK SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES 3 (when reactor VERIFY RR pump pressure less operating or RHR N/A SAT than SDC 66 pump operating in pressure SDC permissive)

(e)

(e) APPLY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure less than shutdown cooling permissive pressure.

KEY

ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

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

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

INITIATING CUE:

You have been directed to perform the daily Jet Pump (6.LOG.601 Attachment 12) and Recirc Pump Flow Check (6.LOG.601 Attachment 13) as part of the routine shift activities. Notify the CRS when the task is complete.

ATTACHMENT 12 JET PUMP OPERABILITY ATTACHMENT 2

2. IF in single loop operation, THEN MARK idle loop N/A.

JET PUMP P (%)

PARAMETERS LOOP A LOOP B JP # LOOP B JP # LOOP A A Core Flow (106 lb/hr) NBI-FRDPR-95 1 11 B RR Pump Flow (103 gpm) RR-FR-163 2 12 C RRMG Set Speed (%) RRFC-SI-1A/B 3 13 D JP Flow (106 lb/hr) NBI-FI-92A/B 4 14 5 15 6 16 7 17 8 18 9 19 10 20 LOOP B LOOP Avg A Avg OPERABILITY APPLICABLE ATT. 22 CHECKS SAT UNSAT LIMIT MODE NOTE VERIFY B and C values Loop A:

1 within curve limits (a) Loop B:

VERIFY C and D values Loop A:

2 SAT within curve limits (a) Loop B:

Loop A: 1 (b), 2 (b) 50 VERIFY Jet Pump P differ 3

by 20% from curves (a,c) Loop B:

VERIFY Checks 1 and 2 SAT or Loop A:

SAT Check 3 SAT Loop B:

(a)

REFER to Jet Pump Operability Curves maintained in Control Room.

(b)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after greater than 25% RTP and 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, THEN immediately NOTIFY Reactor Engineering.

ATTACHMENT 13 RECIRC PUMP FLOW ATTACHMENT 2 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 CHECKS SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES VERIFY Core Flow, MCO NBI-DPR/FR-95, value is not in Stability SAT 1, 2 49 Exclusion Region of Power to Flow Map (a)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is SAT 1 (c)

,2 (c) 48 7.35x106 lbs/hr at

< 51.45x106 lbs/hr Rated Core Flow (b,d)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is SAT 1 (c)

,2 (c) 48 3.67x106 lbs/hr at 51.45x106 lbs/hr Rated Core Flow (b,d)

(a)

REFERENCE power to flow map in Procedure 2.1.10.

(b)

NOTE - Per Technical Specification, a recirculation loop is considered not in operation when total jet pump flows of the two loops is greater than mismatch limits.

3. IF greater than mismatch limit, THEN ENSURE loop with lower flow considered not in operation.

(c)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops in service.

(d)

IF flow mismatch results in one loop considered not in operation, THEN CONTACT FRED to manually insert single loop operation limits into GARDEL.

ATTACHMENT 2 ATTACHMENT 13 RECIRC PUMP FLOW 0700-1000 1900-2200 READING READING OPERABILITY APPLICABLE ATT. 22 CHECK SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES 3 (when reactor VERIFY RR pump pressure less operating or RHR SAT than SDC 66 pump operating in pressure SDC permissive)

(e)

(e) APPLY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure less than shutdown cooling permissive pressure.

CNS NRC Examin3/2017 JPM A1 Rev 0 Page 16 of 16 ATTACHMENT 2 PLANT DATA Indicated core flow (106 lb/hr) Recorder NBI-DPR/FR-95 (Blue Pen) = 62 Rx Recirc pump flow (103 gpm) RR-FR-163 for Pumps A & B; A = 38.5; B =38.5 RRMG Set speed RRFC-SI-1A for RRMG A = 79 RRFC-SI-1B for RRMG B = 80 Jet Pump Flow NBI-FI-92A for LOOP A = 31 NBI-FI-92B for LOOP B = 31 Jet Pump Differential Pressure NBI-FI-78A through NBI-FI-78Z 1 = 31 11 = 32 2 = 32 12 = 30 3 = 31 13 = 20 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

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 1 of 16 Cooper Nuclear Station Job Performance Measure NRC A2

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 2 of 16 (JPM A2)

Perform SLC Operability Checks Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: RO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A: 2.1.20 (4.6/4.6)

General

References:

1. Procedure 6.LOG.601, Daily Surveillance Log - Modes 1, 2 and 3 General Tools and Equipment:

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

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determined SLC Squib continuity was SAT, determined SLC tank level, determined SLC tank level Operability Limit, and completed appropriate sections of Procedure 6.LOG.601, Attachment 9.

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 3 of 16 (JPM A2)

Directions to Examiner:

1. This JPM evaluates the applicant's ability to perform SLC operability checks including determining the tank volume from boron concentration.

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2 and 3. NOTE: Attachment 4 will be handed out when requested during the performance of the JPM.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 4 of 16 Read the following to the JPM performer.

General Conditions:

1. The plant is in Mode 1.

2. Local and remote SLC tanks levels are the same.

Initiating Cue(s):

You have been directed to perform day shift SLC Operability checks, including MCO readings, per 6.LOG.601. Turn in your paperwork to the evaluator after the checks are complete.

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 5 of 16 PERFORMANCE:

Start Time: ____________

NOTE to Examiner: Provide Attachment 2 (1 Page) and Attachment 3 (2 Pages) to applicant to begin.

1. Procedure Step: Verify Continuity of SLC-SQUIB-1106A Explosive Charge Standard Circled ON in Table.

Cue Notes Results SAT UNSAT

2. Procedure Step: Verify Continuity of SLC-SQUIB-1106B Explosive Charge Standard Circled ON in Table.

Cue Notes Results SAT UNSAT

3. Procedure Step: Record SLC Boron Solution Tank Volume Standard Recorded 73% volume.

Cue Notes Allowable range is 72% to 73.9%.

Results SAT UNSAT

4. Procedure Step: (c)1. Mark one of following check boxes: Data Obtained from SLC-LI-66 (PNL 9-5 preferred).

Standard [] Data obtained from SLC-LI-66 (PNL 9-5, preferred).

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 6 of 16

5. Procedure Step: (c) 2 Step is N/A Standard Cue Notes Results SAT UNSAT

6. Procedure Step: (d) 1 If OPERABILITY LIMIT exceeded, THEN DETERMINE calculated Lr per Attachment 21, Section 3.

Standard Transitioned to Attachment 21, Section 3.

Cue Notes Results SAT UNSAT NOTE to Examiner: When asked for 6.SLC.601, provide Attachment 4 (3 Pages) to applicant.

7. Procedure Step: Attachment 21, Step 3.1.1 Perform one of following: CONTACT Surveillance Coordinator to obtain SLC tank concentration from latest performed Procedure 6.SLC.601.

Standard Requested 6.SLC.601 from Surveillance Coordinator.

Cue Notes Results SAT UNSAT

8. Procedure Step: 3.2 Record SLC tank concentration.

Standard Recorded 15.5 Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A2 Rev 0 Page 7 of 16

9. Procedure Step: 3.3 Using recorded SLC tank concentration and TS Figure 3.1.7-1 OBTAIN Minimum Net Tank Volume (gallons).

Standard Obtained volume in gallons from TS Figure 3.1.7-1 Cue Notes Results SAT UNSAT

10. Procedure Step: 3.4 RECORD obtained gallons. _________Minimum Net Tank Volume.

Standard Recorded 3250 gallons.

Cue Notes Acceptable range is 3150 to 3350 gallons.

Results SAT UNSAT

11. Procedure Step: 3.5 CALCULATE SLC tank level OPERABILITY LIMIT (Lr):

Standard Calculated 3250/45.65=71.2%

Cue Notes Acceptable range is 69% to 73.4%.

Results SAT UNSAT

12. Procedure Step: 3.6 RECORD Lr in Attachment 9, OPERABILITY LIMIT cell of SLC Boron Solution Tank Volume table.

Standard Recorded 71.2% in OPERABILITY LIMIT.

Cue Notes Acceptable range is 69% to 73.4%.

Results SAT UNSAT

13. Procedure Step: Step is N/A Standard Handed completed documents to evaluator.

Cue Notes Results SAT UNSAT Stop Time: __________

Attachment 9 SLC ATTACHMENT 1 KEY ATTACHMENT 9 SLC 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 LOC CHECK READING READING LIMIT MODES NOTES VERIFY MCO PNL Continuity of (a)

ON/OFF ON/OFF Light ON 1, 2, 3 109 9-5 SLC-SQUIB-1106A Explosive Charge VERIFY MCO PNL Continuity of (b)

ON/OFF ON/OFF Light ON 1, 2, 3 109 9-5 SLC-SQUIB-1106B Explosive Charge (a)

IF white SQUIB VALVE READY DS-3A light off, THEN ENSURE SLC-MREL-67A (back of Panel 9-5) indicates 3 to 5 milliamps.

(b)

IF white SQUIB VALVE READY DS-3B light off, THEN ENSURE SLC-MREL-67B (back of Panel 9-5) indicates 3 to 5 milliamps OPERABILITY APPLICABLE ATT. 22 PARAMETER VALUE LIMIT MODES NOTES 74% (d)

SLC Boron Solution or calculated 1, 2, 3 3 Tank Volume (c) 73 Lr

[ 71.2 %]

(c)

1. MARK one of following check boxes:

[ ] Data obtained from SLC-LI-66 (PNL 9-5, preferred).

[] Data obtained from SLC-LI-46 (R-976-E, Rack 25-19), Reactor Building Log.

IF data from Reactor Building Log, THEN SM/CRS ENSURE data transferred correctly.

SM/CRS Initials:

(d)

NOTE - SLC OPERABILITY LIMIT in table based on SLC Tank level administrative limits.

1. IF OPERABILITY LIMIT exceeded, THEN DETERMINE calculated Lr per Attachment 21, Section Error!

Reference source not found..

PROCEDURE 6.LOG.601 REVISION 123 PAGE 23 OF 80

Attachment 21 CONTINGENCY ACTIONS ATTACHMENT 1 Key ATTACHMENT 21 CONTINGENCY ACTIONA

3. CALCULATING SLC TANK LEVEL OPERABILITY LIMIT 3.1 PERFORM one of following:

3.1.1 CONTACT Surveillance Coordinator to obtain SLC tank concentration from latest performed Procedure 6.SLC.601.

3.1.2 ACCESS latest performed Procedure 6.SLC.601 as follows:

3.1.2.1 At Control Room key depository, OBTAIN Key

27.

3.1.2.2 ACCESS Surveillance Coordinator office.

3.1.2.3 Using Key #27, OBTAIN SLC tank concentration from latest performed Procedure 6.SLC.601.

3.2 RECORD SLC tank concentration.

15.5% % Sodium Pentaborate Solution by Weight 3.3 Using recorded SLC tank concentration and TS Figure 3.1.7-1, OBTAIN Minimum Net Tank Volume (gallons).

3.4 RECORD obtained gallons. 3250___ Minimum Net Tank volume 3.5 CALCULATE SLC tank level OPERABILITY LIMIT (Lr):

NOTE - SLC tank conversion from gallons to % equals 45.65 gallons/%.

Lr = 3250 /45.65 = 71.2 %

Minimum Net Tank Volume 3.6 RECORD Lr in Attachment 9, OPERABILITY LIMIT cell, of SLC Boron Solution Tank Volume table.

PROCEDURE 6.LOG.601 REVISION 123 PAGE 58 OF 80

(JPM A2)

ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is in Mode 1.

2. Local and remote SLC tanks levels are the same.

INITIATING CUE:

You have been directed to perform day shift SLC Operability checks, including MCO readings, per 6.LOG.601. Turn in your paperwork to the evaluator after the checks are complete.

ATTACHMENT 3 (Page 1 of 2)

CNS NRC Examin3/2017 JPM A2 Rev 0 Page 12 of 16 ATTACHMENT 3 (PAGE 2 OF 2)

ATTACHMENT 9 SLC 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 LOC CHECK READING READING LIMIT MODES NOTES VERIFY MCO PNL Continuity of (a)

ON/OFF ON/OFF Light ON 1, 2, 3 109 9-5 SLC-SQUIB-1106A Explosive Charge VERIFY MCO PNL Continuity of (b)

ON/OFF ON/OFF Light ON 1, 2, 3 109 9-5 SLC-SQUIB-1106B Explosive Charge (a)

IF white SQUIB VALVE READY DS-3A light off, THEN ENSURE SLC-MREL-67A (back of Panel 9-5) indicates 3 to 5 milliamps.

(b)

IF white SQUIB VALVE READY DS-3B light off, THEN ENSURE SLC-MREL-67B (back of Panel 9-5) indicates 3 to 5 milliamps OPERABILITY APPLICABLE ATT. 22 PARAMETER VALUE LIMIT MODES NOTES 74% (d)

SLC Boron Solution or calculated Lr 1, 2, 3 3 Tank Volume (c)

[ %]

(c)

1. MARK one of following check boxes:

[] Data obtained from SLC-LI-66 (PNL 9-5, preferred).

[] Data obtained from SLC-LI-46 (R-976-E, Rack 25-19), Reactor Building Log.

IF data from Reactor Building Log, THEN SM/CRS ENSURE data transferred correctly.

SM/CRS Initials:

(d)

CNS NRC Examin3/2017 JPM A2 Rev 0 Page 13 of 16 NOTE - SLC OPERABILITY LIMIT in table based on SLC Tank level administrative limits.

2. IF OPERABILITY LIMIT exceeded, THEN DETERMINE calculated Lr per Attachment 21, Section Error! Reference source not found..

ATTACHMENT 4 (Page 1 of 3)

ATTACHMENT 21 CONTINGENCY ACTIONA

3. CALCULATING SLC TANK LEVEL OPERABILITY LIMIT 3.1 PERFORM one of following:

3.1.1 CONTACT Surveillance Coordinator to obtain SLC tank concentration from latest performed Procedure 6.SLC.601.

3.1.2 ACCESS latest performed Procedure 6.SLC.601 as follows:

3.1.2.1 At Control Room key depository, OBTAIN Key #27.

3.1.2.2 ACCESS Surveillance Coordinator office.

3.1.2.3 Using Key #27, OBTAIN SLC tank concentration from latest performed Procedure 6.SLC.601.

3.2 RECORD SLC tank concentration.

% Sodium Pentaborate Solution by Weight 3.3 Using recorded SLC tank concentration and TS Figure 3.1.7-1, OBTAIN Minimum Net Tank Volume (gallons).

3.4 RECORD obtained gallons. ___ Minimum Net Tank volume

CNS NRC Examin3/2017 JPM A2 Rev 0 Page 14 of 16 3.5 CALCULATE SLC tank level OPERABILITY LIMIT (Lr):

NOTE - SLC tank conversion from gallons to % equals 45.65 gallons/%.

Lr = /45.65 = %

Minimum Net Tank Volume 3.6 RECORD Lr in Attachment 9, OPERABILITY LIMIT cell, of SLC Boron Solution Tank Volume table.

CNS NRC Examin3/2017 JPM A2 Rev 0 Page 15 of 16 ATTACHMENT 4 (Page 2 of 3)

CNS NRC Examin3/2017 JPM A2 Rev 0 Page 16 of 16 ATTACHMENT 4 ( Page 3 of 3)

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 1 of 10 Cooper Nuclear Station Job Performance Measure NRC A3

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 2 of 10 (JPM A3)

Determine Isolation Boundaries Revision Statement: This is a simplified format of bank JPM SKL034-50-57 Rev 1 Additional Program Information:

1. Appropriate Performance Location: Classroom

2. Appropriate Trainee level: RO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A: 2.2.13 (4.1/4.3)

General

References:

1. Copy of Procedure 0.9, Tagout General Tools and Equipment:

1. B & R drawing 2040, Sheet 1

2. B & R drawing 2031, Sheet 2 Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant identified the MINIMUM mechanical/piping isolation boundaries required to isolate RHR Pump A for removal per Procedure 0.9, Tagging. The vent and drain paths should be identified but are not critical steps since isolation of RHR Pump A is the required task.

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 3 of 10 (JPM A3)

Directions to Examiner:

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

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2 and B & R drawings 2040 sheet 1 and 2031 sheet 2.

Notes: ______________________________________________________________________

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 4 of 10 Read the following to the JPM performer.

General Conditions:

1. The plant is shut down in a refueling outage.

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

3. Reactor coolant temperature is 120°F.

Initiating Cue(s):

You have been directed to determine the minimum mechanical/piping boundaries to isolate RHR Pump A for removal. RHR A Motor has already been tagged and removed. Highlight all valves required to be included in the clearance order to support this task. Indicate the valve position (open/closed) and whether the valve shall contain a clearance order tag (Tag/No Tag).

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 5 of 10 PERFORMANCE:

Start Time: ____________

1. Procedure Step: May reference Procedure 0.9 Standard Referenced Procedure 0.9 Cue Notes Results SAT UNSAT NOTE to Examiner: Following steps can be completed in any order.

2. Procedure Step: References B&R 2040 Sheet 1.

Standard Reviewed B&R 2040 Sheet 1 for RHR Pump A.

Cue Notes Results SAT UNSAT

3. Procedure Step: Highlights RHR-MO-15A Standard RHR-MO-15A was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

4. Procedure Step: Highlights RHR-MO-13A Standard RHR-MO-13A was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 6 of 10

5. Procedure Step: Highlights RHR-98 Standard RHR-98 was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

6. Procedure Step: Highlights RHR-58 Standard RHR-58 was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

7. Procedure Step: Highlights RHR-11 Standard RHR-11 was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

8. Procedure Step: Highlights drain path Standard RHR-35, RHR 36 and RHR 34 were highlighted and marked No Tag and open.

Cue Notes Results SAT UNSAT

9. Procedure Step: B&R 2031 Sheet 2 Standard Reviewed B&R 2031 Sheet 2 for REC tie in with RHR Pump A.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 7 of 10

10. Procedure Step: Highlights REC-75 Standard REC-75 was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

11. Procedure Step: Highlights REC-76 Standard REC-76 was highlighted and marked tagged and closed.

Cue Notes Results SAT UNSAT

12. Procedure Step: Highlights drain path.

Standard REC-87, 397 and 83 were highlighted No Tag and open.

Cue Notes Results SAT UNSAT

13. Procedure Step: Hands marked up drawings to examiner.

Standard Gave highlighted drawings to the examiner.

Cue Notes Results SAT UNSAT Stop Time: __________

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 8 of 10 DO NOT GIVE TO APPLICANTS Tagged and Closed No Tag and Open Tagged and Closed ATTACHMENT 1 ANSWER KEY B&R 2040 Sheet 1

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 9 of 10 DO NOT GIVE TO APPLICANTS Tagged and Closed No Tag and Open ATTACHMENT 1 ANSWER KEY B&R 2031 Sheet 2

CNS NRC Exam 3/2017 JPM A3 Rev 0 Page 10 of 10 (JPM A3)

ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is shut down in a refueling outage.

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

3. Reactor coolant temperature is 120°F.

INITIATING CUE:

You have been directed to determine the minimum mechanical/piping boundaries to isolate RHR Pump A for removal. RHR A Motor has already been tagged and removed. Highlight all valves required to be included in the clearance order to support this task. Indicate the valve position (open/closed) and whether the valve shall contain a clearance order tag (Tag/No Tag).

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 1 of 11 Cooper Nuclear Station Job Performance Measure NRC A4

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 2 of 11 (JPM A4)

Radiation Protection Table Top Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: RO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A: 2.3.14 (3.4/3.8)

General

References:

1. Procedure 9.ALARA.1, Personnel dosimetry and Occupational Radiation Exposure Program General Tools and Equipment:

1. Calculator

2. RCA Survey Map Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determined the projected dose each worker would receive and that CNS Administrative dose limitations would be exceeded requiring management authorization for Worker #1, in accordance with Procedure 9.ALARA.1.

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 3 of 11 (JPM A4)

Directions to Examiner:

1. This JPM evaluates the applicant's ability to determine of work can be performed within the confines of CNS radiation protection procedures.

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2 (4 Pages).

Notes: ______________________________________________________________________

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 4 of 11 Read the following to the JPM performer.

General Conditions:

1. The plant is operating at power.

2. RHR-MO-25B has developed a leak.

3. Two workers are required to perform the leak repair.

4. Worker #1 has accumulated 600 mrem this year (CNS dose).

5. Worker #2 has accumulated 1415 mrem this year (CNS dose).

6. The following times for each worker have been estimated for task performance:

a. Staging time in area directly outside the CA, at the Step-off Pad, of Rx 903 Angle Valve room:

45 min.

b. Staging time in area directly inside CA, just inside the Step-ff Pad, of Rx 903 Angle Valve room:

15 min.

c. Work time directly on RHR-MO-25B: 90 min.

7. Following completion of the job, an additional 15 mrem per worker will be received during de-staging activities.

Initiating Cue(s):

You have been directed to use the information above and the radiological survey provided, evaluate the leak repair job for possible performance under current plant conditions to include:

1. Determine the total projected dose each worker would receive to pre-stage, perform the leak repair and de-stage. (Assume the same task times for both workers)

2. Based upon the projected total accumulated dose for this job, determine if any CNS administrative dose limitations will be exceeded requiring management authorization prior to performing job.

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 5 of 11 PERFORMANCE:

START TIME:

1. Procedure Step: Determines dose for area directly outside the Rx 903 Angle Valve room.

Standard Determined dose for Staging-Staging time in access area directly outside the Rx 903 Angle Valve room (45 min) 0.75 Hr X 5 mr/hr = 3.75 mrem (Accept 3.37 to 4.12 mrem)

Cue Notes Results SAT UNSAT

2. Procedure Step: Determines dose for area directly inside the Rx 903 Angle Valve room Standard Determined dose for Staging-Staging time in area directly inside the Rx 903 Angle Valve room step off pad.

(15 min) 0.25 Hr X 16 mr/hr = 4 mrem (Accept 3.6 to 4.4 mrem)

Cue Notes Results SAT UNSAT

3. Procedure Step: Determines dose for work in RHR-MO-25B area.

Standard Determined dose for work-Work time at RHR-MO-25B 1.5 Hr X 325 mr/hr = 487.5 mrem (Accept 438.7 to 536.2 mrem)

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 6 of 11

4. Procedure Step: Determines de-staging dose.

Standard Determined dose for de-staging-An additional 15 mrem will be accumulated for de-staging activities.

Cue Notes Results SAT UNSAT

5. Procedure Step: Determines total projected dose for each worker Standard Determined total dose for job-Total = 3.75 + 4 + 487.5 + 15 = 510.25 mrem (Accept 459.2 to 561.2 mrem)

Cue Notes Results SAT UNSAT

6. Procedure Step: Determines Worker admin limit authorizations.

Standard Determined Worker #1 would require authorization to exceed 1000 mrem dose.

Cue Notes Results SAT UNSAT

7. Procedure Step: Determines Worker admin limit authorizations.

Standard Determined Worker #2 would NOT require authorization due to already having greater than 1000 mrem and not reaching next admin limit of 2000 mrem.

Cue Notes Results SAT UNSAT Stop Time: __________

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 7 of 11 DO NOT GIVE TO APPLICANTS ATTACHMENT 1 ANSWER KEY WORKER #1 WORKER #2 INITIAL DOSE 600 mrem 1415 mrem PROJECTED DOSE 510.25 mrem 510.25 mrem PROJECTED TOTAL DOSE 1110.25 mrem 1925.25 mrem CNS ADMIN DOSE LIMIT AUTHORIZATION YES NO YES NO REQUIRED (Circle Yes or No)

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 8 of 11 (JPM A4)

ATTACHMENT 2 (Page 1 of 4)

DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is operating at power.

2. RHR-MO-25B has developed a leak.

3. Two workers are required to perform the leak repair.

4. Worker #1 has accumulated 600 mrem this year (CNS dose).

5. Worker #2 has accumulated 1415 mrem this year (CNS dose).

6. The following times for each worker have been estimated for task performance:

a. Staging time in area directly outside the CA, at the Step-off Pad, of Rx 903 Angle Valve room:

45 min.

b. Staging time in area directly inside CA, just inside the Step-ff Pad, of Rx 903 Angle Valve room:

15 min.

c. Work time directly on RHR-MO-25B: 90 min.

7. Following completion of the job, an additional 15 mrem per worker will be received during de-staging activities.

INITIATING CUE(S):

You have been directed to use the information above and the radiological survey provided, evaluate the leak repair job for possible performance under current plant conditions to include:

1. Determine the total projected dose each worker would receive to pre-stage, perform the leak repair and de-stage. (Assume the same task times for both workers)

2. Based upon the projected total accumulated dose for this job, determine if any CNS administrative dose limitations will be exceeded requiring management authorization prior to performing job.

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 9 of 11 ATTACHMENT 2 (Page 2 of 4)

DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is operating at power.

2. RHR-MO-25B has developed a leak.

3. Two workers are required to perform the leak repair.

4. Worker #1 has accumulated 600 mrem this year (CNS dose).

5. Worker #2 has accumulated 1415 mrem this year (CNS dose).

6. The following times for each worker have been estimated for task performance:

a. Staging time in area directly outside the CA, at the Step-off Pad, of Rx 903 Angle Valve room:

45 min.

b. Staging time in area directly inside CA, just inside the Step-ff Pad, of Rx 903 Angle Valve room:

15 min.

c. Work time directly on RHR-MO-25B: 90 min.

7. Following completion of the job, an additional 15 mrem per worker will be received during de-staging activities.

INITIATING CUE:

You have been directed to use the information above and the radiological survey provided, evaluate the leak repair job for possible performance under current plant conditions to include:

1. Determine the total projected dose each worker would receive to pre-stage, perform the leak repair and de-stage. (Assume the same task times for both workers)

2. Based upon the projected total accumulated dose for this job, determine if any CNS administrative dose limitations will be exceeded requiring management authorization prior to performing job.

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 10 of 11 ATTACHMENT 2 (Page 3 of 4)

WORKER #1 WORKER #2 INITIAL DOSE PROJECTED DOSE PROJECTED TOTAL DOSE CNS ADMIN DOSE LIMIT AUTHORIZATION YES NO YES NO REQUIRED (Circle Yes or No)

Print Name/Signature

CNS NRC Exam 3/2017 JPM A4 Rev 0 Page 11 of 11 ATTACHMENT 2 (Page 4 of4)

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 1 of 7 Cooper Nuclear Station Job Performance Measure JPM A5

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 2 of 7 (JPM A5)

Determine if Mode Change is Allowed Revision Statement: This is a simplified format of bank JPM SKL034-50-70 Rev 1.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A: 2.1.20 (4.6/4.6), 2.2.35 (3.6/4.5), 2.2.40 (3.4/4.7)

General

References:

1. Technical Specifications General Tools and Equipment:

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

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determines a Reactor Mode changes is not permitted with RCIC inoperable in accordance with Technical Specifications LCO 3.5.3 and Bases.

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 3 of 7 (JPM A5)

Directions to Examiner:

1. This JPM evaluates the applicant's ability to determine if a Reactor Mode change is allowed in accordance with Technical Specifications.

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 1 and a copy of Technical Specifications and Bases..

Notes: ______________________________________________________________________

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 4 of 7 Read the following to the JPM performer.

General Conditions:

1. The Plant is in MODE 3 with Reactor Pressure 600 psig.

2. RCIC is inoperable and is in day 10 of a 14 day LCO.

Initiating Cue(s):

You have been directed to determine if a Reactor Mode change to Mode 2 is allowed in accordance with Technical Specifications. Record your findings on Attachment 1.

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 5 of 7 Start Time: ____________

NOTE to Examiner: Operations Department expectation is that both Technical Specifications and applicable Bases are referenced when making Technical Specification calls. This will not result in failure but should be noted if both are not referenced during the performance of this JPM.

1. Procedure Step: Technical Specifications LCO 3.5.3 RCIC System referenced.

Standard Referred to TS LCO 3.5.3 RCIC.

Cue Notes Results SAT UNSAT

2. Procedure Step: Reviews the NOTE in the Actions section of LCO 3.5.3.

Standard Reviewed the NOTE concerning LCO 3.0.4b.

Cue Notes Results SAT UNSAT

3. Procedure Step: Technical Specifications 3.5.3 Bases referenced.

Standard Reviewed 3.5.3 Bases.

Cue Notes Results SAT UNSAT

4. Procedure Step: Technical Specifications section 3.0 reviewed.

Standard Reviewed LCO 3.0.4 Bases Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 6 of 7

5. Procedure Step:

Standard Determined LCO 3.0.4b not applicable to RCIC and Mode of Applicability applied.

Cue Notes Results SAT UNSAT

6. Procedure Step:

Standard Determined a Mode Change to MODE 2 is not allowed and recorded determination on Attachment 1.

Cue Notes Results SAT UNSAT Stop Time: __________

CNS NRC Exam 3/2017 JPM A5 Rev 0 Page 7 of 7 (JPM A5)

ATTACHMENT 1 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The Plant is in MODE 3 with Reactor Pressure 600 psig.

2. RCIC is inoperable and is in day 10 of a 14 day LCO.

INITIATING CUE(S):

You have been directed to determine if a Reactor Mode change to Mode 2 is allowed in accordance with Technical Specifications. Record your findings on Attachment 1.

NOTE: Circle the appropriate choice (IS / IS NOT) below.

Change to Mode 2: IS or IS NOT allowed.

Signature

CNS NRC Exam 3/2017 JPM A6 Rev o Page 1 of 10 Cooper Nuclear Station Job Performance Measure JPM A6

CNS NRC Exam 3/2017 JPM A6 Rev o Page 2 of 10 (JPM A6)

Reportable Occurrences to the NRC (#8)

Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A: 2.1.18 (3.6/3.8), 2.1.20 (4.6/4.6), 2.4.30 (2.7/4.1)

General

References:

1. Conduct of Operations Procedure 2.0.5, Reports to NRC Operations Center

2. NUREG 1022 General Tools and Equipment:

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

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determined a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report to the NRC is required and NRC Form 361 contains the correct technical information (50.72 non-emergency classification; 10 CFR 50.72(b)(2)(iv)(A) ECCS Discharge to RCS) per Procedure 2.0.5.

CNS NRC Exam 3/2017 JPM A6 Rev o Page 3 of 10 (JPM A6)

Directions to Examiner:

1. This JPM evaluates the applicant's ability to determine an NRC Reportable Occurrence has occurred and correctly complete NRC Form 361.

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2 (3 Pages) and a copy of NUREG 1022.

Notes: ______________________________________________________________________

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A6 Rev o Page 4 of 10 Read the following to the JPM performer.

General Conditions:

1. The plant is in Mode 3, shutting down for a refueling outage.

2. RPV Pressure is 500 psig.

3. RPV level is + 35 inches.

4. A drywell steam leak causes drywell pressure to rise to 2 psig.

5. RPS Actuates as designed.

6. ECCS initiation occurs as designed.

7. HPCI injected into the RPV and the operators took manual control and stopped HPCI from injecting.

8. The Shift Manager's phone number is 402-825-5253.

Initiating Cue(s):

You have been directed to determine what notification requirements exist for the NRC and complete the required sections of NRC Form 361. Return the completed form(s) to the evaluator when you have finished.

CNS NRC Exam 3/2017 JPM A6 Rev o Page 5 of 10 PERFORMANCE:

START TIME:

1. Procedure Step: Refers to Procedure 2.0.5 Standard Referred to body of Procedure 2.0.5, Step 4.4 and Attachments 1, 2 and 8.

Cue Notes Results SAT UNSAT NOTE to Examiner: Only a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is required.

2. Procedure Step: Determine appropriate reporting category per NUREG 1022 Standard Determined a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report is required.

Cue Notes Results SAT UNSAT

3. Procedure Step: 4.4.2.1 Ensure the Preparer completes the Event Notification Worksheet (Attachment 8), with exception of the Event Number and Notification Time (these are provided by the Headquarters Operations Officer at the NRC Operations Center when the notification is made).

Standard Filled out the NRC Form 361 with info provided.

Cue Notes Results SAT UNSAT

4. Procedure Step: Submits NRC Form 361 Standard Provided NRC Form 361 to the Evaluator.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A6 Rev o Page 6 of 10

5. Procedure Step: 3.1.4 Ensure the Event Notification Worksheet, Attachment 8 (available at R:\cnsprocs\FORMS\NRC Form 361.doc), is properly filled out when not in a declared emergency.

Standard The information contained in the Form submitted by the student matches the Technical Information (bolded) provided in the Key.

Cue Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM A6 Rev o Page 7 of 10 DO NOT GIVE TO APPLICANTS ATTACHMENT 1 ANSWER KEY NRC FORM 361 U.S. NUCLEAR REGULATORY COMMISSION OPERATIONS CENTER REACTOR PLANT EVENT NOTIFICATION WORKSHEET EN #

NRC OPERATION TELEPHONE NUMBER: PRIMARY - 301-816-5100 or 800-532-3469*, BACKUPS - [1st] 301-951-0550 or 800-449-3694*

[2nd] 301-415-0550 and [3rd]301-415-0553 *Licensees who maintain their own ETS are provided these telephone numbers FACILITY OR ORGANIZATION UNIT NOTIFICATION TIME NAME OF CALLER CALL BACK #

Cooper Nuclear Station 1 Applicants Name 402-825-5253 Event time and zone Event date Power/mode before Power/mode after Current Time Todays Date Subcritical / Mode 3 Subcritical / Mode 3 Event classification 1-Hr. Non-Emergency 10 CFR 50.72(b)(1) (v)(A) Safe S/D Capability AINA GENERAL EMERGENCY GEN/AAEC TS Deviation ADEV (v)(B) RHR Capability AINB SITE AREA EMERGENCY SIT/AAEC 4-Hr. Non-Emergency 10 CFR 50.72(b)(2) (v)(C) Control of Rad Release AINC ALERT ALE/AAEC (i) TS Required S/D ASHU (v)(D) Accident Mitigation AIND UNSUAL EVENT UNU/AAEC X (iv)(A) ECCS Discharge to RCS ACCS (xii) Offsite Medical AMED X 50.72 NON-EMERGENCY (See Next Column) (iv) (B) RPS Actuation (Scram) ARPS (xiii) Loss Comm/Asmt/Resp ACCM PHYSICAL SECURITY (73.71) DDDD (xi) Offsite Notification APRE 60-Day Optional 10 CFR 50.73(a)(1)

MATERIAL/EXPOSURE B??? 8-Hr. Non-Emergency 10 CFR 50.72(b)(3) Invalid Specified System Actuation AINV FITNESS FOR DUTY HRT (ii)(A) Degraded Condition ADEG Other Unspecified Requirement (Identify)

OTHER UNSPECIFIED REQMT. (See Last Column) (ii)(B) Unanalyzed Condition AUNA NONR INFORMATION ONLY NNF (iv)(A) Specified System Actuation AESF NONR DESCRIPTION Include: Systems affected, actuations and their initiating signals, causes, effect of event on plant, actions taken or planned, etc. (continue on back)

The HPCI system started on a valid initiation signal and injected into the RPV.

NOTIFICATIONS YES NO WILL BE ANYTHING UNUSUAL OR NOT YES (Explain above) NO NRC RESIDENT X UNDERSTOOD?

STATE(s) X DID ALL SYSTEMS FUNCTION YES AS REQUIRED? NO (Explain above)

LOCAL X OTHER GOV AGENCIES X MODE OF OPERATION ESTIMATED ADDITIONAL INFO ON BACK RESTART DATE: Date + 4 or YES NO MEDIA/PRESS RELEASE X UNTIL CORRECTED: 4 unknown at this time NOTE to Evaluator: The RPS Actuation is an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months notification because the reactor was subcritical. ECCS initiation without injection is an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months report, but since HPCI injected, it is a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report.

CNS NRC Exam 3/2017 JPM A6 Rev o Page 8 of 10 (JPM A6)

ATTACHMENT 2 (Page 1 of 3)

DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is in Mode 3, shutting down for a refueling outage.

2. RPV Pressure is 500 psig.

3. RPV level is + 35 inches.

4. A drywell steam leak causes drywell pressure to rise to 2 psig.

5. RPS Actuates as designed.

6. ECCS initiation occurs as designed.

7. HPCI injected into the RPV and the operators took manual control and stopped HPCI from injecting.

8. The Shift Manager's phone number is 402-825-5253.

INITIATING CUE(S):

You have been directed to determine what notification requirements exist for the NRC and complete the required sections of NRC Form 361. Return the completed form(s) to the evaluator when you have finished.

CNS NRC Exam 3/2017 JPM A6 Rev o Page 9 of 10 ATTACHMENT 2 (Page 2 of 3)

NRC FORM 361 U.S. NUCLEAR REGULATORY COMMISSION OPERATIONS CENTER REACTOR PLANT EVENT NOTIFICATION WORKSHEET EN #

NRC OPERATION TELEPHONE NUMBER: PRIMARY - 301-816-5100 or 800-532-3469*, BACKUPS - [1st] 301-951-0550 or 800-449-3694*

[2nd] 301-415-0550 and [3rd]301-415-0553 *Licensees who maintain their own ETS are provided these telephone numbers NOTIFICATION TIME FACILITY OR ORGANIZATION UNIT NAME OF CALLER CALL BACK #

Event time and zone Event date Power/mode before Power/mode after Event classification 1-Hr. Non-Emergency 10 CFR 50.72(b)(1) (v)(A) Safe S/D Capability AINA GENERAL EMERGENCY GEN/AAEC TS Deviation ADEV (v)(B) RHR Capability AINB SITE AREA EMERGENCY SIT/AAEC 4-Hr. Non-Emergency 10 CFR 50.72(b)(2) (v)(C) Control of Rad Release AINC ALERT ALE/AAEC (i) TS Required S/D ASHU (v)(D) Accident Mitigation AIND UNSUAL EVENT UNU/AAEC (iv)(A) ECCS Discharge to RCS ACCS (xii) Offsite Medical AMED 50.72 NON-EMERGENCY (See Next Columns) (iv) (B) RPS Actuation (Scram) ARPS (xiii) Loss Comm/Asmt/Resp ACCM PHYSICAL SECURITY (73.71) DDDD (xi) Offsite Notification APRE 60-Day Optional 10 CFR 50.73(a)(1)

MATERIAL/EXPOSURE B??? 8-Hr. Non-Emergency 10 CFR 50.72(b)(3) Invalid Specified System Actuation AINV FITNESS FOR DUTY HRT (ii)(A) Degraded Condition ADEG Other Unspecified Requirement (Identify)

OTHER UNSPECIFIED REQMT. (See Last Column) (ii)(B) Unanalyzed Condition AUNA NONR INFORMATION ONLY NNF (iv)(A) Specified System Actuation AESF NONR DESCRIPTION Include: Systems affected, actuations and their initiating signals, causes, effect of event on plant, actions taken or planned, etc. (continue on back)

NOTIFICATIONS YES NO WILL BE ANYTHING UNUSUAL OR NOT NRC RESIDENT UNDERSTOOD? YES (Explain above)

NO STATE(s) DID ALL SYSTEMS FUNCTION YES NO (Explain above)

LOCAL AS REQUIRED?

OTHER GOV AGENCIES MODE OF OPERATION ESTIMATED ADDITIONAL INFO ON BACK UNTIL CORRECTED: RESTART DATE: YES NO MEDIA/PRESS RELEASE

CNS NRC Exam 3/2017 JPM A6 Rev o Page 10 of 10 ATTACHMENT 2 (Page 3 of 3)

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 1 of 11 Cooper Nuclear Station Job Performance Measure JPM A7

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 2 of 11 (JPM A7)

Review of Completed Jet Pump Operability Check (SRO Version 2)

Revision Statement: This is a modified version of bank JPM SKL034-50-65 Rev 2.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 18 minutes

5. NRC K/A: 2.2.12 (3.7/4.1), 2.2.42 (3.9/4.6)

General

References:

1. Procedure 6.LOG.601, Daily Surveillance Log - Modes 1, 2, And 3 General Tools and Equipment:

1. Calculator

2. Jet Pump Operability Curves

3. Reactor Recirculation pump curves

4. Procedure 2.1.10, (Attachment 1) Power-To-Flow map.

Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant reviewed the recorded Jet Pump, RR, core flow parameters, averaged Jet Pump Ps and determined Recirc Pump Flow loop flow mismatch for >51.45 mlbm/hr was inappropriately N/Ad and is UNSAT in accordance with Procedure 6.LOG.601.

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 3 of 11 (JPM A7)

Directions to Examiner:

1. This JPM evaluates the applicant'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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2 (5 Pages).

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 4 of 11 Read the following to the JPM performer.

General Conditions:

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

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

Initiating Cue(s):

You have been directed to perform a review of the daily Jet Pump Operability (6.LOG.601 Attachment

12) and Recirc Pump Flow Checks (6.LOG.601 Attachment 13) 0700-1000 readings as part of the routine shift activities to ensure they have been performed correctly. Record your conclusions on Page 5 of Attachment 2.

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 5 of 11 PERFORMANCE:

START TIME:

1. Procedure Step: N/A Standard Reviewed Attachment 12 Jet Pump Operability Parameters A, B, C, D and Checks 1, 2, 3 and verified Operability Limit on Checks 1 and 2, or 3 are SAT for Loop A and B.

Cue Notes Results SAT UNSAT

2. Procedure Step: N/A Standard Reviewed Attachment 13 checks.

Cue Notes Results SAT UNSAT

3. Procedure Step:

Standard Recorded comments on Attachment 2, Page 5.

Cue Notes Results SAT UNSAT

4. Procedure Step: N/A Standard The comments recorded on Attachment 2 match the intent of the Answer Key (Attachment 1).

Cue Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 6 of 11 DO NOT GIVE TO APPLICANTS ATTACHMENT 1 ANSWER KEY Identified loop flow mismatch for >51.45 mlbm/hr was inappropriately N/Ad and check for <51.45 mlbm/hr was inappropriately filled in and should have been N/Ad.

Noted loop flow mismatch for >51.45 mlbm/hr is not met should be checked UNSAT.

ATTACHMENT 13 RECIRC PUMP FLOW RECIRC PUMP FLOW 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 CHECKS SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES VERIFY Core Flow, MCO NBI-DPR/FR-95, value is not in Stability SAT SAT 1, 2 49 Exclusion Region of Power to Flow Map (a)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B SAT flow mismatch is SAT 1 (c)

,2 (c) 48 Should be 7.35x106 lbs/hr at N/A

< 51.45x106 lbs/hr Rated Core Flow (b,d)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B N/A flow mismatch is SAT 1 (c)

,2 (c) 48 Should be 3.67x106 lbs/hr at UNSAT 51.45x106 lbs/hr Rated Core Flow (b,d)

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 7 of 11 (JPM A7)

ATTACHMENT 2 (PAGE 1 OF 5)

GENERAL CONDITIONS:

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

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

INITIATING CUE(S):

You have been directed to perform a review of the daily Jet Pump Operability (6.LOG.601 Attachment

12) and Recirc Pump Flow Checks (6.LOG.601 Attachment 13) 0700-1000 readings as part of the routine shift activities to ensure they have been performed correctly. Record your conclusions on Page 5 of Attachment 2.

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 8 of 11 ATTACHMENT 12 JET PUMP OPERABILITY ATTACHMENT 2 (PAGE 2 OF 5)

1. IF in single loop operation, THEN MARK idle loop N/A.

JET PUMP P (%)

PARAMETERS LOOP A LOOP B JP # LOOP B JP # LOOP A 6

A Core Flow (10 lb/hr) NBI-FRDPR-95 66 1 39 11 29 B RR Pump Flow (103 gpm) RR-FR-163 39.9 41.7 2 39 12 28 C RRMG Set Speed (%) RRFC-SI-1A/B 81.0 84.8 3 39 13 29 D JP Flow (106 lb/hr) NBI-FI-92A/B 31 35 4 43 14 31 5 45 15 34 6 42 16 32 7 40 17 31 8 39 18 30 9 40 19 33 10 40 20 32 LOOP B LOOP 40.6 30.9 Avg A Avg OPERABILITY APPLICABLE ATT. 22 CHECKS SAT UNSAT LIMIT MODE NOTE VERIFY B and C values Loop A:

1 within curve limits (a) Loop B:

VERIFY C and D values Loop A:

2 SAT within curve limits (a) Loop B:

Loop A: 1 (b), 2 (b) 50 VERIFY Jet Pump P differ 3

by 20% from curves (a,c) Loop B:

VERIFY Checks 1 and 2 SAT or Loop A:

SAT Check 3 SAT Loop B:

(a)

REFER to Jet Pump Operability Curves maintained in Control Room.

(b)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after greater than 25% RTP and 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, THEN immediately NOTIFY Reactor Engineering.

Procedure 6.LOG.601 Revision 123 Page 28 of 80

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 9 of 11 ATTACHMENT 13 RECIRC PUMP FLOW ATTACHMENT 2 (Page 3 of 5) 0700-1000 1900-2200 OPERABILITY APPLICABLE ATT. 22 CHECKS SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES VERIFY Core Flow, MCO NBI-DPR/FR-95, value is not in Stability SAT SAT 1, 2 49 Exclusion Region of Power to Flow Map (a)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is SAT SAT 1 (c)

,2 (c) 48 7.35x106 lbs/hr at

< 51.45x106 lbs/hr Rated Core Flow (b,d)

VERIFY JP Flow, MCO NBI-FI-92A/B, values for Loop A and Loop B flow mismatch is N/A SAT 1 (c)

,2 (c) 48 3.67x106 lbs/hr at 51.45x106 lbs/hr Rated Core Flow (b,d)

(a)

REFERENCE power to flow map in Procedure 2.1.10.

(b)

NOTE - Per Technical Specification, a recirculation loop is considered not in operation when total jet pump flows of the two loops is greater than mismatch limits.

IF greater than mismatch limit, THEN ENSURE loop with lower flow considered not in operation.

(c)

APPLY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops in service.

(d)

IF flow mismatch results in one loop considered not in operation, THEN CONTACT FRED to manually insert single loop operation limits into GARDEL.

Procedure 6.LOG.601 Revision 123 Page 29 of 80

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 10 of 11 ATTACHMENT 2 (Page 4 of 5)

ATTACHMENT 13 RECIRC PUMP FLOW 0700-1000 1900-2200 READING READING OPERABILITY APPLICABLE ATT. 22 CHECK SAT/UNSAT SAT/UNSAT LIMIT MODES NOTES 3 (when reactor VERIFY RR pump pressure less operating or RHR N/A SAT than SDC 66 pump operating in pressure SDC permissive)

(e)

(e) APPLY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reactor steam dome pressure less than shutdown cooling permissive pressure.

Procedure 6.LOG.601 Revision 123 Page 30 of 80

CNS NRC Exam 3/2017 JPM A7 Rev 0 Page 11 of 11 ATTACHMENT 2 (Page 5 of 5)

==

Conclusion:==

__________________________/_________________________

Print Name/Signature

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 1 of 8 Cooper Nuclear Station Job Performance Measure JPM A8

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 2 of 8 (JPM A8)

Authorize Stable Iodine Thyroid Blocking Revision Statement: This is a simplified format of bank JPM SKL034-50-20 Rev 2.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 8 minutes

5. NRC K/A: 2.3.14 (3.4/3.8)

General

References:

1. Procedure 5.7.14, Stable Iodine Thyroid Blocking (KI)

2. Procedure 5.7.2, Emergency Director EPIP General Tools and Equipment:

1. Marked up Procedure 5.7.2, Attachment 1 (EPIP 5.7.2 ALERT).

Special Conditions, References, Tools, Equipment:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determined Stable Iodine Thyroid Blocking is authorized for the affected personnel per Procedure 5.7.14.

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 3 of 8 (JPM A8)

Directions to Examiner:

1. This JPM evaluates the applicant's ability to determine the need to authorize stable iodine thyroid blocking.

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachment 2, and tell them to begin.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 4 of 8 Read the following to the JPM performer.

General Conditions:

1. The Reactor is shutdown in a refueling outage.

2. An accident due to a failure of the refuel floor crane has resulted in personnel injuries to two refuel floor workers and severe damage to several fuel bundles.

3. The two injured refuel floor workers have no immediate life threatening injuries but they are unable to leave the area on their own. No other personnel are currently present on the refueling floor.

4. The Emergency Director has declared an ALERT.

5. RMA-RA-1, FUEL POOL AREA, indicates 60 Rem/hr. and RMA-RA-2 FUEL POOL AREA is upscale.

6. No survey data or air samples are presently available from the refuel floor.

7. A Team of EMTs one of which is an RP (and will be performing the RP functions) are standing by to evacuate the injured workers.

8. RPs estimate 15 rem per person for the lifesaving efforts.

Initiating Cue(s):

You are the Emergency Director; Procedure 5.7.2 Emergency Director EPIP, Attachment 1 has been initiated and is completed through the following steps:

A-1 is complete

A-2 is in progress.

STAFF AUGMENTATION/ON-SITE NOTIFICATION - A/A-1 through A/A-6 are complete.

OFF-SITE NOTIFICATION - A/N-1 through A/N-5 are complete.

RAD RELEASE AND KI - A/R-1 needs to be addressed.

UPGRADE/TERMINATION - A/T-1 is being performed.

ANY ANNOUNCEMENTS ARE TO BE OMMITTED.

As Shift Manager (Emergency Director) in procedure 5.7.2, address Step A/R-1 and determine the requirements for the EMTs to complete the evacuation of the injured workers. Hand any completed forms or attachments to the evaluator.

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 5 of 8 PERFORMANCE:

START TIME:

1. Procedure Step: Procedure 5.7.2, RAD RELEASE AND KI leg, Table 1 IODINE THYROID BLOCKING IS REQUIRED IF ANY OF THE FOLLOWING CONDITIONS EXIST: 3.

Emergency workers being dispatched to areas where high levels of radio-iodine are suspected AND no current air sample data is available, Standard Determined that Stable Iodine Thyroid Blocking was indicated for the emergency workers involved in the rescue and Procedure 5.7.14 was required to be entered.

Cue Notes Results SAT UNSAT NOTE to Examiner: When asked, provide Procedure 5.7.14 to Applicant.

2. Procedure Step: A/R-1 Enter Procedure 5.7.14 Standard Obtained current revision of Procedure 5.7.14, Stable Iodine Thyroid Blocking.

Cue Notes Results SAT UNSAT NOTE: This JPM is not intended to evaluate the Applicant distributing KI, ONLY to authorize the distribution of KI, Section 4 and Attachment 1.

NOTE: The applicant may or may not authorize distribution of KI to the injured personnel on the refuel floor.

3. Procedure Step: 4.1.1 Log decision to authorize KI.

Standard Kept a rough log of decision to authorize KI.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 6 of 8 NOTE to Examiner: The applicant may consider the RP on the EMT Team or select Other on the KI Distribution/Notification Checklist.

4. Procedure Step: 4.1.2 Determine affected personnel and complete Attachment 1, KI Distribution/Notification Checklist.

Standard Determined the affected personnel and completed Attachment 1, KI Distribution/Notification Checklist, by marking in the first table [ ] Emergency Medical Technicians.

Cue Notes Results SAT UNSAT NOTE to Examiner: Plant announcements are to be omitted as stated in Initiating Cue.

5. Procedure Step: 4.1.3 Announce authorization of KI to the facility.

Standard Step is N/A.

Cue Notes Results SAT UNSAT NOTE to Examiner: Plant announcements are to be omitted as stated in Initiating Cue.

6. Procedure Step: 4.1.4 Direct the following plant announcement be made: 4.1.4.1 "Attention all plant personnel, the Emergency Director has authorized the use of KI for designated Emergency Workers". (Repeat)

Standard Step is N/A Cue Notes Results SAT UNSAT

7. Procedure Step: N/A Standard Provided the completed worksheet to the designated individual who is to manage the distribution of the KI per Attachment 2 and verified that KI was being distributed to affected personnel.

Cue Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 7 of 8 ATTACHMENT 1 DO NOT GIVE TO APPLICANTS KEY 0.ATTACHMENT 1 KI DISTRIBUTION/NOTIFICATION CHECKLIST

1. Determine groups selected to be authorized KI and check the selected groups.

[ ] Control Room Personnel [ ] OSC (ALL including OSC Mission Teams)

[ ] Station Operators [ ] TSC (ALL)

[ ] Downwind Field Monitoring Teams [ ] EOF (ALL)

[ ] Security Officers [X] Emergency Medical Technicians (NPPD)

[ ] Fire Brigade Personnel [ ] West Warehouse Addition (ERO Personnel)

[ ] Other:

NOTE - KI may also be provided to non-NPPD emergency response organizations for distribution to their emergency workers. Administration of KI to non-NPPD personnel shall be the responsibility of the organizations to which these personnel belong.

N/A2. Determine non-NPPD emergency response organizations to be contacted to inform them of the decision by NPPD to authorize KI.

[ ] Federal Agencies [ ] MO State Personnel

[ ] Fire Brigade/Mutual Aid [ ] NE State Personnel

[ ] State Police [ ] Off-Site Ambulance Crews

[ ] Local Law Enforcement [ ] Other:

3. Provide completed worksheet to individual designated to manage the distribution of KI per Attachment 2.

CNS NRC Exam 3/2017 JPM A8 Rev 0 Page 8 of 8 ATTACHMENT 2 DIRECTIONS TO APPLICANT:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The Reactor is shutdown in a refueling outage.

2. An accident due to a failure of the refuel floor crane has resulted in personnel injuries to two refuel floor workers and severe damage to several fuel bundles.

3. The two injured refuel floor workers have no immediate life threatening injuries but they are unable to leave the area on their own. No other personnel are currently present on the refueling floor.

4. The Emergency Director has declared an ALERT.

5. RMA-RA-1, FUEL POOL AREA, indicates 60 Rem/hr. and RMA-RA-2 FUEL POOL AREA is upscale.

6. No survey data or air samples are presently available from the refuel floor.

7. A Team of EMTs one of which is an RP (and will be performing the RP functions) are standing by to evacuate the injured workers.

8. RPs estimate 15 rem per person for the lifesaving efforts.

INITIATING CUE:

You are the Emergency Director; Procedure 5.7.2 Emergency Director EPIP, Attachment 1 has been initiated and is completed through the following steps:

A-1 is complete

A-2 is in progress.

STAFF AUGMENTATION/ON-SITE NOTIFICATION - A/A-1 through A/A-6 are complete.

OFF-SITE NOTIFICATION - A/N-1 through A/N-5 are complete.

RAD RELEASE AND KI - A/R-1 needs to be addressed.

UPGRADE/TERMINATION - A/T-1 is being performed.

ANY ANNOUNCEMENTS ARE TO BE OMMITTED.

Continue actions as Emergency Director in procedure 5.7.2. Address Step A/R-1 and determine the requirements for the EMTs to complete the evacuation of the injured workers. Hand any completed forms or attachments to the evaluator.

Record any entries below:

Signature: ______________________

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 1 of 8 Cooper Nuclear Station Job Performance Measure JPM A9

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 2 of 8 (JPM A9)

EAL Table Top 9 Revision Statement: This is a simplified format of bank JPM SKL034-50-77 Rev 1.

Additional Program Information:

1. Appropriate Performance Locations: Classroom

2. Appropriate Trainee level: SRO

3. Evaluation Method: Perform

4. Performance Time: 15 minutes

5. NRC K/A: 2.4.41 (2.9/4.6)

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 steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant determined an ALERT classification and EAL Number (SA2.1) was required in accordance with Procedure 5.7.1. The classification was determined within 15 minutes of reading the General Conditions.

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 3 of 8 (JPM A9)

Directions to Examiner:

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

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 Unsatisfactory, state why in the notes section below.

4. Give the trainee Attachments 1, 2, EAL Hard Cards and Procedure 5.7.1.

Notes: ______________________________________________________________________

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 4 of 8 Read the following to the JPM performer.

General Conditions:

The plant has experienced the events listed on Attachment 2.

Initiating Cue(s):

This is a time critical JPM.

You have been directed to use the provided attachment to classify the event at the highest EAL attained.

Record the emergency classification and EAL number on Attachment 2.

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 5 of 8 PERFORMANCE:

START TIME:

1. Procedure Step: 2.1.1 AFTER recognition of off-normal event, THEN SM shall COMPARE event to EALs on EAL Classification Matrix.

Standard Referred to EAL Classification Matrix, Attachment 4.

Cue Notes Results SAT UNSAT

2. Procedure Step: 2.1.2 IF more than one EAL of different classification levels is reached (i.e., EAL for ALERT and EAL for SITE AREA EMERGENCY), THEN SELECT EAL for most severe emergency classification.

Standard Determined the most severe emergency classification.

Cue .

Notes Results SAT UNSAT

3. Procedure Step: 2.1.3 IF event appears to meet an EAL and time permits, THEN REFER to Attachment 2 or 3 for further explanation and guidance.

Standard Referred to Attachment 2 or 3 if time permitted.

Cue Notes Results SAT UNSAT

4. Procedure Step: 2.1.4 IF determined EAL is met, THEN PERFORM following DECLARE emergency within 15 minutes.

Standard The classification was determined within 15 minutes Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 6 of 8

5. Procedure Step: 2.1.4.1 DECLARE emergency.

Standard Classified the event as an ALERT and EAL Number SA2.1 based on EPIP 5.7.1.

Cue Notes Results SAT UNSAT

6. Procedure Step: N/A Standard Returned the completed Attachment 2 to the examiner.

Cue Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 7 of 8 ATTACHMENT 1 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

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

INITIATING CUE(S):

This is a time critical JPM.

You have been directed to use the provided attachment to classify the event at the highest EAL attained.

Record the emergency classification and EAL number on Attachment 2.

CNS NRC Exam 3/2017 JPM A9 Rev 0 Page 8 of 8 (JPM A9)

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 inches 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.

Notify Examiner you have read the plant status and are ready to determine the classification.

Highest Classification: __________________

EAL Number____________________

Signature of Operator: _____________________________________________

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 1 of 9 Cooper Nuclear Station Job Performance Measure NRC S1

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 2 of 9 (JPM S1)

Secure SDG from Control Room Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 8 minutes

5. NRC K/A 264000 A4.04 (3.7/3.7)

6. PSA Applicability: Top 10 Risk Significant System - Emergency AC Power

References:

1. Procedure 2.2.99, Supplemental Diesel Generator System General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant transferred Bus 480S to the 12.5 kV system, opened SDG output breaker and stopped the SDG in accordance with Procedure 2.2.99, Supplemental Diesel Generator System.

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 3 of 9 Directions to Examiner:

1. This JPM evaluates the trainees ability to shut down the Supplemental Diesel Generator (SDG) from the control room.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 4 of 9 Read the following to the JPM performer.

General Conditions:

1. The plant is shutdown following a loss of electrical power.

2. SDG is powering Bus 4160S.

3. The Emergency Transformer has been restored to service.

4. The critical buses 1F and 1G are powered from the ESST.

Initiating Cue:

You have been directed to secure the SDG from the Control Room per Procedure 2.2.99, Supplemental Diesel Generator System, Section 14.

Inform the CRS when the task is complete.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 5 of 9 PERFORMANCE:

START TIME:

1. Procedure Step: 14.1 Ensure BUS 1S TIE BKR 1SS is in NORMAL AFTER TRIP (green flagged).

Standard Checked BUS 1S TIE BKR 1SS is in NORMAL AFTER TRIP (green flagged) position.

Cue None Notes Results SAT UNSAT

2. Procedure Step: 14.2 IF Bus 480S is powered from SDG, THEN transfer Bus 480S to 12.5 kV System using Switchgear Display screen: 14.2.1 Open Breaker 480S-(4160S) by pressing TRIP button twice.

Standard On SDG HMI, pressed Breaker 480S-(4160S) TRIP button twice.

Cue None Notes Results SAT UNSAT

3. Procedure Step: 14.2.2 Close Breaker 480S-(12.5) by pressing CLOSE button twice.

Standard On SDG HMI, pressed Breaker 480S-(12.5) CLOSE button twice.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 6 of 9

4. Procedure Step: 14.3 Open SDG OUTPUT BKR SG1 and check following:

14.3.1 Switch spring returns to NORMAL AFTER TRIP (green flagged).

Standard Rotated SDG OUTPUT BKR SG1 switch counter-clockwise to Trip position and released allowing it to spring return to NORMAL AFTER TRIP (green flagged).

Cue None Notes Results SAT UNSAT

5. Procedure 14.3.2 4160V BUS 1S BUS ENERGIZED light goes out.

Step:

Standard Checked 4160V BUS 1S BUS ENERGIZED light goes out Cue None Notes Results SAT UNSAT NOTE to Examiner: Time compression is allowed in this step.

6. Procedure 14.4 Ensure SDG has been unloaded 15 minutes.

Step:

Standard Allowed 15 minutes to pass.

Cue After examinee reads procedure step, relay that 16 minutes has passed.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 7 of 9 Note to Examiner: SDG has a programmed auto cooldown of ~2 minutes. SDG stops after the cooldown is complete.

7. Procedure 14.5 Place SUPPLEMENTAL DIESEL GENERATOR switch to Step: STOP.

Standard Rotated SUPPLEMENTAL DIESEL GENERATOR switch counter-clockwise to STOP position and released.

Cue None Notes Results SAT UNSAT

8. Procedure 14.6 At SDG-LCP-GEN1, SDG GEN LOCAL CONTROL Step: PANEL (SDG Engine Room), press ALARM SILENCE, if required.

Standard Directed operator in SDG Engine room to press Alarm Silence on local panel.

Cue NLO in SDG Engine Room has pressed Alarm Silence on local panel.

Notes Results SAT UNSAT

9. Procedure N/A Step:

Standard Reported to CRS that SDG has been secured per Procedure 2.2.99.

Cue CRS Acknowledges report.

Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 8 of 9 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 158 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None

2. Malfunctions

3. Remotes None N/A N/A

4. Overrides None N/A N/A

a. Insert Malfunction ED05 (Loss of SSST)

b. Insert Malfunctions DG06A and DG06B. (DGs fail to start)

c. Scram reactor.

d. Place Mode Switch in Shutdown.

5. Panel Setup e. Ensure 4160V Bus 1F and 1G transfer to ESST.

f. Trip both DGs.

g. Start SDG and energize Bus 4160S Per 2.2.99 (Section 5 and Section 8 through Step 8.6.4.

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

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 9 of 9 ATTACHMENT 2 DIRECTIONS TO APPLICANT:

Read the following and inform the evaluator when you are ready to begin.

GENERAL CONDITIONS:

1. The plant is shutdown following a loss of electrical power.

2. SDG is powering Bus 4160S.

3. The Emergency Transformer has been restored to service.

4. The critical buses 1F and 1G are powered from the ESST.

INITIATING CUE:

You have been directed to secure the SDG from the Control Room per Procedure 2.2.99, Supplemental Diesel Generator System, Section 14.

Inform the CRS when the task is complete.

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 1 of 9 Cooper Nuclear Station Job Performance Measure NRC S2

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 2 of 9 (JPM S2)

Defeat RPS Logic Trips During Failure To Scram (5.8.3) (Restoration)

Revision Statement: Simplified format of bank JPM SKL034-21-79.

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 6 minutes

5. NRC K/A 295037 EA1.01 (4.6/4.6)

6. PSA Applicability: Top 10 Risk Significant System - RPS

References:

1. Procedure 5.8.3, Alternate Rod Insertion Methods General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant removed RPS bypass PTMs and reset the reactor scram in accordance with Procedure 5.8.3, Alternate Rod Insertion Methods.

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 3 of 9 Directions to Examiner:

1. This JPM evaluates the applicant's ability to remove RPS plant temporary modification (PTM) jumpers and reset the reactor scram.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 4 of 9 Read the following to the JPM performer.

General Conditions:

1. An ATWS condition was present.

2. All control rods have been inserted using alternate control rod insertion methods.

Initiating Cue:

You are to remove RPS Logic Trip Bypass jumpers and reset the scram logic per procedure 5.8.3, Alternate Rod Insertion Methods Section 5.

Inform the CRS when the task is complete.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 5 of 9 PERFORMANCE:

START TIME:

1. Procedure Step: 5.6.1 Prior to removing RPS LOGIC TRIP BYPASS jumpers, insert a manual scram.

Standard At 9-5 Panel, depressed RX SCRAM CHAN A and CHAN B buttons.

Cue None Notes Results SAT UNSAT

2. Procedure Step: 5.6.2 Remove RPIS LOGIC TRIP BYPASS jumper between Terminals DD-41 and DD-42 (BAY-1, PNL 9-15).

Standard In Panel 9-15 BAY-1, removed jumper between Terminals DD-41 and DD-42.

Cue None Notes Results SAT UNSAT

3. Procedure Step: 5.6.3 Remove RPIS LOGIC TRIP BYPASS jumper between Terminals BB-41 and BB-42 (BAY-3, PNL 9-15).

Standard In Panel 9-15 BAY-3, removed jumper between Terminals BB-41 and BB-42.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 6 of 9

4. Procedure Step: 5.6.4 Remove RPIS LOGIC TRIP BYPASS jumper between Terminals DD-41 and DD-42 (BAY-1, PNL 9-17).

Standard In Panel 9-17 BAY-1, removed jumper between Terminals DD-41 and DD-42.

Cue None Notes Results SAT UNSAT

5. Procedure 5.6.5 Remove RPIS LOGIC TRIP BYPASS jumper between Step: Terminals BB-41 and BB-42 (BAY-3, PNL 9-17).

Standard In Panel 9-17 BAY-3, removed jumper between Terminals BB-41 and BB-42.

Cue None Notes Results SAT UNSAT

6. Procedure 5.7 Reset RPS manual scram (Panel 9-5) by placing Step: REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 and 3, then back to NORM.

Standard At Panel 9-5, turned REACTOR SCRAM RESET switch counter-clockwise to Group 1 and 4, clockwise to Group 2 and 3, then counter-clockwise back to NORM.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 7 of 9

7. Procedure 5.8 Inform CRS that RPS logic has been restored.

Step:

Standard Cue None Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 8 of 9 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 161 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None

2. Malfunctions N/A N/A

3. Remotes None N/A N/A

4. Overrides None N/A N/A

a. Place Simulator in RUN.

b. Scram the reactor.

c. Place Mode switch in SHUTDOWN.

d. Install low RPV water level MSIV isolation jumpers per 5.8.20, Section 4.

e. Install jumpers to defeat RPS logic trips per 5.8.3, Section 5.

f. Bypass RWM at Panel 9-5.

g. Let RPV water level recover to normal band.

5. Panel Setup h. Trip one RFP.

i. Trip 2 CBPs and CPs.

j. Reset the scram.

k. Place SDIV high water bypass switch to BYPASS.

l. Drain the SDV.

m. Close RR-AO-740 and 741.

n. Run simulator long enough to ensure RPV level will not lower below low RPV level scram setpoint and SDIV NOT Drained alarms are clear.

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

CNS NRC Exam 3/2017 JPM S1 Rev 0 Page 9 of 9 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. An ATWS condition was present.

2. All control rods have been inserted using alternate control rod insertion methods.

Initiating Cue:

You are to remove RPS Logic Trip Bypass jumpers and reset the scram logic per procedure 5.8.3, Alternate Rod Insertion Methods Section 5.

Inform the CRS when the task is complete.

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 1 of 13 Cooper Nuclear Station Job Performance Measure NRC S3

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 2 of 13 (JPM S3)

Conduct Alternate Pressure Control Using Reactor Feed Pumps Revision Statement: Simplified format of bank JPM SKL034-20-81, Rev 6.

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 259001 A4.02 (3.9/3.7)

6. PSA Applicability: N/A

References:

1. Procedure 5.8.1, RPV Pressure Control Systems (Table 1)

General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant reset RFPT A trip, performed a QUICK RESTART, and raised RFPT A speed between 2400 and 2600 rpm in accordance with Emergency Operating Procedure 5.8.1.

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 3 of 13 Directions to Examiner:

1. This JPM evaluates the applicant's ability to conduct alternate pressure control using a Reactor Feed Pump.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 4 of 13 Read the following to the JPM performer.

General Conditions:

1. The plant had been operating at 100% on day 90 of a 2 year cycle.

2. The turbine tripped followed by a reactor scram.

3. EOPs have been entered.

4. There are no known issues with RFPT 1A turning gear.

5. RFPT 1A has not been stationary.

Initiating Cue:

You are to conduct alternate pressure control using Reactor Feed Pump A as directed in EOP support procedure 5.8.1, Section 8. RPV injection with Reactor Feed Pump 1A is NOT required at this time. Inform the CRS when RFP A is operating between 2400 and 2600 rpm.

Inform the CRS when the task is complete.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 5 of 13 PERFORMANCE:

START TIME:

1. Procedure Step: Obtained procedure.

Standard Obtained procedure 5.8.1, RPV Pressure Control Systems (Table 1).

Cue None Notes Results SAT UNSAT

2. Procedure Step: 8.1 Ensured RF-MO-29, RFP A DISCHARGE VLV, is closed (PANEL A).

Standard Verified RF-MO-29 Green indicating light ON and Red indicating light OFF.

Cue None Notes Results SAT UNSAT

3. Procedure Step: 8.2 Ensured RF-MO-30, RFP B DISCHARGE VLV, is closed (PANEL A).

Standard Verified RF-MO-30 Green indicating light ON and Red indicating light OFF.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 6 of 13

4. Procedure Step: 8.3 At RFPT/RVLC HMI on MAIN CONTROL screen, ensure RFPT-1A controller is in MDVP with OUTPUT at 0.0%

Standard At RFPT/RVLC HMI, selected MAIN CONTROL screen, verified RFPT-1A controller in MDVP with OUTPUT at 0.0%.

Cue None Notes Results SAT UNSAT

5. Procedure 8.4 At RFPT/RVLC HMI, select STARTUP VALVE screen.

Step:

Standard At RFPT/RVLC HMI, selected STARTUP VALVE screen.

Cue None Notes Results SAT UNSAT

6. Procedure 8.5 On STARTUP VALVE screen, press EMER CLOSE Step: button on either FCV-11AA or FCV-11BB and confirm

'YES' to pop-up box.

Standard Selected STARTUP VALVE screen; selected EMER CLOSE button and selected YES on pop-up box.

Cue None Notes EMER CLOSE boxes for both FCV-11AA and FCV-11BB turn yellow.

Results SAT UNSAT

7. Procedure 8.6 Ensure Feedwater Startup Flow Controllers FCV-11AA and Step: FCV-11BB OUTPUTs are at 0.0%.

Standard Verified FCV-11AA and FCV-11BB OUTPUT indicating 0.0%

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 7 of 13

8. Procedure 8.7 On STARTUP VALVE screen, press EMER CLOSE button Step: on either FCV-11AA or FCV-11BB and confirm 'YES' to pop-up box.

Standard Selected STARTUP VALVE screen; selected EMER CLOSE button and selected YES on pop-up box.

Cue None Notes EMER CLOSE boxes for both FCV-11AA and FCV-11BB turn green. Emergency close signal removed with this step in case RPV injection is desired.

Results SAT UNSAT

9. Procedure 8.8 N/A Step:

Standard Cue None Notes No known deficiency with turning gear stated in General Conditions information.

Results SAT UNSAT

10. Procedure 8.9 N/A Step:

Standard Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 8 of 13

11. Procedure 8.10 Depress RFPT A trip RESET button until RFPT A HP Step: and LP STOP valves are open (PANEL A).

Standard On bench board A, depressed RFPT A RESET pushbutton and verified RFPT A HP and LP STOP valve indicating lights are: RED light illuminated and Green light extinguished, then released pushbutton.

Cue None Notes Results SAT UNSAT

12. Procedure 8.11 Verify that minimum flow valve RF-FCV-11A is open.

Step:

Standard On bench board A, verified RF-FCV-11A RED light illuminated and GREEN light extinguished.

Cue None Notes Results SAT UNSAT

13. Procedure 8.12 On FEEDPUMP 1A screen, select QUICK RESTART.

Step:

Standard Selected FEEDPUMP 1A screen and selected QUICK RESTART and verified QUICK RESTART box turned yellow.

Cue None Notes Only Bold portion of the Standard is critical.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 9 of 13

14. Procedure 8.12.1 IF LP steam is available, THEN select LP START; Step: otherwise, select HP START.

Standard Pressed HP START box and verified box turned yellow.

Cue None Notes Results SAT UNSAT

15. Procedure 8.12.2 Press green START button in turbine latch box.

Step:

Standard Selected green START in turbine latch box.

Cue None Notes Results SAT UNSAT

16. Procedure 8.12.3 Confirm starting turbine by pressing START button Step: in pop-up box.

Standard Pressed green START in pop-up box.

Cue None Notes Display cycles through ACCELERATE TO IDLE and ends at MINIMUM GOVERNOR. At ~ 000 rpm, CONTINUE box turns green.

Results SAT UNSAT

17. Procedure 8.12.4 WHEN RFPT-1A reaches MINIMUM GOVERNOR Step: TARGET SPEED, THEN press green CONTINUE button.

Standard Pressed green CONTINUE button.

Cue Notes None Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 10 of 13

18. Procedure 8.13 Control RFPT-1A speed using UP/DOWN arrows or Step: using SPEED TARGET SETPOINT to control reactor pressure.

Standard Used UP/DOWN to raise speed between 2400 and 2600 rpm or selected SPEED TARGET and entered a speed between 2400 and 2600 rpm in pop-up box.

Cue Notes Results SAT UNSAT

19. Procedure 8.14 N/A Step:

Standard Cue None Notes Injection not required for this task.

Results SAT UNSAT

20. Procedure 8.15 N/A Step:

Standard Cue None Notes Injection not required for this task.

Results SAT UNSAT

21. Procedure 8.16 N/A Step:

Standard Cue None Notes Automatic control of startup valves not required.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 11 of 13

22. Procedure 8.17 Inform CRS that RFPT A is being utilized as a RPV Step: Pressure Control System.

Standard Informed CRS RFPT 1A operating in RPV Pressure Control at

~2500 rpm.

Cue None Notes Injection not required for this task.

Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 12 of 13 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 157 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None

2. Malfunctions N/A N/A

3. Remotes None N/A N/A

4. Overrides None N/A N/A

a. Place Simulator in RUN.

b. Trip main turbine.

c. Scram Rx.

5. Panel Setup d. Place Mode switch to SHUTDOWN

e. Trip both feed pumps.

f. Ensure 1 Condensate Booster and 1 Condensate pump running.

g. Ensure RPV level between 15 and 35 inches.

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

CNS NRC Exam 3/2017 JPM S3 Rev 0 Page 13 of 13 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The plant had been operating at 100% on day 90 of a 2 year cycle.

2. The turbine tripped followed by a reactor scram.

3. EOPs have been entered.

4. There are no known issues with RFPT 1A turning gear.

5. RFPT 1A has not been stationary.

Initiating Cue:

You are to conduct alternate pressure control using Reactor Feed Pump A as directed in EOP support procedure 5.8.1, Section 8. RPV injection with Reactor Feed Pump 1A is NOT required at this time. Inform the CRS when RFP A is operating between 2400 and 2600 rpm.

Inform the CRS when the task is complete.

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 1 of 10 Cooper Nuclear Station Job Performance Measure NRC S4

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 2 of 10 (JPM S4)

Level Recovery During Shutdown Conditions Using LPCI ALTERNATE PATH Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

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

6. PSA Applicability: Top 10 Risk Significant System RHR

References:

1. Procedure 2.2.69.1, RHR LPCI Mode, Attachment 1 [RHR Systems Operation Hard Card]

General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

2. Alternate path denoted by.

Task Standard:

Upon completion of this JPM, the applicant resets the RHR injection valve isolation, opens the injection valve, starts a Loop A pump and opens the other injection valve in accordance with Procedure 2.2.69.1, RHR LPCI Mode, Attachment 1 [RHR Systems Operation Hard Card].

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 3 of 10 Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. RHR injection valve MO-25A(B) for the RHR Loop, A or B, first attempted to be placed into operation will fail to open, requiring the applicant to transition to the opposite RHR Loop.

1. This JPM evaluates the applicant's ability to recover RPV level using the LPCI Mode of RHR after a Shutdown Cooling isolation.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 4 of 10 Read the following to the JPM performer.

General Conditions:

1. The plant is in cold shutdown.

2. Level dropped to approximately -50 inches on fuel zone indicators NBI-LI-91A and NBI-LI-91B before a valve lineup error was corrected.

3. RHR Loop A and B is in a LPCI standby lineup with suction from the Torus.

Initiating Cue:

You have been directed to raise RPV water level to +10 inches narrow range using RHR Loop A or B per procedure 2.2.69.1, RHR LPCI Mode, Attachment 1 [RHR Systems Operation Hard Card].

Report to the CRS when RPV level is at or above +10 inches narrow range.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 5 of 10 PERFORMANCE:

START TIME:

1. Procedure Step: 1.1 Ensure RR-MO-53A(B) is closed.

Standard Verified green light ON and red light OFF for RR-MO-53A(B),

PUMP DISCHARGE VALVE on panel 9-4 Cue None Notes Results SAT UNSAT

2. Procedure Step: 1.2 Ensure PCIS Group 2 SDC RHR-MO-25A isolation reset by pressing:

1.2.1 SDC ISOL RESET VLV 25A Standard Depressed SDC ISOL RESET VLV 25A push-button on Panel 9-3.

Cue None Notes Results SAT UNSAT

3. Procedure Step: 1.2 Ensure PCIS Group 2 SDC RHR-MO-25(B) isolation reset by pressing:

1.2.2 SDC ISOL RESET VLV 25B Standard Depressed SDC ISOL RESET VLV 25B push-button on Panel 9-3.

Cue None Notes Results SAT UNSAT Note to Examiner: The applicant may initially choose to operate either RHR Loop A or RHR Loop B for level restoration. The respective injection valve, MO-25A(B), will fail to open for whichever RHR Loop is first attempted.

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 6 of 10

4. Procedure Step: 1.3 Ensure RHR-MO-25A(B) is open.

Standard Momentarily placed control switch for RHR-MO-25A(B),

INBD INJECTION VLV. to OPEN. Determined valve will not open. Green light ON, Red light OFF on Panel 9-3.

Cue None Notes Only bolded part of standard is critical.

Results SAT UNSAT

5. Procedure N/A.

Step:

Standard Reported RHR-MO-25A(B) will not open to CRS.

Cue CRS acknowledges.

Notes Results SAT UNSAT

4. Procedure Step: 1.1 Ensure RR-MO-53A(B) is closed.

Standard Verified green light ON and red light OFF for RR-MO-53A(B), PUMP DISCHARGE VALVE on panel 9-4 Cue None Notes Results SAT UNSAT

5. Procedure Step: 1.3 Ensure RHR-MO-25A(B) is open.

Standard Momentarily placed control switch for RHR-MO-25A(B),

INBD INJECTION VLV. to OPEN. Verified red light ON, green light OFF on Panel 9-3.

Cue None Notes Only bolded part of standard is critical.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 7 of 10

6. Procedure 1.4 Start RHR pump(s).

Step:

Standard Placed control switch for RHR Pump A or C (B or D) to START. Verified pump started, red light ON, green light OFF on Panel 9-3.

Cue None Notes Only bolded part of standard is critical.

Results SAT UNSAT

7. Procedure 1.5 Throttle RHR-MO-27A(B), as required.

Step:

Standard Rotated control switch for RHR-MO-27A(B), A(B) OUTBD INJECTION VLV, clockwise to OPEN and verified Red light ON and Green light ON. Verified system flow at desired value on RHR-FI-133A(B) on Panel 9-3.

Cue None Notes Only bolded part of standard is critical.

Results SAT UNSAT

8. Procedure Step 1.6 is N/A.

Step:

Standard N/A Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 8 of 10

9. Procedure 1.7 If PCIS Group 6 lights lit on Panel 9-5, THEN ensure one Step: of the following open:

1.7.1 REC-MO-711 Standard Verified REC-MO-711 open, red light ON, green light OFF on VBD-M.

Cue None Notes Results SAT UNSAT

10. Procedure 1.7 If PCIS Group 6 lights lit on Panel 9-5, THEN ensure one Step: of the following open:

1.7.2 REC-MO-714 Standard OR Verified REC-MO-714 open, red light ON, green light OFF on VBD-M.

Cue None Notes Results SAT UNSAT

11. Procedure N/A Step:

Standard Informed the CRS RHR Loop A(B) is injecting to the reactor vessel and level is > +10 inches on the narrow range.

Cue None Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 9 of 10 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 158 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers zdirhrsws8a[2]==1 3

dmf rh04b zdirhrsws8b[2]==1 4

dmf rh04a

2. Malfunctions N/A

3. Remotes None N/A N/A

4. Overrides None RHR-MO-25B an:p1255 OFF Valve overload RHR-MO-25A an:p1134 OFF Valve overload

a. At VBD-M, ensure REC-MO-711 or REC-MO-714, associated with the in-service REC HX is open.

5. Panel Setup b. Place Simulator in RUN and insert RR20A and RR20B at 100%.

c. Allow RPV water level to lower to ~ -50 inches on NBI-LI-91A (takes approximately 10 minutes), then remove the malfunctions.

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

CNS NRC Exam 3/2017 JPM S4 Rev 0 Page 10 of 10 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The plant is in cold shutdown.

2. Level dropped to approximately -50 inches on fuel zone indicators NBI-LI-91A and NBI-LI-91B before a valve lineup error was corrected.

3. RHR Loop A and B are in a LPCI standby lineup with suction from the Torus.

Initiating Cue:

You have been directed to raise RPV water level to +10 inches narrow range using RHR Loop A or B per procedure 2.2.69.1, RHR LPCI Mode, Attachment 1 [RHR Systems Operation Hard Card].

Report to the CRS when RPV level is at or above +10 inches narrow range.

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 1 of 11 Cooper Nuclear Station Job Performance Measure NRC S5

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 2 of 11 (JPM S5)

Perform 6.TG.303 Testing OPC Overspeed Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 14 minutes

5. NRC K/A 241000 A3.12 (2.9/2.9), 241000 A4.19 (3.5/3.4)

6. PSA Applicability: N/A

References:

1. Procedure 6.TG.303, Main Turbine Overspeed Trip Test General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant enabled OPC 20-1 and OPC 20-2 test on DEH HMI and selected OPC 20-1 and 20-2 test in accordance with Surveillance Procedure 6.TG.303.

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 3 of 11 Directions to Examiner:

1. This JPM evaluates the applicant's ability to perform OPC 20-1 and OPC 20-2 overspeed test surveillance.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 4 of 11 Read the following to the JPM performer.

General Conditions:

1. The plant is shut down.

Initiating Cue:

You are to perform turbine overspeed testing per Surveillance Procedure 6.TG.303, Section 5. Another operator will perform the Restoration and Acceptance Criteria Sections.

Inform the CRS when the task is complete.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 5 of 11 PERFORMANCE:

START TIME:

1. Procedure Step: 5.1.1 Test OPC 20-1 by performing following: On OPC/OS TEST screen, OPC control, push TEST ENABLE button.

Standard Ensured Group 2 selected. Selected OPC/OS TEST Screen and pushed TEST ENABLE button.

Cue None Notes Results SAT UNSAT

2. Procedure Step: 5.1.2 Push YES to enable OPC TEST.

Standard Pushed YES on the Enable OPC TEST pop up.

Cue None Notes Results SAT UNSAT

3. Procedure Step: 5.1.3 Select OPC 20-1 TEST.

Standard Selected OPS 20-1 TEST.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 6 of 11

4. Procedure Step: 5.1.4 Push YES to enact OPC 20-1 TEST.

Standard Pushed YES on the TEST OPC 20-1 popup.

Cue None Notes Results SAT UNSAT

5. Procedure 5.1.5 Monitor turbine speed lowering, and governor valves Step: and intercept valves closing.

Standard Selected Group 1 and Main Display and monitored turbine speed and governor valves and intercept valves changing.

Cue None Notes Results SAT UNSAT

6. Procedure 5.1.6 AC Check governor (< 2%) and interceptor valves Step: close.

Standard Checked governor valves <2% open and interceptor valves closed.

Cue None Notes Results SAT UNSAT

7. Procedure 5.1.7 Monitor turbine speed returning to 1800 rpm and OPC Step: 20-1 TEST button returning to shadow condition.

Standard Selected Group 2 and OPC/OS TEST Screen and monitored turbine speed returning to 1800 rpm and OPT 20-1 TEST button returning to shadow condition.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 7 of 11

8. Procedure 5.1.8 Verify TEST ENABLE no longer backlit yellow.

Step:

Standard Observed TEST ENABLE yellow backlight extinguishes.

Cue None Notes Results SAT UNSAT

9. Procedure 5.2.1 Test OPC 20-2 by performing following: On OPC/OS Step: TEST screen, OPC control, push TEST ENABLE button.

Standard Ensured Group 2 selected. Selected OPC/OS TEST Screen and pushed TEST ENABLE button.

Cue None Notes Results SAT UNSAT

10. Procedure 5.2.2 Push YES to enable OPC TEST.

Step:

Standard Pushed YES on the Enable OPC TEST pop up.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 8 of 11

11. Procedure 5.2.3 Select OPC 20-2 TEST.

Step:

Standard Selected OPS 20-2 TEST.

Cue None Notes Results SAT UNSAT

12. Procedure 5.2.4 Push YES to enact OPC 20-2 TEST.

Step:

Standard Pushed YES on the TEST OPC 20-1 popup.

Cue Notes Results SAT UNSAT

13. Procedure 5.2.5 Monitor turbine speed lowering, and governor valves Step: and intercept valves closing.

Standard Selected Group 1 and Main Display and monitored turbine speed and governor valves and intercept valves changing.

Cue None Notes Results SAT UNSAT

14. Procedure 5.2.6 AC Check governor (< 2%) and interceptor valves Step: close.

Standard Checked governor valves <2% open and interceptor valves closed.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 9 of 11

15. Procedure 5.2.7 Monitor turbine speed returning to 1800 rpm and OPC Step: 20-2 TEST button returning to shadow condition.

Standard Selected Group 2 and OPC/OS TEST Screen and monitored turbine speed returning to 1800 rpm and OPT 20-1 TEST button returning to shadow condition.

Cue None Notes Results SAT UNSAT

16. Procedure 5.2.8 Verify TEST ENABLE no longer backlit yellow.

Step:

Standard Observed TEST ENABLE yellow backlight extinguishes.

Cue None Notes Results SAT UNSAT

17. Procedure Step:

Standard Reported to the CRS that 6.TG.303, Section 5 is complete.

Cue CRS acknowledges the report.

Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 10 of 11 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 160 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None

2. Malfunctions N/A

3. Remotes None N/A N/A

4. Overrides None N/A N/A

a. Mark up 6.TG.303 with signatures giving permission to perform

5. Panel Setup surveillance.

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

CNS NRC Exam 3/2017 JPM S5 Rev 0 Page 11 of 11 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The plant is shut down.

Initiating Cue:

You are to perform turbine overspeed testing per Surveillance Procedure 6.TG.303, Section 5. Another operator will perform the Restoration and Acceptance Criteria Sections.

Inform the CRS when the task is complete.

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 1 of 10 Cooper Nuclear Station Job Performance Measure NRC S6

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 2 of 10 (JPM S6)

Perform Standby Gas Treatment System Decay Heat Removal Revision Statement: Simplified format of bank JPM SKL034-20-70 (Rev 12).

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 10 minutes

5. NRC K/A 261000 A3.04 (3.0/3.1), 261000 A4.03 (3.0/3.0)

6. PSA Applicability: N/A

References:

1. Procedure 2.2.73, Standby Gas Treatment System General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

Upon completion of this JPM, the applicant places SGT B in service, secures SGT A and opens SGT A dilution air valve in accordance with Procedure 2.2.73, Section 10.

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 3 of 10 Directions to Examiner:

1. This JPM evaluates the applicant's ability to place SGT A train into decay heat removal mode.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 4 of 10 Read the following to the JPM performer.

General Conditions:

1. The plant is in the post-LOCA mode of operation.

2. SGT A is in service. SGT B is in STANDBY.

3. A Group 6 isolation signal is present.

Initiating Cue:

You are to perform Standby Gas Treatment system decay heat removal for SGT A per Procedure 2.2.73 Section 10.

Inform the CRS when SGT A Decay Heat Removal is in progress.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 5 of 10 PERFORMANCE:

START TIME:

1. Procedure Step: 10.1.1 IF SGT Subsystem A requires decay heat removal, THEN perform following:

10.1.1.1 Ensure EF-R-1F, SGT B EXHAUST FAN, switch in RUN and check following: a. EF-R-1F starts.

Standard Rotated EF-R-1F control switch clockwise to RUN position and released. Checked EF-R-1F red light ON green light OFF.

Cue None Notes Only bolded portion is critical Results SAT UNSAT

2. Procedure Step: b. SGT-AO-250, SGT B INLET, opens.

Standard Checked SGT-AO-250 red light ON green light OFF.

Cue None Notes Results SAT UNSAT

3. Procedure Step: c. SGT-AO-252, SGT B DISCHARGE, opens Standard Checked SGT-AO-252 red light ON green light OFF.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 6 of 10

4. Procedure Step: 10.1.1.2 Place EF-R-1E, SGT A EXHAUST FAN, switch to OFF and check following: a. EF-R-1E stops.

Standard Rotated EF-R-1E control switch counter-clockwise to OFF and released. Checked green light ON and red light OFF.

Cue None Notes Only bolded portion is critical Results SAT UNSAT

5. Procedure b. SGT-AO-249, SGT A INLET, closes Step:

Standard Checked SGT-AO-249 green light ON and red light OFF.

Cue None Notes Results SAT UNSAT

6. Procedure c. SGT-AO-251, SGT A DISCHARGE, closes.

Step:

Standard Checked SGT-AO-251 green light ON and red light OFF.

Cue None Notes Results SAT UNSAT

7. Procedure 10.1.1.3 Open SGT-AO-270, SGT A DILUTION AIR, by Step: placing the switch to AUTO.

Standard Rotated SGT-AO-270 control switch clockwise to the AUTO position.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 7 of 10

8. Procedure 10.1.1.4 AFTER allowing time for pressure to stabilize, THEN Step: check pressure on HV DPR 835, RX BLDG/ATMOS DP (VBD-R), being maintained at 0.25" wg.

Standard Observed HV-DPR-835 indicating < 0.25" wg.

Cue None Notes Results SAT UNSAT

9. Procedure 10.1.1.5 Continue to operate EF-R-1F until temperature on Step: SGT-TI-537A, CARBON OUTLET TEMP, can be maintained below 200°F.

Standard Observed SGT-TI-537A temperature.

Cue None Notes Results SAT UNSAT

10. Procedure Step:

Standard Informed the CRS SGT train A Decay Heat Removal is in progress.

Cue None Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 8 of 10 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 161 C. Batch File jpm/342070A and jpm/342070 D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers 10 trgset 10 "zdisgtswefre[1]==1" (SGT A c/s in OFF) 12 trgset 12 "zdisgtswefre[1]==1" (SGT A c/s in OFF)

2. Malfunctions This loads with batch file jpm/342070:

MS01a A 20 Steam Leak inside PC MSL-A

3. Remotes None N/A N/A

4. Overrides This loads with batch file 18A2M06 SGT TI jpm/342070: A 205 537A ZAOSGTTI537A This loads with batch file 18A2M08 SGT TI jpm/342070: A 205 547 ZAOSGTTI547 This loads with batch file K-1/A-1 SGT A jpm/342070: high outlet high on an:p4540 temp This loads with batch file 18A2M06 SGT TI jpm/342070A: 10 190 1:00 537A ZAOSGTTI537A This loads with batch file 18A2M08 SGT TI jpm/342070A: 12 190 1:00 547 ZAOSGTTI547

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 9 of 10

a. Load batch file jpm/342070.

b. Wait until the Group 6 comes in, then proceed to Step c.

c. Perform Procedure 2.2.73, Section 4 for RESPONSE TO AUTOMATIC INITIATION through Step 4.7.

d. Ensure SGT A running and control switch in RUN.

5. Panel Setup

e. Ensure SGT B NOT running an d control switch in STANDBY.

f. Ensure SGT-AO-270 is closed.

g. Ensure batch file jpm/342070A is in the batch file folder.

h. Ensure Reactor Building dP < -25 WG on setup.

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

Batch file 342070a (Batch file loads when SGT A fan control switch is taken to OFF and simulates cooling SGT A via Decay Heat removal activity).

mor zaosgtti537a 190 1:00 mor zaosgtti547 190 1:00

CNS NRC Exam 3/2017 JPM S6 Rev 0 Page 10 of 10 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The plant is in the post-LOCA mode of operation.

2. SGT A is in service. SGT B is in STANDBY.

3. A Group 6 isolation signal is present.

Initiating Cue:

You are to perform Standby Gas Treatment system decay heat removal for SGT A per Procedure 2.2.73 Section 10.

Inform the CRS when SGT A Decay Heat Removal is in progress.

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 1 of 12 Cooper Nuclear Station Job Performance Measure NRC S7

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 2 of 12 (JPM S7)

Withdrawal Of Control Rod From Position 00 (Alternate Path 2)

ALTERNATE PATH Revision Statement: This is a new 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 201003 A2.01 (3.4/3.6)

6. PSA Applicability: N/A

References:

1. Procedure 2.2.8, Control Rod Drive Hydraulic System General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

2. Alternate path denoted by.

Task Standard:

Upon completion of this JPM, the applicant raised CRD drive water dP and withdrew control rod 10-11 to position 02 utilizing the RMCS in accordance with Procedure 2.2.8, Control Rod Drive Hydraulic System.

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 3 of 12 Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. Control rod 10-11 cannot be withdrawn from position 00 normally. CRD Drive water pressure must be raised above 290 psid, then the control rod can be withdrawn.

1. This JPM evaluates the applicant's ability to withdraw a control rod that is temporarily stuck at position 00.

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 2.

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 4 of 12 Read the following to the JPM performer.

General Conditions:

1. Plant startup is in progress.

2. Rod Sequence is: Rod Group 9/1, Step 1.

3. Normal attempts to withdraw control rod 10-11 from position 00 have failed.

4. Control rod 10-11 is selected.

Initiating Cue:

You are directed to withdraw control rod 10-11 to position 02 per Procedure 2.2.8, Control Rod Drive Hydraulic System, Section 32. Reactor Engineering has determined double notching is allowed.

Inform the CRS when the control rod has been withdrawn to position 02.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 5 of 12 PERFORMANCE:

START TIME:

1. Procedure Step: 32.1 Reactor Engineering should be consulted for additional guidance if the control rod should move past its intended target position when performing this section.

Standard Reviewed the step.

Cue None Notes Results SAT UNSAT

2. Procedure Step: 32.2.1 Place ROD MOVEMENT CONTROL switch to OUT NOTCH and HOLD.

Standard Rotated ROD MOVEMENT CONTROL switch clockwise to the OUT NOTCH position and held switch.

Cue None Notes Control rod does not withdraw.

Results SAT UNSAT

3. Procedure Step: 32.2.2 Place EMERGENCY NOTCH OVERRIDE switch to EMER ROD IN and release.

Standard Rotated EMERGENCY NOTCH OVERRIDE switch clockwise to the EMER ROD IN position and released.

Cue None Notes Control rod does not withdraw.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 6 of 12

4. Procedure Step: 32.2.3 IF first attempt is unsuccessful, THEN repeat Step 32.2.2 several times or until outward rod movement occurs.

Standard Repeated steps multiple times.

Cue After step is repeated once, cue applicant several is met.

Notes Control rod does not withdraw.

Results SAT UNSAT

5. Procedure 32.2.4 Release ROD MOVEMENT CONTROL switch.

Step:

Standard Released ROD MOVEMENT CONTROL switch allowing it to return to the OFF position.

Cue None Notes Control rod does not withdraw.

Results SAT UNSAT

6. Procedure 32.3.1 Place EMERGENCY NOTCH OVERRIDE switch to Step: EMER ROD IN and hold for several seconds.

Standard Rotated EMERGENCY NOTCH OVERRIDE switch clockwise to the EMER ROD IN position and held for several seconds.

Cue None Notes Control rod does not withdraw.

Results SAT UNSAT

7. Procedure 32.3.2 Simultaneously release EMERGENCY NOTCH Step: OVERRIDE switch and place ROD MOVEMENT CONTROL switch to OUT NOTCH.

Standard Released EMERGENCY NOTCH OVERRIDE switch and simultaneously placed ROD MOVEMENT CONTROL switch to OUT NOTCH.

Cue None

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 7 of 12 Notes Control rod does not withdraw.

Results SAT UNSAT

8. Procedure 32.3.3 IF first attempt is unsuccessful, THEN repeat Steps Step: 32.3.1 and 32.3.2 several times.

Standard Repeated previous two steps several times.

Cue After step is repeated once, cue applicant several is met.

Notes Control rod does not withdraw.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 8 of 12 NOTE to Examiner: When trainee reads next step inform trainee double-notching cannot be tolerated.

9. Procedure 32.3.4 Step is N/A.

Step:

Standard Step not performed.

Cue Double-notching cannot be tolerated.

Notes Results SAT UNSAT

10. Procedure 32.3.5.1 Place EMERGENCY NOT OVERRIDE switch to Step: EMER ROD IN and hold for several seconds.

Standard Rotated EMERGENCY NOTCH OVERRIDE switch clockwise to the EMER ROD IN position and held for several seconds.

Cue None Notes Results SAT UNSAT

11. Procedure 32.3.5.2 Simultaneously place EMERGENCH NOTCH Step: OVERRIDE switch to OVERRIDE and place ROD MOVEMENT CONTROL switch to OUT NOTCH.

Standard Released EMERGENCY NOTCH OVERRIDE switch and simultaneously placed ROD MOVEMENT CONTROL switch to OUT NOTCH.

Cue Notes Control rod does not withdraw.

Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 9 of 12

12. Procedure 32.3.5.3 If outward rod movement occurs, THEN immediately Step: release EMERGENCY NOTCH OVERRIDE and ROD MOVEMENT CONTROL switches.

Standard After observing no rod movement, released EMERGENCY NOTCH OVERRIDE and ROD MOVEMENT CONTROL switches.

Cue Notes Results SAT UNSAT

13. Procedure 32.3.5.4 If first attempt is unsuccessful, THEN repeat Steps Step: 32.3.5.1, 32.3.5.2, and 32.3.5.3 several times.

Standard Repeated steps 32.3.5.1 through 32.3.5.3 several times.

Cue After step is repeated once, cue applicant several is met.

Notes Results SAT UNSAT NOTE to Examiner: Either way of obtaining 290 to 310 psig are allowed.

14. Procedure 32.4 IF Step 32.2 is unsuccessful, THEN 32.4.1 Raise drive Step: water P to 290 to 310 psid.

Standard Manipulated DRIVE PRESS CONT VALVE MO 20 counter-clockwise CLOSE (to raise)/clockwise to OPEN (to lower) as necessary to obtain 290 to 310 psid on CR WTR DP, CRD-DPI-303.

OR If CRD-FC-301 is in BAL, then adjust setpoint thumbwheel to obtain 290 to 310 psid on CR WTR DP, CRD, DPI-303.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 10 of 12

15. Procedure 32.4.2 Attempt rod withdrawal using normal not out Step: method.

Standard Rotated ROD MOVEMENT CONTROL switch clockwise to the OUT NOTCH position and released.

Cue None Notes Results SAT UNSAT

16. Procedure N/A Step:

Standard Informed CRS control rod 10-11 had been withdrawn to position 02.

Cue None Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 11 of 12 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 160 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

Event Action: ZAOCRDDPI303 .GT.290

1. Triggers Command: dmf rd121011

2. Malfunctions Control Rod 10-11 RD121011 N/A A 0 N/A N/A N/A stuck

3. Remotes None N/A N/A

4. Overrides None Stuck rod individual scram N/A N/A switch.

a. Insert trigger information into Event Trigger popup.

b. Ensure correct Control Rod Sequence book at 9-5.

c. Ensure Control Rod Sequence book marked indicating current control

5. Panel Setup rod movement attempt. (Rod Group 9/1, Step 1)

d. With draw control rods per sequence package until control rod 10-11 is the next rod to be withdrawn.

e. Select control rod 10-11.

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

CNS NRC Exam 3/2017 JPM S7 Rev 0 Page 12 of 12 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. Plant startup is in progress.

2. Rod Sequence is: Rod Group 9/1, Step 1.

3. Normal attempts to withdraw control rod 10-11 from position 00 have failed.

4. Control rod 10-11 is selected.

Initiating Cue:

You are directed to withdraw control rod 10-11 to position 02 per Procedure 2.2.8, Control Rod Drive Hydraulic System, Section 32. Reactor Engineering has determined double notching is allowed.

Inform the CRS when the control rod has been withdrawn to position 02.

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 1 of 14 Cooper Nuclear Station Job Performance Measure NRC S8

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 2 of 14 (JPM S8)

Verify Group 2 Isolation (Alt Path TIP Shear)

ALTERNATE PATH Revision Statement: Simplified format of bank JPM SKL034-21-66, Rev 6..

Additional Program Information:

1. Appropriate Performance Locations: SIM

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Perform

4. Performance Time: 14 minutes

5. NRC K/A 223002 A4.01 (3.6/3.5), 223002 A4.06 (3.6/3.7)

6. PSA Applicability: Top 10 Risk Significant System - Primary Containment (Isolation)

References:

1. Procedure 2.1.22, Recovering From A Group Isolation

2. Procedure 4.1.4, Traversing In-Core Probe System General Tools and Equipment:

1. Simulator set up IAW Attachment 1 Critical Elements:

1. Critical steps denoted in bold.

2. Alternate path denoted by.

Task Standard:

Upon completion of this JPM, the applicant verified a PCIS Group 2 Isolation, recognized TIP A Ball Valve failed to close, and attempted to close the Ball Valve in accordance with Procedure 2.1.22, Recovering From A Group Isolation, and isolated the penetration by firing the TIP A Shear Valve, in accordance with Procedure 4.1.4, Traversing In-Core Probe System.

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 3 of 14 Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. The TIP Ball valve will have failed to close. The applicant will identify the failure and actuate the TIP tubing shear valve to isolate the penetration.

Note: DO NOT give Procedure 4.1.4 to applicant until requested at JPM step 4.

1. This JPM evaluates the applicant's ability to verify a PCIS Group 2 Isolation and determine the TIP Ball valve failed to close and effect isolation of the penetration.

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 2. (DO NOT give Procedure 4.1.4 to applicant.)

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

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 4 of 14 Read the following to the JPM performer.

General Conditions:

1. The plant Scrammed on Low Reactor Water level.

2. The low level resulted in a PCIS Group 2 isolation.

3. There are indications of a coolant leak inside the drywell in progress.

Initiating Cue:

You have been directed to perform the actions associated with verifying the Groups 2 and 6 Isolation using Procedure 2.1.22 Attachment 1, Group Isolation Hard Card. Verify Group 2 first and then Group 6.

Notify the CRS when you have completed the required actions.

NOTE: Ensure the Simulator is set up IAW Attachment 1 and is in RUN, and tell the trainee to begin.

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 5 of 14 PERFORMANCE:

START TIME:

1. Procedure Step: N/A Standard Obtained GROUP ISOLATION HARD CARD Cue None Notes Results SAT UNSAT NOTE to Examiner: Shutdown Cooling was NOT in service prior to the event.

2. Procedure Step: 5.2.1 Step is N/A Standard Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 6 of 14

3. Procedure Step: 5.2.2.1 through 5.2.2.36 Standard Ensured the following Group 2 valves close: Green light ON Red light OFF.

RHR-MO-920 ____ RHR-MO-921 ____ RW-AO-82 ____ RW-AO-83 ____

RHR-MO-274A ____ RHR-MO-274B ____ RW-AO-94 ____ RW-AO-95 ____

RHR-MO-25A ____ RHR-MO-25B ____ PC-MO-1306____ PC-MO-1305 ____

RHR-MO-18 ____ RHR-MO-17 ____ PC-MO-1304____ PC-MO-1303 ____

RHR-SSV-60 ____ RHR-SSV-61____ PC-MO-1311____ PC-MO-1310 ____

RHR-SSV-95 ____ RHR-SSV-96____ PC-MO-1302____ PC-MO-1312 ____

RHR-MO-57 ____ RHR-MO-67 ____ PC-MO-1308____ PC-MO-1301 ____

RMV-AO-10 ____ RMV-AO-11 ____ TIP BALL VALVES ____

RMV-AO-12 ____ RMV-AO-13 ____

Cue None Notes Refer to attached Figure 1 and 2 showing Group 2 valves highlighted yellow on Hard Card.

Results SAT UNSAT

4. Procedure N/A Step:

Standard Obtained Procedure 4.1.4.

Cue None Notes Results SAT UNSAT NOTE to Examiner: This step was performed using the Hard Card.

5. Procedure 6.1 lF containment ISOLATION VALVE POSITIONS (Panel 9-Step: 3) display red TIP VALVES LIGHT REMAINS on, THEN at Panel 9-13, check ball valve positions to determine which valve remain in open position.

Standard Checked TIP Machine A ball valve indicating open- red light ON.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 7 of 14

6. Procedure 6.2 For affected TIP drives:

Step: 6.2.1 Manually retract TIP drives as follows:

6.2.1.1 Place MODE switch to MAN A [ ]

Standard At TIP Drive Control Panel A, rotated MODE switch to MAN position.

Cue None Notes Results SAT UNSAT

7. Procedure 6.2.1.2 Place MAN. VALVE CONTROL switch to OPEN:

Step:

Standard At TIP Drive Control Panel A, verified MAN. VALVE CONTROL switch in OPEN and verified ball valve was open.

Cue None Notes Results SAT UNSAT

8. Procedure 6.2.1.3 Verify red BALL VALVE OPEN light on VALVE Step: CONTROL MONITOR on.

A[ ]

Standard Observed red BALL VALVE OPEN light illuminated.

Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 8 of 14

9. Procedure 6.2.1.4 Place MANUAL switch to REV.

Step:

Standard Rotated MANUAL switch to the REV. position.

Cue None Notes Results SAT UNSAT

10. Procedure 6.2.1.5 Check lN-SHIELD light status. A [ ].

Step:

Standard Observed IN-SHIELD light Off.

Cue None Notes Results SAT UNSAT

11. Procedure a IF IN-SHIELD light remains OFF and there are indications of Step: reactor coolant leak in the drywell, THEN PROCEED to Step 6.3.

Standard Proceeded to Step 6.3 Cue None Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 9 of 14

12. Procedure 6.3 IF attempt to close affected ball valves failed and there Step: are indications of a reactor coolant leak in the drywell, THEN, PERFORM following for affected TIP Drives:

6.3.1 PLACE TIP SQUIB VALVE to fire. A [ ]

Standard Inserted a key in TIP SQUIB VALVE keylock and rotated switch clockwise to the FIRE position.

Cue Notes Results SAT UNSAT

13. Procedure 6.3.2 VERIFY amber SQUIB MONITOR light on A [ ].

Step:

Standard Verified amber SQUIB MONITOR light illuminated.

Cue Notes Results SAT UNSAT

14. Procedure 6.3.3 VERIFY amber SHEAR VLV MONITOR light on. A [ ].

Step:

Standard Verified amber SHEAR VLV MONITOR light illuminated.

Cue Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 10 of 14

15. Procedure N/A Step:

Standard Notified the CRS that the TIP Shear valve has been fired and the line is isolated.

Cue Tell trainee the JPM is complete.

Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 11 of 14 Examiner KEY - DO NOT GIVE TO APPLICANT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 12 of 14 Examiner KEY - DO NOT GIVE TO APPLICANT

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 13 of 14 ATTACHMENT 1 SIMULATOR SET-UP A. Materials Required None B. Initialize the Simulator in IC 157 C. Run Batch File None D. Change the simulator Number Title Tgr TD Sev Ramp Initial conditions as follows:

1. Triggers None

2. Malfunctions RR20a A RR Loop leak N/A N/A 20 N/A N/A NM10a TIP stuck in core N/A N/A N/A N/A N/A

3. Remotes None N/A N/A

4. Overrides

a. Place simulator in Run.

b. Place TIP Machine A ON

c. MODE switch in Auto.

d. Place toggle switch for the in-shield limit switches above the TIP machines to ON.

e. Place MAN. VALVE CONTROL switch to OPEN.

5. Panel Setup f. Drive TIP into the core area and insert malfunction NM10a and verify the detector stops within the core region.

g. Put in malfunction RR20a and ensure a Group 2 is in.

h. Place Reactor Mode Switch in SHUTDOWN.

i. Place HPCI AOP in PTL.

j. Ensure PCIS Display for TIP Ball Valves indicates OPEN and that TIP Ball Valve A indicates OPEN.

CNS NRC Exam 3/2017 JPM S8 Rev 0 Page 14 of 14 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The plant Scrammed on Low Reactor Water level.

2. The low level resulted in a PCIS Group 2 isolation.

3. There are indications of a coolant leak inside the drywell in progress.

Initiating Cue:

You have been directed to perform the actions associated with verifying the Groups 2 and 6 Isolation using Procedure 2.1.22 Attachment 1, Group Isolation Hard Card. Verify Group 2 first and then Group 6.

Notify the CRS when you have completed the required actions.

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 1 of 8 Cooper Nuclear Station Job Performance Measure NRC P1

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 2 of 8 (JPM P1)

Securing Fire Pump C Locally (Alternate Path)

ALTERNATE PATH Revision Statement: This is a new JPM.

Additional Program Information:

1. Appropriate Performance Locations: PLANT

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 10 minutes

5. NRC K/A 286000 A4.05 (3.3/3.3), A1.05 (3.2/3.2)

6. PSA Applicability: N/A

References:

1. Procedure 2.2.30, Fire Protection System General Tools and Equipment:

1. none Critical Elements:

1. Critical steps denoted in bold.

2. Alternate path denoted by .

Task Standard:

Upon completion of this JPM, the applicant attempted to secure Fire Pump C locally using the normal method, and secured and electrically isolated the pump by opening the circuit breaker and isolating means disconnect switch, in accordance with Procedure 2.2.30, Fire Protection System.

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 3 of 8 Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. Fire Pump C will fail to stop using the normal local control push button. The applicant will then have to open the associated circuit breaker and disconnect switch.

1. This JPM evaluates the applicant's ability secure Fire Pump C locally using an alternate method.

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.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 4 of 8 Read the following to the JPM performer.

General Conditions:

1. Fire Pump C is running.

2. Attempts to secure Fire Pump C from the Control Room have failed.

Initiating Cue:

You have been directed to secure Fire Pump C locally per Procedure 2.2.30, Fire Protection System, Section 9.

Report completion of the task to the control room.

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 5 of 8 PERFORMANCE:

START TIME:

1. Procedure Step: 9.1.2 At FP-PNL-C (SW Pump Room), press STOP button.

Standard Pressed the STOP button.

Cue There is no reduction of noise in the room.

Notes Results SAT UNSAT

2. Procedure Step: 9.2 IF pump has not stopped using above methods, THEN perform following at FP-PNL-C:

9.2.1 Place CIRCUIT BREAKER (DISCONNECTING MEANS) to OFF.

Standard Rotated CIRCUIT BREAKER counter-clockwise to OFF.

Cue With writing instrument, point to mimic breaker pointing to OFF.

Notes Results SAT UNSAT

3. Procedure Step: 9.2.2 Press RELEASE button and hold.

Standard Depressed ISOLATING MEANS RELEASE button and held.

Cue Button depresses and remains depressed.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 6 of 8

4. Procedure Step: 9.2.3 Place ISOLATING MEANS DISCONNECT to OFF.

Standard Rotated ISOLATING MEANS DISCONNECT counter-clockwise to OFF.

Cue With writing instrument, point to mimic disconnect pointing to OFF.

Notes Results SAT UNSAT

5. Procedure 9.2.4 Release the RELEASE button.

Step:

Standard Released RELEASE button.

Cue Button moved outward and Fire Pump C noise is rapidly lowering.

Notes Results SAT UNSAT

6. Procedure Inform control room Fire Pump C has been secured locally.

Step:

Standard Contacted control room and reported Fire Pump C has been secured locally.

Cue Control room acknowledges the report.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 7 of 8

10. Procedure N/A Step:

Standard Cue None Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM P1 Rev 0 Page 8 of 8 ATTACHMENT 1 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. Fire Pump C is running.

2. Attempts to secure Fire Pump C from the Control Room have failed.

Initiating Cue:

You have been directed to secure Fire Pump C locally per Procedure 2.2.30, Fire Protection System, Section 9.

Report completion of the task to the control room.

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 1 of 11 Cooper Nuclear Station Job Performance Measure NRC P2

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 2 of 11 (JPM P2)

Place in Service/Remove From Service Hydrogen Cylinders/Tanks ALTERNATE PATH Revision Statement: Simplified format of bank JPM SKL034-40-21, Rev 5.

Additional Program Information:

1. Appropriate Performance Locations: PLANT

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 20 minutes

5. NRC K/A 245000 A3.08 (2.5/2.6)

6. PSA Applicability: N/A

References:

1. Procedure 2.2.51, Hydrogen Gas System General Tools and Equipment:

1. none Critical Elements:

1. Critical steps denoted in bold.

2. Alternate path denoted by .

Task Standard:

Upon completion of this JPM, the applicant valves out the active hydrogen cylinders, valves in the reserve hydrogen cylinders, determines the manifold pressure is too high, bleeds down the manifold, and adjusts the pressure control valve to bring manifold pressure into specifications in accordance with Procedure 2.2.55, Hydrogen Gas System.

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 3 of 11 Directions to Examiner:

NOTE: THIS IS AN ALTERNATE PATH JPM. The reserve hydrogen cylinders supply pressure will be too high when it is valved in. The applicant will then have to adjust the hydrogen pressure regulating valve output to within specification.

1. This JPM evaluates the applicant's ability place the reserve hydrogen cylinders in service and place the active hydrogen cylinders in standby.

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 2.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 4 of 11 Read the following to the JPM performer.

General Conditions:

1. Plant is at 100% power.

2. The active hydrogen bulk storage vessels are about to be depleted.

3. H2-PRV-A, ACTIVE PRESSURE REGULATOR, is in service.

Initiating Cue:

You have been directed to place the reserve Hydrogen Cylinders in service and remove the Main Cylinders from service per Procedure 2.2.51, Hydrogen Gas System, Section 8.

Report completion of the task to the control room.

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 5 of 11 PERFORMANCE:

START TIME:

NOTE To Examiner: Attachment 1 (2 Pages) contains a picture and diagram for information. DO NOT give to applicant.

1. Procedure Step: 8.2 Place reserve vessels in service and place active vessels in STANDBY by performing following: 8.2.1 Ensure H2-85, ACTIVE VESSEL MANIFOLD OUTLET VALVE (H2 House), is closed.

Standard Rotated H2-85, ACTIVE VESSEL MANIFOLD OUTLET VALVE fully clockwise until closed.

Cue Handle rotates clockwise and resistance is felt.

Notes Results SAT UNSAT

2. Procedure Step: 8.2.2 Open H2-84, RESERVE VESSEL MANIFOLD OUTLET VALVE (H2 House).

Standard Rotated H2-84, RESERVE VESSEL MANIFOLD OUTLET VALVE fully counter-clockwise until open.

Cue Handle rotates clock-wise and resistance is felt.

Notes Results SAT UNSAT

3. Procedure Step: 8.2.3 Ensure outlet pressure for in service pressure regulator indicates 125 psig to 150 psig, adjusting per Section 7 if necessary.

Standard Observed H2-PRV-A pressure gauge.

Cue Using writing instrument, indicate H2-PI-12, H2-PRV-A OUTLET is indicating 180 psig.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 6 of 11

4. Procedure Step: 7.1.1 IF H2-PRV-A outlet pressure is > 150 psig, THEN perform following: 7.1.1.1 Close H2-90, ACTIVE PRESSURE REGULATOR INLET (H2 House Pressure Regulating Cabinet).

Standard Rotated H2-90, ACTIVE PRESSURE REGULATOR INLET fully clockwise until closed.

Cue Handle rotates clockwise and little or no spring tension is felt.

Notes Results SAT UNSAT

5. Procedure 7.1.1.2 Adjust H2-PRV-A to minimum (fully counter-Step: clockwise).

Standard Rotated handle counter-clockwise until at minimum.

Cue Handle rotates counter-clockwise and no spring tension is felt.

Notes Results SAT UNSAT

6. Procedure 7.1.1.3 Open H2-118, H2 SUPPLY FROM BOTTLE Step: MANIFOLD (H2 House).

Standard Rotated H2-118, H2 SUPPLY FROM BOTTLE MANIFOLD counter-clockwise until open.

Cue Handle rotates counter-clockwise and resistance is felt.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 7 of 11

7. Procedure 7.1.1.4 Throttle open H2-93, H2 TRUCK FILL STATION Step: VENT (H2 House), until H2-PRV-A outlet pressure is 120 psig, then close H2-93.

Standard Rotated H2-93, H2 TRUCK FILL STATION VENT counter-clockwise until H2-PRV-A outlet pressure was 120 psig and then clockwise until closed.

Cue Using writing instrument, indicate H2-PI-12, H2-PRV-A OUTLET is indicating 120 psig AND H2-93 stops turning.

Notes Results SAT UNSAT

8. Procedure 7.1.1.5 Close H2-118.

Step:

Standard Rotated H2-118 clockwise until closed.

Cue Handle rotates clockwise and resistance is felt.

Notes Results SAT UNSAT

10. Procedure 7.1.1.6 Open H2-90.

Step:

Standard Rotated H2-90 counter-clockwise until open.

Cue Handle rotates counter-clockwise and resistance is felt.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 8 of 11

11. Procedure 7.1.2 Slowly adjust H2-PRV-A clockwise until outlet Step: pressure indicates 125 psig to 150 psig, allowing time for outlet pressure to stabilize between adjustments.

Standard Rotated H2-PRV-A handle clockwise until outlet pressure indicated 125 psig to 150 psig.

Cue Handle rotates clockwise, spring tension is felt, and pressure indicates 140 psig.

Notes Results SAT UNSAT

12. Procedure 8.2.4 Inform Control Room to update H2 VESSELS IN Step: SERVICE placard on PANEL B that Reserve vessels are in service and Active vessels are in STANDBY.

Standard Informed CRS to update placard on Panel B and the Reserve vessels are in service and the Active vessels are in standby.

Cue CRS acknowledges.

Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 9 of 11 Examiner Aid - DO NOT give to applicant ATTACHMENT 1 PRV-A OUTLET PRESSURE PRV-A H2-90

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 10 of 11 Examiner Aid - DO NOT give to applicant ATTACHMENT 1 (Continued)

CNS NRC Exam 3/2017 JPM P2 Rev 0 Page 11 of 11 ATTACHMENT 2 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. Plant is at 100% power.

2. The active hydrogen bulk storage vessels are about to be depleted.

3. H2-PRV-A, ACTIVE PRESSURE REGULATOR, is in service.

Initiating Cue:

You have been directed to place the reserve Hydrogen Cylinders in service and remove the Main Cylinders from service per Procedure 2.2.51, Hydrogen Gas System, Section 8.

Report completion of the task to the control room.

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 1 of 9 Cooper Nuclear Station Job Performance Measure NRC P3

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 2 of 9 (JPM P3)

Alternate Shutdown-Locally Operate SW-MO-89B for Starting Torus Cooling Revision Statement: Simplified format of bank JPM SKL034-11-15, Rev 0.

Additional Program Information:

1. Appropriate Performance Locations: PLANT

2. Appropriate Trainee level: RO / SRO

3. Evaluation Method: Simulate

4. Performance Time: 11 minutes

5. NRC K/A 219000 A1.08 (3.7/3.6), 295016 AK2.02 (4.0/4.1)

6. PSA Applicability: Top 10 Risk Significant System - RHR/SPC

References:

1. Procedure 5.1ASD, Alternate Shutdown General Tools and Equipment:

1. Key for SW-SW-LASP IS-MO89B at MCC-Y Local Aux. Shutdown Panel Critical Elements:

1. Critical steps denoted in bold.

Task Standard:

On completion of this JPM, the applicant opened SW-MO-89B MCC-Y Local Auxiliary Shutdown Panel to achieve SWBP current of 100 amps to 136 amps IAW 5.1ASD, Alternate Shutdown, Attachment 5, Section 3.

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 3 of 9 Directions to Examiner:

1. This JPM evaluates the applicant's ability to locally operate components for placing Torus cooling in operation IAW 5.1ASD Attachment 5.

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.

Notes:

Total Time: ___________

Trainee: Examiner:

Pass Fail Examiner Signature: Date:

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 4 of 9 Read the following to the JPM performer.

General Conditions:

1. The Control Room was evacuated due to toxic fumes.

2. Torus cooling is required.

3. The ASD panel is manned.

4. The Control Building Operator is standing by the breaker for SW Booster Pump B.

Initiating Cue:

The Control Room Supervisor directs you to operate SW-MO-89B, HX-B SW DISCH VLV 89B, locally IAW 5.1ASD, Alternate Shutdown, Attachment 5, Section 3 to place Torus cooling in operation.

Inform the CRS when SW-MO-89B has been aligned with a SW Booster Pump in operation to support Torus cooling.

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 5 of 9 PERFORMANCE:

START TIME:

1. Procedure Step: N/A Standard Obtained Key for SW-SW-LASP IS-MO89B.

Cue When the applicant describes where a key would be obtained, indicate the key is in the applicants possession.

Notes Results SAT UNSAT Note All step references are from 5.1ASD Attachment 5.

2. Procedure Step: 3.1 At MCC-Y Local Aux. Shutdown Panel, insert key and place SW-SW-LASP IS-MO89B for SW-MO-89B to ISOL.

Standard Inserted key into control switch SW-SW-LASP IS-MO89B for SW-MO-89B, HX-B SW DISCH VLV 89B, and rotated control switch clockwise to ISOL position.

Cue Indicate switch is aligned with ISOL position.

Notes Results SAT UNSAT

3. Procedure Step: 3.2 Remove key from SW-SW-LASP IS-MO89B.

Standard Pulled key from SW-SW-LASP IS-MO89B and retained key.

Cue Indicate the applicant has the key.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 6 of 9

4. Procedure Step: 3.3 At MCC-Y Local Aux. Shutdown Panel, ensure SW-MO-89B is closed.

Standard Observed position indicating lights at SW-MO-89B control switch and determined valve is NOT closed.

Cue Indicate the red light is On and green light is Off.

Notes Results SAT UNSAT

5. Procedure 3.3.1 IF SW-MO89B is not fully closed, THEN place and Step: hold SW-SW-LASP CS-MO89B to CLOSE until SW-MO-89B is closed.

Standard Closed SW-MO-89B, HX-B SW DISCH VLV 89B, by rotating control switch SW-SW-LASP CS-MO89B counter-clockwise to CLOSE and holding until red light extinguished, then released.

Cue When the applicant describes rotating the control switch to CLOSE, indicate the green light is On, indicate 60 seconds have elapsed, and now, the red light is Off. Indicate the switch spring-returned to NORM when released.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 7 of 9

6. Procedure 3.4 At MCC-Y Local Aux. Shutdown Panel, throttle open SW-Step: MO-89B by holding SW-SW-LASP CS-MO89B to OPEN for 15 seconds.

Standard Throttled open SW-MO-89B, HX-B SW DISCH VLV 89B, by rotating control switch SW-SW-LASP CS-MO89B clockwise to OPEN and holding for approximately 15 seconds, then released.

Cue When the applicant describes rotating the control switch to OPEN, after approximately 5 seconds indicate the red light illuminated. When the applicant states the switch has been released, indicate the switch spring-returned to NORM.

Notes Results SAT UNSAT

7. Procedure 3.4 Inform ASD Operator SW-MO-89B is positioned for SW Step: Booster pump start.

Standard Contacted ASD Operator via suitable communications method and informed SW-MO-89B, HX-B SW DISCH VLV 89B has been throttled open for SW Booster pump start .

Cue As ASD operator acknowledge the report. Then, state the Control Building Operator has started SW Booster Pump B and is ready for the applicant to perform step 3.6. State the Control Building Operator will monitor SWBP B motor amps.

Notes Results SAT UNSAT

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 8 of 9 Note to Examiners: The applicant may intermittently place the control switch to OPEN in the following step, or may hold it in OPEN.

8. Procedure 3.6 WHEN SW Booster pump is running, THEN Step: coordinate with Control Building Operator to open SW-MO-89B, limiting pump amps to < 136 amps.

Standard Throttled open SW-MO-89B, HX-B SW DISCH VLV 89B, by rotating control switch SW-SW-LASP CS-MO89B clockwise to OPEN and releasing when pump amps are between 100 to 136 amps.

Cue When applicant places the control switch to open, as the Control Building Operator report SWBP B pump amps rising and indicate ascending values followed by brief pauses (e.g.

90 amps 100 amps 110 amps 120 amps 130 amps).

Notes Results SAT UNSAT

9. Procedure N/A Step:

Standard Informed CRS SW-MO-89B has been aligned to support Torus cooling Cue As CRS, acknowledge the report.

Notes Results SAT UNSAT STOP TIME:

CNS NRC Exam 3/2017 JPM P3 Rev 0 Page 9 of 9 ATTACHMENT 1 DIRECTIONS TO TRAINEE:

Read the following and inform the evaluator when you are ready to begin.

General Conditions:

1. The Control Room was evacuated due to toxic fumes.

2. Torus cooling is required.

3. The ASD panel is manned.

4. The Control Building Operator is standing by the breaker for SW Booster Pump B.

Initiating Cue:

The Control Room Supervisor directs you to operate SW-MO-89B, HX-B SW DISCH VLV 89B, locally IAW 5.1ASD, Alternate Shutdown, Attachment 5, Section 3 to place Torus cooling in operation.

Inform the CRS when SW-MO-89B has been aligned with a SW Booster Pump in operation to support Torus cooling.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 1 of 44 Facility: Cooper Nuclear Station Scenario No.: 1 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Shift CRD Stabilizing valves.

2. Lower reactor power using RR pumps.

3. Respond to Reactor Bldg to Torus Vacuum Breaker PC-AO-243 failing open.

4. Respond to RPV flange leakage.

5. Respond to trip of RPS A EPAs with failure of RMV-AO-10 to close.

6. Respond to loss of multiple REC pumps.

7. ATWS Level Power control

8. Respond to RHR SPC valve failing to open.

Initial Conditions: Plant operating at 100% power.

Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement.

Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Shift CRD Stabilizing valves per Procedure 2.2.8.

Lower power to 95% with RR Pumps per Procedure 2.1.10.

Electrical Maintenance working on replacing HPCI AOP motor.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 2 of 44 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Shift CRD stabilizing valves 2 N/A R (ATC, CRS) Lower Reactor power by lowering RR pump speed.

(or) Reactor Building to Torus vacuum breaker fails open.

3 TS (CRS) zdipcswcs243av[2] CRS declares vacuum breaker inoperable.

C (BOP,CRS) Respond to reactor vessel flange seal leak alarm, 4 rr21 enter Procedure 4.6.3, and cycle the flange leak-off A (CREW) drain valves.

I rp03a (BOP,ATC,CRS) RPS EPA Breaker 1A1/1A2 trip, (half scram and half 5 PCIS group isolations) RMV-AO-10 fails to isolate.

(rf) rh32a A (CREW) CRS declares valve inoperable.

TS (CRS) sw 11a C (BOP, ATC) REC Pump A trip. Start another REC pump. REC 6

sw11b A(CREW) Pump B trip. Manual scram due to loss of REC.

Hydraulic block ATWS > 3% power (EOP-1A, 3A, 6A, 6B, 7A)

(CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

(CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

7 rd02a,b (CT-3) During failure to scram conditions with M (CREW) power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -

60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

(CT-4) When control rods fail to scram and energy is discharging to the primary containment, crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

(rf) rh29(A) First RHR loop to be put into suppression pool 8 C (BOP,CRS)

(rf) rh30(A) cooling has RHR-MO-39A(B) fail to open.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 3 of 44 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after EOP entry 1-2 1 RHR-MO-39A(B) fails to open.

RPV flange seal leak Abnormal Events 2-4 3 RPS EPA trip Loss of multiple REC Pumps Major Transients 1-2 1 ATWS EOP-3A EOP entries requiring substantive action 1-2 2 EOP-6A EOP contingencies requiring substantive 0-2 1 EOP- 7A action (CT-1) When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

(CT-2) Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

(CT-3) During failure to scram conditions with EOP based Critical power >3%, stop and prevent injection from all 2-3 4 Tasks sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -

60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

(CT-4) When control rods fail to scram and energy is discharging to the primary containment crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Normal Events N/A 1 Shift CRD Stabilizing valves.

Reactivity Manipulations N/A 1 Lower power using Reactor Recirculation pumps RPV Flange leak RPS A EPA Breaker trip Instrument/

Component Failures N/A 5 Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open RPV Flange leak RPS a EPA Breaker trip Total Malfunctions N/A 5 Loss of REC pump A Loss of REC pump B RHR-MO39A(B) valve fails to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System (Event 5)

Residual Heat Removal System in Suppression Pool Cooling Mode (Event 8)

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 4 of 44 SCENARIO

SUMMARY

The plant is operating at 100% power. HPCI Auxiliary Oil Pump motor replacement is taking place.

Event 1 After the crew takes the watch, the ATC shifts the CRD Stabilizing valves per Procedure 2.2.8. (Event 1)

Event 2 After shifting stabilizing valves, the ATC lowers power ~5% per Load Dispatcher schedule.

Event 3 (Triggered by Lead Examiner)

After lowering power the Reactor to Torus vacuum breaker PC-AO-243 fails open. The CRS enters LCO 3.6.1.7, Condition A and declares the vacuum breaker inoperable.

Event 4 (Triggered by Lead Examiner)

The RPV inner seal develops a small leak requiring the BOP cycle the leak-off isolation valves from the control room to clear the alarm.

Event 5 (Triggered by Lead Examiner)

After actions for RPV flange leakage are complete, RPS EPAs on Division I trip causing a half reactor scram and half Group 1, 2, 7, and full Group 3, 6 isolations.

RMV-AO-10 fails to isolate on the loss of RPS. The CRS enters LCO 3.6.1.3, Condition A for RMV-AO-10 failing to isolate and determines a potential LCO for TS 3.3.8.2 Condition A is required for the EPA breaker. LCO 3.3.8.2 entry is not required, since the EPA is no longer supplying RPS.

Event 6 (Triggered by Lead Examiner)

After RPS A power has been restored from the alternate supply and RRMG cooling restored, and the half scram is reset, REC Pump A trips requiring the BOP to start the standby pump per alarm procedures. Shortly after the standby pump is started, REC Pump B trips requiring entry into Emergency Procedure 5.2REC. The ATC will insert a reactor scram. The CRS will not have time to enter Technical Specifications for the REC pumps.

Event 7 (No Trigger required)

When the reactor is scrammed, a low power ATWS occurs due to hydraulic block of both scram discharge volumes, and EOP-6A and 7A are entered via EOP-1A.

Reactor power is above 3%. The crew injects SLC and/or installs the necessary PTMs to bypass interlocks and insert control rods individually via RMCS (CT-1, CT-4).

ADS is manually inhibited to prevent automatic operation (CT-2). Stop and Prevent is Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 5 of 44 required because reactor power is above 3%. RPV level is intentionally lowered below -60 inches wide range in order to lower core inlet subcooling and lower reactor power (CT-3). Only 1 Main Turbine Bypass valve is available to control RPV pressure. SRVs have to be used to supplement pressure control. Feedwater injection is available for RPV level control.

Event 9 (Automatically Triggered when opening the first MO39A(B) is attempted)

After the crew has stabilized conditions following the scram, the selected RHR suppression pool cooling loop cannot be placed into service because RHR-MO39A (B) fails to open. The BOP transfers to the other division of RHR and places it into suppression pool cooling.

After the scram has been reset twice, the control rods are allowed to be fully inserted with the next scram. The CRD transitions from ATWS to non-ATWS flowcharts, SLC injection is halted and RPV level restoration is directed.

The exercise ends when control rods are inserted, and RPV water level is being maintained between -183 inches and +54 inches.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 6 of 44 Critical Tasks (CT-1) When control rods fail to scram, (CT-2) Inhibit ADS prior to automatic ADS crew injects SLC and/or inserts control valve opening during a failure to Scram.

rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

EVENT 7 7 Safety Failure to effect shutdown of the reactor With a Reactor Scram required, reactor not significance when a RPS setting has been exceeded shut down, and conditions for ADS blowdown would unnecessarily extend the level of are met, INHIBIT ADS to prevent an degradation of the safety of the plant. This uncontrolled RPV depressurization and cold could further degrade into damage to the water injection from low pressure sources to principle fission product barriers if left prevent causing a significant power excursion.

unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail.

Cueing Manual scram is initiated and numerous ADS Timer initiated alarm on panel 9-3-1/A-1 control rods indicate beyond position 00 and reactor power not downscale on panel 9-5 indications.

Performance Operator manipulates keylocked switches for Manipulation of ADS A and ADS B Inhibit indicator SLC B pump to START on panel 9-5. switches on panel 9-3 vertical section.

Operator selects individual control rods by depressing the respective pushbutton on the panel 9-5 matrix and inserts the rod by manipulating the emergency in switch on panel 9-5.

Performance SLC B pump red light illuminated, SLC Inhibit switches click into the vertical, inhibit feedback discharge pressure rising, and SLC tank level position on panel 9-3.

lowering on panel 9-5.

Receipt of ADS inhibited alarm panel 9-3-1/D-Operator selecting and inserting control rods 1.

indicated by rod position decreasing to 00 for selected rod on panel 9-5.

Justification There is no time limit for effecting complete The 105 second ADS timer allows sufficient for the chosen reactor shutdown via boron injection or time for the crew to recognize and override performance control rod insertion. For the timeframe of automatic operation of the system. As long as limit this scenario, containment limits are not ADS is inhibited before ADS valves open, closely challenged and power oscillations are reactor pressure will not be reduced to the not experienced. However, if the failure to shutoff heads of high volume, cold water scram EOP were to be exited, other systems.

procedures would not provide the guidance necessary to achieve reactor shutdown.

Before exiting EOP-6A ensures guidance to effect reactor shutdown is not removed.

BWR Owners App. B, step RC/Q-6,RC/Q-7 App. B, step RC/Q-6 Group Appendix Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 7 of 44 Critical Tasks (CT-3) During failure to scram conditions (CT-4) When control rods fail to scram and with power >3%, stop and prevent energy is discharging to the primary injection from all sources (except boron, containment, crew injects SLC before CRD, RCIC) as necessary to lower RPV exceeding the Boron Injection Initiation water level to below -60 CFZ (or LL, as Temperature (BIIT) curve.

applicable) and control between -60 (or LL, as applicable) to -183 CFZ prior to exiting EOP-7A.

EVENT 7 7 Safety Regarding lowering level below -60 CFZ, to Failure to effect shutdown of the reactor when significance prevent or mitigate the consequences of any a RPS setting has been exceeded would large irregular neutron flux oscillations unnecessarily extend the level of degradation induced by neutronic/thermal-hydraulic of the safety of the plant. This could further instabilities, RPV water level is lowered degrade into damage to the principle fission sufficiently below the elevation of the product barriers if left unmitigated. Action to feedwater sparger nozzles. This places the shut down the reactor is required when RPS feedwater spargers in the steam space and control rod drive systems fail.

providing effective heating of the relatively cold feedwater and eliminating the potential The Boron Injection Initiation Temperature for high core inlet subcooling. For conditions (BIIT) is the greater of:

that are susceptible to oscillations, the initiation and growth of oscillations is

The highest suppression pool temperature at principally dependent upon the subcooling at which initiation of boron injection will permit the core inlet; the greater the subcooling, the injection of the Hot Shutdown Boron Weight of more likely oscillations will commence and boron before suppression pool temperature increase in magnitude. exceeds the Heat Capacity Temperature Limit.

24" below the lowest nozzle in the feedwater

The suppression pool temperature at which a sparger has been selected as the upper reactor scram is required by plant Technical bound of the RPV water level control band. Specifications.

This water level is sufficiently low that steam heating of the injected water will be at least The BIIT is a function of reactor power. If 65% to 75% effective (i.e., the temperature of boron injection is initiated before suppression the injected water will be increased to 65% to pool temperature reaches the BIIT, emergency 75% of its equilibrium value in the steam RPV depressurization may be precluded at environment). This water level is sufficiently lower reactor power levels. At higher reactor high that the capability to bypass the low power levels, however, the suppression pool RPV water level MSIV isolation should be heatup rate may become so high that the Hot able to control RPV water level with Shutdown Boron Weight of boron cannot be feedwater pumps to preclude the isolation. injected before suppression pool temperature reaches the Heat Capacity Temperature Limit Regarding lowering level below LL, the even if boron injection is initiated early in the combination of high reactor power (above the event. Since failure-to-scram conditions may APRM downscale trip), high suppression pool present severe plant safety consequences, the temperature (above the Boron Injection requirement to initiate boron injection is Initiation Temperature), and an open SRV or independent of any anticipated success of high drywell pressure (above the scram control rod insertion. When attempts to insert setpoint) are symptomatic of heat being control rods satisfactorily achieve reactor rejected to the suppression pool at a rate in shutdown, the requirement for boron injection excess of that which can be removed by the no longer exists. (Control rod insertion is Suppression Pool Cooling System. Unless directed under Step RC/Q-7 concurrently with mitigated, these conditions ultimately result in Step RC/Q-6.)

loss of NPSH for ECCS pumps taking suction on the suppression pool, containment over-pressurization, and (ultimately) loss of primary containment integrity, which in turn could lead to a loss of adequate core cooling and uncontrolled release of radioactivity to the environment. The conditions listed, combined with the inability to shut down the Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 1 Page 8 of 44 reactor through control rod insertion, dictate a requirement to promptly reduce reactor power since, as long as these conditions exist, suppression pool heatup will continue.

If torus water temperature was allowed to exceed the HCTL prior to commencing the lowering of level, a RPV depressurization would be required. Failure to completely stop RPV injection flow (with the exception of CRD and SLC) prolongs the elevated reactor power condition; thus, depositing more energy than necessary into the suppression pool.

Cueing Manual scram is initiated and numerous Manual scram is initiated and numerous control rods indicate beyond position 00 and control rods indicate beyond position 00 and reactor power >3% on panel 9-5 indications reactor power not downscale on panel 9-5 and SPDS and RPV level is >-60CFZ on indications.

SPDS.

Suppression Pool temperature rising on panel 9-3 indication.

Performance Operator manipulates Feedwater HMIs on Operator manipulates keylocked switches for indicator panel 9-5 or panel A as necessary to stop SLC A(B) pump to START on panel 9-5.

FW injection until RPV level goes below -

60CFZ.

Operator manipulates HPCI controls on panel 9-3 to stop HPCI injection until RPV level is below -60CFZ.

Performance Feedwater flow indication on panel 9-5 SLC A(B) pump red light illuminated, SLC feedback indicate zero. discharge pressure rising, SLC tank level lowering on panel 9-5.

HPCI flow indication on panel 9-3 indicates zero and/or HPCI injection MOV indicates closed.

Justification Applicability for this CT is during EOP-7A If boron injection is initiated before suppression pool for the chosen conditions where it is necessary to lower level to temperature reaches the BIIT, emergency RPV performance control power with Table 17 condition NOT met depressurization may be precluded at lower reactor (i.e. no high energy input into primary power levels. At higher reactor power levels, limit containment). There is no time limit for this however, the suppression pool heatup rate may lowering level, but it establishes margin to become so high that the Hot Shutdown Boron conditions where fuel damaging power oscillations Weight of boron cannot be injected before may theoretically occur. Before exiting EOP-7A suppression pool temperature reaches the Heat was chosen because Capacity Temperature Limit even if boron injection is other procedures would not provide the guidance initiated early in the event. Since failure-to-scram necessary to establish margin for power oscillation conditions may present severe plant safety mitigation. Before exiting EOP-7A ensures consequences, the requirement to initiate boron guidance to effect this control is not removed. injection is independent of any anticipated success of control rod insertion.

NOTE This critical task must be evaluated carefully based on the level changes. If power is reduced significantly below 3%, reactor water level may continue to rise above -60" with only CRD and SLC while driving rods this would not result in an UNSAT on this critical task.

BWR Owners App. B, Contingency #5 App. B, step RC/Q-6 Group Appendix Rev. 1

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 9 of 44 SIMULATOR SET-UP A. Materials Required None B. Initialize the simulator in IC 20, 100% power (EOL) (IC-152)

Batch File Name - None C. Change the simulator conditions as follows:

1. Auto Triggers Number File Name/Variable Description 6 zdirhrsws14b(2)==1 Place RHR-MO-39B switch to open and modify remote mrf rh29a ENER function for RHR-MO-39A to function.

7 zdirhrsws14a(2)==1 Place RHR-MO-39A switch to open and modify remote mrf rh30a ENER function for RHR-MO-39B to function.

8 zloslcsws1a[2]==1 When SLC Pump A starts, the main turbine trips after a 5 imf tc01 0:05 second time delay

2. Malfunctions Number Title Trigger TD Severity Ramp Initial rr21 Vessel Head Inner Seal Leakage 1 N/A 100 N/A N/A rp03a EPA 1A1\1A2 trip 4 N/A N/A N/A N/A sw11a REC Pump A trip 5 N/A N/A N/A N/A sw11b REC Pump B trip 5 2.5 min N/A N/A N/A rd02a ATWS south rods A N/A 45 N/A N/A rd02b ATWS north rods A N/A 35 N/A N/A tc01 Main Turbine trip 8 5 sec N/A N/A N/A Main Turbine bypass valve A tc07a A N/A 0 N/A N/A fails closed.

Main Turbine bypass valve B tc07b A N/A 0 N/A N/A fails closed.

Main Turbine bypass valve C N/A tc07c A N/A 0 N/A fails closed.

yp:lpmisor PMIS Point N977 (RMV-AO-10) 4 N/A OPEN N/A N/A (768) position Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 10 of 44

3. Remotes Number Title Trigger Value TD Ramp rr18 Un-isolate air to NBI-736/737AV 2 OPEN N/A N/A rh29a RHR-MO-39A 7 DE-ENER 3 sec N/A rh30A RHR-MO-39B 6 DE-ENER 3 sec N/A

4. Overrides Instrument Tag Trigger TD Value Ramp HPCI AOP C/S zdihpcisws20[1] N/A 0 PTL N/A HPCI AOP Green light zlophcisws20[1] N/A N/A OFF N/A HPCI Inop on SSSP zdieeswhp[1] N/A N/A PUSH-IN N/A PC-AO-243 RB to Torus zdipcswcs243av[2] 3 N/A OPEN N/A vacuum breaker RMV-AO-10 red light zlormvao10[2] 4 N/A ON N/A D. Panel Setup

1. Ensure PMIS IDTs are blank

2. Ensure RR Controllers are selected to P.

3. Place HPCI Auxiliary Oil Pump control switch to P-T-L.

4. Place caution tag on HPCI Auxiliary Oil Pump control switch.

5. Ensure copy of Procedure 2.2.8 available for turnover.

6. Ensure copy of Procedure 2.1.10 available for turnover.

7. At Panel 9-4-3, set SSST Y Voltage Adjust to Tap 2.

8. At Panel C, set SSST X Voltage Adjust to Tap 5.

9. On STARTUP TRANSFORMER BACKUP VOLTAGE BUS, placard:

TAP POSITION: 5 MAX 4473 MIN 4372 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 11 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 1 Event

Description:

Shift CRD Stabilizing Valves Time Position Applicants Action or Behavior CRS Directs ATC to shift CRD Stabilizing valve per Procedure 2.2.8, Section 17.

Role Play: The isolation valves (CRD-32 and CRD-33) are normally open Booth valves. When sent to ensure the valves are open, wait 2 minutes and Operator report they are open.

17.2 IF placing STABILIZER VALVE B in service, THEN perform following; N/A if placing STABILIZER VALVE A in service:

17.2.1 Ensure CRD-32, STABILIZING VALVE ASSEMBLY 25B INLET ISOLATION (R-903-SE), is open.

17.2.2 Ensure CRD-33, STABILIZING VALVE ASSEMBLY 25B ATC OUTLET ISOLATION (R-903-SE), is open.

17.2.3 At Panel 9-5, place STABILIZER VALVES A & B switch to B.

17.2.4 Check stabilizing flow at ~ 6 gpm on CRD-FI-216, CRD SYSTEM STABILIZER FLOW (R-903-SE near CRD-MO-20).

Booth Role Play: When asked for reading on CRD-FI-216, CRD SYSTEM Operator STABILIZER FLOW, report it is indicating ~6 gpm.

BOP Provides peer check of ATC actions.

END OF EVENT Notes Booth Proceed to next event after valve shift is complete.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 12 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 2 Event

Description:

Lower power to 95% using Reactor Recirculation flow control Time Position Applicants Action or Behavior CRS Directs ATC to lower power using Recirculation flow to 95% IAW 2.1.10.

Booth As soon as crew proceeds to reduce power, Insert Trigger 1 to cause Operator RPV flange inner seal leak to begin.

NOTE to Examiners: It takes ~ 10 minutes for the RPV flange inner seal leak (Event 4) to cause an alarm.

Role Play: As Rx Building NLO, when requested to monitor RRMG lube Booth oil temps and maintain 110-130°F, respond you will monitor RRMG lube Operator oil temps and maintain them in band.

Lowers power using Recirculation flow IAW 2.1.10:

Selects S on RR flow controllers on panel 9-4 7.4 Lowers RR pump flow (by turning speed demand counter-clockwise on one speed controller at a time and allowing conditions to stabilize before ATC adjusting other controller).

Closely monitors scoop tube position/RR pump flow response.

Closely monitors reactor power on APRMs and Main Turbine output on DEH HMI.

Repeats until desired thermal power (target 95%) achieved.

BOP Provides peer check of ATC actions.

END OF EVENT Notes Booth Proceed to next event when power reduction is complete and the RPV Flange Operator leakage alarm (9-4-1/F-5) comes in.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 13 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 3 Event

Description:

Reactor Building to Torus Vacuum Breaker fails open Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 3, causing Reactor Operation Building to Torus Vacuum Breaker fails open.

Respond to alarm H-1/A-5 SUPPR CHAMBER VACUUM RELIEF 243AV OPEN and report to CRS.

1.1 Check valve position.

1.2 Check suppression chamber and secondary containment pressures.

NOTE - When torus pressure greater than Reactor Building pressure, BOP PC-DPIS-516A indication will be positive.

1.3 Dispatch Operator to monitor PC-DPIS-516A, ATMOSPHERE TO TORUS DIFFERENTIAL PRESSURE SWITCH (R-903-B RHR HX ROOM east wall).

1.4 Check pneumatic and electrical supplies to PC-SOV-SPV243, PILOT VALVE FOR PC-AO-243.

Report to CRS the valve will not close with its control switch.

BOP Report Torus and Drywell pressures.

Role Play: If directed to verify air valves PC-569 and PC-570 closed, Booth wait 2 minutes and report valves are closed.

Operator If sent to PC-DPIS-516, report it is indicating 0.3 psig.

NOTE to Examiners: It takes ~ 10 minutes for the RPV flange inner seal leak (Event 4) to cause an alarm.

When the CRS begins to review TS for PC-AO-243 or as directed by Booth Lead Examiner, Insert Trigger 1 to cause RPV flange inner seal leak Operator to begin.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 14 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 3 Event

Description:

Reactor Building to Torus Vacuum Breaker fails open Time Position Applicants Action or Behavior Refers to LCO 3.6.1.7, Condition A and declares PC-AO-243 inoperable for closing.

CRS Required Action is to close the open vacuum breaker with a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

END OF EVENT Notes Proceed to the next event when directed by the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 15 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 4 Event

Description:

Respond to RPV Flange Leakage Time Position Applicants Action or Behavior NOTE to Examiners: RPV flange inner seal leak is already active.

Respond to Annunciator 9-4-1/F-5, VESSEL FLANGE SEAL LEAK BOP 1.1 Enter Procedure 4.6.3.

CRS Direct entry into Procedure 4.6.3.

Booth Role Play: As Rx Bldg NLO perform steps 5.1 through 5.4 of Procedure 4.6.3. Wait 2 minutes and Insert Trigger 2 to open the listed valves.

Operator Report the valves are open.

Direct NLO perform following Procedure 4.6.3 steps:

5.1 Remove seal and open PC-559, NBI-AO-737AV ISOLATION (R-903-SE).

5.2 Remove seal and open PC-560, NBI-AO-737AV SUPPLY (R-903-SE).

BOP 5.3 Remove seal and open PC-565, NBI-AO-736AV ISOLATION (R-881-NW Torus Area).

5.4 Remove seal and open PC-566, NBI-AO-736AV SUPPLY (R-881-NW Torus Area).

5.5 Place REACTOR FLANGE LEAKOFF switch (Panel 9-4) to OPEN and verify following:

5.5.1 INLET NBI-736AV (NORMALLY OPEN) closes.

5.5.2 DRAIN NBI-737AV (NORMALLY CLOSED) opens.

5.5.3 Annunciator 9-4-1/F-5 clears.

BOP 5.5.4 NBI-PI-101 (R-931-NW LR 25-5) indicates 0 pressure.

5.6 (Independent Verification) Place REACTOR FLANGE LEAKOFF switch (Panel 9-4) to CLOSE.

5.6.1 Verify following:

5.6.1.1 INLET NBI-736AV (NORMALLY OPEN) opens.

5.6.1.2 DRAIN NBI-737AV (NORMALLY CLOSED) closes.

Booth Action: When directed to close the PC valves, Delete Remote Function rr18 and Delete Malfunction rr21.

Operator Role Play: When directed to NBI-PI-101, as Rx Bldg NLO, wait 1 minute Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 16 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 4 Event

Description:

Respond to RPV Flange Leakage Time Position Applicants Action or Behavior and report pressure is 0 psig.

Direct NLO perform following Procedure 4.6.3 steps:

5.7 (Independent Verification) Close and seal PC-559.

BOP 5.8 (Independent Verification) Close and seal PC-560.

5.9 (Independent Verification) Close and seal PC-565.

5.10 (Independent Verification) Close and seal PC-566.

Booth Role Play: As Rx Bldg NLO, wait 2 minutes and report steps 5.7 through 5.10 are complete and you will come up to the control room and Operator complete step 5.11.

END OF EVENT Notes Booth Proceed to next event when directed by the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 17 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 5 Event

Description:

RPS EPA 1A1/1A2 trip, RMV-AO-10 fails to fully close.

Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 4 causing a trip of Operator RPS EPA 1A1/1A2 and failure of RMV-AO-10 to close.

Responds per alarm C-1/F-1, RPS PWR PANEL 1A VOLTAGE FAILURE:

2.1 IF a RPS power supply is available, THEN transfer RPS A to available source.

2.2 Reset RPS Channel "A" half scram per Procedure 2.1.5.

BOP 2.3 Reset Group Isolations per Procedure 2.1.22.

2.4 Dispatch Operator to RPS A Room to check EPAs and RPS MG Set A to determine cause of failure.

2.5 WHEN cause of failure has been determined and corrected, THEN place RPS A on desired source per Procedure 2.2.22.

Reports status of reactor and status of PCIS Group isolations:

Half Group 1 Half Group 2 ATC Full Group 3 Full Group 6 Half Group 7 Directs NLO to RPS MG Set A to investigate loss of power.

Booth Role Play: as building operator at RPS MG Set A, wait 3 minutes, then Operator report EPAs 1A1/1A2 are open.

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 transfers RPS A to alternate power by placing RPS BUS A POWER BOP TRANSFER switch to ALT FEED on panel 9-16.

Verifies Red ALT SOURCE ON light illuminated.

Reset half scram per Procedure 2.1.5: (Section 4) 4.1 Place REACTOR SCRAM RESET switch to Group 1 and 4, Group 2 ATC and 3, then back to NORM.

4.2 Ensure eight SCRAM GROUP lights (Panels 9-15 and 9-17) or SCRAM INDICATIONS GROUP A and GROUP B lights are on.

CRS Direct BOP to verify Group Isolations per Procedure 2.1.22.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 18 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 5 Event

Description:

RPS EPA 1A1/1A2 trip, RMV-AO-10 fails to fully close.

Time Position Applicants Action or Behavior NOTE to Examiners: Group Isolation verification Hard Card is attached for reference Reviews TS LCO 3.3.8.2. Enters Condition A. Required Action to remove CRS associated in service power supply from service within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

CREW May report small rise in drywell pressure and temperature.

May enter Abnormal Procedure 2.4PC due to drywell pressure and CRS temperature rising.

CRS Enter Abnormal Procedure 2.4HVAC Per 2.4HVAC, Attachment 1

1. IF RRMG Set Ventilation System has failed, THEN perform following:

NOTE - RRMG Set motor and generator winding temperatures are monitored on RRMG-TR-26, RX RECIRC DRIVE MTR GEN WINDING TEMPERATURE RECORDER (PNL-21).

BOP 1.1 Reduce reactor power per Procedure 2.1.10 to maintain affected RRMG Set motor and generator winding temperatures below 250°F.

1.1.1 IF winding temperature cannot be maintained below 250°F, THEN trip affected RRMG and concurrently enter Procedure 2.4RR.

1.1.2 IF both RRMGs are tripped, THEN SCRAM and concurrently enter Procedure 2.1.5.

Verifies isolations IAW 2.1.22 (May use the Hard Card), including:

Crack open RWCU-MO-74, DEMIN SUCTION BYPASS VLV, on panel 9-4 BOP Checks SGT for proper operation Determines RMV-AO-10 is indicating dual position.

Reports RMV-AO-10 status to CRS.

Direct NLO to investigate valve.

Booth Role Play: as building operator sent to investigate RMV-AO-10, wait 3 Operator minutes, then report you see nothing wrong with the valve.

Directs BOP to reset isolations and restore Reactor Recirc MG ventilation CRS IAW 2.1.22.

Checks RR MG temperatures on recorders RRMG-TR-25 and 26 on BOP Panel 9-21. Reports RRMG temperatures rising.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 19 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 5 Event

Description:

RPS EPA 1A1/1A2 trip, RMV-AO-10 fails to fully close.

Time Position Applicants Action or Behavior Resets Group 6 isolation and restores RB ventilation IAW 2.1.22:

9.8.1 Place following Reactor Building HVAC switches to OFF:

9.8.1.1 SF-R-1A-A, SUPPLY FAN.

9.8.1.2 SF-R-1A-B, SUPPLY FAN.

9.8.1.3 EF-R-1A, EXHAUST FAN.

9.8.1.4 EF-R-1B, EXHAUST FAN.

9.8.1.5 BF-R-1A, EXH BSTR FAN.

9.8.1.6 BF-R-1B, EXH BSTR FAN.

9.8.2 Place following Reactor Building HVAC isolation valve switches to CLOSE:

9.8.2.1 HV-MO-272, HV-R-1A DISCH VLV.

9.8.2.2 HV-AO-257, HV-R-1A DISCH VLV.

9.8.2.3 HV-AO-259, EXH FANS DISCH VLV.

9.8.2.4 HV-AO-261, EXH FANS DISCH VLV.

9.8.2.5 HV-MO-258, EXH FANS DISCH VLV.

9.8.2.6 HV-MO-260, EXH FANS DISCH VLV 9.8.3 IF isolation caused by high radiation in Reactor Building, THEN have Chemistry and Radiation Protection sample and evaluate situation.

9.8.4 IF Reactor Building Vent Monitors tripped and Reactor Building radiation levels normal, THEN perform following:

9.8.5 IF required, turn Group ISOL RESET CHANNEL A and CHANNEL B switches (Panel 9-5) to right RESET position, and THEN release to NOR.

9.8.6 At VBD-R, ensure SGT-DPIC-546, RX BLDG/SGT DP, in MANUAL.

9.8.7 Adjust SGT-DPIC-546, RX BLDG/SGT DP, Parameter V to 0%.

9.8.8 Ensure switch for operating SGT fan, EF-R-1F, SGT B EXHAUST FAN, in RUN (VBD-K).

9.8.9 Simultaneously press PCIS GROUP 6 DIV 1 and DIV 2 ISOLATION RESET buttons (VBD-K).

9.8.10 Perform one of following:

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 20 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 5 Event

Description:

RPS EPA 1A1/1A2 trip, RMV-AO-10 fails to fully close.

Time Position Applicants Action or Behavior 9.8.10.1 Check PCIS GROUP 6 DIV 1 ISOLATION and PCIS GROUP 6 DIV 2 ISOLATION lights turn on (VBD-K).

9.8.10.2 Check Group 6, CHANNEL A and Group 6, CHANNEL B Isolation indicating lights on (Panel 9-5).

9.9 Check following RRMG ventilation isolation valves open:

9.9.1 HV-AO-263, MG SET-1A INLET VLV.

9.9.2 HV-MO-262, MG SET-1A INLET VLV.

9.9.3 HV-AO-267, MG SET-1A OUTLET VLV.

9.9.4 HV-MO-266, MG SET-1A OUTLET VLV.

9.9.5 HV-AO-265, MG SET-1B INLET VLV.

9.9.6 HV-MO-264, MG SET-1B INLET VLV.

9.9.7 HV-AO-269, MG SET-1B OUTLET VLV.

9.9.8 HV-MO-268, MG SET-1B OUTLET VLV.

9.10 Check following RRMG exhaust fan starts:

9.10.1 EF-R-1C, EXH FAN (BOTTOM).

9.10.2 EF-R-1D, EXH FAN (TOP).

Checks RR MG temperatures lowering on RRMG-TR-25 and 26, panel 9-21.

NOTE to Examiners: It is not expected that the crew restore all systems within the time frame of the scenario for this event before proceeding to the next event, only that the half scram is reset, the RMV-AO-10 valve failure is recognized, and possibly, RRMG ventilation is restored. (Normal recovery time for all of Group 6 isolation is 15 minutes).

Resets Group 1 isolation IAW 2.1.22:

4.6.2 Reset Group 1 Isolation by turning Group ISOL RESET, CHANNEL A and CHANNEL B, switches (Panel 9-5) to left RESET position and then releasing to NOR.

BOP 4.6.3 Check Group 1, CHANNEL A Isolation lights turn on (Panel 9-5).

4.6.4 Check Group 1, CHANNEL B Isolation lights turn on (Panel 9-5).

4.6.5 Ensure MS-MO-74 open.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 21 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 5 Event

Description:

RPS EPA 1A1/1A2 trip, RMV-AO-10 fails to fully close.

Time Position Applicants Action or Behavior NOTE - If all main steam lines in service and main steam line flow through each steam line is > 1.2x106 lbm/hr, Step 4.6.6 may be N/A'd.

4.6.6 Ensure MS-MO-77 open.

Refers to TS LCO 3.6.1.3, Condition A and declares RMV-AO-10 INOPERABLE.

Required Action is to isolate the penetration with a Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Enters TS LCO 3.4.5 and enter Condition B due to isolation of the drywell CRS atmosphere Inboard Sample Return Valve RMV-AO-12 on panel 9-4.

Required Actions are to analyze grab samples of DW atmosphere with a Completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND restore required DW atmospheric monitoring system to Operable status with a Completion Time of 30 days.

May enter TS LCO 3.6.4.1, Condition A due to Reactor Building dP momentarily exceeding -0.25 wg. Required Action is to restore secondary containment to Operable status with a Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

END OF EVENT Notes Proceed to the next event at direction of the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 22 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 6 Event

Description:

REC Pump A and B trip Time Position Applicants Action or Behavior Booth When directed by Lead Examiner insert Trigger 5, REC Pump A trips Operation followed by REC Pump B trip after 2.5 minutes.

Respond to alarm M-1/B-1, REC PUMP A FAILURE.

1. OPERATOR OBSERVATION AND ACTION BOP 1.1 Start another REC pump.

1.2 Monitor REC pump discharge pressures and ensure valve line-up is correct.

1.3 For multiple loss of REC pumps, enter Procedure 5.2REC.

Note to Examiners: If operator starts another REC pump before the system low pressure isolation times out (~40 seconds), the following BOP actions are N/A. He may perform the actions to start another pump from memory as immediate operator actions of Abnormal Procedure 5.2REC or he may perform the steps from the Annunciator card.

If alarm M-1/A-1, REC SYSTEM LOW PRESSURE, remains in for > 40 seconds (REC system will isolate), operator will respond per alarm card:

2. OPERATOR OBSERVATION AND ACTION 2.1 If available, start additional REC pumps.

2.2 Ensure REC-MO-711, NORTH CRITICAL LOOP SUPPLY, BOP or REC-MO-714, SOUTH CRITICAL LOOP SUPPLY (associated with an in service HX), is open to obtain critical subsystem pressure indication.

2.3 If REC System header pressure on REC-PI-452, REC HEADER PRESSURE, remains 62 psig, enter Procedure 5.2REC.

2.4 If REC HX or Drywell header isolated and restoration desired, take action per REC Restoration Hard Card (2.2.65.1).

If REC system isolated restore per REC Restoration Hard Card:

BOP 1.1 Ensure low pressure isolation not due to leakage or leak isolated.

1.2 Ensure two REC pumps are running.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 23 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 6 Event

Description:

REC Pump A and B trip Time Position Applicants Action or Behavior 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.

CAUTION - Restoring REC flow to drywell FCUs with drywell temperature > 260F could result in a breach of FCU tubing.

1.4 If drywell temperature 260F on PC-TI-505A through PC-TI-505E, place DRYWELL REC ISOL VALVE CONTROL switch to OPEN.

NOTE - REC-MO-712 and REC-MO-713 are throttle open only.

If REC HX OUTLET PRESSURE alarm is received, REC-MO-712 or REC MO 713 must be fully closed prior to recommencing pressurization of REC non-critical header.

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. NOTE to Examiners: Step is N/A.

1.7 Throttle open REC HX outlet valve, as necessary, to obtain following conditions:

1.7.1 REC CRIT LOOP SUPPLY PRESS 62 psig.

1.7.2 REC HEADER PRESSURE in top 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 If REC-AO-710, RWCU NON-REGEN HX INLET, not closed for leak isolation, open REC-AO-710.

1.11 If REC-MO-1329, 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 Respond to alarm M-1/B-2, REC PUMP B FAILURE.

BOP 1.3 For multiple loss of REC pumps, enter Procedure 5.2REC Report trip of REC Pump B.

CRS Enter Procedure 5.2REC. Assign actions to BOP.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 24 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 6 Event

Description:

REC Pump A and B trip Time Position Applicants Action or Behavior Enter Procedure 5.2REC.

3. IMMEDIATE OPERATOR ACTIONS 3.1 IF REC HEADER PRESSURE 62 psig, THEN start available REC pumps.

3.2 IF REC HEADER PRESSURE not restored, THEN close following valves:

3.2.1 REC-AO-710, RWCU NON-REGEN HX INLET.

3.2.2 REC-MO-1329, AUGMENTED RADWASTE SUPPLY.

4.3 IF REC HEADER PRESSURE not restored after completing Immediate Operator Actions, THEN perform following:

BOP 4.3.1 SCRAM and enter Procedure 2.1.5.

4.3.2 Stop both Reactor Recirc pumps and enter Procedure 2.4RR.

4.3.3 Stop running CRD pump.

NOTE - Securing all AC lube oil pumps first will cause DC lube oil pumps to start unless DC oil pump control switches are first taken to STOP and allowed to spring return to their normal positions.

4.3.4 WHEN Recirc MG Sets have stopped, THEN perform following:

4.3.4.1 Momentarily place respective control switches for both DC lube oil pumps to STOP and allow them to spring return to their normal positions (R-958-NW at 125 VDC Reactor Bldg Starter Rack).

4.3.4.2 Shut down all AC lube oil pumps.

END OF EVENT Notes Proceed to the next event for the scram and ATWS.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 25 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior NOTE to Examiners: Malfunctions RD02a and RD02b, set at 45% (ATWS) are already active.

CRS Direct the reactor scrammed per 5.2REC.

Depress both Manual Scram pushbuttons on Panel 9-5.

ATC 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. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 26 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior Place Reactor Mode Switch in SHUTDOWN.

Initiate ARI.

Report reactor power level to CRS.

ATC Run RR pumps back to minimum when directed by CRS. NOTE to Examiners, the RR Pumps might already be tripped from 5.2REC actions..

Trip RR pumps when directed by CRS.

Insert control rods per EOP5.8.3 The ATC inserts approximately 10-15 control rods manually prior to bypassing the RPS trips.

ATC NOTE to Examiners: The crew may elect to trip the running CRD pump per 5.2REC direction so no rod insertion takes place with RMCS.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 27 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior Place both CRD Pumps in service or trip running CRD pumps due to loss of ATC REC cooling (5.2REC direction).

Ensure CRD-FC-301 is in Manual to maintain drive water d/p approximately ATC 265 psid.

Selects the rods starting in the center and works out in a spiral pattern using ATC the 5.8.3 Board depicted below.

ATC EOP 7A RPV Level CRS Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 28 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior (CT-1): When control rods fail to scram, crew injects SLC and/or inserts control rods before exiting EOP-6A. (All control rods do not have to be fully inserted to satisfy this critical task; this only requires that the crew is making progress to achieving all rods in by fully inserting at least 5 control rods using RMCS.)

Direct inhibiting ADS and installing PTMs for any open MSIV.

CRS EOP 7A RPV Level (CT-2): Inhibit ADS prior to automatic ADS valve opening during a failure to Scram.

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.

4.3.1 Install EOP PTM Number 57 jumper between Terminals DD-1 and DD-2 (BAY-1, PNL 9-15).

4.3.2 Install EOP PTM Number 58 jumper between Terminals BB-1 BOP and BB-2 (BAY-3, PNL 9-15).

4.3.3 Install EOP PTM Number 59 jumper between Terminals DD-1 and DD-2 (BAY-1, PNL 9-17).

4.3.4 Install EOP PTM Number 60 jumper between Terminals BB-1 and BB-2 (BAY-3, PNL 9-17).

CRS Direct RPV level lowered below -60 inches by using stop and prevent.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 29 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior EOP 7A RPV Level (CT-3): During failure to scram conditions with power >3%, stop and prevent injection from all sources (except boron, CRD, RCIC) as necessary to lower RPV water level to below -60 CFZ (or LL, as applicable) and control between -60 (or LL, as applicable) to -

183 CFZ prior to exiting EOP-7A.

Perform Stop and Prevent per EOP 5.8 HARD CARD:

STOP INJECTION 1.1 Stop HPCI by performing one of following:

1.1.1 IF HPCI is not running, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

1.1.2 IF HPCI is running, THEN perform one of following:

1.1.2.1 Place HPCI controller to MANUAL and lower on controller to maintain > 100 psig below low-end of RPV pressure band, and

a. Maintain turbine speed > 2050 rpm.

1.1.2.2 Trip HPCI turbine by performing following:

a. Press and hold TURBINE TRIP button.

b. WHEN turbine is at zero rpm, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

c. Release TURBINE TRIP button.

BOP 1.2 Stop feedwater by performing following:

1.2.1 At a RVLC/RFPT HMI, select STARTUP VALVE screen, press Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 30 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior EMER CLOSE button, and confirm "YES" in pop-up box.

1.2.2 IF two RFPs are running, THEN ensure non-preferred RFP is tripped.

1.2.3 IF both RF-MO-29 and RF-MO-30 are not fully closed, THEN ensure operating pump is run back to ~ 2800 rpm by performing following:

1.2.3.1 On MAIN CONTROL screen, place MASTER LEVEL in MANUAL.

1.2.3.2 Depress FAST and use arrows to lower DEMAND to ~

0%.

1.2.3.3 Ensure RF-MO-29 and RF-MO-30 are closed.

1.2.4 IF RF-MO-29 or RF-MO-30 are open and cannot be closed from Control Room, THEN perform following:

1.2.4.1 Trip both RFP's.

1.2.4.2 Trip all operating condensate booster pumps.

BOP Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 31 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior CAUTION - If Core Spray and RHR pumps are placed in PULL-TO-LOCK before system flow is reduced to minimum, draining of system may occur.

1.3 Place both core spray pumps in PULL-TO-LOCK.

1.4 Stop RHR by ensuring one of following:

1.4.1Both RHR Systems secured with pumps in PULL-TO-LOCK.

1.4.2 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

1.4.3 IF RPV pressure is maintained 500 psig, THEN operate RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3.

PREVENT INJECTION 2.1 Prevent RHR by performing one of following:

2.1.1Both RHR Systems secured with pumps in PULL-TO-LOCK.

2.1.2 IF RPV pressure is maintained 500 psig, THEN operate RHR aligned to suppression pool cooling and/or containment spray per Procedure 2.2.69.3; and 2.1.2.1 Bypass RHR outboard injection valves automatic open signal per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

2.1.3 RHR outboard injection valves automatic open signal bypassed per Procedure 5.8.20 (PTMs 97 through 100) with injection valves closed.

2.2 Prevent feedwater by performing following:

2.2.1 If required, at a RVLC/RFPT HMI, select STARTUP VALVE screen, press EMER CLOSE button, and confirm "YES" in pop-up box.

2.2.2 Ensure RF-MO-29 is closed.

2.2.3 Ensure RF-MO-30 is closed.

2.2.4 Trip condensate and condensate booster pump(s), as required.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 32 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior 2.3 Prevent CS by performing following:

2.3.1 Ensure CS-MO-12A is closed.

2.3.2 Ensure CS Pump A control switch in PULL-TO-LOCK.

2.3.3 Ensure CS-MO-12B is closed.

2.3.4 Ensure CS Pump B control switch in PULL-TO-LOCK.

2.4 Prevent HPCI by performing following:

2.4.1 IF HPCI is not running, THEN ensure AUXILIARY OIL PUMP switch is in PULL-TO-LOCK.

2.4.2 IF HPCI is running, THEN trip HPCI turbine by performing following:

2.4.2.1 Press and hold TURBINE TRIP button.

2.4.2.2 WHEN turbine is at zero rpm, THEN place AUXILIARY OIL PUMP switch to PULL-TO-LOCK.

2.4.2.3 Release TURBINE TRIP button.

Direct NLO to install EOP PTMs 97 through 100 for RHR injection valve control.

BOP Role Play: When directed by BOP to install EOP PTMs97-100, wait 3 Booth minutes then put in the overrides for the PTMs. Report back to BOP Operator when PTMs installed.

(CT-4): When control rods fail to scram and energy is discharging to the primary containment (e.g. SRVs, LOCA), crew injects SLC before exceeding the Boron Injection Initiation Temperature (BIIT) curve.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 33 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior ATC When directed by the CRS inject SLC.

1.1 Place both keys in SLC PUMP A and SLC PUMP B keylock switches on Panel 9-5 and place switches to START.

1.2 Check both SLC pumps start.

1.3 Check white SQUIB VALVE READY DS-3A (1106A) and SQUIB VALVE READY DS-3B (1106B) lights turn off (Panel 9-5).

1.4 Check pressure on SLC-PI-65, PUMP PRESSURE (Panel 9-5), is greater than reactor pressure.

1.5 Check Annunciator 9-5-2/G-7, LOSS OF CONT TO SQUIB VLVS, alarms.

1.6 Ensure RWCU-MO-15, INBD ISOL VLV (Panel 9-4), is closed.

1.7 Ensure RWCU-MO-18, OUTBD ISOL VLV (Panel 9-4), is closed.

1.8 Ensure both RWCU pumps are off (Panel 9-4).

1.9 Ensure RWCU-MO-74, DEMIN SUCTION BYPASS VLV (Panel 9-4), is throttled open.

ATC Provide CRS initial SLC tank level.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 34 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior 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.

EOP 7A RPV Level BOP Direct BOP stabilize RPV pressure below 1050 psig:

CRS EOP 6A RPV Pressure Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 35 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 36 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior ATC Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 37 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior ATC ATC Resets the scram and repeats scrams as rod motion is verified.

END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 38 of 44 Op-Test No.: 1 Scenario No.: 1 Event No.: 7 Event

Description:

40% Power ATWS Time Position Applicants Action or Behavior Proceed to the next event Suppression Pool Cooling valve failure.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 39 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 7 Event

Description:

RHR SPC valve failure Time Position Applicants Action or Behavior Monitor and report torus level and temperature as SRVs are utilized for pressure control. Announce EOP 3A entry conditions as necessary.

BOP Direct suppression pool cooling placed into service when torus temperature CRS rises above 95°F.

NOTE to Examiners: RHR-MO-39A(B) fails on the first loop placed into Suppression Pool Cooling. The other loop is available and the trainee re-performs procedure steps below.

Use RHR and place into suppression pool cooling (after EOP PTMs97-100 are installed) per the HARD CARD.

1.10 Place RHR SW System in service:

1.10.1 Start SWBP(s).

1.10.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

1.11 If required, with CRS permission, place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD.

1.12 If required, place CONTMT COOLING VLV CONTROL BOP PERMISSIVE switch to MANUAL.

1.13 Open RHR-MO-39A(B).

1.14 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.

1.15 Ensure RHR PUMP running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

1.16 Throttle RHR-MO-34A(B), as required to obtain desired cooling flow.

1.17 Throttle RHR-MO-66A(B), as required to obtain desired cooling rate.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 1 Page 40 of 44 Op-Test No.: CNS 15-01 Scenario No.: 1 Event No.: 7 Event

Description:

RHR SPC valve failure Time Position Applicants Action or Behavior 1.18 If PCIS Group 6 lights lit on Panel 9-5, ensure one of following open:

1.18.1 REC-MO-711; or 1.18.2 REC-MO-714.

If additional cooling required, initiate cooling in non-running RHR Loop and start additional pumps.

BOP Report failure of RHR-MO39A(B) and use other loop of RHR for SPC.

After the scram has been reset for the second time, or as directed by the Booth Lead Examiner, delete both RD02A and RD02B malfunctions while the Operator scram is reset to allow the control rods to fully insert on the next scram.

ATC Report all control rods fully inserted.

CRS Order SLC injection stopped.

ATC Place both SLC Pump control switches to OFF.

Exits EOP 6A and 7A and enters EOP 1A.

CRS RC/L-3 Direct restoring and maintaining RPV water level between +3 inches and +54 inches.

BOP Raise RPV injection and raise level between +3 inches and +54 inches.

END OF SCENARIO Notes Booth When directed by the Lead Examiner, place the simulator in freeze and Operator tell the crew to stop operating.

Rev. 0

0.ATTACHMENT 1 GROUP ISOLATION HARD CARD FRONT PANEL GROUP ISOLATION HARD CARD Re viis ion 0 R HR-9 21 R HR-9 2 0 4 HPCI-1 6 HPCI-15 MS-80A MS-86A 1 MS-80B MS-86B 5 R C IC-1 6 RCIC-15 MS-80C MS-86C R H R-9 5 /96 MS-80D MS-86D R R -7 4 0 7 MS-74 MS-77 R H R-6 0 /61 R R -74 1 RWCU-15 RWCU-18 3 RHR -5 7 R H R-6 7 R H R -1 7 RHR-18

2 TIP VALVES RW-83 RW-82 RMV-10 RMV -1 1 RW-95 RW-94 RMV-12 R MV-13 RHR-25 A

1 R H R-274A RHR-25B

2 *1 RHR-274B

1 - U n le s s LP CI in itia t io n s ig n a l e xis ts

2

2 - Va lve n o rm a lly de e n e rg ize d c lo s e d Figure 1 ADDITIONAL ACTIONS GROUP 3

[ ] Crack open RWCU-MO-74, DEMIN SUCTION BYPASS VLV, allowing a path for mini purge flow to reactor to prevent over-pressurization.

GROUP 4

[ ] Ensure HPCI-MO-58 is CLOSED if HPCI-MO-17 is full open.

[ ] Ensure HPCI-AO-70 and HPCI-AO-71 are CLOSED.

[ ] Ensure HPCI turbine has tripped.

GROUP 5

[ ] Ensure RCIC turbine has tripped.

Rev. 0

BACK PANEL GROUP ISOLATION/SGT HARD CARD Figure 2 RESPONSE TO AUTOMATIC SGT INITIATION

[ ] IF pressure on HV-DPR-835, RX BLDG/ATMOS DP (VBD-R), is being maintained at

-0.25" wg, THEN perform following:

[ ] Place switch for preferred SGT fan, EF-R-1E, SGT A EXHAUST FAN, or EF-R-1F, SGT B EXHAUST FAN, to RUN.

[ ] Place switch for SGT fan to be selected for standby, EF-R-1E or EF-R-1F, to OFF, then to STANDBY, and check following:

[ ] EF-R-1E or EF-R-1F stops.

[ ] SGT-AO-249, SGT A INLET, or SGT-AO-250, SGT B INLET, closes.

[ ] SGT-AO-251, SGT A DISCHARGE, or SGT-AO-252, SGT B DISCHARGE, closes.

[ ] Close SGT-AO-270, SGT A DILUTION AIR.

[ ] Close SGT-AO-271, SGT B DILUTION AIR.

Rev. 0

INITIAL CONDITIONS A. Plant Status:

1. 100% power, near End of Cycle.

2. Rod Sequence Information: RO to provide B. Tech. Spec. Limitations in effect:

1. LCO 3.5.1 C.1 restore HPCI to Operable status in 14 days

2. Currently on day 2 of LCO 3.5.1.

3. RCIC is verified OPERABLE.

C. Significant problems/abnormalities:

1. HPCI Auxiliary Oil Pump Motor replacement.

D. PRA Risk:

Green E. Evolutions/maintenance for the on-coming shift:

1. After taking the shift, the ATC is to shift CRD Stabilizing valves per Procedure 2.2.8, Section 17.

2. Lower power to 95% per Load Dispatchers request, with RR Pumps per Procedure 2.1.10, Section 7.

3. Electrical Maintenance working on replacing the HPCI AOP motor.

Rev. 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 1 of 31 Facility: Cooper Nuclear Station Scenario No.: 2 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Shift REC Pumps.

2. CRD FCV auto function fails requiring manual control.

3. One outboard MSIV fails closed.

4. Partial loss of main condenser vacuum requiring manual scram.

5. Electric ATWS

6. FW line break inside PC with loss of RCIC.

7. RWCU fails to auto isolate

8. Emergency Depressurize on low RPV level.

9. Low pressure injections valves fail to automatically open.

Initial Conditions: Plant operating at 100% power.

Inoperable Equipment: HPCI inoperable. Auxiliary Oil pump motor replacement.

Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Maintain present power level.

Electrical Maintenance working on installing HPCI AOP motor.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 2 of 31 Event Malf. No. Event Type Event No. Description N (BOP, CRS) 1 N/A Shift REC pumps TS (CRS)

OR: I (ATC,CRS) Auto function on CRD FCV fails causing manual control 2

zaicrdfc301 A (CREW) to be used C

ms09e (ATC,BOP,CRS) 3 OR: Outboard MSIV 86A closes but leaks by A (CREW) zdipcissws4a TS (CRS)

C (ATC,CRS) 4 mc01 Partial loss of condenser vacuum-scram A (CREW) rp01 (a-d)

OR: Electrical ATWS 5 zdirpssws1 M(CREW) (CT-1 When RPS fails to scram the reactor on a manual zdirpssws3a scram signal, within two minutes initiate the ARI System.

zdirpssws3b FW A line break inside primary containment.

RCIC spurious isolation fw18a (CT-2) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and 6 rr20a M (CREW) insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize rc02 by opening the first of six SRVs before RPV level lowers below -183 CFZ.) (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

7 rp12 C (BOP,CRS) RWCU fails to automatically isolate.

ECCS system valves fail to auto open.

cs02a C (CT-3 )When operating injection systems cannot maintain cs02b 8 (ATC,BOP,CRS) RPV level and ECCS systems fail to automatically align for rh04a injection, crew manually aligns ECCS systems for injection:

rh04b

For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 3 of 31 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after Low Pressure ECCS injection valves fail to open.

1-2 2 EOP entry RWCU fails to auto isolate Outboard MSIV closure Abnormal Events 2-4 2 Partial loss of condenser vacuum ATWS Major Transients 1-2 2 FW line break EOP entries requiring EOP-1A 1-2 2 substantive action EOP-3A EOP contingencies requiring substantive 0-2 1 EOP-2A action (CT-1) When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

(CT-2) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection: For low pressure ECCS systems, prior to RPV pressure EOP based Critical 2-3 lowering below 200 psig.

3 Tasks (CT-3) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.) (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

Normal Events N/A 1 Shift REC Pumps Reactivity Manipulations N/A 1 None CRD FCV controller failure Outboard MSIV closure Instrument/

Component Failures N/A 5 Condenser in-leakage loss of vacuum RWCU fails to isolate.

LP ECCS injection valves fail to open CRD FCV controller failure Outboard MSIV closure Total Malfunctions N/A 5 Condenser in-leakage rise RWCU fails to isolate.

LP ECCS injection valves fail to open Top 10 systems and operator actions important to risk that are tested:

Reactor Protection System (Event 5)

ADS/SRV (Event 6)

Residual Heat Removal System in LPCI Mode (Event 8)

Operator fails to depressurize with SRVs (Event 6)

Operator fails to initiate ADS and initiate ECCS early (Event 6)

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 4 of 31 SCENARIO

SUMMARY

The plant is operating at 100% power at the end of the operating cycle. HPCI Auxiliary Oil Pump motor is removed and a replacement is being installed.

Event 1 After the crew takes the watch, the BOP operator shifts REC pumps by starting B and securing A. The CRS is required to declare REC Div I subsystem inoperable per LCO 3.7.3, Condition B, Event 2 (Triggered by Lead Examiner)

After TS are addressed for the REC pump shift, the CRD FCV automatic setpoint fails downscale requiring the ATC to take manual control and return CRD cooling water flow and pressure to normal.

Event 3 (Triggered by Lead Examiner)

After the CRD system flows are returned to normal in manual, outboard MSIV 86A partially closes. The crew enters Abnormal Procedure 2.4MSIV and the RO rapidly lowers reactor power to <70%. The BOP places the effected MSIV control switch to CLOSE to prevent reopening. The CRS enters LCO 3.6.1.3, Condition A and declares the PCIV inoperable.

Event 4 (Triggered by Lead Examiner)

After TS are addressed for the partially closed MSIV, condenser in-leakage rises requiring reactor power to be lowered to maintain vacuum > 23 inches mercury.

Condenser vacuum continues to lower requiring the reactor scram.

Event 5 (No Trigger required)

On the manual reactor scram, the crew recognizes the ATWS is an electric block ATWS. Manual ARI initiation successfully inserts the control rods (CT-1).

Event 6 (Automatically triggered when RFP Discharge Valve automatically closes,

~2 minutes after ARI is initiated)

After the control rods are inserted, Feedwater A line break inside the PC commences and the CRS enters EOP 3A. The torus and drywell are sprayed to control containment pressure and temperature. RPV water level continues to drop. RCIC will be unavailable due to a spurious isolation signal.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 5 of 31 Event 7 (No Trigger required)

RWCU fails to isolate on low RPV level. Manual isolation from the control room is required.

Event 8 (No Trigger required)

RPV level lowers to TAF requiring the crew to emergency depressurize (CT-2). As RPV level and pressure lower, RHR injection valves fail to open and cannot be opened. The CS injection valves fail to open and can be opened from the control room (CT-3).

The exercise ends when emergency depressurization is complete and RPV level restoration is being controlled.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 6 of 31 Critical Tasks (CT-1) When RPS fails to scram the (CT-2) When RPV level lowers to -158 CFZ reactor on a manual scram signal, within (TAF) and cannot be maintained above -

two minutes initiate the ARI System. 183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ. (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

EVENT 5 6 Safety RPS initiates a reactor scram when one or The MSCWL is the lowest RPV water level at significance more monitored parameters exceed their which the covered portion of the reactor core specified limits to preserve the integrity of the will generate sufficient steam to preclude any fuel cladding and the reactor coolant clad temperature in the uncovered portion of pressure boundary (RCPB) and minimize the the core from exceeding 1500F. When water energy that must be absorbed following a level decreases below MSCWL with injection, loss of coolant accident (LOCA). Failure to clad temperatures may exceed 1500F.

effect shutdown of the reactor when a RPS setting has been exceeded, even at low power, would unnecessarily extend the level of degradation of the safety of the plant. This could further degrade into damage to the principle fission product barriers if left unmitigated. The crew is authorized and required by Conduct of Operations to take mitigating actions when automatic safety systems fail to perform their intended function. Action to shut down the reactor is required when RPS and control rod drive systems fail.

Cueing Annunciators 9-5-2/A-1 (A-2) RX SCRAM Corrected Fuel Zone indication (SPDS) falls to CHANNEL A (B) in alarm with RPS -158 and lowering trend continues, and, remaining energized. before -158 CFZ is reached, initial conditions, field reports, and control room indications convey that adequate high pressure injection cannot be restored before level falls below -

183 CFZ Performance Operator depresses both manual scram Manipulation of any six SRV controls on panel indicator pushbuttons, or places the Reactor Mode 9-3:

Switch to SHUTDOWN on panel 9-5. SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance RPS Group lights de-energized on panel 9-5. Crew will observe SRV light indication go from feedback Control Rod full -in indication on panel 9-5. green to red, amber pressure switch lights Reactor power trend on nuclear illuminate, reactor pressure lowering on SPDS instrumentation on panel 9-5. and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Procedure 2.0.3, Conduct of Operations There is no time limit for effecting complete for the chosen requires upon recognition of a failure of The MSCWL (-183 CFZ) is the lowest RPV performance automatic action, the CRO shall manually water level at which the covered portion of the limit perform those actions necessary to fulfill the reactor core will generate sufficient steam to Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 7 of 31 safety function and report the completion of preclude any clad temperature in the the manual action to the CRS as soon as uncovered portion of the core from exceeding possible. Failure of RPS to automatically 1500F. Emergency depressurization is function would involve multiple sensor and allowed when level goes below TAF (-158 sensor relay failures. The complexity of an CFZ) and should be performed, if in the automatic RPS failure would necessarily judgment of the CRS, level cannot be require a short amount of time to diagnose maintained above -183 CFZ. Since it is and validate using control room indications. intended for the scenario supporting this CT to, Two minutes is a reasonable time for early in the event, clearly indicate no high operators to recognize a scram signal, verify pressure injection systems can be made the condition is valid, communicate available to reverse the lowering level trend, conditions to the crew, and insert a manual the crew will have time to communicate and scram, without unnecessarily extending the open 6 SRVs before -183 CFZ.

level of degradation to plant safety.

BWR Owners App. B, step RC-1 App. B, Contingency#1 Group Appendix Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 2 Page 8 of 31 Critical Tasks (CT-3) When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection:

For low pressure ECCS systems, prior to RPV pressure lowering below 200 psig.

EVENT 8 Safety Failure to recognize the auto valve alignment significance not occurring, and failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.

Cueing Indication ECCS valves are not opening with initiation conditions present:

Green light on and Red lamp extinguished at respective injection handswitch on panel 9-3 or 9-4.

Indication of Drywell Pressure 1.83 psig Indication of RPV water level -113 RPV pressure below injection valve open permissive setpoint Performance Manipulation of controls as required to open indicator the affected ECCS injection valve(s) or pump turbine controls from panel 9-3 or 9-4:

Operator places affected ECCS injection valve(s) control switch(es) to OPEN on panel 9-3 or 9-4.

Performance Red light illuminates and Green light feedback extinguishes for the affected ECCS injection valve(s), as applicable, on panel 9-3 or 9-4.

RCIC or HPCI turbine speed and flow rate rises, as applicable, on panel 9-3 or 9-4.

Justification Attempting to align high pressure ECCS for the chosen systems must be performed to determine performance their availability by the time TAF is reached in limit order to properly implement EOP-1A decision steps regarding restoring and maintaining RPV level. Attempting to align low pressure ECCS systems can only be done one RPV pressure falls below the injection valve RPV pressure permissive and will only be effective once RPV pressure falls below the shutoff head of the respective ECCS pump. The reduction in RPV pressure will normally be via Emergency Depressurization, which is a separate critical task bounded by a minimum RPV level.

BWR Owners App. B, Contingency 1, step C1-1 Group Appendix Rev. 1

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 9 of 31 SIMULATOR SET-UP A. Materials Required None B. Initialize the simulator in IC 153, 100% power (EOL)

Batch File Name - None C. Change the simulator conditions as follows:

1. Auto Triggers Number File Name/Variable Description 5 zlorfsw29mv[2]==0 RFP A discharge valve red light off (valve closed) 3 zlopcissws1a[1]==1 Close MSIV 80A and green light illuminating deletes MSIV dmf fw13a leakage through 86A malfunction..

2. Malfunctions Number Title Trigger TD Severity Ramp Initial rp01a RPS Group 1 A N/A N/A N/A N/A rp01b RPS Group 2 A N/A N/A N/A N/A rp01c RPS Group 3 A N/A N/A N/A N/A rp01d RPS Group 4 A N/A N/A N/A N/A ms09e MSIV 86A slower closing time 2 N/A 0 N/A N/A fw13a Represents MSIV 86A not fully closed 2 10 97 N/A N/A mc01 Main Condenser air in-leakage 4 N/A 50 60 N/A fw18a FW line A break inside PC 5 N/A 20 N/A N/A RR loop break to replicate break rr20a downstream of the feedwater check 5 N/A 20 N/A N/A valve.

RHR-MO-25A mechanically bound rh04a A N/A N/A N/A N/A closed RHR-MO-25B mechanically bound rh04b A N/A N/A N/A N/A closed cs02a CS-MO-12A fails to auto open A N/A N/A N/A N/A cs02b CS-MO-12B fails to auto open A N/A N/A N/A N/A Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 10 of 31 rp12 RWCU fails to isolate A N/A N/A N/A N/A rc02 RCIC turbine trip 5 3:00 N/A N/A N/A

3. Remotes Number Title Trigger Value TD Ramp rd12 CRD Filter B in service 6 open N/A N/A rd13 Close CRD-13 & 14 (min flow) 7 0 N/A N/A

4. Overrides Instrument Tag Trigger TD Value Ramp zdirpssws1 Rx Mode Switch A N/A RUN N/A zdirpssws3a Man Scram A3 A N/A PUSH-OUT N/A zdirpssws3b Man Scram B3 A N/A PUSH-OUT N/A zaicrdfc301[2] CRD FC set point 1 N/A 0 N/A zdipcissws4a MSIV A Test PB 2 N/A ON N/A HPCI AOP C/S zdihpcisws20[1] N/A 0 PTL N/A HPCI AOP Green light zlophcisws20[1] N/A N/A OFF N/A HPCI Inop on SSSP zdieeswhp[1] N/A N/A PUSH-IN N/A MSIV 86A red light zlomslt86a[1] N/A ON N/A MSIV 86A red light Zlopcissws2a[2] N/A ON N/A D. Panel Setup

1. Ensure PMIS IDTs are blank

2. Ensure RR Controllers are selected to P.

3. Ensure REC Pump A is running and Selector Switch is in STANDBY.

4. Ensure copy of Procedure 2.2.65.1 ready for review during turnover.

5. Place HPCI Aux Oil Pump c/s to P-T-L.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 11 of 31

6. Place Caution Tag on HPCI AOP c/s.

7. Ensure HPCI INOP displayed on Safety System Status Panel.

8. At Panel 9-4-3, set SSST Y Voltage Adjust to Tap 2.

9. At Panel C, set SSST X Voltage Adjust to Tap 5.

10. On STARTUP TRANSFORMER BACKUP VOLTAGE BUS, placard:

TAP POSITION: 5 MAX 4473 MIN 4372 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 12 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 1 Event

Description:

Shift REC Pumps A and B Time Position Applicants Action or Behavior CRS Directs BOP to Shift Rec Pumps A and B per Procedure 2.2.65.1, Section 8.

NOTE 1 - When starting REC pump, it is normal for associated pump low pressure alarm to come in and immediately clear.

NOTE 2 - For better temperature control, it is preferred to run two REC pumps aligned to in-service HX or HX with its TCV in Auto.

8.1 Start desired idle pump.

8.2 Secure selected running pump.

8.3 Check system pressure > 65 psig on REC-PI-452, REC HEADER PRESSURE.

8.4 Check drywell flow remains in green band on REC-FI-453, SUPPLY HEADER FLOW.

8.5 IF REC pump NORMAL/STANDBY selector switches must be BOP manipulated in step 8.6, THEN perform one of following:

8.5.1 IF both NORMAL/STANDBY selector switches in one division will be in NORMAL during swap, THEN inform SM that associated REC Subsystem is inoperable.

8.5.2 IF both NORMAL/STANDBY selector switches in one division will be in STANDBY during swap, THEN inform SM that associated Diesel Generator is inoperable. (Not expected to be performed.)

8.6 (Independent Verification) For each REC division, ensure selector switch for one running REC pump in STANDBY and other selector switch in NORMAL.

8.7 Monitor REC temperature and adjust REC HX outlet temperature per Section 5 or 6, as required, to maintain REC temperature stable.

NOTE to Examiner: Depending upon how the REC Pump Selectors switches are manipulated determines whether the Diesel Generator or the REC subsystem is inoperable for a short duration. (See Procedure steps 8.5.1 and 8.5.2 above)

CRS Enter LCO for REC Subsystem Inoperable.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 13 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 1 Event

Description:

Shift REC Pumps A and B Time Position Applicants Action or Behavior LCO 3.7.3 Condition B, Restore REC Subsystem to an operable status in 30 days.

END OF EVENT Notes Booth Proceed to next event when directed by the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 14 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 2 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 Trigger 1, to fail CRD-FC-301 output to zero flow.

Operator Report CRD system flows and pressures abnormal. Respond to alarm ATC 9-5-2/E-6, CRD CHARGING HEADER HIGH PRESSURE.

NOTE to Examiner: Crew may respond per the alarm card and not enter Abnormal Procedure 2.4CRD.

Enter Abnormal Procedure 2.4CRD, direct ATC to 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. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 15 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 2 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. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 16 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 3 Event

Description:

Outboard MSIV 86A fails partially closed Time Position Applicants Action or Behavior Booth When directed by the Lead Examiner, insert Trigger 2 to cause Main Operator Steam Line A outboard valve, MSIV 86A, to fail partially closed.

NOTE to Examiners: The APRM Upscale/Rod Block alarms may annunciate.

A small amount of steam Flow is indicated on RFC-FI-88B on Panel 9-5.

Recognizes and reports power rise on all channels of APRM recorders.

If applicable, refers to alarm cards for the following possible alarms:

ATC 9-5-1/A-4 Rod Withdrawal Block 9-5-1/B-8 APRM Upscale 9-5-2/F-4 RVLC SYSTEM LOGIC INITIATED 9-5-2/G-4 RVLC SYSTEM TROUBLE Checks panel indications and recognize pressure transient on recorders.

CREW Diagnose and recognize MSIV 86A indicates intermediate and some flow is still present on RFC-FI-88B on Panel 9-5.

Enters 2.4MSIV:

CRS Directs BOP to perform remaining action of 2.4MSIV.

May enter 5.1BREAK as directed by alarm 9-5-2/F-4.

Performs rapid power reduction IAW 2.1.10 hard card:

NOTE - Power reduction may be stopped at any point when determined to be no longer needed.

1.1 IF power change is going to be > 10% and OWC Injection System operating in Operator Flow Control Mode, THEN place OWC INJECTION SYS ENABLE switch to SHUTDOWN (Panel A).

CAUTION - When reducing core flow from high power, rod line could ATC exceed 118.0%.

NOTE 1 - If conditions exist where RR flow reduction cannot be reduced rapidly, rod insertion per Section 9 may be required.

NOTE 2 - Core flow reduction may result in entry into LCO 3.4.1 due to recirculation loop flow mis-match.

1.2 While monitoring rod line and feedwater flow, reduce core flow to 40x106 lbs/hr using Reactor Recirculation.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 17 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 3 Event

Description:

Outboard MSIV 86A fails partially closed Time Position Applicants Action or Behavior 1.2.1 IF RRMG is being controlled locally, THEN operate per Procedure 2.4RR.

1.2.2 Before rod line exceeds 118.0%, go to Section 9.

1.3 WHEN core flow is ~ 40x106 lbs/hr, THEN go to Section 9 and perform remaining applicable steps of rapid power reduction.

Booth If crew closes MSIV 80A, ensure Trigger 3 activates and DELETES Operator malfunction fw13a.

Performs 2.4MSIV actions as directed:

4.2 Performed by ATC.

4.3 Places control switch for affected MSIV(s) to CLOSE.

4.4 IF only one line is isolated or conditions did not result in a scram, THEN maintain MSIV(s) closed until Engineering Evaluation has been performed.

NOTE - MS-MO-78, OUTBD THROTTLE VLV, stroke time is ~ 24 seconds.

BOP 4.5 Ensure main steam line drain is in service as follows (PNL 9-4):

4.5.1 Ensure MS-MO-79, RO BYPASS VLV, is closed.

4.5.2 Throttle open MS-MO-78, OUTBD THROTTLE VLV, to intermediate position.

4.5.3 Fully open MS-MO-78, OUTBD THROTTLE VLV.

4.5.4 Open MS-MO-77, OUTBD ISOL VLV.

Booth Role Play: If sent as building operator to investigate MSIV SOLENOID Operator CURRENT METERS, wait 2 minutes, then report all meters in Green band.

Completes power reduction to 70% by inserting control rods IAW 2.1.10:

9.5.1 Insert Emergency Power Reduction Rods per Procedure 10.13:

8. EMERGENCY POWER REDUCTION 8.1 Obtain Attachment 6 from the Control Rod Sequence Package.

ATC 8.2 Insert rods in the order listed on Attachment 6 from the current position to 00 using continuous insert. Do not stop at intermediate positions.

8.3 Operator and Concurrent Verifier initials are not required on this attachment.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 18 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 3 Event

Description:

Outboard MSIV 86A fails partially closed Time Position Applicants Action or Behavior Reviews Technical Specifications:

TS LCO 3.6.1.3 Condition A for MSIV 86A inoperable.

Required Actions to isolate the penetration by at least one closed and de-activated automatic valve with a Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV AND verify affected penetration flow path is isolated with a Completion Time of once CRS per 31 days.

Determines the penetration is isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

May refer to 3.6.1.3 Condition D for MSIV leakage.

Required Action is to restore leakage within limits with a Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

END OF EVENT Notes Booth Proceed to the next event at direction of the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 19 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 4 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 4 to cause a loss of Main Condenser vacuum.

Operator BOP Report main condenser vacuum going away.

Enter Abnormal Procedure 2.4VAC and direct BOP to perform subsequent CRS operator actions.

Take ownership of 2.4VAC Scram actions:

3.1.2 IF vacuum cannot be maintained 23" Hg, THEN:

ATC 3.1.2.1 IF Annunciator 9-5-2/C-4 clear, THEN SCRAM and enter Procedure 2.1.5.

Take action per Abnormal Procedure 2.4VAC:

3. IMMEDIATE OPERATOR ACTIONS ATC 3.1 For lowering condenser vacuum:

3.1.1 Reduce power per Procedure 2.1.10 to maintain vacuum 23" Hg.

Take action per Abnormal Procedure 2.4VAC:

3. IMMEDIATE OPERATOR ACTIONS 3.2 IF vacuum cannot be maintained 23" Hg, THEN:

3.2.1 IF Annunciator 9-5-2/C-4 clear, THEN SCRAM and enter BOP Procedure 2.1.5.

3.2.2 Trip Main Turbine.

3.2.3 IF reactor not scrammed, THEN enter Procedure 2.2.77.

Provide vacuum critical parameter at which the reactor is to be scrammed.

Procedure 2.1.5 Attachments 1, 2, and 3 and scram are ATCs responsibility.

CRS Procedure 2.1.5 Attachments 4 and 5 and trip Main Turbine and close MSIVs are BOPs responsibility.

Provide BOP the critical parameter for closing MSIVs.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 20 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 4 Event

Description:

Air in-leakage causes loss of main condenser vacuum.

Time Position Applicants Action or Behavior When 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.

1.2 Place REACTOR MODE switch to REFUEL.

1.3 IF reactor power > 3%, THEN perform following:

ATC 1.3.1 Place REACTOR MODE switch to SHUTDOWN.

1.3.2 Initiate ARI.

Provides scram report:

Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

END OF EVENT Notes Note the next event (ATWS) is already active.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 21 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 5 Event

Description:

Electrical ATWS Time Position Applicants Action or Behavior NOTE to Examiners: ATWS Malfunctions and RWCU isolation failures are already active.

ATC 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).

CREW Recognize the ATWS is an electrical block.

EOP 6A Power CRS Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 22 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 5 Event

Description:

Electrical ATWS Time Position Applicants Action or Behavior NOTE to Examiners: If the crew recognizes the ATWS is an electronic block, they may not enter EOP 7A and take those actions.

EOP 7A RPV Level CRS Booth Role Play: If directed to install PTM 103 to defeat RCIC trip, wait 2 Operator minutes and report you cannot get the door open to lift the wire.

Place Reactor Mode Switch in SHUTDOWN.

Initiate ARI.

ATC Report reactor power level to CRS.

Report ARI was successful in inserting all control rods full in.

(CT-1): When RPS fails to scram the reactor on a manual scram signal, within two minutes initiate the ARI System.

CRS Transition from EOPs 6A and 7A to EOP 1A after all control rods full in.

EOP-1A step RC/L-1:

Directs BOP/ATC to verify CRS PCIS isolations per 2.1.22 ECCS initiations DG initiations END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 23 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 5 Event

Description:

Electrical ATWS Time Position Applicants Action or Behavior The next event goes active when Reactor Feed Pump A discharge valve (RF-MO-29) is closed.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 24 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior Trigger 5 causes Feedwater Line A to rupture in the drywell. (A RR Booth break is also added to represent a leak downstream of the feedwater Operator check valve).

CREW Recognize and report high drywell pressure, LOCA.

CRS Enters EOP-1A and EOP-3A on drywell pressure high.

ATC Recognizes and reports Feedwater Line break inside containment.

CRS Enters and directs ATC to perform 2.4MC-RF for Feedwater line break.

Perform the actions of 2.4MC-RF to secure Feedwater and condensate:

4.2 IF system piping not intact, THEN perform following:

4.2.1 IF break is endangering personnel or equipment necessary for safe operation, THEN perform following:

4.2.1.1 Concurrently enter Procedure 2.1.5.

4.2.1.2 Ensure RFPs tripped.

4.2.1.3 Ensure RFP discharge valves closed.

4.2.1.4 At a RFPT/RVLC HMI, perform following:

a. Select RFPT-1A or RFPT-1B System.

ATC

b. Select STARTUP VALVE screen.

c. Press EMER CLOSE button.

d. Confirm pop-up screen.

4.2.1.5 Ensure condensate booster pumps tripped.

4.2.1.6 IF necessary to stop the leak, THEN trip condensate pumps.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 25 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior Checks low pressure ECCS systems have initiated Checks low pressure ECCS pumps have started on panel 9-3:

o RHR Pump A o RHR Pump C BOP o Core Spray Pump B o RHR Pump B o RHR Pump D Reports ECCS status to CRS.

BOP Checks DG1 and DG2 are operating, with Service Water.

BOP Per EOP 1A, stabilize RPV pressure below 1050 psig - verifies Low-Low Set controlling pressure on panel 9-3.

If ADS timer has initiated THEN inhibit ADS by placing ADS A and B inhibit switches to INHIBIT on panel 9-3.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 26 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior CRS Per EOP 3A, directs Torus Sprays and then Drywell Sprays into service..

Per EOP 3A, directs Drywell pressure maintained between +2 psig and +10 psig.

Role Play: When directed by BOP to install EOP PTMs97-100, wait 3 Booth minutes then put in the overrides for the PTMs. Report back to BOP Operator when PTMs installed.

2. Containment Sprays (RHR Hard Card) 2.1 IF required, with CRS permission, THEN place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL BOP OVERRD.

2.2 IF required, THEN place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

2.3 Ensure RHR-MO-39 A(B) open.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 27 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior 2.4 IF reactor pressure 300 psig and injection not desired, THEN close RHR-MO-27B, OUTBD INJECTION VLV.

2.5 Ensure RHR PUMP(s) running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

2.6 Throttle RHR-MO-38 A(B) to maintain desired containment pressure.

2.7 Throttle RHR-MO-66 A(B) to obtain desired cooling rate.

2.8 IF Drywell Spray required, THEN perform following:

2.8.1 Open RHR-MO-31 A(B).

2.8.2 Throttle RHR-MO-26 A(B) to maintain desired containment pressure.

2.9 IF PCIS Group 6 lights lit on Panel 9-5, THEN ensure one of following open:

2.9.1 REC-MO-711; or 2.9.2 REC-MO-714.

2.10 Place RHR SW System in service:

2.10.1 Start SWBP(s).

2.10.2 Adjust SW-MO-89 A(B) to maintain flow between 2500 and 4000 gpm.

2.11 Throttle RHR-MO-66 A(B) to maintain desired cooling rate.

Maintain Drywell pressure between +2 psig and +10 psig.

CRS Assigns RPV water level as critical parameter.

CRS Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 28 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior Per EOP 1A, directs ATC to maximize CRD flow Per EOP 1A, directs ATC to initiate SLC for level control Role Play: If directed to manually close CRD-63, wait 3 minutes and report it is closed. (Valve is not modeled)

Role Play: If directed to place CRD Filter B in service, wait 2 minutes Booth and report filter is in service. (Valves are not modeled)

Operator Role Play: If directed to place Flow Control Valve B in service, wait 2 minutes and Insert Trigger 6.

Role Play: If directed to close CRD-13 and CRD-14, wait 2 minutes and Insert Trigger 7.

Maximizes CRD flow as directed IAW 5.8.4:

9.3 Start second CRD Pump A by placing its control switch to start on panel 9-5.

9.4 Close CRD-63, NBI CONTINUOUS BASKFILL SHUTOFF. (R-903-SE)

ATC 9.5 Place Standby CRD discharge filter in service (R-903-SE):

9.5.2 FILTER B 9.5.2.1 Ensure CRD-21, CRD FILTER 1B OUTLET, is closed.

9.5.2.2 Ensure CRD-20, CRD FILTER 1B INLET, is open.

9.5.2.3 Vent Filter B with CRD-22, CRD FILTER B VENT.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 29 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior 9.5.2.4 Throttle open CRD-21 until pressure equalized, then fully open.

9.6 Place Standby CRD System FCV in service to operate in parallel with in service FCV as follows (R-903-SE):

9.6.2 FLOW CONTROL VALVE B 9.6.2.1 Ensure CRD-26, FLOW CONT. VLV. AO19B INLET, is open.

9.6.2.2 Place CRD-MA-245B, SYSTEM FLOW CONTROL MANUAL/AUTO STATION, to AUTO.

9.6.2.3 Throttle open CRD-27, FLOW CONT. VLV. AO19B OUTLET, until pressure equalized, then fully open.

9.7 Close CRD-13, CRD PUMP A MINIMUM FLOW (R-881-SE QUAD).

9.8 Close CRD-14, CRD PUMP B MINIMUM FLOW (R-881-SE QUAD).

9.9 Places CRD-FC-301 to MAN and throttle open in-service FCV.

Informs CRS CRD injection has been maximized.

Sends building operator to perform local steps of 5.8.4 section 9.

Initiates SLC as directed IAW 5.8.4:

6.2 WHEN CRS directs, THEN commence Alternate RPV Injection as follows (PANEL 9-5):

6.2.1 Place following keylock switches to START:

6.2.1.1 SLC PUMP A.

6.2.1.2 SLC PUMP B.

6.2.2 Verify red indicating lights for each pump energize.

6.2.3 Verify SLC-14A, LOOP A SQUIB VALVE, has fired by observing that SQUIB VALVE READY Light 1106A has extinguished.

6.2.4 Verify SLC-14B, LOOP B SQUIB VALVE, has fired by observing that SQUIB VALVE READY Light 1106B has extinguished.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 30 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior 6.2.5 Observe SLC pump discharge pressure rises above RPV pressure as indicated on SLC-PI-65, PUMP PRESS.

6.2.6 Inform CRS that Alternate RPV Injection with SLC from boron tank has commenced.

Verifies PCIS Group Isolations IAW Procedure 2.1.22 or Hard Card.

ATC/BOP Recognize failure of RWCU to isolate on low RPV level or SLC Pump CREW actuation.

Place RWCU-MO-15 and 18 control switches to CLOSE.

BOP Report valves closed.

CRS Per EOP 1A, when RPV water level cannot be restored and maintained above -150 inches CFZ, directs ATC to line up Core Spray and LPCI for injection.

Verifies Core Spray is lined up for injection with pump running on panel 9-3.

Ensures at least one LPCI loop is aligned for injection with at least one pump running on panel 9-3.

ATC If RHR loop was in containment spray mode, on panel 9-3:

o Ensures MO-26A(B) is closed o Ensures MO-39A(B) is closed o Ensures MO-27A(B) is open Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 31 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior CRS When RPV level goes below -158 inches CFZ and cannot be maintained above -183 inches CFZ, emergency depressurization is required.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 32 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior Enters EOP-2A:

Verifies suppression pool water level is >6 feet PC-LRPR1A (Panel 9-3-1)

Directs BOP to open 6 SRVs (CT-2): When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below

-183 CFZ. (Momentary shrink below -183 due to automatic SRV closure does not constitute failure of this critical task).

Opens 6 of the following SRVs by placing control switches to OPEN on panel 9-3.

SRV-71A SRV-71B SRV-71E SRV-71G BOP SRV-71H SRV-71C SRV-71D SRV-71F Verifies red solenoid light and amber SRV tailpipe pressure switch light illuminate and green solenoid light extinguishes Verifies RPV pressure falls (CT-3): When operating injection systems cannot maintain RPV level and ECCS systems fail to automatically align for injection, crew manually aligns ECCS systems for injection:

For low pressure ECCS systems, prior to RPV pressure lowering below200 psig.

Verifies low pressure ECCS injection valves open on panel 9-3 when pressure goes below the injection valve auto open permissive (approximately 400 psig).

Core Spray - INBD INJ THROTTLE VLV MO 12A(B)

LPCI A(B) - INBD INJECTION VLV MO 25A(B)

ATC Reports Core Spray and LPCI injection valves not opening.

Reports RHR-MO-25A(B) cannot open with control switch.

Opens CS-MO-12A(B) with control switch.

Core Spray A(B) - PUMP FLOW CS-FI-50A(B)

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 2 Page 33 of 31 Op-Test No.: CNS 15-01 Scenario No.: 2 Event No.: 6, 7 & 8 Event

Description:

FW Line A break in DW (LOCA), ED on low RPV level RWCU fails to isolate on low RPV level, & Low pressure ECCS injection valves fail to open Time Position Applicants Action or Behavior Role Play: If directed to manually open RHR injection valves, wait 5 Booth minutes and report you cannot get either valve to come off its closed Operator seat.

ATC Reports CS injection flow and level rising.

CRS Directs ATC to restore and control level +3 inches to +54 inches.

When level rises above -158 inches CFZ, controls injection from CS A(B),

ATC SLC, and CRD by throttling valves and/or cycling pumps to raise and maintain level +3 inches to +54 inches.

NOTE to Examiners: Scenario objectives have been met when the crew has emergency depressurized and level is being raised in a controlled manner to restore it to +3 inches to

+54 inches.

END OF SCENARIO Notes Booth When directed by the Lead Examiner, place the simulator in freeze and Operator tell the crew to stop operating.

Rev. 0

INITIAL CONDITIONS A. Plant Status:

1. 100% power, near End of Cycle.

2. Rod Sequence Information: RO to provide B. Tech. Spec. Limitations in effect:

1. LCO 3.5.1 C.1 restore HPCI to Operable status in 14 days

2. Currently on day 2 of LCO 3.5.1.

C. Significant problems/abnormalities:

1. HPCI Auxiliary Oil Pump motor replacement.

D. PRA Risk:

Green E. Evolutions/maintenance for the on-coming shift:

1. Shift REC Pumps A and B so B is running and A is secured per Procedure 2.2.65.1, Section 8 so mechanical maintenance can take vibration readings.

AVOID entry into TS 3.8.1 for DG1 inoperable during the pump shift.

2. Electrical Maintenance working on replacing the HPCI AOP motor.

Rev. 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 1 of 30 Facility: Cooper Nuclear Station Scenario No.: 3 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Secure Standby Gas Treatment after nitrogen makeup to the drywell and torus.

2. Raise reactor power with Reactor Recirculation to 95%.

3. Respond to Battery Room Exhaust Fans failing.

4. Respond to a stuck open SRV.

5. Respond to TG bearing 9 high vibration.

6. Respond to a leak in the torus.

7. Respond to failure of 6 SRVs to open when emergency depressurizing, use of alternate ED systems.

Initial Conditions: Plant operating at 90%power.

Inoperable Equipment: None Turnover:

The plant is at 90% power.

Planned activities for this shift are:

Secure SGT A from nitrogen purge operation.

Raise power to 95% with Reactor Recirculation.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 2 of 30 Event Malf. No. Event Type Event No. Description Secure SGT after nitrogen makeup to the torus and 1 N/A N (BOP,CRS) drywell.

2 N/A R (ATC,CRS) Raise power to 95% with Reactor Recirculation OR C (BOP,CRS)

ZDIHVSWEFCI Running Battery Room Exhaust Fan trip, failure of 3 C[1]

standby fan to run, manual start of Essential Ventilation.

ZDIHVSWEFCI TS (CRS)

A[1]

C SRV fails open.

(BOP,ATC,CRS) 4 ad06c * (CT-1) When a SRV fails open, close the SRV or prior A (CREW) to torus bulk temperature reaching 110F, initiate a TS (CRS) Reactor Scram C (ATC, CRS) 5 tu3i Main Turbine Bearing #9 high vibration.

A (CREW)

Torus water leak-Emergency Depressurization

(CT-2) When torus water level cannot be maintained above 11', prevent HPCI operation prior to torus water level lowering below 11.0.

M * (CT-3) When torus water level cannot be maintained 6 pc08 above 9.6', scram the reactor prior to torus water level (CREW) falling below 9.6.

(CT-4) When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

OR zdimssws1a zdimssws1b zdimssws1a Only 2 SRVs open on emergency depressurization use 7 C (BOP,CRS) zdimssws4a of alternate ED systems.

zdimssws1d zdimssws1e (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 3 of 30 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1-2 1 All but 2 SRVs fail to open on emergency EOP entry depressurization SRV fails open.

Abnormal Events 2-4 3 MT Bearing #9 Hi Vibs Major Transients 1-2 1 Torus water leak EOP entries requiring EOP-1A 1-2 2 substantive action EOP-3A EOP contingencies requiring substantive 0-2 1 EOP-2A action (CT-1) When a SRV fails open, close the SRV or prior to torus bulk temperature reaching 110F, initiate a Reactor Scram (CT-2) When torus water level cannot be maintained above 11', prevent HPCI operation prior to torus water EOP based Critical level lowering below 11.0.

2-3 4 (CT-3) When torus water level cannot be maintained Tasks above 9.6', scram the reactor prior to torus water level falling below 9.6 (CT-4) When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

Normal Events N/A 1 Secure SGT after nitrogen makeup evolution Reactivity Manipulations N/A 1 Raise power 5% with Reactor Recirculation SRV sticks open.

Instrument/ Battery Room Exhaust fans fail.

N/A 4 Component Failures MT Bearing #9 Hi Vibs 6 SRVs fail to open on emergency depressurization SRV sticks open Battery Room Exhaust fans fail.

Total Malfunctions N/A 4 MT Bearing #9 Hi Vibs 6 SRVs fail to open on emergency depressurization Top 10 systems and operator actions important to risk that are tested:

ADS/SRV (Events 4,7)

Operator failure to depressurize with SRVs (Event 7)

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 4 of 30 SCENARIO

SUMMARY

The plant is at 90% power.

Event 1 After the crew takes the watch, Standby Gas Treatment fan is secured following nitrogen makeup evolution.

Event 2 After SGT is secured, reactor power is raised 5% using Reactor Recirculation pumps per Procedure 2.1.10.

Event 3 (Triggered by Lead Examiner)

The running Battery Room Exhaust Fan trips and the standby fan cannot be started.

The Essential Control Building Ventilation system is required to be placed into service.

The CRS determines TLCO 3.8.1 is not met and declares both Battery Room Exhaust Fans inoperable.

Event 4 (Triggered by Lead Examiner)

After the TRM is addressed for the Battery Room Exhaust fans, SRV C sticks open.

The crew enters the abnormal procedure 2.4SRV and lowers power to below 90% with Reactor Recirculation. The crew inhibits ADS, which closes the SRV (CT-1). The CRS declares the valve inoperable per LCO 3.5.1. Condition E.

Event 5 (Triggered by Lead Examiner)

The Main Turbine bearing #9 develops high vibrations requiring the crew to lower reactor power with Reactor Recirculation flow and control rod insertion. The power drop lowers bearing vibrations.

Event 6 (Triggered by Lead Examiner)

After the Main Turbine #9 bearing high vibration is addressed, the torus develops a water leak on the bottom beyond makeup capability. Due to lowering Torus level, the crew prevents HPCI operation (CT-2), scrams the reactor (CT-3), and emergency depressurizes (CT-4).

Event 7 (No Trigger required)

Only 2 SRVs can be open for Emergency Depressurization, and alternate ED systems are used.

The exercise ends when emergency depressurization is complete and RPV level recovery is under control.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 5 of 30 Critical Tasks (CT-1) When a SRV fails open, close the (CT-2) When torus water level cannot be SRV or prior to torus bulk temperature maintained above 11', prevent HPCI reaching 110F, initiate a Reactor Scram. operation prior to torus water level lowering below 11.0.

EVENT 4 6 Safety Closing the SRV or shutting down the reactor Operation of the HPCI System with its exhaust significance before 110°F in the Suppression Pool discharge device not submerged will directly ensures containment design limits due to pressurize the torus. HPCI operation is heat addition to the suppression pool will not therefore secured, as required, to preclude the be exceeded. 110°F in the Suppression Pool occurrence of this condition. The is both the Technical Specification limit and consequences of not doing so may extend to EOP-3A limit for effecting a reactor scram. failure of the primary containment from over-Tech Spec 3.6.2.1 requires that the Reactor pressurization, and thus, HPCI must be Scram be inserted at 110F. This secured irrespective of adequate core cooling requirement ensures that the unit will be shut concerns.

down at > 110F. The pool is designed to No comparable task regarding RCIC operation absorb decay heat and sensible heat but is provided because:

could be heated beyond design limits by the The exhaust flow rate of RCIC is no greater steam generated if the reactor is not shut than the steam generated by decay heat after down (TS Basis). Per PSTGs, the lowest reactor shutdown. The basis for determining temperature of the Boron Injection Initiation Primary Containment Pressure Limit assumes Temperature (BIIT) is specified as the action the operability of a containment vent capable level (110°F). A single value instead of a of removing decay heat 10 minutes after graph implements the BIIT in this step to reactor shutdown. Thus, any steam simplify the guideline. The BIIT specifies the discharged by RCIC into the torus airspace suppression pool temperature before which can be removed through the primary boron injection must be started. It is the containment vent and will not cause torus greater of: pressure to exceed PCPL even if the RCIC

The highest suppression pool temperature exhaust is not submerged.

at which initiation of boron injection will Elevated torus pressure will cause the RCIC permit injection of the Hot Shutdown Boron turbine to trip much sooner than the HPCI Weight of boron before suppression pool turbine.

temperature exceeds the Heat Capacity Temperature Limit.

The suppression pool temperature at which a reactor scram is required by plant Technical Specifications.

The BIIT is a function of reactor power. It is utilized to establish a requirement for boron injection following a failure-to-scram. If boron injection is initiated before suppression pool temperature reaches the BIIT, emergency RPV depressurization may be precluded at lower reactor power levels. At higher reactor power levels, however, the suppression pool heatup rate may become so high that the HSBW cannot be injected before suppression pool temperature reaches the Heat Capacity Temperature Limit even if boron injection is initiated early in the event. Entering the RPV Control guideline at Step RC-1 ensures that, if possible, the reactor is scrammed before boron injection is required and in anticipation of possible RPV depressurization in Step SP/T-3.

Cueing SRV open indications (solenoid lights, Lowering Torus water level, tailpipe pressure light, tailpipe temperature). approaching 11, as indicated on Step reduction in turbine generator load and SPDS and panel 9-3 indicators PC-steam flow. LRPR-1A and PC-LI-10.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 6 of 30 Rising suppression pool temperatures on panel 9-3.

Performance Operator depresses both manual scram Crew stops and prevents HPCI by one of the indicator pushbuttons or places the Reactor Mode following on panel 9-3:

Switch to SHUTDOWN on panel 9-5, prior to

Depressing and holding the HPCI trip exceeding 110°F in the Suppression Pool; pushbutton, and placing HPCI Aux Oil Pump or the operator closes the SRV IAW 2.4SRV control switch in PTL without exceeding 110°F in the Suppression

Placing HPCI-MO-16, STM SUPP Pool. OUTBD ISOL VLV control switch in CLOSE

Depressing HPCI MANUAL ISOLATION pushbutton, if initiation signal present Performance RPS Group lights de-energized on panel 9-5. HPCI speed lowers to zero on HPCI-SI-2792 feedback Reactor Power trend. on panel 9-3 Control Rod full-in indication. HPCI flow lowers to zero on HPCI-FIC-108 on SRV tailpipe pressure, steam flow, solenoid panel 9-3 lights, step increase in turbine generator load Steam supply isolation valve HPCI-MO-15 and steam flow and/or HPCI-MO-16 control switch Greed light Anytime when a SRV fails open and the illuminated and Red light extinguished on actions addressed in Procedure 2.4SRV panel 9-3.

would be effective in closing the valve OR EOP-3A conditions when actions taken IAW 2.4SRV are ineffective or not attempted.

Justification 110F is both the EOP-3A step SP/T-2 limit If torus water level cannot be restored and for the chosen and the TS 3.6.2.1 limit for reactor shutdown maintained above 11 feet is the EOP-3A, step performance to limit heat addition to the suppression pool. SP/L-10 criteria for preventing HPCI operation limit Closing the failed open SRV would also to ensure HPCI exhaust does not directly terminate heat addition to the suppression impinge on the torus air space.

pool.

BWR Owners App. B, step SP/T-2 App. B, step SP/L-2.2.

Group Appendix Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 3 Page 7 of 30 Critical Tasks (CT-3) When torus water level cannot be (CT-4) When torus water level cannot be maintained above 9.6', scram the reactor maintained above 9.6', crew Emergency prior to torus water level falling below 9.6 Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6.

EVENT 6 6 Safety Energy in the RPV should be discharged The RPV is not permitted to remain at significance outside the primary containment, if possible, pressure if suppression of steam discharged and thereby reduce or limit the energy added from the RPV into the drywell cannot be to the suppression pool if emergency RPV assured. When the downcomer vent openings depressurization becomes necessary. Entry are not adequately submerged, any steam to the RPV Control guideline is therefore discharged from the RPV into the drywell may specified before reaching the elevation of the not condense in the suppression pool before downcomer openings so that the override torus pressure reaches unacceptable levels.

before Step RC/P-1 can be used to anticipate RPV depressurization is required at or before emergency RPV depressurization and rapidly the point at which this low water level condition depressurize the RPV, irrespective of the occurs. This reduces the amount of energy resulting cooldown rate. that may be discharged directly to the torus air Entering the EOP-1A at Step 1 assures that, space to as low as possible.

if possible, the reactor is scrammed and shutdown is assured by control rod insertion before RPV depressurization is initiated.

Entry into the EOP-1A must be explicitly stated because conditions requiring entry into the EOP-3A do not necessarily require entry into EOP-1A. Therefore, a scram may not yet have been initiated. Directing that EOP-1A be entered, rather than explicitly stating here "Initiate a Reactor Scram", coordinates actions currently being executed if the EOP-1A has already been entered. In addition, entry to EOP-1A must be made because it is through EOP-1A that EOP-2A "Emergency RPV Depressurization", is performed.

Cueing Lowering Torus water level, approaching 9.6, Lowering Torus water level, approaching 9.6, as indicated on SPDS and panel 9-3 as indicated on SPDS and panel 9-3 indicators PC-LRPR-1A and PC-LI-10. indicators PC-LRPR-1A and PC-LI-10.

Performance Operator depresses both manual scram Manipulation of any six SRV controls on panel indicator pushbuttons, or places the Reactor Mode 9-3:

Switch to SHUTDOWN on panel 9-5. SRV-71A SRV-71B SRV-71E SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance RPS Group lights de-energized on panel 9-5. Crew will observe SRV light indication go from feedback Control Rod full -in indication on panel 9-5. green to red, amber pressure switch lights Reactor power trend on nuclear illuminate, reactor pressure lowering on SPDS instrumentation on panel 9-5. and panel 9-3 and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Before torus water level drops to 9.6 is the Inability to maintain torus water level above for the chosen EOP-3A, step SP/L-11 criteria for 9.6 is the EOP-3A, step SP/L-12 criteria for performance transitioning to EOP-1A to shut down the transitioning to emergency depressurization.

limit reactor.

BWR Owners App. B, step SP/L-2.1 App. B, step SP/L-2.1 Group Appendix Rev. 1

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 8 of 30 SIMULATOR SET-UP A. Materials Required None B. Initialize the simulator in IC 154, 100% power (EOL)

Batch File Name - none C. Change the simulator conditions as follows:

1. Auto Triggers Number File Name/Variable Description 8 zdimssws3a==0 ADS Inhibit switch A to INHIBIT deletes stuck open relief dmf ad06c valve malfunction.

9 zdimssws3b==0 ADS Inhibit switch B to INHIBIT deletes stuck open relief dmf ad06c valve malfunction.

2. Malfunctions Number Title Trigger TD Severity Ramp Initial tu03i MT Bearing #9 Hi Vibs 2 N/A 45 90 N/A ad06c SRV C fails open 4 N/A 100 N/A pc08 Torus water leak 6 N/A 33 3 min N/A

3. Remotes Number Title Trigger Value TD Ramp ad04 SRV C fuses 5 OUT N/A N/A Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 9 of 30

4. Overrides Instrument Tag Trigger TD Value Ramp Battery Rm Exh Fan 1C ZDIHVSWEFC1C[1] 3 N/A OFF N/A Battery Rm Exh Fan 1A ZDIHVSWEFC1A[1] N/A N/A OFF N/A zdimssws1a SRV A control switch N/A N/A AUTO N/A zdimssws1b SRV B control switch N/A N/A AUTO N/A zdimssws4a SRV D control switch N/A N/A AUTO N/A zdimssws1d SRV E control switch N/A N/A AUTO N/A zdimssws1e SRV G control switch N/A N/A AUTO N/A D. Panel Setup

1. Ensure PMIS IDTs are blank

2. Ensure RR Controllers are selected to P.

3. Start SGT train A and place drywell on nitrogen purge per Procedure 2.2.60, Step 9.1.

4. Marked up copy of Procedure 2.2.60, complete through step 9.1.9.5.

5. Ensure Battery Room Exhaust Fan 1C is running.

6. At Panel 9-4-3, set SSST Y Voltage Adjust to Tap 2.

7. At Panel C, set SSST X Voltage Adjust to Tap 5.

8. On STARTUP TRANSFORMER BACKUP VOLTAGE BUS, placard:

TAP POSITION: 5 MAX 4473 MIN 4372 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 10 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 1 Event

Description:

Secure from purging the drywell with N2 Time Position Applicants Action or Behavior CRS Direct BOP to secure from nitrogen purging Per Procedure 2.2.60.

9.1.10 WHEN purge is complete, THEN perform following:

9.1.10.1 (Independent Verification) Close PC-MO-306.

9.1.10.2 (Independent Verification) Close PC-AO-246 and return control switch to AUTO.

9.1.10.3 (Independent Verification) Open PC-AO-240.

NOTE - Securing SGT or adjusting SGT-DPIC-546, RX BLDG/SGT DP, too rapidly can cause Secondary Containment pressure transient and entry into TS LCO 3.6.4.1.

9.1.11 At VBD-R, ensure SGT-DPIC-546, RX BLDG/SGT DP, in MANUAL.

9.1.12 Ensure Parameter V displayed on SGT-DPIC-546, RX BLDG/SGT DP, by pressing D pushbutton, as necessary.

9.1.13 While maintaining secondary containment pressure -.25" wg, adjust SGT-DPIC-546, RX BLDG/SGT DP, to 100.

BOP 9.1.14 At VBD-K, check SGT heaters are off.

9.1.15 Secure running SGT fan.

9.1.15.1 SGT A

a. (Independent Verification) Place EF-R-1E, SGT A EXHAUST FAN, switch to AUTO and check following:

1. EF-R-1E stops.

2. (Independent Verification) SGT-AO-249, SGT A INLET, closes.

3. (Independent Verification) SGT-AO-251, SGT A DISCHARGE, closes.

9.1.15.2 NOTE to Examiners: Step and substeps are N/A.

9.1.16 (Independent Verification) SGT-DPIC-546, RX BLDG/SGT DP, in MANUAL.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 11 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 1 Event

Description:

Secure from purging the drywell with N2 Time Position Applicants Action or Behavior 9.1.17 (Independent Verification) SGT-DPIC-546 at 100.

Report to CRS the nitrogen purge has been secured and SGT A is in standby lineup.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 12 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 2 Event

Description:

Raise reactor power 5% with Reactor Recirculation Time Position Applicants Action or Behavior Directs ATC to raise power using Recirculation flow to 95% IAW CRS 2.1.10.

Booth Role Play: As Rx Building NLO, when requested to monitor RRMG lube oil temps and maintain 110-130°F, respond you will Operation monitor RRMG lube oil temps and maintain them in band.

Raises power using Recirculation flow IAW 2.1.10:

Selects S on RR flow controllers on panel 9-4 6.4 Raise power by raising RR pump flow as follows:

6.4.1 IF thermal power 2413 (or 2375 if power limited to 2381 MWt), THEN raise power by raising RR pump flow.

6.4.1.1 Maintain rate of power change consistent with system capabilities as determined by Load Dispatcher and TG limits.

6.5 Monitor core thermal limits (MFLCPR, MFLPD, and MAPRAT),

per Procedure 6.LOG.601, to ensure compliance with Technical Specifications Section 3.2.

Closely monitors scoop tube position/RR pump flow response.

Closely monitors reactor power on APRMs and Main Turbine output on DEH HMI.

Repeats until desired thermal power (target 95%) achieved.

BOP Provides peer check of ATC actions.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 13 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 3 Event

Description:

Battery Room Exhaust Fan trip failure of standby to run Time Position Applicants Action or Behavior Booth When directed by Lead Examiner insert Trigger 3, trip of running Operator Battery Room exhaust fan.

Report alarm R-1/D-6, BATTERY RM A VENTILATION FLOW FAILURE 1.1 NOTE to Examiners: Step is N/A.

1.2 IF failure is not due to Essential Control Building Ventilation System operation, THEN perform following:

1.2.1 Start standby fan 1.2.2 Check belts on fan that was in RUN.

1.2.3 Check inlet and outlet dampers for fan in RUN did not fail closed.

1.2.4 Check exhaust duct fire damper in room did not fail closed.

BOP 1.2.5 Check for possible duct work failure in room or Exhaust Fan room.

WARNING - Essential Control Building Ventilation must be started within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to prevent hydrogen concentrations in Battery Rooms from exceeding limits.

1.2.6 If flow cannot be established, start Essential Control Building Ventilation per Procedure 2.2.38.

Report failure of standby fan to operate.

Role Play: When dispatched to investigate the Battery Room exhaust fan 1C failure, report the fan is not running but looks normal.

If asked, report the belts are on the fan.

Booth If asked, report the inlet and outlet dampers are closed.

Operator If asked, report the fire damper in room is open. If asked, report ductwork all looks intact.

If asked, report Battery Room exhaust fan 1A is not running but appears normal.

Direct Essential Ventilation placed in service per Procedure 2.2.38.

CRS Enter Procedure 2.2.38:

NOTE 1 - Both Essential Control Building Ventilation Subsystems may be running at same time but this should only be done to maintain area BOP temperatures below 120F.

NOTE 2 - Fire Boundary Door R208, CRITICAL SWITCHGEAR 1F ROOM NORTH, may not close properly with Essential ventilation in service.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 14 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 3 Event

Description:

Battery Room Exhaust Fan trip failure of standby to run Time Position Applicants Action or Behavior NOTE 3 - Non-Essential Control Building HVAC System fan failure to run and Battery Room ventilation flow failure annunciators may alarm due to fans tripping when EF-SWGR-1F (EF-SWGR-1G) starts.

5.1 At VBD-R, place EF-SWGR-1F (EF- SWGR-G), EXHAUST FAN, switch to START and check:

5.1.1 Annunciator R-1/C-4, CONTROL BLDG ESSENTIAL H&V FANS RUNNING, alarms.

5.1.2 EF-SWGR-1F (EF-SWGR-1G), EXHAUST FAN, starts.

5.1.3 SF-SWGR-1F (SF-SWGR-1G), SUPPLY FAN, starts.

5.1.4 Check dampers close:

5.1.4.1 HV-AD-AD1405, EXH ISOL DAMPER.

5.1.4.2 HV-AD-AD1406, EXH ISOL DAMPER.

5.1.4.3 HV-AD-AD1407, SUPP ISOL DAMPER.

5.1.4.4 HV-AD-AD1408, SUPP ISOL DAMPER.

5.1.4.5 HV-AD-AD1409, ISOL DAMPER.

5.1.4.6 HV-AD-AD1410, ISOL DAMPER.

5.1.5 Check fans off:

5.1.5.1 SF-C-1A-A, SUPPLY FAN.

5.1.5.2 SF-C-1A-B, SUPPLY FAN.

5.1.5.3 RF-C-1A, RECIRC FAN.

5.1.5.4 RF-C-1B, RECIRC FAN.

5.1.5.5 EF-C-1A, BATT RM EXH FAN.

5.1.5.6 EF-C-1C, BATT RM EXH FAN 5.2 At VBD-R, place fan switches to OFF:

5.2.1 SF-C-1A-A, SUPPLY FAN.

5.2.2 SF-C-1A-B, SUPPLY FAN.

5.2.3 RF-C-1A, RECIRC FAN.

5.2.4 RF-C-1B, RECIRC FAN.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 15 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 3 Event

Description:

Battery Room Exhaust Fan trip failure of standby to run Time Position Applicants Action or Behavior Determine TLCO 3.8.1 not met.

Declare EF-C-1C and 1A inoperable.

Entered Conditions A and B.

For Condition A, Required Action is to restore Battery Room fan operable CRS with a Completion Time of 7 days.

For Condition B, Required Action is to initiate action to establish battery room ventilation with a Completion Time of immediately.

When essential ventilation established, exited Condition B.

Direct battery room ventilation to be established immediately.

END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 16 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 4 Event

Description:

SRV C Fails Open Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 4 failing Operator SRV C Open.

Report indication of SRV C being open.

Reports following alarms to CRS:

9-3-1/A-2 RELIEF VALVE OPEN 2.1 Check amber lights on Panel 9-3 to determine which valve(s) has opened.

BOP 2.2 Enter Procedure 2.4SRV if valve does not close.

9-3-1/C-1 SAFETY/RELIEF VALVE LEAKING 1.1 Check MS-TR-166 to determine leaking valve.

1.2 Monitor suppression pool temperature.

1.3 Enter Procedure 2.4SRV.

(CT-1): When a SRV fails open, close the SRV or prior to torus bulk temperature reaching 110°F, initiate a Reactor Scram BOP Reports event is an entry condition for Procedure 2.4SRV.

CRS Enters 2.4SRV and assigns subsequent actions to BOP.

NOTE to Examiners: Placing the first ADS Inhibit Switch to INHIB deletes malfunction and allows the valve to close.

Enters 2.4SRV 4.2 Place stuck open relief valve (SORV) control switch to OPEN.

4.3 Rapidly reduce reactor power to 90%.

4.4 BEFORE average suppression pool temperature reaches 110F, BOP THEN SCRAM and concurrently enter Procedure 2.1.5.

4.5 Monitor reactor power while attempting to close relief valve.

4.6 Inform Shift Manager to review EAL SU6.1 for applicability.

4.7 NOTE to Examiner: Step is N/A.

4.8 IF an ADS valve affected, THEN perform following:

4.8.1 Place ADS INHIBIT A and B to INHIB.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 17 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 4 Event

Description:

SRV C Fails Open Time Position Applicants Action or Behavior 4.8.2 (Independent Verification) Place SORV control switch to AUTO.

4.9 NOTE to Examiners: Step is N/A.

Report the SRV has closed.

4.10 Concurrently place RHR Subsystem closest to SORV in Suppression BOP Pool Cooling per Procedure 2.2.69.3.

Role Play: If directed to the Auxiliary Relay Room to standby and pull fuses in Panel 9-45, report you are on your way.

Booth Operator When directed to pull fuses for SRV C, wait 3 minutes and insert Trigger 5, report you have pulled the fuses.

4.11 In Panel 9-45, Terminal Board CC, pull fuses for affected valve (Auxiliary Relay Room).

4.11.3 SRV C BOP 4.11.3.1 2E-F3C 4.11.3.2 2E-F11C 4.12 (Independent Verification) Ensure ADS INHIBIT in AUTO.

Enter TS LCO 3.5.1, Condition E. Required Action is to restore the ADS valve to an Operable status with a Completion Time of 14 days.

When RHR placed in suppression pool cooling, entered TS LCO 3.5.1, Condition A. Required Action is to restore LPCI to operable status with a Completion Time of 7 days.

CRS Enter TS LCO 3.5.1 Condition F. Required Actions is to restore ADS valve or the LPCI function to operable status with a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

If torus high level comes in due to SRV being open, entered TS LCO 3.6.2.2, Condition A. Required Action is to restore suppression pool level in limits with a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 18 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 4 Event

Description:

SRV C Fails Open Time Position Applicants Action or Behavior END OF EVENT Notes Proceed to the next event at direction of the lead examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 19 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 5 Event

Description:

Main Turbine Bearing #9 high vibration Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 2 to cause Exciter Operator bearing vibrations to rise.

Report rising Main Turbine bearing vibration.

Report bearing #9 only bearing vibration rising.

Respond to annunciator B-1/C-2, TG ROTOR HI VIBRATION NOTE - 7 mils causes the indicator bar on TGI-M-DUA and TGI-M-DUB to turn yellow, 10 mils turns indicator bars red.

1.1 Validate alarm by checking other bearings on Monitors TGI-M-DUA, TGI VIB/INST MONITOR CHANNEL A, and/or TGI-M-DUB, TGI VIB/INST MONITOR CHANNEL B, and/or observing locally:

1.1.1 Select Turbine Mimic and/or Generator Mimic to determine which bearing is alarming.

1.1.2 Select alarmed bearing screen, as required.

1.2 Note to Examiners: Step is N/A.

BOP 1.3 IF bearing vibration greater than or equal to Danger level (10 mils), THEN enter Procedure 2.4TURB.

NOTE - Bypassing Alert level alarm will allow for Danger alarm to cause audible and viable warning as both affect same alarm point.

1.4 IF vibration remains greater than Alert (7 mils) and less than Danger (10 mils), and is stable, THEN consider bypassing Alert level alarm per Procedure 2.2.77.

1.5 IF Bearing 9 is > 10 mils, THEN take action per Procedure 2.2.52, OPERATION OF H2 SEAL OIL SYSTEM WITH GENERATOR BEARING 9 > 10 MILS.

1.6 Note to Examiners: Step is N/A.

Procedure 2.2.52, Section 10 Note to Examiners: Section 10 performed locally by NLOs.

CRS Enter Abnormal Procedure 2.4TURB, assign actions to BOP.

Enter 2.4TURB Attachment 1:

BOP NOTE - 7 mils causes the indicator bar on TGI-M-DUA and TGI-M-DUB to turn yellow, 10 mils turns TGI-M-DUA and TGI-M-DUB red.

1. Validate vibration by observing vibration on several bearings as Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 20 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 5 Event

Description:

Main Turbine Bearing #9 high vibration Time Position Applicants Action or Behavior read on TGI-M-DUA, TGI VIB/INST MONITOR CHANNEL A, and/or TGI-M-DUB, TGI VIB/INST MONITOR CHANNEL B.

1.1 Select Turbine Mimic and/or Generator Mimic to determine which bearing is alarming.

1.2 Select alarmed bearing screen, as required.

1.3 IF time permits, THEN locally observe turbine for vibration (i.e.,

visual indication or feeling of abnormal vibration of turbine casing, bearing casing, or attached piping; abnormal sounds; or if local instrumentation is installed, indication of abnormal vibrations indicated).

2. NOTE to Examiners-Step is N/A.

3. For Bearing 9 only, IF rotor vibration 14 mils, THEN reduce power to maintain < 14 mils.

4. IF any bearing vibration is 10 mils, THEN immediately contact Turbine Engineering Group for data analysis and recommendation, and Vendor support.

Direct ATC lower power to lower bearing #9 vibration, if vibs are above CRS 14 mils.

Lower power with RR per Procedure 2.1.10 Selects S on RR flow controllers on panel 9-4 7.4 Lowers RR pump flow (by turning speed demand counter-ATC clockwise on one speed controller at a time and allowing conditions to stabilize before adjusting other controller).

Closely monitors scoop tube position/RR pump flow response.

Repeats until desired thermal power achieved.

BOP Provide peer check for lowering power.

END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 21 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 5 Event

Description:

Main Turbine Bearing #9 high vibration Time Position Applicants Action or Behavior Proceed to the next event at direction of the lead examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 22 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior Booth When directed by lead examiner, insert Trigger 6 causing a Torus water leak.

Operator BOP Report lowering Suppression Pool level.

CRS Enter EOP 5A on high reactor building sump level.

Role Play: If sent to investigate RHR in the Reactor Building, wait 4 Booth minutes and call back and report you find nothing wrong with the Operator RHR system.

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: If sent to investigate damage in the Reactor Building, wait Booth 4 minutes and call back and report that the Suppression Pool has a Operator leak near the bottom and water is pouring out.

Role Play: If directed to maintain CST level per Section 10 of Booth Emergency Procedure 5.8.14, respond you will complete Demin Water Operator transfer per Step 10.3.1.

BOP Enter Procedure 5.8.14:

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 23 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior If using HPCI and/or RCIC, the system Minimum flow valve will be opened:

HPCI-MO-25 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.

(CT-2): When torus water level cannot be maintained above 11', prevent HPCI operation prior to torus water level lowering below 11.0.

CRS Direct stop and prevent with HPCI.

Booth Action: When HPCI operation prevented or 11 in Suppression Pool, Operator modify PC08 to 44% to raise leak.

Place HPCI Aux Oil Pump control switch in Pull-To-Lock.

BOP May order Emergency Depressurization is Anticipated and direct BOP to CRS fully open the main turbine bypass valves.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 24 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior If directed to fully open bypass valves, per DEH Hard Card (Procedure 2.2.77.1):

1.1 On BYPASS VALVE POSITION control, press OPEN to access controls.

BOP 1.2 Press MANUAL button and check it backlights yellow.

1.3 On BYPASS VALVE POSITION control, use UP/DOWN, JOG UP/JOG DOWN and FAST/SLOW controls to adjust BYPASS VALVE POSITION to desired value.

(CT-3): When torus water level cannot be maintained above 9.6', scram the reactor prior to torus water level falling below 9.6.

CRS Transition to EOP 1A.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 25 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior CRS Direct ATC to scram the reactor.

Press both RX SCRAM buttons.

Place REACTOR MODE switch to REFUEL.

Provides scram report:

ATC Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

CRS Direct pressure controlled 800 psig to 1050 psig per EOP 1A Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 26 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior BOP Allow Main Turbine bypass valves to control RPV pressure.

CREW Report PCIS Group isolations Per EOP 1A, direct ATC to control RPV level in band.

ATC Operate RFPs to control RPV level in band directed by CRS.

CRS May Anticipate Emergency Depressuriation.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 27 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior If directed to fully open BPVs, (Perform one of following)

1. ADJUSTING PRESSURE SETPOINT CONTROL IN ACTUAL 1.1 On PRESSURE SETPOINT control, perform following:

1.1.1 Press ACTUAL pushbutton and verify if highlights yellow.

1.1.2 Change ACTUAL SP as follows:

1.1.2.1 Utilize UP/DOWN, FAST/SLOW until ACTUAL SP is desired value.

BOP 1.1.2.2 Push TARGET pushbutton and verify it highlights yellow.

3. MANUAL BPV CONTROL 3.1 On BYPASS VALVE POSITION control, press OPEN to access controls.

3.2 Press MANUAL button and check it backlights yellow.

3.3 On BYPASS VALVE POSITION control, use UP/DOWN, JOG UP/JOG DOWN and FAST/SLOW controls to adjust BYPASS VALVE POSITION to desired value.

END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 28 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 6 Event

Description:

Torus leak and Scram Time Position Applicants Action or Behavior The next event (Emergency Depressurization) is a continuation.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 29 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 7 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 (CT-4): When torus water level cannot be maintained above 9.6', crew Emergency Depressurizes by opening 2 SRVs prior to torus water level falling below 9.6 Direct ED per EOP 2A CRS When directed to ED, verify PC water level is above 6 ft. and open 6 SRVS BOP by taking their control switches to OPEN.

Report SRV A, B, D, E, and G failed to open.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 30 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 7 Event

Description:

Emergency Depressurize on low PC water level.

Time Position Applicants Action or Behavior Direct Alternate Depressurization systems be placed in service per EOP CRS 5.8.2.3 Direct BOP to fully open Main Turbine bypass valves.

NOTE to Examiners: If Emergency Depressurization was Anticipated, the bypass valves are already open.

Utilize Procedure 2.2.77.1 Hard Card to take manual control of the BPVs and fully open them.

3. MANUAL BPV CONTROL 3.1 On BYPASS VALVE POSITION control, press OPEN to access controls.

ATC 3.2 Press MANUAL button and check it backlights yellow.

3.3 On BYPASS VALVE POSITION control, use UP/DOWN, JOG UP/JOG DOWN and FAST/SLOW controls to adjust BYPASS VALVE POSITION to desired value.

Fully open BPVs.

Direct ATC maintain RPV level band of -110 inches to -60 inches FZ CRS during ED.

Utilize low pressure systems to control an injection flow rate of a minimum ATC 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 drywell pressure report ED is BOP complete.

CRS Directs ATC to restore and control level +3 inches to +54 inches.

Controls injection from CS, RHR, or Condensate by throttling valves and/or ATC cycling pumps to raise and maintain level +3 inches to +54 inches.

NOTE to Examiners: Scenario objectives have been met when the crew has emergency depressurized and level is being raised in a controlled manner to restore it to +3 inches to +54 inches.

END OF SCENARIO Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 3 Page 31 of 30 Op-Test No.: CNS 15-01 Scenario No.: 3 Event No.: 7 Event

Description:

Emergency Depressurize on low PC water level.

Time Position Applicants Action or Behavior Notes Booth When directed by the lead evaluator, place the simulator in freeze and Operator tell the crew to stop operating.

Rev. 0

INITIAL CONDITIONS A. Plant Status:

1. 90% power, near End of Cycle.

2. Rod Sequence Information: RO to provide B. Tech. Spec. Limitations in effect:

None C. Significant problems/abnormalities:

None D. PRA Risk:

Green E. Evolutions/maintenance for the on-coming shift:

1. Secure SGT from nitrogen purge operation per 2.2.60, at Step 9.1.10.

2. Raise power to 95% with Reactor Recirculation per the load schedule.

Rev. 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 1 of 42 Facility: Cooper Nuclear Station Scenario No.: 4 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Shift RRMG oil pumps.

2. Place RHR SP cooling in service.

3. Respond to CRD pump trip.

4. Respond to a RR Pump A #1 seal failure and subsequent pump trip.

5. Respond to a RR Pump A #2 seal failure. Vent PC.

6. Respond to a FW line break inside PC.

7. Respond to failure of HPCI to automatically start.

8. Respond to loss of RPV level indication, flood the RPV to the MSLs.

Initial Conditions: Plant operating at 100%power.

Inoperable Equipment: None Turnover:

The plant is at 100% power.

Planned activities for this shift are:

Shift RRMG oil pumps.

Place RHR SPC in service.

Maintain present power level.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 2 of 42 Event Malf. No. Event Type Event No. Description 1 N/A N (ATC,CRS) Shift RRMG lube oil pumps N (BOP,CRS) 2 rf rh14 Place RHR loop B in SPC, Min flow valve de-energizes open.

TS (CRS) 3 rd08b C (ATC,CRS) CRD Pump B trip.

C (BOP) 4 rr10a A (CREW) RR Pump A seal #1 leak and RR Pump A trip.

TS (CRS)

C rr04b 5 (BOP,ATC,CRS) RR Pump A seal #2 leak, vent PC.

rr11a A (CREW)

FW Line B break in PC-Scram M (CT-1) Initiate drywell sprays when torus pressure 6 fw18b exceeds 10 psig, prior to drywell temperature reaching (CREW) 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

7 hp01 C (BOP,CRS) HPCI fails to automatically start.

Loss of RPV level instruments, RPV flooding.

(CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

(CT-3) When RPV level cannot be determined and the reactor has been depressurized below the shutoff head of the respective pump(s), inject into the RPV to flood M

8 NBI various to the Main Steam Lines before drywell radiation (CREW) reaches 150 R/hr or entering PC Flooding.

(CT-4) When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines.

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Critical Task (As defined in NUREG 1021 Appendix D)

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 3 of 42 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target Malfunctions after 1-2 1 HPCI fails to automatically start.

EOP entry RR pump trip.

Abnormal Events 2-4 2 RR seal leakage FW line break inside PC Major Transients 1-2 2 Loss of all RPV level instruments EOP-1A EOP entries requiring substantive action 1-2 2 EOP-3A EOP contingencies requiring substantive 0-2 1 EOP- 2B action (CT-1) Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

(CT-2) When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

EOP based Critical (CT-3) When RPV level cannot be determined and the 2-3 4 reactor has been depressurized below the shutoff head Tasks of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding.

(CT-4) When RPV level cannot be determined andat least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines.

Shift RRMG oil pumps Normal Events N/A 2 Place RHR Suppression Pool Cooling in service Reactivity Manipulations N/A 0 N/A CRD Pump trip.

Instrument/ RR pump A seal #1 failure with RR pump trip.

N/A 4 Component Failures RR pump A seal #2 failure.

HPCI fails to automatically start CRD Pump trip.

RR pump A seal #1 failure with RR pump trip.

Total Malfunctions N/A 4 RR pump A seal #2 failure.

HPCI fails to automatically start Top 10 systems and operator actions important to risk that are tested:

Nuclear Boiler Instrumentation (Event 8)

Residual Heat Removal in Containment Spray Mode (Event 6)

HPCI (Event 7)

ADS/SRV Operator fails to depressurize with SRVs Operator fails to initiate ADS and initiate ECCS early.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 4 of 42 SCENARIO

SUMMARY

The plant is at 100% power.

Event 1 After the crew takes the watch, the ATC shifts RRMG oil pumps B1 and B3 per procedure 2.2.68.1. The oil pump shift is in preparation for tagging out the oil pump later in the shift.

Event 2 The BOP then places RHR in Suppression Pool Cooling in preparation a HPCI run the next shift. As the system's minimum flow valve starts to close it de-energizes in an intermediate position. The CRS declares the LPCI subsystem inoperable per LCO 3.5.1, Condition A. The valve is declared inoperable per LCO 3.6.2.3, Condition A.

Event 3 (Triggered by Lead Examiner)

After Technical Specifications are addressed for LPCI inoperable, the operating CRD pump trips requiring the ATC to start the standby pump.

Event 4 (Triggered by Lead Examiner)

After the CRD pump trip is addressed, RR pump A develops a #1 seal failure. The crews responds to rising seal temperatures and lowers RR pump speed.

Subsequently the RR pump trips, placing plant operation near the buffer region of the power to flow map. The CRS enters TS LCO 3.4.1.

Event 5 (Triggered by Lead Examiner)

After the RR pump trip is addressed, the pump's #2 seal develops a leak requiring the pump to be isolated and the PC to be vented with Standby Gas Treatment.

Event 6 (Triggered by Lead Examiner)

After the #2 seal failure is addressed, FW line B develops a leak inside PC. The reactor scrams on high drywell pressure. The crew initiates Torus and Drywell Sprays (CT-1).

Event 7 (No Trigger required)

HPCI fails to automatically start on high drywell pressure and must be started manually.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 5 of 42 Event 8 (Triggered by Lead Examiner)

All RPV level instrumentation is lost and the crew emergency depressurizes (CT-2).

Steam lines are isolated (CT-3) and the crew uses injection systems to flood the RPV to the bottom of the Main Steam Lines (CT-4).

The exercise ends when emergency depressurization is complete and RPV level is maintained at the bottom of the MSLs.

Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 6 of 42 Critical Tasks (CT-1) Initiate drywell sprays when torus (CT-2) When RPV level cannot be pressure exceeds 10 psig, prior to drywell determined and torus level is above 6',

temperature reaching 280F and prior to open six SRVs before drywell radiation torus pressure exceeding the Pressure reaches 150 R/hr or entering PC Flooding.

Suppression Pressure (PSP) curve.

EVENT 6 8 Safety Drywell sprays are initiated in two legs of Depressurization of the RPV is necessary to significance EOP-3A: Temperature and Pressure control. perform the RPV flooding actions for the following reasons:

The open SRVs establish a path from Regarding drywell temperature, if operation the RPV capable of rejecting energy of all available drywell cooling is unable to in excess of decay heat to ensure the terminate increasing drywell temperature RPV flooding actions are successful.

before the structural design temperature limit Reduced RPV pressure results in of 280 F is reached, drywell sprays are increased injection flow rates, initiated to affect the required drywell reducing the total time required to flood the RPV.

temperature reduction status of the DSIL and Reduced RPV pressure reduces the adequate core cooling permitting. Spray water inventory loss through operation effects a drywell pressure and non-isolable leaks and breaks.

temperature reduction through the combined Dynamic loading on the SRVs and effects of evaporative cooling and convective downstream piping is minimized as cooling. RPV water level reaches and is discharged through these valves.

Regarding drywell pressure, operation of RPV depressurization can be most easily and drywell sprays reduces primary containment rapidly accomplished by opening SRVs. The pressure by condensing any steam that may ADS valves are used first since they are the be present and by absorbing heat from the most reliable, considering component containment atmosphere through the qualifications, pneumatic supply systems, combined effects of evaporative and initiation circuitry, and control power. In convective cooling. Drywell sprays are addition, the relative locations of the ADS initiated when torus pressure exceeds the valve discharges provide uniform distribution of Torus Spray Initiation Pressure (10# torus the heat load around the suppression pool.

pressure) to preclude chugging the cyclic The direction to open all ADS valves requires condensation of steam at the downcomer manual action, even if the valves are already openings of the drywell vents. When a steam open on high pressure. Automatic valve bubble collapses at the exit of the operation in the relief or safety mode does not downcomers, the rush of water drawn into accomplish the objective of this step, even if the downcomers to fill the void induces low-low set logic has actuated. RPV flooding stresses at the junction of the downcomers conditions are defined based on steam flow and the vent header in Mark I containments through the SRVs. Direct manual control must and at the junction of the downcomers. be established to ensure that the valves Repeated application of such stresses could remain open as RPV pressure decreases.

cause fatigue failure of these joints; thereby, SRVs may be opened only if suppression pool creating a direct path between the drywell water level is above the elevation of the top of and torus. When drywell sprays are initiated, the discharge devices. If the SRVs were the resulting pressure reduction opens the opened with the discharge devices exposed, vacuum breakers, drawing non-condensable steam would pass directly into the suppression from the torus back into the drywell. This chamber airspace, bypassing the suppression condition defines the Torus Spray Initiation pool. The resulting pressure increase could Pressure. As the drywell atmosphere is exceed the maximum pressure capability of purged to the torus and replaced by steam, the primary containment.

torus pressure increases. The SCSIP is the Failing to depressurize could prevent recovery lowest torus pressure which can occur when of RPV level above MSCRWL, resulting in core 95% of the non-condensable in the drywell damage have been transferred to the torus. Since the failure mode is based on fatigue failure, a precise time limit or pressure cannot be provided. Therefore, prompt initiation of drywell sprays is required based on existing Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 7 of 42 EOP priorities.

Cueing Rising torus pressure indicated on SPDS and Erratic or inconsistent indication on all RPV panel 9-3 recorder PC-LRPR-1A. level indications, and CRS declares RPV level cannot be determined.

Cursor approaching unsafe boundary on PSP graph display on SPDS.

Performance Aligns torus spray on panel 9-3 using RHR Manipulation of any six SRV controls on panel indicator loop A and/or B: 9-3:

SRV-71A places CONTMT COOLING 2/3 CORE SRV-71B VALVE CONTROL PERMISSIVE SRV-71E switch to MANUAL OVERRD SRV-71G SRV-71H opens RHR-MO-39B, if closed SRV-71C SRV-71D closes close RHR-MO-27B, OUTBD SRV-71F INJECTION VLV, if necessary starts RHR PUMP(s), if not running For drywell spray, opens RHR-MO-31B Performance On panel 9-3, RHR pump/valve control Crew will observe SRV light indication go from feedback switch light indication consistent with green to red, amber pressure switch lights intended operation (Red - open/running, illuminate, reactor pressure lowering on SPDS Green - closed/stopped). and panel 9-3 and 9-5 meters and recorders, RHR flow rate rises on recorder RHR-FR-143 and SRV tailpipe temperatures rise on and indicator RHR-FI-133A(B.) recorder MS-TR-166.

Torus/drywell pressure stabilizes/lowers on SPDS and panel 9-3 recorder PC-LRPR-1A.

Justification When torus pressure cannot be maintained Before 150R/hr in the drywell was chosen for the chosen below PSP is the EOP-3A, step PC/P-4 because this is an indicator of loss of RPV performance criteria requiring transition to emergency level and the shielding effect of the water, limit depressurization. indicating core exposure, yet it is much lower than the 2500R/hr trigger point during RPV Flooding that indicates gross cladding failure is in progress. Before exiting to PC Flooding was chosen because the design of the scenario provides the crew with the means to restore and maintain adequate core cooling IAW EOP-2B or 7B, and exiting to SAGs is neither required nor authorized.

BWR Owners App. B, step PC/P-1. App. B, Contingency#4 Group Appendix Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 8 of 42 Critical Tasks (CT-3) When RPV level cannot be (CT-4) When RPV level cannot be determined and the reactor has been determined and at least 1 SRV is manually depressurized below the shutoff head of opened, isolate MSIVs, MSL drains, HPCI the respective pump(s), inject into the steam supply, and RCIC steam supply, RPV to flood to the Main Steam Lines prior to RPV water level rising to the bottom before drywell radiation reaches 150 R/hr of the main steam lines.

or entering PC Flooding.

EVENT 8 8 Safety Once the SRVs have been opened to Steam lines connected to the RPV are isolated significance depressurize the RPV, injection systems are prior to initiating action to flood the RPV to aligned to flood the RPV and establish core preclude damage which may occur from cold cooling by submergence. The list of flooding water coming in contact with the hot metal, methods includes all motor-driven systems excessive loading of lines or hangers not capable of injecting into the RPV. Any or all designed to accommodate the weight of water, of these systems may be used, as and flooding of steam driven equipment (RCIC necessary, to flood the RPV to the elevation turbine, main turbine, etc.). Isolation is of the main steam lines. Steam-driven performed, however, only if the status of SRVs systems are not listed since, with SRVs open assures the RPV will remain depressurized and the reactor shut down, the RPV will during the flooding evolution. For non-ATWS, depressurize to below the turbine stall only one SRV open is required to meet this pressures. Failing to raise RPV level to and condition.

observable point could prevent recovery of RPV level above MSCRWL, resulting in core damage.

Cueing Erratic or inconsistent indication on all RPV Erratic or inconsistent indication on all RPV level indications, and CRS declare RPV level level indications, and CRS declares RPV level undetermined. cannot be determined, and SRVs have been Six ADS valves have been manually manually opened IAW EOP-2B or EOP-7B for opened. RPV depressurization.

Performance Crew establishes injection flow by Crew places the following valve control indicator manipulating controls as required to start the switches to CLOSE:

associated pumps and align system valves Inboard MSIVs on panel 9-3 for injection using at least two pumps of the MSL Drains on panel 9-4 following systems: HPCI steam supply on panel 9-3 Main condensate/booster pumps on panel A RCIC steam supply on panel 9-4 RHR/LPCI loop A and/or B on panel 9-3 Core spray A and/or B

[Operator places affected ECCS pump(s) control switch(es) to START and valve control switches to OPEN (or CLOSE, if necessary)]

Performance Indication that the RPV is flooded to the main Indication for applicable isolation valves Green feedback steam lines may include one or more of the light illuminates and Red light extinguishes.

following indication on panels 9-3, 9-4, 9-5 or field reports by the booth operator:

Rising RPV pressure

If a main steam line is not isolated, field report of two-phase flow conditions audible in the vicinity of the steam tunnel, main steam equalizing header, or main turbine stop and bypass valves

Actuation of HPCI, RCIC or main steam line high flow logic

Field report of water leakage from HPCI or Rev. 1

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 4 Page 9 of 42 RCIC turbine shaft seals

HPCI/RCIC STM DRAIN POT LEVEL HI alarms

SRVs re-open and stay open at RPV pressures below 50 psig above torus pressure

If injection sources are aligned with torus suction, torus water level:

- decreases as RPV and steam lines are flooded

- stabilizes when steam lines are full

Local torus water temperatures near open SRVs Justification LOCA severity should result in a near linear Equipment damage due to cold water cannot for the chosen RPV level reduction that gives the crew an occur until water level reaches the main steam performance initial trend on all level instruments. Failing lines.

limit all of the level instruments should occur within about 30 seconds and should yield inconsistent indications such that there is no doubt level cannot be determined (e.g. LOCA conditions with operation in the possible boiling region of the RPVST curve, minimal RPV injection, level slowly lowering to -100 CFZ, then all level instruments fail upscale within 10 seconds, simulating all reference legs flashing).

BWR Owners App. B, Contingency #4. App. B, Contingency#4, step C4-2.2 Group Appendix Rev. 1

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 10 of 42 SIMULATOR SET-UP A. Materials Required None B. Initialize the simulator in IC 155, 100% power (EOL)

Batch File Name - none C. Change the simulator conditions as follows:

1. Auto Triggers Number File Name/Variable Description 1 zlorhrsws16b(1)==1 RHR-MO-16 Green light ON irf rh14 DE-ENER 9 zlorpsds1b==0 When RPS de-energizes, the feedwater leak goes to 100%

mmf fw18b 100 3:00 after 3 minute time delay

2. Malfunctions Number Title Trigger TD Severity Ramp rd08b CRD Pump B trip 2 N/A N/A N/A rr04b RR Pump A trip 3 4 min N/A N/A rr10a RR Pump A #1 seal failure 3 N/A 100 N/A rr11a RR Pump A #2 seal failure 4 N/A 15 N/A fw18b FW line B break 5 N/A 10 N/A hp01 Failure of HPCI to auto start A N/A N/A N/A rr33A Reference Leg 3A Failure 7 9:00 20% 2:00 rr44 NBI-LT-92 failure 7 9:00 80% 2:00 rr43 NBI- LT- 61(flood up level) 7 9:00 5% 2:00 rr41a NBI-LT-59A (wide range level A) 7 9:00 100% 2:00 rr41b NBI-LT-59B (wide range level B) 7 9:00 10% 2:00 rr41c NBI-LT-59C (wide range level C) 7 10:00 0% N/A rr41d NBI-LT-59D (wide range level D) 7 10:00 100% N/A rr42a NBI-LT-91a (Fuel Zone Level A) 7 9:00 20% 2:00 rr42b NBI-LT-91b (Fuel Zone Level B) 7 9:00 60% 2:00 rr27a NBI-LT-52a (narrow range level A) 7 8:00 0 2:00 rr27b NBI-LT-52b (narrow range level B) 7 8:00 60 2:00 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 11 of 42 rr27c NBI-LT-52c (narrow range level C) 7 8:00 30 2:00

3. Remotes Number Title Trigger Value TD Ramp rh14 de-energize RHR-MO-16B 1 DE-ENER N/A N/A

4. Overrides Instrument Tag Trigger TD Value Ramp 9-3-2/B-5 Rx Low Water Level -

an:p1704 7 8:00 ON N/A 113 9-3-2/B-5 Rx Low Water Level -

an:p1705 7 8:00 OFF N/A 113 9-5-1/B-1 Rx Low Water Level -

an:p2103 7 10:45 ON N/A 113 9-3-2/A-5 Rx Low Water Level -

an:p1701 7 9:30 ON N/A 42 9-3-3/D-2 RHR-16 overload an:p1266 1 N/A ON N/A D. Panel Setup

1. Ensure PMIS IDTs are blank

2. Ensure RR Controllers are selected to P.

3. Ensure RRMG A supplied from the SSST and RRMG B from the NSST

4. Have clean copy of 2.2.68.1 ready for turnover.

5. Mark up Procedure 2.2.69.3 complete through Step 8.18. N/A Steps 8.20

6. Place B Loop RHRSW in service.

7. Ensure CRD Pump B operating.

8. Ensure REC TIC-451B MODE marked with AUTO and 70.

9. At Panel 9-4-3, set SSST Y Voltage Adjust to Tap 2.

10. At Panel C, set SSST X Voltage Adjust to Tap 5.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 12 of 42

11. On STARTUP TRANSFORMER BACKUP VOLTAGE BUS, placard:

TAP POSITION: 5 MAX 4473 MIN 4372 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 13 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 1 Event

Description:

Shift RRMG set oil pumps Time Position Applicants Action or Behavior Direct the ATC to shift RRMG Set B Lube Oil Pumps B1 and B3 per CRS Procedure 2.2.68.1, section 20.

20. SHIFTING RRMG SET OIL PUMPS NOTE - Annunciators 9-4-3/F-2(F-6), RRMG A(B) FLUID DRIVE OIL LOW PRESSURE, may momentarily alarm during pump shifting.

20.1 Place and hold standby oil pump switch to START.

ATC 20.2 Place one operating oil pump's switch to STOP.

20.3 WHEN standby oil pump has started, THEN release standby oil pump switch and check it spring returns to NORMAL.

20.4 (Independent Verification) Place secured oil pump's switch to NORMAL.

END OF EVENT Notes Proceed to the next event.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 14 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 2 Event

Description:

Place RHR Loop B in Suppression Pool Cooling (Procedure 2.2.69.3)

Time Position Applicants Action or Behavior Directs BOP to place RHR Loop B in Suppression Pool Cooling per CRS 2.2.69.3, continuing at Step 8.19 Booth Role Play: If contacted, report suction line drain flush for hot spots Operator is not required.

BOP 8.19 Inform CRS LPCI Mode of RHR Subsystem B inoperable.

NOTE to Examiners: Annunciator 9-3-1/G-1, ADS AUX COOLING INTERLOCK, is an expected alarm when starting an RHR pump.

8.20 If not spot flush required, THEN perform following: Note to Examiner, Step is N/A.

8.21 IF required, with CRS permission, THEN place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL OVERRD. Note to Examiner, Step is N/A.

8.22 IF required, THEN place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL. Note to Examiner, Step is N/A.

8.23 Open RHR-MO-39B, SUPPR POOL COOLING/TORUS SPRAY VLV.

8.24 IF reactor pressure 300 psig and injection not desired, THEN close RHR-MO-27B, OUTBD INJECTION VLV. Note to Examiner, Step is N/A.

8.25 Start RHR Pump B or D.

BOP 8.26 Throttle open RHR-MO-34B, SUPPR POOL COOLING INBD THROTTLE VLV, to obtain rated cooling flow or as directed by Control Room Supervisor.

8.27 Ensure RHR-MO-16B (min flow valve) closed.

8.28 Perform one of the following:

8.28.1 Close CM-38, LOOP B INJECTION LINE PRESSURE MAINTENANCE SHUTOFF (R-958-SW).

8.28.2 Maintain RHR Subsystem B pressure greater than Condensate Transfer System pressure to prevent filling Torus.

8.29 Throttle closed RHR-MO66B, HX BYPASS VLV, to obtain desired cooling rate.

8.30 IF PCIS Group 6 lights lit on Panel 9-5, at VBD-M, THEN ensure Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 15 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 2 Event

Description:

Place RHR Loop B in Suppression Pool Cooling (Procedure 2.2.69.3)

Time Position Applicants Action or Behavior REC-MO-711 or REC-MO-714, CRITICAL LOOP SUPPLY (associated with an in service REC HX), open.

8.31 F additional cooling required, THEN ensure RHR Pump A or C running in Suppression Cooling Mode, if available. Note to Examiner, Step is N/A.

Enter Technical Specification LCO 3.5.1 Condition A and declare RHR CRS LPCI Loop B inoperable. Required Action is to restore the subsystem to an operable status within 7 days.

BOP Respond to alarm 9-3-3/D-2, RHR B VALVE OVLD Declare RHR-MO-16B inoperable per LCO 3.6.2.3, Condition A.

Required Action is to restore it to operable status within 7 days.

Refer to TS LCO 3.6.1.9 Condition A.

CRS Required Action is to restore to operable status within 7 days.

Refer to TS LCO 3.5.1 Condition A.

Required Action is to restore to operable status within 7 days.

NOTE to Examiners: Crew may elect to secure Suppression Pool Cooling and remain in TS LCO 3.5.1 due to minimum flow valve inoperability.

END OF EVENT Notes Proceed to the next event when directed by the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 16 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 3 Event

Description:

Operating CRD pump trips Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 2 causing CRD Operator Pump B trip.

Respond to alarm 9-5-2/C-6, CRD PUMP B BREAKER TRIP ATC Report trip of CRD Pump B.

Direct starting standby CRD Pump.

CRS Assign CRDM temperatures to BOP as monitored parameter.

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 ATC A.

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.

Respond to alarm C-3/G-9 480V Bus 1G GROUND.

Direct investigation for identifying possible ground per Procedure 2.0.1, BOP Plant Operations Policy, Monitor CRDM temperatures on PMIS if directed.

END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 17 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 3 Event

Description:

Operating CRD pump trips Time Position Applicants Action or Behavior Proceed to the next event at direction of the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 18 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 4 Event

Description:

Reactor Recirc Pump A #1 seal failure and pump trip Time Position Applicants Action or Behavior Booth When directed by LeadExaminer, insert Trigger 3 to cause RR Pump A #1 Operator seal failure and RR Pump A trip in 4 minutes.

Recognizes and reports alarm:

9-4-3/A-3, Recirc Pump A Seal Trouble ATC Checks seal cavity pressures on panel 9-4. Diagnoses #1 seal failure based on seal cavity 2 pressure equalizing with seal cavity 1 pressure.

Refers to alarm procedure.

Enters 2.4RR.

Directs monitoring Recirc Pump A temperatures, seal parameters, drywell CRS pressure.

As necessary, directs reducing RR B speed to maintain seal cavity temperatures

<200°F IAW alarm card 9-4-3/E-4.

Lowers RR pump B speed, as directed, by selecting S on RR B speed controller ATC on panel 9-4 and rotating knob counter-clockwise to lower speed demand (not less than 20% demand).

Monitors pump seal temperatures on RR-TR-31, RX RECIRC PUMPS AND BOP MOTOR TEMPERATURE RECORDER on panel 9-21.

Reports seal cavity #1 temp >200°F.

Recognizes and reports alarm due to seal cavity temperature high:

ATC 9-4-3/E-4, RECIRC A/B PUMP MOTOR HI/LOW TEMP PMIS alarm (1854) RECIRC A/B PUMP SEAL CAVITY TEMP HIGH IAW alarm procedure 9-4-3/A-3 step 1.4 and 9-4-3/E-4, step 1.5:

CRS IF Annunciator 9-4-3/E-4, (1854) RECIRC A/B PUMP SEAL CAVITY TEMP HIGH, alarms for RR Pump A concurrently with seal cavity temperature 200F, THEN remove RR Pump A from service per Procedure 2.2.68.1 or 2.4RR.

Identifies A RRMG TRIP by observing the RRMG breakers and parameter ATC/BOP indications and alarms on Panel 9-4.

Directs the RO to monitor for instabilities, Updates the crew that this is an entry CRS condition into 2.4RR and assigns it to the RO.

ATC Updates the crew with Scram Actions of 2.4RR.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 19 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 4 Event

Description:

Reactor Recirc Pump A #1 seal failure and pump trip Time Position Applicants Action or Behavior 1.1 If both RR pumps are tripped and reactor power > 1% rated thermal.

SCRAM.

Enter Procedure 2.1.5.

1.2 If abnormal neutron flux oscillations are observed while operating in the Stability Exclusion Region:

SCRAM.

Enter Procedure 2.1.5.

Monitor various independent/redundant parameters and power indications for proper plant response during all power changes; utilize the list below (as a minimum) as dictated by plant conditions:

Reactor Water Level.

Reactor Steam Pressure and Flow.

CRS/ATC Reactor Power, APRMs, RBMs, IRMs, or SRMs, as required.

Reactor Recirc Speed, Jet Pump, and Loop Flows.

Total Core Flow and Core Support Plate DP.

Reactor Feed Pump Flow and Speed.

Main Generator Output (Gross and Net).

AOP 2.4RR Attachment 1:

1. If one RR pump trips, perform following:

NOTE 1 - Core flow may indicate higher than actual if an RR pump is tripped and reverse core flow summer is not operating; the following indicate summer is operating:

Annunciator 9-4-3/E-3 (9-4-3/E-7), RECIRC LOOP A (B) OUT OF SERVICE, alarming.

Indicated core flow is approximately equal to difference between NBI-FI-92A and NBI-FI-92B, JP LOOP FLOW.

BOP NOTE 2 - It takes ~ 1 minute from time pump has tripped for indicated core flow to stabilize.

1.1 If operation in Stability Exclusion Region, concurrently enter Attachment 3.

1.2 For tripped RR pump, ensure RRMG Set A(B) GEN FIELD BKR open.

1.3 For tripped RR pump, close RR-MO-53A(B), PUMP DISCHARGE VLV.

1.4 Continue with remaining steps in this attachment while waiting to open Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 20 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 4 Event

Description:

Reactor Recirc Pump A #1 seal failure and pump trip Time Position Applicants Action or Behavior RR-MO-53A(B).

1.5 After RR-MO-53A(B) has been closed for 5 minutes, open valve.

1.6 Ensure operating RRMG is transferred to Startup Transformer, if available, per Procedure 2.2.18.

1.7 Throttle REC-49(51), MG SET A(B) OIL HX OUTLET (R-931-NW), to maintain oil outlet temperature 90F to 130F on RRLO-TI-2626A(B), MG SET HX A(B) OUTLET TEMPERATURE (R-931-NW NEAR HXs), for tripped RRMG.

1.8 Monitor loop cooldown rate on RR-TR-165, RR SUCTION &

FEEDWATER TEMP. If loop cooldown rate exceeds 100F/hr, initiate Condition Report and evaluate excessive cooldown rate prior to restoration of system normal operation.

1.9 Enter Single Loop Operation per Procedure 2.2.68.1.

2. Dispatch Operators to R-976-W and Non-Critical Switchgear Room to record lockout relays and targets for tripped pump.

3. Align RRMG H&V System per Procedure 2.2.85.

Role Play: If dispatched to R-976-W and Non-Critical Switchgear Room to record lockout relays and targets for tripped pump., tell them that you are on your way. In a few minutes call back and tell them that you have OVERLOAD Booth GROUND targets.

Operator If called to monitor and adjust Reactor Recirc lube oil temperature, tell them that you are on your way and will call them when you are ready. In a few minutes call back and tell them that you are standing by.

Displays the Power to Flow map on the CRT and evaluates the location of ATC operation. It will take approximately 1 minute for the screen to update real time data.

Recognizes and reports operation NOT in the Stability Exclusion Region of the ATC Power-to-Flow Map.

Address Tech Specs and finds that with one RR Pump out of service, CRS LCO 3.4.1 Condition B Required Action B.1 Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 21 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 4 Event

Description:

Reactor Recirc Pump A #1 seal failure and pump trip Time Position Applicants Action or Behavior Satisfy the requirements of the LCO within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Notifies work control of the RR pump failure and need to contact Reactor CRS Engineering to insert GARDEL thermal limits for single loop operation.

Direct RRMG B power supply transferred to the Startup Station transformer.

If directed, transfer RRMG B drive motor supply to SSST. Per Procedure 2.2.18:

17.2 TRANSFERRING BUS 1D FROM NORMAL TRANSFORMER TO STARTUP TRANSFORMER 17.2.1 Ensure Startup Transformer energized.

CAUTION - Maximum load on Startup Transformer is 27 MWe.

NOTE 1 - Switch for Breaker 1DS must be held to START to allow time for synchro check Relay 25/1DN to pick up auxiliary Relay 25X/1DN which permits breaker to close.

NOTE 2 - Breaker 1DN will automatically trip when Breaker 1DS closes.

17.2.2 Close Breaker 1DS by placing and holding switch to START.

17.2.3 AFTER Breaker 1DS has closed, THEN release switch and check following:

ATC 17.2.3.1 Switch spring returns to NORMAL AFTER START (red flagged).

17.2.3.2 Breaker 1DN, DRIVE MOTOR BKR, has automatically tripped.

17.2.3.3 Amber tripped indicating light above Breaker 1DN switch turns on.

17.2.3.4 Annunciator 9-4-3/A-6, RRMG B BKR 1DN TRIP, alarms.

17.2.4 Place switch for Breaker 1DN to STOP and check following:

17.2.4.1 Switch spring returns to NORMAL AFTER STOP (green flagged).

17.2.4.2 Amber tripped indicating light above Breaker 1DN switch turns off.

17.2.4.3 Annunciator 9-4-3/A-6 clears.

Booth Role Play: As the work control center, respond to the report and let the CRS Operator know that a work order will be initiated and a team put together to investigate RR pump and Reactor Engineering will be informed of the need Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 22 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 4 Event

Description:

Reactor Recirc Pump A #1 seal failure and pump trip Time Position Applicants Action or Behavior to insert GARDEL thermal limits for single loop operation.

END OF EVENT Notes Booth Proceed to the next event at direction of the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 23 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 5 Event

Description:

Reactor Recirculation Pump A seal #2 failure Time Position Applicants Action or Behavior Booth When directed by the Lead Examiner, insert Trigger 4 causing RR Operator Pump A #2 seal to fail.

Recognizes and reports RR pump A seal cavity #2 pressure lowering, ATC diagnoses seal #2 failure.

Recognizes and reports drywell pressure rising.

CREW Directs ATC to perform prompt isolation of RR pump A IAW 2.4RR Att.

CRS 2.

Secures and isolates RR pump A IAW 2.4RR Att. 2:

1. IF both seals have failed and affected RR pump requires prompt isolation, THEN perform following:

1.1 Ensure DRIVE MOTOR BKR 1CN, 1CS is tripped.

1.2 Close RR-MO-43A, PUMP SUCTION VLV.

1.3 Close RR-MO-53A(B), PUMP DISCHARGE VLV.

1.4 Close CRD-50, REACTOR RECIRCULATION PUMP A ATC SEAL FLOW REGULATOR 46A INLET (R-903-SE).

1.5 Following steps may be performed concurrently:

1.5.1 Enter Single Loop Operation section of Procedure 2.2.68.1.

1.5.2 Ensure RRMG Set A GEN FIELD BKR open.

1.5.3 Ensure RRMG B is powered by Startup Transformer per Procedure 2.2.18.

Enters 2.4PC.

CRS 4.3 Directs BOP to vent containment per 2.2.60 to maintain 0.25 to 0.45 psig.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 24 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 5 Event

Description:

Reactor Recirculation Pump A seal #2 failure Time Position Applicants Action or Behavior Vent the drywell per the Hard Card (2.2.60, Att.1) at VBD-H:

1.1 Ensure PC-AD-R-1B is open and PC-AD-R-1A is closed.

1.2 Start preferred SGT fan on VBD-K.

1.3 Open SGT-DPCV-546A(B) valve on VBD-K.

NOTE - Steps 1.4 and 1.5 may be performed in any order or concurrently, depending on plant conditions.

1.4 Vent Torus by performing following at VBD-H:

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, THEN close PC-MO-305MV.

1.4.5 Close PC-AO-245AV.

BOP 1.4.6 Place switch for PC-AO-245AV to AUTO.

1.5 Vent Drywell by performing following at VBD-H:

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 1.5.3 WHEN Drywell pressure ~ 0.25 psig, THEN 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 at VBD-K.

1.7 Place switch for SGT-DPCV-546A(B) to AUTO at VBD-K.

Checks and reports drywell pressure lowering.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 25 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 5 Event

Description:

Reactor Recirculation Pump A seal #2 failure Time Position Applicants Action or Behavior NOTE to Examiners: If Drywell pressure exceeds 0.75 psig, TS 3.6.1.4 will apply. Due to scenario timing, the CRS may not refer to TS 3.6.1.4, so this may be asked as a follow-up question.

If Drywell pressure rises above 0.75 psig during this event, enters TS CRS LCO 3.6.1.4 Condition A.

END OF EVENT Notes Proceed to the next event at the direction of the Lead Examiner Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 26 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, activate trigger 5 (FW leak line Operator break inside primary containment and HPCI auto start failure).

Respond to the reactor scram:

Provides scram report:

ATC Reactor Power:______________

Reactor Water Level and controlling system:______________

Reactor Pressure and controlling system:_______________

Direct ATC to perform Procedure 2.1.5, Attachments 1, 2, and 3 and control RPV level +3 inches to +54 inches.

CRS Direct BOP to Perform Procedure 2.1.5, Attachments 4 and 5 and control RPV pressure between 800 psig and 1000 psig.

Performs Procedure 2.1.5, Attachment 3 Reactor Water Level Control, actions:

1.1 IF STARTUP FCVs in MAN, THEN perform following; N/A if FCVs in AUTO:

1.2 After FW Sequence has reached Mode 2 or level has stabilized, place RFC-SW-S1, SETPOINT SETDOWN, switch to DISABLE/RESET.

1.3 Maintain RPV level in prescribed band using following systems, as required, based on plant conditions:

1.3.1 Verify preferred RFP is controlling level in FW Sequence ATC Mode 2 with controlling RFP in RX PRESS FOLLOW Mode.

1.3.1.1 Note to Examiner, Step is N/A.

1.3.1.2 If EMER CLOSE button is yellow, press EMER CLOSE button on either FCV-11AA or FCV-11BB.

1.3.1.3 Ensure following controllers are in AUTO:

a FCVs 11AA and 11BB.

b STARTUP MASTER CONTROL.

1.3.1.4 IF RX PRESS FOLLOW is not desired or cannot be obtained, THEN maintain RPV level in prescribed band using following systems based on plant conditions: Note to Examiner, Step is N/A.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 27 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior 1.3.1.5 Adjust STARTUP MASTER controller using UP/DOWN arrows or RAMP FUNCTION to adjust LEVEL SETPOINT as desired.

a During plant cooldown, for further guidance, adjust Reactor Feedwater System/Condensate System per Procedure 2.2.28.1/2.2.6.

1.3.2 HPCI per Procedure 2.2.33.1.

1.3.3 RCIC per Procedure 2.2.67.1.

1.4 Trip non-preferred RFP, if not needed, or minimum flow is isolated.

1.5 Trip all but one condensate booster pump.

1.6 Trip all but one condensate pump.

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.

BOP 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 BOP 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 IF main turbine does not trip, THEN perform following (N/A if not Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 28 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior performed): Note to Examiner, Step is N/A.

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.

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).

Direct the actions of 2.4MC-RF, CONDENSATE AND FEEDWATER ABNORMAL Determine the B Feedwater line has failed and RCIC will need to be CRS used to restore Reactor Water Level.

Enter EOP 3A, Primary Containment Control and direct actions.

Direct Drywell FCUs placed in OVERRIDE.

NOTE to Examiners: RCIC injects into feedwater Line A.

HPCI injects into feedwater Line B. (Failure is in FW line B)

Places Drywell FCUs in OVERRIDE as directed.

Perform the actions of 2.4MC-RF, CONDENSATE AND FEEDWATER ABNORMAL as follows:

4.1 If system piping not intact, perform following:

4.4.1 If break is in Turbine Building, concurrently enter ATC Procedure 5.1BREAK.

4.4.2 If break is endangering personnel or equipment necessary for safe operation:

4.4.2.1 Concurrently enter Procedure 2.1.5.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 29 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior 4.4.2.2 Ensure RFPs tripped.

4.4.2.3 Ensure RFP discharge valves closed.

4.4.2.4 At a RFPT/RVLC HMI, perform following:

4.4.2.4.1 Select RFPT-1A or RFPT-1B System.

4.4.2.4.2 Select STARTUP VALVE screen.

4.4.2.4.3 Press EMER CLOSE button.

4.4.2.4.4 Confirm pop-up screen.

4.4.2.5 Ensure condensate booster pumps tripped.

4.4.2.6 If necessary, trip condensate pumps.

4.4.3 Notify Plant personnel to stay clear of affected area via Gaitronics.

CREW Determine leak in FW Line B.

Determine failure of HPCI to initiate.

Start HPCI using the Hard Card:

1.1 Start Gland Seal Condenser Blower.

1.2 If necessary, depress REACTOR HI WTR LEVEL SIGNAL RESET pushbutton.

1.3 IF AUXILIARY OIL PUMP running NOTE Step is N/A BOP 1.4 Open HPCI-MO-14.

1.5 Start AUXILIARY OIL PUMP.

1.6 Open HPCI-MO-19.

1.7 Adjust HPCI-FIC-108, HPCI Flow Controller, to maintain level.

1.8 IF HPCI is needed for RPV Pressure Control and RPV Injection is required, THEN perform following:

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 30 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior 1.8.1 Ensure HPCI Initiation signal is clear.

1.8.2 Open HPCI-MO-24.

1.8.3 Throttle HPCI-MO-21 to control HPCI pump discharge pressure.

1.8.4 Adjust HPCI-FIC-108, HPCI Flow Controller, to maintain turbine speed.

1.9 If available, ensure REC-MO-711 or REC-MO-714 is open.

1.10 If available, ensure SGT System is in service.

Report HPCI failure to CRS.

CREW Update Group 6 isolation.

Direct transfer of reactor level control from Condensate/Feedwater to CRS RCIC.

When RFPs/condensate booster pumps are tripped, use RCIC per BOP Procedure 2.2.67.1 to maintain RPV level.

CRS CRS Direct Torus Sprays placed into service.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 31 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior Direct maintaining containment pressure between 2 psig and 10 psig.

Role Play: When directed by BOP to install EOP PTMs97-100, wait Booth 3 minutes then put in the overrides for the PTMs. Report back to Operator BOP when PTMs installed.

2. CONTAINMENT SPRAYS 2.1 IF required, with CRS permission, THEN place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch BOP to MANUAL OVERRD.

2.2 IF required, THEN place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 32 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior 2.3 Ensure RHR-MO-39A(B) open.

2.4 IF reactor pressure 300 psig and injection not desired, THEN close RHR-MO-27A(B), OUTBD INJECTION VLV.

2.5 Ensure RHR PUMP(s) running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

2.6 Throttle RHR-MO-38A(B) to maintain desired containment pressure.

2.7 Throttle RHR-MO-66A(B) to obtain desired cooling rate.

BOP Report containment pressure trend.

(CT-1): Initiate drywell sprays when torus pressure exceeds 10 psig, prior to drywell temperature reaching 280F and prior to torus pressure exceeding the Pressure Suppression Pressure (PSP) curve.

CRS Direct Drywell Sprays placed into service.

Direct maintaining containment pressure between 2 psig and 10 psig.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 33 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior CRS 2.8 IF Drywell Spray required, THEN perform following:

BOP 2.8.1 Open RHR-MO-31A(B).

2.8.2 Throttle RHR-MO-26A(B) to maintain desired Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 34 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 6&7 Event

Description:

FW Line B Break Inside the Drywell & HPCI auto start failure Time Position Applicants Action or Behavior containment pressure.

2.9 IF PCIS Group 6 lights lit on Panel 9-5, THEN ensure one of following open:

2.9.1 REC-MO-711; or 2.9.2 REC-MO-714.

2.10 Place RHR SW System in service:

2.10.1 Start SWBP(s).

2.10.2 Adjust SW-MO-89A(B) to maintain flow between 2500 and 4000 gpm.

2.11 Throttle RHR-MO-66A(B) to maintain desired cooling rate.

Monitor and control torus water level between +2 in. and -2 in. (refer to SOP 2.2.69.3 prior to discharging water) (SP/L-1).

BOP Monitor average torus water temperature, EOP 5.8.9, and control below 95°F using available suppression pool cooling (SP/T-1).

END OF EVENT Notes Proceed to the next event at direction of the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 35 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 7 to cause loss of Operator RPV level instrumentation CREW Recognize RPV level instruments are diverging.

CRS May direct BOP to enter Procedure 2.4RXLVL.

If directed to enter 2.4RXLVL, goes to Attachment 3:

NOTE - Below guidance will aid in determining which level indicator(s) is reading accurately.

1. WHEN any wide range level indication at -30", THEN check status of following annunciators:

1.1 9-5-2/D-7, ATWS RPT CHAN A/B LEVEL TRIP.

1.2 9-3-2/A-5, RX LOW WATER LEVEL -42".

1.3 If both annunciators clear, actual level > -33".

1.4 IF both annunciators alarming, THEN check following alarm points in alarm:

1.4.1 (3074) ATWS RPT CHAN A LEVEL TRIP.

1.4.2 (3075) ATWS RPT CHAN B LEVEL TRIP.

1.4.3 (1700) RX LOW WATER LEVEL -42" ALARM (NBI-LIS-BOP 72A).

1.4.4 (1701) RX LOW WATER LEVEL -42" ALARM (NBI-LIS-72B).

1.4.5 (1702) RX LOW WATER LEVEL -42" ALARM (NBI-LIS-72C).

1.4.6 (1703) RX LOW WATER LEVEL -42" ALARM (NBI-LIS-72D).

1.4.7 If all alarm points in alarm, actual level < -33".

2. WHEN any wide range level indication at -100", THEN check status of following annunciators:

NOTE - RX low water level alarms on Panel 9-5 and associated annunciator points alarm when the respective RPS Bus is lost.

2.1 9-5-1/B-1, RX LOW LEVEL CHANNEL A -113".

2.2 9-5-1/B-2, RX LOW LEVEL CHANNEL B -113".

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 36 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior 2.3 9-3-2/B-5, RX LOW WATER LEVEL -113".

2.4 If all three annunciators clear, actual level > -104".

2.5 IF all annunciators alarming, THEN check following alarm points in alarm:

2.5.1 (2102) RX LOW LEVEL -113" CHANNEL A2 TRIP.

2.5.2 (2103) RX LOW LEVEL -113" CHANNEL A1 TRIP.

2.5.3 (2114) RX LOW LEVEL -113" CHANNEL B2 TRIP.

2.5.4 (2115) RX LOW LEVEL -113" CHANNEL B1 TRIP.

2.5.5 (1704) RX LOW WATER LEVEL -113" ALARM (NBI-LIS-72A).

2.5.6 (1705) RX LOW WATER LEVEL -113" ALARM (NBI-LIS-72B).

2.5.7 (1706) RX LOW WATER LEVEL -113" ALARM (NBI-LIS-72C).

2.5.8 (1707) RX LOW WATER LEVEL -113" ALARM (NBI-LIS-72D).

2.5.9 If all alarm points in alarm, actual level < -104".

BOP Report to CRS RPV level cannot be determined.

BOP/ATC Send Reactor Building operator to check local racks.

Role Play- If directed as a building operator to verify level indications on racks 25-5 and 25-6, wait about 3 minutes and then report as follows:

Rack 25-5 All indicators pegged high.

Rack 25-6 NBI-LIS-58A-Downscale NBI-LIS-58B-Downscale Booth Operator NBI-LIS-72B-Upscale NBI-LIS-72D-(-50 inches)

NBI-LIS-101C-Downscale NBI-LIS-101D-Downscale Rack 25-51 NBI-LITS-73A indicators pegged high.

Rack 25-52 NBI-LITS-73B pegged high.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 37 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior Enter EOP 2B CRS Per Procedure 5.8.19:

BOP 4.2 Verify that Annunciator 9-3-2/C-5, RX LOW PRESS 291-436 PSIG, is alarming prior to continuing with this section.

Request current PC water level:

CRS (CT-2): When RPV level cannot be determined and torus level is above 6', open six SRVs before drywell radiation reaches 150 R/hr or entering PC Flooding.

When directed to ED, verify PC water level is above 6 ft. and open 6 ATC/BOP SRVS by taking their control switches to OPEN.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 38 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior CRS (CT-3): When RPV level cannot be determined and the reactor has been depressurized below the shutoff head of the respective pump(s), inject into the RPV to flood to the Main Steam Lines before drywell radiation reaches 150 R/hr or entering PC Flooding Flood the RPV per Procedure 5.8.6

4. RPV FLOODING WITH CORE SPRAY LOOP A NOTE - EOP Flowcharts authorize exceeding CS System NPSH and vortex limits, if necessary, for RPV flooding.

4.1 Perform following (PNL 9-3):

BOP 4.1.1 Ensure CS-MO-5A, MIN FLOW BYP VLV, is open.

4.1.2 Ensure CS-MO-11A, OUTBD INJECTION VLV, is open.

4.1.3 Ensure CS PUMP A is running.

4.1.3.1 Ensure CS-MO-5A, MIN FLOW BYP VLV, Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 39 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior remains open.

4.1.4 WHEN Annunciator 9-3-2/C-5, RX LOW PRESS 291-436 PSIG, alarms, THEN perform following:

4.1.4.1 Throttle open CS-MO-12A, INBD INJ THROTTLE VLV, as directed by EOPs.

4.1.4.2 Ensure CS-MO-5A, MIN FLOW BYP VLV, closes at > 1370 gpm system flow.

4.2 Inform CRS that CS Loop A is injecting as a RPV Flooding System.

5. RPV FLOODING WITH CORE SPRAY LOOP B NOTE - EOP Flowcharts authorize exceeding CS System NPSH and vortex limits, if necessary, for RPV flooding.

5.1 Perform following (PNL 9-3):

5.1.1 Ensure CS-MO-5B, MIN FLOW BYP VLV, is open.

5.1.2 Ensure CS-MO-11B, OUTBD INJECTION VLV, is open.

5.1.3 Ensure CS PUMP B is running.

5.1.3.1 Ensure CS-MO-5B, MIN FLOW BYP VLV, remains open.

5.1.4 WHEN Annunciator 9-3-2/C-5, RX LOW PRESS 291-436 PSIG, alarms, THEN perform following:

5.1.4.1 Throttle open CS-MO-12B, INBD INJ THROTTLE VLV, as directed by EOPs.

5.1.4.2 Ensure CS-MO-5B, MIN FLOW BYP VLV, closes at > 1370 gpm system flow.

5.2 Inform CRS that CS Loop B is injecting as a RPV Flooding System.

6. RPV FLOODING WITH LPCI LOOP A WITH HX IN SERVICE NOTE 1 - EOP Flowcharts authorize exceeding RHR System NPSH and Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 40 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior vortex limits, if necessary, for RPV flooding.

NOTE 2 - Step 6.1 may be performed concurrently with Steps 6.2 and 6.3.

6.1 Defeat automatic open interlocks for RHR-MO-27A, OUTBD INJECTION VLV, by performing following:

6.1.1 Install EOP PTM # 97, lift Wire RH197-13 from GG-84 (BAY-2, PNL 9-32).

6.1.1.1 EOP PTM # 97 installed (wire lifted).

6.1.2 Install EOP PTM # 98, jumper between Terminals GG-85 and GG-86 (BAY-2, PNL 9-32).

6.1.2.1 EOP PTM # 98 installed (jumper installed).

6.1.3 Inform CRS that automatic open interlocks for RHR-MO-27A are bypassed for EOPs.

6.2 Ensure RR Pump A secured (PNL 9-4).

6.3 Ensure RR-MO-53A, PUMP DISCHARGE VLV (PNL 9-4),

closed.

6.4 Perform following at (PNL 9-3):

6.4.1 Close RHR-MO-27A, OUTBD INJECTION VLV.

6.4.2 Open RHR-MO-25A, INBD INJECTION VLV.

6.4.3 Ensure RHR-MO-16A, LOOP A MIN FLOW BYP VLV, is open.

NOTE - Unless plant conditions warrant, RHR pump operation at minimum flow should be limited to < 15 minutes or pump damage may result.

6.4.4 Ensure one or both RHR pump(s) running.

6.4.4.1 RHR Pump A.

6.4.4.2 RHR Pump C.

6.4.5 WHEN Annunciator 9-3-2/C-5, RX LOW PRESS 291-436 PSIG, alarms, THEN perform following:

6.4.5.1 Throttle open RHR-MO-27A, OUTBD INJECTION VLV, as directed by EOPs.

6.4.5.2 Ensure RHR-MO-16A, LOOP A MIN FLOW Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 41 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior BYP VLV, closes when RHR flow raises to >

2107 gpm.

CAUTION - Maximum SWBP amps are 136 amps (PNL 9-3).

6.5 Start SW BOOSTER PUMP A or SW BOOSTER PUMP C (PNL 9-3).

6.6 Throttle SW-MO-89A, HX-A SW DISCH VLV, until a flow rate of 4000 gpm is seen on SW-FI-132A, SW FLOW (PNL 9-3).

6.7 WHEN conditions allow, THEN close RHR-MO-66A, HX BYPASS VLV (PNL 9-3).

6.8 Inform CRS that RHR Loop A is injecting for RPV flooding.

7. RPV FLOODING WITH LPCI LOOP B WITH HX IN SERVICE NOTE 1 - EOP Flowcharts authorize exceeding RHR System NPSH and vortex limits, if necessary, for RPV flooding.

NOTE 2 - Step 7.1 may be performed concurrently with Steps 7.2 and 7.3.

7.1 Defeat automatic open interlocks for RHR-MO-27B, OUTBD INJECTION VLV, by performing following:

7.1.1 Install EOP PTM # 99, lift Wire RH22-13 from GG-84 (BAY-2, PNL 9-33).

6.1.1.1 EOP PTM # 99 installed (wire lifted).

7.1.2 Install EOP PTM # 100, jumper between Terminals GG-85 and GG-86 (BAY-2, PNL 9-33).

7.1.2.1 EOP PTM # 100 installed (jumper installed).

7.1.3 Inform CRS that automatic open interlocks for RHR-MO-27B are bypassed for EOPs.

7.2 Ensure RR Pump B secured (PNL 9-4).

7.3 Ensure RR-MO-53B, PUMP DISCHARGE VLV (PNL 9-4),

closed.

7.4 Perform following at (PNL 9-3):

7.4.1 Close RHR-MO-27B, OUTBD INJECTION VLV.

7.4.2 Open RHR-MO-2BA, INBD INJECTION VLV.

7.4.3 Ensure RHR-MO-16B, LOOP B MIN FLOW BYP VLV, is open.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 42 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior NOTE - Unless plant conditions warrant, RHR pump operation at minimum flow should be limited to < 15 minutes or pump damage may result.

7.4.4 Ensure one or both RHR pump(s) running.

7.4.4.1 RHR Pump B.

7.4.4.2 RHR Pump D.

7.4.5 WHEN Annunciator 9-3-2/C-5, RX LOW PRESS 291-436 PSIG, alarms, THEN perform following:

7.4.5.1 Throttle open RHR-MO-27B, OUTBD INJECTION VLV, as directed by EOPs.

7.4.5.2 Ensure RHR-MO-16B, LOOP B MIN FLOW BYP VLV, closes when RHR flow raises to >

2107 gpm.

CAUTION - Maximum SWBP amps are 136 amps (PNL 9-3).

7.5 Start SW BOOSTER PUMP B or SW BOOSTER PUMP D (PNL 9-3).

7.6 Throttle SW-MO-89B, HX-B SW DISCH VLV, until a flow rate of 4000 gpm is seen on SW-FI-132B, SW FLOW (PNL 9-3).

6.7 WHEN conditions allow, THEN close RHR-MO-66B, HX BYPASS VLV (PNL 9-3).

6.8 Inform CRS that RHR Loop B is injecting for RPV flooding.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 43 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior Role Play: If sent into reactor building to check for flooding Booth conditions, depending on where sent, report there is water leakage Operator from the HPCI / RCIC turbine shaft seals.

Monitor SRV tailpipe temperatures on MS-TR-166 on Panel 9-21.

Monitor torus water level on Panel 9-4-1; level lowers as RPV and ATC/BOP steam lines flood and stabilizes when steam lines are full.

Monitor RPV pressure lowering and rising when the MSLs are full CREW Report to CRS when RPV flooded to the Main Steam Lines.

(CT-4): When RPV level cannot be determined and at least 1 SRV is manually opened, isolate MSIVs, MSL drains, HPCI steam supply, and RCIC steam supply, prior to RPV water level rising to the bottom of the main steam lines.

CRS Direct steam lines connect to the RPV isolated.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 4 Page 44 of 42 Op-Test No.: CNS 15-01 Scenario No.: 4 Event No.: 8 Event

Description:

Loss of RPV level instrumentation, RPV flooding to MSLs Time Position Applicants Action or Behavior As directed by the CRS, Close:

MSIVs ATC/BOP Main Steam line drain valves.

HPCI isolation valves RCIC isolation valves.

ATC/BOP Control RPV injection as low as practicable.

END OF SCENARIO Notes Booth When directed by the Lead Examiner, place the simulator in freeze Operator and tell the crew to stop operating.

Rev. 0

INITIAL CONDITIONS A. Plant Status:

1. 100% power, near End of Cycle.

2. Rod Sequence Information: RO to provide B. Tech. Spec. Limitations in effect:

1. None C. Significant problems/abnormalities:

1. None D. PRA Risk:

Green E. Evolutions/maintenance for the on-coming shift:

1. After taking the watch the ATC is to shift B RRMG Oil Pumps B1 and B3 per Procedure 2.2.68.1, Section 20. The oil pump shift is to allow the oil pump to be tagged out later in the shift.

2. After RRMG oil pump shift, the BOP is to complete placing RHR Loop B in Suppression Pool Cooling in service per Procedure 2.2.69.3 which is complete to Step 8.19.

3. A hot spot flush is NOT required for RHR Loop B.

4. A HPCI run is planned for the next shift.

Rev. 0

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 1 of 33 Facility: Cooper Nuclear Station Scenario No.: 5 Op-Test No.: CNS 15-01 Examiners: ____________________________ Operators: _____________________________

Objectives: To evaluate the candidates ability to operate the facility in response to the following evolutions:

1. Withdraw control rods to establish 20-25% bypass valve position.

2. Respond to a control rod drifting out.

3. Respond to APRM A INOP failure.

4. Respond to a spurious RCIC initiation.

5. Respond to a HPCI steam line break and failure to isolate.

6. Respond to a LOOP.

7. Respond to failure of DG 2 to automatically start.

8. Respond to a LOCA.

9. Respond to RHR Loop A and Core Spray A pumps failure to automatically start.

10. Emergency Depressurize on low RPV level.

Initial Conditions: Plant operating at 5% power and in RUN.

Inoperable Equipment: None Turnover:

The plant is in RUN at 5% power.

Planned activities for this shift are:

Continue the startup.

Scenario Notes:

This is a new scenario.

Validation Time: 75 minutes

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 2 of 33 Event Malf. No. Event Type Event No. Description 1 N/A R (ATC,CRS) Withdraw control rods to establish 20-25% bypass valve position.

C (ATC,CRS) 2 rd10 (22-23) A (CREW) Control rod 22-23 drifts out.

TS (CRS) 3 nm14a I (ATC,CRS) APRM A INOP failure (half scram)

C (BOP,CRS) 4 rc05 A (CREW) Spurious RCIC initiation TS (CRS) hp06 I (BOP,CRS) 5 HPCI steam line break and failure to isolate hp09 TS (CRS) ed05 6 M (CREW) Loss of off-site power.

ed06 DG 2 Fails to automatically start.

C (BOP,CRS) (CT-1) When high pressure injection systems cannot maintain RPV level and low pressure ECCS 7 dg06b A systems fail to automatically start due to loss of AC power, crew manually starts DG 2 to energize LP (CREW) ECCS systems prior to RPV water level falling below -158 CFZ (TAF).**

LOCA, ED on low RPV level.

(CT-2) When RPV level lowers to -158 CFZ (TAF) rr20a M and cannot be maintained above -183 CFZ 8 (MSCWL) and insufficient high pressure injection (CREW) systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.)

Div 1 low pressure ECCS pumps fail to automatically start rh08a C (CT-3) When high pressure injection systems 9 rh08c cannot maintain RPV level and low pressure ECCS (BOP,CRS) pumps fail to automatically start, crew manually cs06a starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -

158 CFZ (TAF).**

(N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor, (A)bnormal (TS) Tech Spec

Critical Task (As defined in NUREG 1021 Appendix D)

- Performing either CT-1 or CT-3 allows the RPV to be flooded. Only one or the other is required.

CREW notation for Abnormal (A) and Major (M) events denotes ATC, BOP, and CRS are credited.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 3 of 33 Quantitative Attributes Table ES-304-1 Attribute Actual Description Target DG 2 fails to automatically start on LOOP.

Malfunctions after EOP entry 1-2 2 RHR Loop A and Core Spray A pumps fail to auto start post LOCA..

Control Rod drifts outward.

Abnormal Events 2-4 2 RCIC spurious initiation.

LOOP Major Transients 1-2 2 LOCA EOP-1A EOP entries requiring substantive action 1-2 2 EOP-3A EOP contingencies requiring substantive 0-2 1 EOP-2A action (CT-1) When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually starts DG 2 to energize LP ECCS systems prior to RPV water level falling below -158 CFZ (TAF)

(CT-2) When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ EOP based Critical (MSCWL) and insufficient high pressure 2-3 3 injection systems are available to restore level, Tasks crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.)

(CT-3) When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -158 CFZ (TAF).

Normal Events N/A 0 N/A Reactivity Manipulations N/A 1 Withdraw control rods Control rod drifts out APRM failure Spurious RCIC initiation Instrument/

Component Failures N/A 6 HPCI steam line break and failure to isolate.

Diesel Generator 2 fails to automatically start.

RHR Loop A and Core Spray A pumps fail to automatically start.

Control rod drifts out APRM failure Spurious RCIC initiation Total Malfunctions N/A 6 HPCI steam line break and failure to isolate.

Diesel Generator 2 fails to automatically start RHR Loop A and Core Spray A pumps fail to automatically start.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 4 of 33 Top 10 systems and operator actions important to risk that are tested:

Emergency AC power/DGs (Event 7)

HPCI (Event 5)

ADS/SRV (Event 8)

Residual Heat Removal in LPCI injection MODE (Event 9)

Operator fails to depressurize with SRVs (Event 8)

Operator fails to initiate ADS and initiate ECCS early (Event 8)

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 5 of 33 SCENARIO

SUMMARY

The plant is at approximately 5% power during startup in MODE 1.

Event 1 After the crew takes the watch, the ATC withdraws control rods to raise power to establish 20-25% bypass valve position.

Event 2 (Automatically Triggered when Control Rod 22-23 is withdrawn)

While withdrawing control rods, control rod 22-23 begins drifting out. The control rod is fully inserted per abnormal procedure 2.4CRD. The CRS declares the control rod inoperable per TS LCO 3.1.3, Condition C.

Event 3 (Triggered by Lead Examiner)

After the TS call is made for the inoperable control rod, APRM A INOP/TRIP occurs. The crew determines it is an instrument failure and bypasses the APRM. The CRS determines it is a potential LCO and no TS LCO entry is required.

Event 4 (Triggered by Lead Examiner)

After the APRM failure is addressed, RCIC spuriously initiates. The crew enters 2.4CSCS and the BOP trips RCIC. RCIC-MO-131 control switch is overridden OPEN so RCIC cannot be reset and used for injection after the LOCA. The CRS enters TS LCO 3.5.3, Condition A and declares the RCIC inoperable.

Event 5 (Triggered by Lead Examiner)

Once TS are addressed for RCIC, a steam line break in HPCI occurs and HPCI fails to isolate. The BOP manually isolates HPCI from the main control room. The CRS enters TS LCO 3.5.1 Condition C and determines RCIC not operable. The CRS then enters TS LCO 3.5.1 Condition G.

Event 6 (Triggered by Lead Examiner)

After TS are addressed for HPCI, a LOOP occurs. The reactor scrams. Only Diesel Generator 1 connects to its bus. Diesel Generator 2 fails to start (Event 7) and must be manually started so it can automatically load onto its respective bus (CT-1).

Event 8 (Automatically Triggered)

A LOCA occurs two minutes after the LOOP. RHR A and C pumps and Core Spray A pump fail to automatically start and must be manually started (CT-2).

The CRS enters EOP 1A to control RPV parameters and EOP 3A to control PC parameters. The torus and drywell are sprayed to control containment pressure and temperature. RPV level lowers to TAF requiring the crew to emergency depressurize (CT-3). The exercise ends when emergency depressurization is complete and RPV level is being restored.

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 6 of 33 Critical Tasks (CT-1) When high pressure injection (CT-2) When RPV level lowers to -158 CFZ (TAF) systems cannot maintain RPV level and and cannot be maintained above -183 CFZ low pressure ECCS systems fail to (MSCWL) and insufficient high pressure injection automatically start due to loss of AC systems are available to restore level, crew power, crew manually starts DG 2 to begins to Emergency Depressurize by opening energize LP ECCS systems prior to RPV the first of six SRVs before RPV level lowers water level falling below -158 CFZ (TAF). below -183 CFZ.

EVENT 7 8 Safety Failure to recognize the auto start not The MSCWL is the lowest RPV water level at which significance occurring and energizing of the safety bus, the covered portion of the reactor core will generate and failure to take manual action per sufficient steam to preclude any clad temperature in Procedure 5.3EMPWR will result in the uncovered portion of the core from exceeding unavailability of safety-related equipment 1500F. When water level decreases below MSCWL necessary to provide adequate core cooling, with injection, clad temperatures may exceed 1500F.

otherwise resulting in core damage and a large offsite release.

Cueing Indication and/or annunciation that all ac Corrected Fuel Zone indication (SPDS) falls to -158 emergency buses are de-energized and lowering trend continues, and, before -158 CFZ

Bus energized lamps extinguished is reached, initial conditions, field reports, and control

Circuit breaker position room indications convey that adequate high pressure

Bus voltage injection cannot be restored before level falls below -

EDG status 183 CFZ.

Control room lighting dimmed Performance Manipulation of controls as required to Manipulation of any six SRV controls on panel 9-3:

indicator energize Div 1(2) AC emergency bus from SRV-71A panel C: SRV-71B Operator places DIESEL GEN 1(2) BKR SRV-71E EG1(2) to CLOSE on panel C. SRV-71G SRV-71H SRV-71C SRV-71D SRV-71F Performance Crew will observe light indication for Crew will observe SRV light indication go from green feedback equipment powered by Division 1(2) AC to red, amber pressure switch lights illuminate, illuminate on panel 9-3 and bus voltage reactor pressure lowering on SPDS and panel 9-3

~4200V on panel C and 9-5 meters and recorders, and SRV tailpipe temperatures rise on recorder MS-TR-166.

Justification Attempting to start ECCS systems must be There is no time limit for effecting complete for the chosen performed to determine their availability by depressurization. The MSCWL (-183 CFZ) is the performance the time TAF is reached in order to properly lowest RPV water level at which the covered portion limit implement EOP-1A decision steps regarding of the reactor core will generate sufficient steam to restoring and maintaining RPV level. preclude any clad temperature in the uncovered portion of the core from exceeding 1500F.

Emergency depressurization is allowed when level goes below TAF (-158 CFZ) and should be performed, if in the judgment of the CRS, level cannot be maintained above -183 CFZ. Since it is intended for the scenario supporting this CT to, early in the event, clearly indicate no high pressure injection systems can be made available to reverse the lowering level trend, the crew will have time to communicate and open 6 SRVs before -183 CFZ.

BWR Owners App. B, Contingency#1 App. B, Contingency#1 Group Appendix

Appendix D Scenario Outline Form ES-D-1 NRC CNS 15-01 Scenario 5 Page 7 of 33 Critical Tasks (CT-3) When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -158 CFZ (TAF).

EVENT 9 Safety Failure to recognize the auto start not occurring, significance and failure to take manual action per Conduct of Ops will result in unavailability of safety-related equipment necessary to provide adequate core cooling, otherwise resulting in core damage and a large offsite release.

Cueing Indication ECCS pumps are not running with initiation conditions present:

Green light on and Red lamp extinguished at respective pump handswitch on panel 9-3 Indication of Drywell Pressure 1.83 psig Indication of RPV water level -113 Performance Manipulation of controls as required to start the indicator affected ECCS pump(s) from panel 9-3:

Operator places affected ECCS pump(s) control switch(es) to START on panel 9-3 Performance Crew will observe Red light illuminate and Green feedback light extinguish for the affected ECCS pump(s) on panel 9-3 Justification Attempting to start ECCS systems must be for the chosen performed to determine their availability by the performance time TAF is reached in order to properly limit implement EOP-1A decision steps regarding restoring and maintaining RPV level.

BWR Owners App. B, Contingency#1 Group Appendix

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 8 of 33 SIMULATOR SET-UP A. Materials Required None B. Initialize the simulator in IC 156, 5% power (BOL)

Batch File Name -

C. Change the simulator conditions as follows:

1. Auto Triggers Number File Name/Variable Description 14 "zlordlta4b[81]==1" 22-23 scram light lit on core display dmf rd102223 24 "zlordlta4b[81]==1" 22-23 scram light lit on core display dmf rd102223

2. Malfunctions Number Title Trigger TD Severity Ramp Initial rd102223 control rod 22-23 drift out A N/A N/A N/A N/A nm14a APRM A INOP 2 N/A N/A N/A N/A rc05 Inadvertent RCIC initiation 3 N/A N/A N/A N/A hp06 HPCI steam line break 4 N/A 3 N/A N/A hp09 HPCI auto isolation failure A N/A N/A N/A N/A rr20a Coolant leak inside PC 5 2:00 25 N/A N/A ed05 Loss of Startup Transformer 5 N/A N/A N/A N/A ed06 Loss of Emergency Transformer 5 N/A N/A N/A N/A dg06b DG #2 fails to auto start A N/A N/A N/A N/A rh08a RHR Pump A failure to auto start A N/A N/A N/A N/A rh08c RHR Pump C failure to auto start A N/A N/A N/A N/A cs06a CS Pump A failure to auto start A N/A N/A N/A N/A rc08 RCIC spurious isolation 5 2:00 N/A N/A N/A Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 9 of 33 rd06 CRD drive water filter plugging 5 100 N/A N/A

3. Remotes Number Title Trigger Value TD Ramp rc13 RCIC-MO-27 breaker to off 13 DE-ENER N/A N/A Ia17 Air Compressor A breaker control 6 RESET N/A N/A sw15 Air Compressor B cooling to REC 7 REC N/A N/A

4. Overrides Instrument Tag Trigger TD Value Ramp RCIC-MO-131 C/S zdircicsws3(2) 3 5 min OPEN N/A HPCI-MO-15 C/S zdihpcisws2[1] 5 N/A CLOSE N/A HPCI-MO-16 C/S Zdihpcisws1[1] 5 N/A CLOSE N/A D. Panel Setup

1. Ensure PMIS IDTs are blank.

2. Ensure RR Controllers are selected to P.

3. Have Marked-up copy of Procedures 2.1.1, 2.2.28.1 and 2.2.77 ready to continue startup. (For Procedure 2.1.1: Section 4, Step 4.18 is in progress. Steps 4.19, 4,.20, 4.21, 4.22 are complete).

4. Mark up BOL STARTUP rod package to Rod 46-31 at Position 12 (Step 24).

5. Ensure Reactor Mode Switch in RUN.

6. Ensure all APRM recorders selected.

7. Ensure all IRMs withdrawn.

8. Ensure Rod Drift status light and alarm on 9-5 are clear.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 10 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 1 Event

Description:

Raise power to establish 20-25% bypass valve position during startup by withdrawing control rods.

Time Position Applicants Action or Behavior NOTE to Examiners: The beginning of shift brief and reactivity brief should have been conducted during turnover outside of the simulator.

Non-EOP Control Rod Movement Protocol per 2.0.3, Conduct of Operations, is attached for reference.

Withdraws control rods IAW sequence 30-SU-BOL using notch withdrawal:

Selects in-sequence control rod by depressing respective Rod Select push button on panel 9-5 Rod Select matrix.

Observes OUT PERMIT light ON on panel 9-5.

Withdraws control rod one notch at a time by placing rod movement control switch on panel 9-5 to OUT NOTCH, then releasing.

Observes ROD IN, ROD OUT, and ROD SETTLE lights illuminate in ATC sequence and selected rod position indicates control rod withdrew one notch on panel 9-5.

Observes IRM recorder indication on panel 9-5 is consistent with control rod withdrawal.

Repeats notch withdrawal until control rod is at target position specified in sequence 29-SU-BOL, then initials step in 29-SU-BOL.

Repeats process for each successive control rod until bypass valve position indicates 20-25%.

NOTE to Examiners: Control rod 22-23 is the 7th control rod to be withdrawn. Proceed to next event when control rod 22-23 is selected.

Provides peer check for control rod selection, movement direction, and step BOP completion and initials as verifier in 29-SU-BOL.

Monitors bypass valve position on HMI, panel B.

END OF EVENT Notes Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 11 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 1 Event

Description:

Raise power to establish 20-25% bypass valve position during startup by withdrawing control rods.

Time Position Applicants Action or Behavior Booth Proceed to the next event when control rod 22-23 is selected.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 12 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 2 Event

Description:

Control Rod 22-23 drift outward Time Position Applicants Action or Behavior NOTE to Examiners: Control Rod 22-23 drift is active and will drift after the rod is moved.

Recognizes and reports alarm:

ATC 9-5-1/C-4 Rod Drift Checks full core display and determines 22-23 is drifting outward.

Enters 2.4CRD:

CRS Directs ATC to perform 2.4CRD Attachment 1 for single rod drifting out.

NOTE to Examiners: 2.4CRD Attachment 1 flow chart is provided for reference, attached.

Performs 2.4CRD Attachment 1:

Ensures rod 22-23 is selected DO-1 Inserts rod using ROD MOVEMENT CONTROL or EMERGENCY IN control switch on panel 9-5.

DO-2 Determines the rod inserts.

DO-4 Determines the rod did not latch.

ATC DO-5 Applies and maintains continuous insert signal.

DO-7 Establishes communication with BOP at panel 9-16 (scram test panel).

DO-8 Notify BOP, then release continuous insert signal.

DO-9 Notify BOP rod did not latch.

DO-10 Notify BOP to scram 22-23.

DO-11 Notifies BOP to return 22-23 scram test switch to normal.

DO-12 Notifies BOP and CRS rod 22-23 latched following scram.

Booth Ensure rod drift malfunction rd102223 auto deletes when the BOP scrams Operator rod 22-23; otherwise, manually delete mf rd102223.

Assists ATC perform 2.4CRD Attachment 1:

BOP Places scram test switch for rod 22-23 to TEST when directed.

Returns scram test switch for rod 22-23 to NORM when directed.

Role Play: IF sent to CRD HCU 22-23 to investigate, as the building Booth operator wait 5 minutes, then report you do not detect anything unusual Operator at the HCU.

ATC IAW 2.4CRD, Attachment 1, notifies Reactor Engineering and System Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 13 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 2 Event

Description:

Control Rod 22-23 drift outward Time Position Applicants Action or Behavior Engineering rod 22-23 drifted out, would not latch, and has been scrammed in.

NOTE to Examiners: The CRS should suspend power ascension and confer with Reactor Engineering for control rod movement recommendations related to their assessment of the rod drift on rod sequence.

Role Play: If asked to assess the 22-23 rod drift, as Reactor Engineering Booth state that a subsequent rod drift may be considered a single rod drift, Operator since 22-23 is fully inserted.

NOTE to Examiners: If control rod 22-23 does not settle at 00 after it is scrammed, the CRS may enter TS 3.1.3 Condition A instead of Condition C. This is acceptable.

Refers to TS for control rod 22-23 drifting out.

Enters TS 3.1.3 Condition C for control rod 22-23. Required actions are to Fully insert control rod 22-23 with a Completion Time of 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months AND Disarm the associated CRD with a Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

CRS May enter TS 3.1.5 Condition A for HCU accumulator alarm that came in during the rod scram. Required Actions are Declare the associated control rod scram time slow with a Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months OR Declare the associated control rod inoperable with a Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

END OF EVENT Notes Booth Proceed to the next event at direction of the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 14 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 3 Event

Description:

APRM A INOP failure.

Time Position Applicants Action or Behavior Booth When directed by the Lead Examiner, insert Trigger 2, to cause APRM Operator A INOP/TRIP to occur.

Respond to a half scram and alarm 9-5-1/A-7, APRM RPS CH A UPSCALE TRIP OR INOP 2.1 IF reactor has scrammed, THEN refer to Procedure 2.1.5.

2.2 Determine following:

2.2.1 Which APRM tripped.

2.2.2 Cause of trip from remote lights on Panel 9-5.

2.3 IF only one APRM inop, THEN perform following:

ATC 2.3.1 Bypass affected channel.

2.3.2 Reset half scram per Procedure 2.1.5.

2.4 IF APRM readings exceed or approach setpoint, THEN reduce power immediately. NOTE to Examiners. Step is N/A.

2.5 IF only one APRM is upscale and all others are normal, THEN perform following: NOTE to Examiners. Step is N/A.

2.5.1 Bypass affected channel.

2.5.2 Reset half scram per Procedure 2.1.5.

Direct APRM A bypassed.

Refer to TS LCO 3.3.1.1 Condition A. Determine from Table 3.3.1.1-1 that the required number of channels (2 per Trip System) is met and the LCO is a potential LCO.

CRS Refers to TRM TLCO 3.3.1 Condition A for Rod Block Instrumentation.

Determined the required number of channels (4 total) is met and TLCO is a potential.

Refers to TRM TLCO 3.3.3 Condition B for Neutron Monitoring Function 7C..

Determined the required number of channels (1 total) is met and TLCO is a potential.

END OF EVENT Notes Booth Proceed to the next event at direction of the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 15 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 4 Event

Description:

Spurious RCIC initiation Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 3 to cause a spurious Operator RCIC initiation Responds to and reports alarm, 9-4-1/A-1 RCIC LOGIC ACTUATED Verifies by independent indication RCIC initiation is not valid (RPV water level above -42):

BOP Absence of other low water level alarms/conditions (RPS, HPCI, PCIS)

Panel 9-3, 9-4, 9-5 Wide Range and Narrow Range level meters/recorders >-42 Refers to alarm procedure.

Checks panel 9-5 RPV water level indications to confirm misoperation of RCIC.

ATC Monitors Narrow Range level and Feedwater for proper response to RCIC injection.

Enters 2.4CSCS CRS Directs BOP to secure RCIC.

Secures RCIC IAW 2.4CSCS, Attachment 4, or ARP 9-4-1/A-1:

2.1 IF initiation not valid, THEN perform following:

2.1.1 Press and hold TURBINE TRIP button until throttle valve closed.

NOTE - Leaving RCIC-MO-131 open to maintain RCIC shutdown, with an BOP initiation signal present, will cause RCIC-MO-131 to cycle if a high water level trip occurs.

2.1.2 Leave RCIC-MO-131 open so turbine trip will not reset.

2.1.3 Ensure SM informed RCIC inoperable.

If directed to de-energize RCIC-MO-27, insert Remote Function rc-13 DE-Booth ENER.

Operator If directed to de-energize RCIC-MO-14, insert Remote Function rc21a DE-ENER.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 16 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 4 Event

Description:

Spurious RCIC initiation Time Position Applicants Action or Behavior NOTE to Examiners: The crew may shut and de-energize RCIC min flow valve MO-27 per the alarm procedure to prevent the valve from cycling.

If time permitted, the CRS refers to LCO 3.5.3, Condition A. Required Actions CRS are to Verify HPCI Operable with a Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Restore RCIC to Operable status with a Completion Time of 14 days.

END OF EVENT Notes Booth Proceed to the next event at direction of the Lead Examiner.

Operator Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 17 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 5 Event

Description:

HPCI Steam Leak in Secondary Containment and failure to isolate.

Time Position Applicants Action or Behavior When directed by Lead Examiner insert Trigger 4 causing a HPCI steam Booth supply line break.

Operator NOTE To Examiners: HP09, HPCI failure to isolate is already active.

BOP responds to alarm 9-3-1/E-10, AREA HIGH TEMP. Reports high area temperature in the SW Torus Area.

1.1 Dispatch Operator to alarming area to determine cause.

1.2 Attempt to isolate leaks.

1.3 IF a leak is identified to be from through-wall leakage in a Class 1 System (Reactor Coolant Pressure Boundary) and leak cannot be isolated, THEN enter Condition and Required Actions of Technical Specifications LCO 3.4.4.

1.4 Start additional HVAC coolers, as required, to maintain normal building temperatures and humidity.

BOP 1.5 Enter Procedure 2.4MC-RF for feedwater line break. NOTE to Examiners, step is N/A.

1.6 IF high area temperature due to leaks and source cannot be determined or Reactor Building is inaccessible, THEN perform following within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NOTE to Examiners, step is N/A.

1.6.1 Remove RWCU from service per Procedure 2.2.66, Section 19, Rapid Removal From Service.

1.6.2 Close AS-11, AUXILIARY STEAM SUPPLY HEADER ROOT (HEATER BOILER ROOM).

Diagnoses HPCI failed to isolate, if area temperatures are above the isolation BOP set point.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 18 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 5 Event

Description:

HPCI Steam Leak in Secondary Containment and failure to isolate.

Time Position Applicants Action or Behavior Role Play: When contacted to investigate for steam leaks in Reactor Booth Building, wait 3 minutes and report you hear loud hissing noise in torus Operator area. Report the torus area humidity level has risen above normal.

CRS enters EOP 5A to control Secondary Containment temperatures.

Monitors area temperature for any temperature rising above 195°F.

Direct all quad coolers started.

Direct verification of sump pump operation

Direct HPCI be manually isolated.

CRS Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 19 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 5 Event

Description:

HPCI Steam Leak in Secondary Containment and failure to isolate.

Time Position Applicants Action or Behavior Isolates HPCI by performing following at Panel 9-3-1:

Place STM SUPP INBD ISOL VLV MO-15 Control Switch to CLOSE.

Place STM SUPP OUTBD ISOL VLV MO-16 Control Switch to CLOSE.

Places area coolers and available Reactor Building HVAC into service at VBd R.

BOP FC-R-1E SE Quad, Control Switch to RUN FC-R-1F NE Quad, Control Switch to RUN 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 Refers to TS due to HPCI being isolated.

Since RCIC is inoperable, enters TS LCO 3.5.1 Condition G for both HPCI CRS and RCIC being inoperable at the same time. Required Actions are to be in MODE 3 with a Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND Reduce reactor steam dome pressure to 150 psig with a Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

END OF EVENT Notes Proceed to the next event at direction of the Lead Examiner.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 20 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior Booth When directed by Lead Examiner, insert Trigger 5 to cause Recirc Loop Operator A LOCA and Loss of Offsite Power NOTE to Examiners: It takes approximately 23 minutes for RPV level to lower to TAF.

Recognize and report loss of AC power.

CREW Recognize and report drywell pressure rising.

Recognize and report reactor scram.

Perform 2.1.5 Mitigating Task Scram Actions:

1.1 Press both RX SCRAM buttons.

1.2 Place REACTOR MODE switch to REFUEL.

Perform Attachment 2 Reactor Power Control of 2.1.5 as follows:

1 REACTOR POWER CONTROL 1.1 Ensure REACTOR MODE switch is in SHUTDOWN.

1.2 Verify all SDV vent and drain valves are closed.

ATC 1.3 Ensure operating RR pumps have run back to 22% speed.(Off due to LOOP) 1.4 Verify all control rods are fully inserted.

1.5 Observe nuclear instrumentation and perform following:

1.5.1 Insert SRM detectors. (unable due to LOOP) 1.5.2 Insert IRM detectors. (unable due to LOOP) 1.5.3 Change APRM recorders to IRMs. (N/A due to IRMs not inserted) 1.5.4 Range IRMs on scale. (N/A due to not inserted) 1.5.5 Check reactor power is lowering.

Enters EOP-1A and EOP-3A on drywell pressure high.

Ensure each of following initiated IAW EOP-1A step RC/L-1:

PCIS Group 1-7 isolations, SOP 2.1.22 CRS ECCS initiations Enters 5.3GRID and 5.3 EMPWR due to loss of AC power.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 21 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior (CT-1): When high pressure injection systems cannot maintain RPV level and low pressure ECCS systems fail to automatically start due to loss of AC power, crew manually starts DG 2 to energize LP ECCS systems prior to RPV water level falling below

-158 CFZ (TAF).

Performs 5.3EMPWR actions:

Ensures automatic actions occur:

Checks AC power availability.

Recognizes and reports DG 2 failure to start.

Recognizes and reports failure of RHR A and C Pumps and Core Spray BOP Pump A to start.

Manually starts DG 2 by placing control switch DIESEL GEN 2 START/STOP to START on Panel C.

Checks DG 2 comes up to speed and ties on to 4160V Bus 1G.

Starts RHR A and C Pumps and Core Spray A Pump.

Role Play: If directed to place Air Compressor A 480V breaker trip Booth pushbutton, insert Trigger 6 and report breaker pushbutton actuated.

Operator If directed to place Air Compressor B on REC cooling, insert Trigger 7 and report cooling has been swapped.

Performs 5.3EMPWR subsequent actions:

4.3 IF REC System has isolated, THEN perform following:

4.3.1 Ensure two REC pumps are running.

4.3.2 Place DRYWELL REC ISOL VALVE CONTROL switch to OPEN (perform regardless of power loss).

4.3.3 Throttle open REC HX outlet valve for a HX that was in service to maintain REC-PI-452, REC HEADER PRESSURE, in green band.

BOP 4.3.3.2 REC-MO-713, HX B OUTLET VLV.

4.3.4 Start third REC pump, if necessary.

4.3.5 Throttle open REC HX outlet valve to maintain REC HEADER PRESSURE in top of green band.

4.3.5.2 REC-MO-713.

4.3.6 Ensure following valves are closed:

4.3.6.1 REC-AO-701, RRMG SET OIL HX INLET.

4.3.6.2 REC-AO-710, RWCU NON-REGEN HX INLET.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 22 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior 4.3.7 Perform following concurrently:

4.3.7.1 Open REC-MO-700, NON-CRITICAL HEADER SUPPLY.

4.3.7.2 Continue throttling open REC HX outlet valve to maintain REC HEADER PRESSURE in green band.

b. REC-MO-713.

4.3.8 Ensure REC HX outlet valve full open.

4.3.8.2 REC-MO-713.

4.3.9 Place DRYWELL REC ISOL VALVE CONTROL switch to AUTO 4.4 IF SAC(s) not running, THEN perform following:

4.4.1 Place COMPRESSOR 1A control switch to OFF (PNL A).

4.4.2 At 480V, SUBSTATION 1F, press TRIP button on Breaker 4C, SAC 1A (Critical Switchgear Room F).

4.4.3 Place COMPRESSOR 1A control switch to AUTO (PNL A).

4.4.7 Place COMPRESSOR 1B control switch to OFF (PNL A).

4.4.8 At 480V, SUBSTATION 1G, press TRIP button on Breaker 2C, SAC1B (Critical Switchgear Room G).

4.4.9 Place COMPRESSOR 1B control switch to AUTO (PNL A).

Role Play: If contacted, as DCC System Operator report there are Booth widespread grid losses and you have no estimated time for restoration Operator yet. You will contact CNS as soon as you have additional information.

Direct stabilizing RPV pressure below 1050 psig.

CRS If ADS timer has initiated THEN direct BOP to inhibit ADS.

Verifies Low-Low Set controlling pressure on panel 9-3.

If Alarm 9-3-1/A-1 ADS TIMERS ACTUATED, report to the CRS.

BOP If directed, place ADS A and B inhibit switches to INHIBIT on panel 9-3.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 23 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior NOTE to Examiners: The CRS may elect to reserve Division 2 RHR for ECCS based on the downward level trend and delay initiating Containment Sprays until RPV level is controlled with Core Spray B following emergency depressurization.

CRS Directs Containment Sprays initiated per EOP-3A step PC/P-2 and/or PC/P-3.

BEFORE torus pressure reaches 10 psig, Spray Torus

2. Containment Sprays (RHR Hard Card) 2.1 IF required, with CRS permission, THEN place CONTMT COOLING 2/3 CORE VALVE CONTROL PERMISSIVE switch to MANUAL BOP OVERRD.

2.2 IF required, THEN place CONTMT COOLING VLV CONTROL PERMISSIVE switch to MANUAL.

2.3 Ensure RHR-MO-39B open.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 24 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior 2.4 IF reactor pressure 300 psig and injection not desired, THEN close RHR-MO-27B, OUTBD INJECTION VLV.

2.5 Ensure RHR PUMP(s) running.

NOTE - RHR pump operation at minimum flow should be limited to

< 15 minutes or pump damage may result.

2.6 Throttle RHR-MO-38B to maintain desired containment pressure.

2.7 Throttle RHR-MO-66B to obtain desired cooling rate.

Role Play: When directed by BOP to install EOP PTMs97-100, wait 3 Booth minutes then put in the overrides for the PTMs. Report back to BOP Operator when PTMs installed.

CRS If torus pressure exceeds 10 psig, Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 25 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior Verify DWSIL met, and direct Drywell Sprays initiated.

If directed to spray the drywell:

When Torus pressure exceeds 10 PSIG Spray the Drywell 2.8 IF Drywell Spray required, THEN perform following:

2.8.1 Open RHR-MO-31B.

2.8.2 Throttle RHR-MO-26B to maintain desired containment pressure.

BOP 2.9 IF PCIS Group 6 lights lit on Panel 9-5, THEN ensure one of following open:

2.9.1 REC-MO-711; or 2.9.2 REC-MO-714.

2.10 Place RHR SW System in service:

2.10.1 Start SWBP(s).

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 26 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior 2.10.2 Adjust SW-MO-89B to maintain flow between 2500 and 4000 gpm.

2.11 Throttle RHR-MO-66B to maintain desired cooling rate.

Maintain Drywell pressure between +2 & +10 psig.

Role Play: If sent to investigate RHR A, B, C, and D, and Core Spray A Booth pumps failure to start, wait 5 minutes, and report nothing is abnormal Operator locally or at the 4160 Buses.

CRS Assigns RPV water level as critical parameter Directs ATC to maximize CRD flow CRS Directs ATC to initiate SLC for level control NOTE to Examiners: RCIC-MO-131 control switch is broken and will not cycle so RCIC cannot be used for injection.

Attempts to put RCIC into service and recognizes RCIC-MO-131 will not BOP cycle.

Reports RCIC failure to the CRS.

Booth If directed to perform local actions for Maximizing CRD, insert Remote Operator Function rf12 OPEN to put other CRD filter into service.

Maximizes CRD flow as directed IAW 5.8.4:

ATC 9.6 Starts CRD Pump A by placing its control switch to start on panel 9-5.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 27 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior 9.9 Places CRD-FC-301 to MAN and throttle open in-service FCV.

Sends building operator to perform local steps of 5.8.4 section 9.

Initiates SLC as directed IAW 5.8.4:

6.2 WHEN CRS directs, THEN commence Alternate RPV Injection as follows (PANEL 9-5):

6.2.1 Place following keylock switches to START:

6.2.1.1 SLC PUMP A.

6.2.1.2 SLC PUMP B.

6.2.2 Verify red indicating lights for each pump energize.

6.2.3 Verify SLC-14A, LOOP A SQUIB VALVE, has fired by observing that SQUIB VALVE READY Light 1106A has extinguished.

6.2.4 Verify SLC-14B, LOOP B SQUIB VALVE, has fired by observing that SQUIB VALVE READY Light 1106B has extinguished.

6.2.5 Observe SLC pump discharge pressure rises above RPV pressure as indicated on SLC-PI-65, PUMP PRESS.

6.2.6 Inform CRS that Alternate RPV Injection with SLC from boron tank has commenced.

CRS When RPV water level cannot be restored and maintained above -150 inches CFZ, directs ATC to line up LPCI B and Core Spray B for injection.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 28 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior (CT-3): When high pressure injection systems cannot maintain RPV level and low pressure ECCS pumps fail to automatically start, crew manually starts pumps to align LP ECCS systems for injection prior to RPV water level falling below -158 CFZ (TAF).

Verifies Core Spray is lined up for injection with pump running on panel 9-3.

Ensures LPCI B is aligned for injection with at least one pump running on panel 9-3.

ATC If RHR loop B was in containment spray mode, on panel 9-3:

o Ensures MO-26B is closed o Ensures MO-39B is closed o Ensures MO-27B is open Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 29 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior CRS When RPV level goes below -158 inches CFZ and cannot be maintained above -183 inches CFZ, emergency depressurization is required.

Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 30 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior Enters EOP-2A Emergency Depressurization:

Verifies suppression pool water level is >6 ft on SPDS Directs BOP to open 6 SRVs (CT-2): When RPV level lowers to -158 CFZ (TAF) and cannot be maintained above -183 CFZ (MSCWL) and insufficient high pressure injection systems are available to restore level, crew begins to Emergency Depressurize by opening the first of six SRVs before RPV level lowers below -183 CFZ.)

Opens 6 of the following SRVs by placing control switches to OPEN on panel 9-3.

SRV-71A SRV-71B SRV-71E SRV-71G BOP SRV-71H SRV-71C SRV-71D SRV-71F Verifies red solenoid light and amber SRV tailpipe pressure switch light illuminate and green solenoid light extinguishes Verifies RPV pressure falls Verifies low pressure ECCS injection valves open on panel 9-3 when pressure goes below the injection valve auto open permissive (approximately 400 psig).

Core Spray A - INBD INJ THROTTLE VLV MO 12A ATC Core Spray B - INBD INJ THROTTLE VLV MO 12B LPCI A - INBD INJECTION VLV MO 25A LPCI B - INBD INJECTION VLV MO 25B Recognizes Core Spray B INBD INJ THROTTLE VLV MO 12B fails to automatically open and reports failure to the CRS.

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Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 31 of 33 Op-Test No.: CNS 15-01 Scenario No.: 5 Event No.: 6, 7, 8 and 9 Event

Description:

Reactor Recirc loop A LOCA, Loss of Offsite Power, DG2 output breaker failure to close, all RHR pump, and Core Spray A failure to auto start Time Position Applicants Action or Behavior Manually opens CS-MO-12B as necessary.

Verifies LPCI and Core Spray injection flow rises on panel 9-3 meters when RPV pressure lowers below the shutoff head of the ECCS pumps (approximately 100 psig).

Core Spray A - PUMP FLOW CS-FI-50A Core Spray B - PUMP FLOW CS-FI-50B LPCI A - RHR A FLOW RHR-FI-133A LPCI B - RHR B FLOW RHR-FI-133B Reports ECCS injection flow and level rising.

CRS Directs ATC to restore and control level +3 inches to +54 inches.

When level rises above -158 CFZ, controls injection from LPCI B, CS B, ATC SLC, and CRD by throttling valves and/or cycling pumps to raise and maintain level +3 inches to +54 inches.

END OF SCENARIO NOTE to Examiners: Scenario objectives have been met when the crew has emergency depressurized and level is being raised to +3 inches to +54 inches in a controlled manner.

Notes Booth When directed by the Lead Examiner, place the simulator in freeze and Operator tell the crew to stop operating Rev. 0

Appendix D Required Operator Actions Form ES-D-2 NRC CNS 15-01 Scenario 5 Page 32 of 33 Rev. 0

INITIAL CONDITIONS A. Plant Status:

1. In Run at 5% power at Beginning of Cycle.

2. Rod Sequence Information: Step 24 complete in sequence 30-SU-BOL B. Tech. Spec. Limitations in effect:

1. None.

C. Significant problems/abnormalities:

1. None D. PRA Risk:

Green E. Evolutions/maintenance for the on-coming shift:

1. Continue to raise power with control rod withdrawal and achieve 20%-25% bypass valve positon.

2. Continue the startup per Procedures:

a. 2.1.1, Steps 4.18 and 5.36

b. 2.2.77, Attachment 1 Step 1.15

c. 2.2.28.1, Step 5.14 Rev. 0

ML17122A080

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