ML14111A404

2014-03-Draft Operating Test

ML14111A404
Person / Time
Site:	River Bend
Issue date:	03/27/2014
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
laura hurley
References
50-458/14-03 50-458/OL-14
Download: ML14111A404 (630)

Text

RJPM-NRC-M14-A1 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Perform Jet Pump Operability OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 30 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: September 30, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-A1 Rev. 0 EXAMINER INFO SHEET Task Standard: Jet Pump Operability Test completed; Items found NOT meeting acceptance criteria are in accordance with the key.

Synopsis: This task will have the applicant perform an operability using STP-053-3001.

Jet Pump Data Sheet will be provided to the applicant.

The applicant will arrive at the conclusion that several items do not meet acceptability requirements.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to complete STP-053-3001, Jet Pump Operability Test, by using the data provided and determine if acceptance criteria are being met.

3) Initial Conditions:

The plant is operating at 100% power and STP-053-3001, Jet Pump Operability Test, is due to be performed.

Reactor Thermal Power 100%

Recirculation Pumps Fast Speed FCV-A Position 61%

FCV-B Position 60%

Recirculation Loop Flows are in compliance with TS 3.4.1

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A1 Rev 0 Page 2 of 8

RJPM-NRC-M14-A1 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Perform Jet Pump Operability 202012001001 G 2.1.20 4.6 G 2.1.25

3.9 REFERENCES

APPLICABLE OBJECTIVES STP-053-3001, Rev 20 RLP-STM-0053, Obj 3, 11 REQUIRED MATERIALS:

STP-053-3001, Rev 20 SIMULATOR CONDITIONS &/or SETUP:

1. Need a marked up copy of the STP up to step 7 circled (not slashed)

2. This is a classroom/Admin JPM - There is no simulator setup

3. Develop the jet pump data sheet for the applicants to use to complete attachment 1.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Jet Pump Operability Test completed; Items found NOT meeting acceptance criteria are in accordance with the key.

RJPM-NRC-M14-A1 Rev 0 Page 3 of 8

RJPM-NRC-M14-A1 Rev. 0 This is a copy of the jet pump data sheet used by ops to complete STP-053-3001 Use this MASTER to fill in the handout on the Operator Cue Sheet JET PUMPS A FCV POSITION ___61___ B FCV POSITION ___60___

A LOOP FLOW __30.2___ B LOOP FLOW ___31.9__

(C51-R614A) (C51-R614B)

TOTAL FLOW __78.7___

(B33-R613)

JET PUMP D/P

1. __50_ 11. __36_

2. __45_ 12. __25_

3. __38_ 13. __24_

4. __32_ 14. __32_

5. __29_ 15. __39_

6. __34_ 16. __32_

7. __37_ 17. __30_

8. __39_ 18. __33_

9. __34_ 19. __46_

10. _25_ 20. __30_

RJPM-NRC-M14-A1 Rev 0 Page 4 of 8

RJPM-NRC-M14-A1 Rev. 0 PERFORMANCE:

START TIME:

7.1 Complete Attachment 1, Jet Pump Operability Test Data Sheet.

1. Procedure Step: 7.1 Complete Attachment 1, Jet Pump Operability Test Data Sheet.

Standard Applicant completed attachment 1 in accordance with the key.

Cue Notes The applicant will use the pictures provided to gather readings.

Results SAT UNSAT

2. Procedure Step: 7.1 through 7.5, PROCEDURE Standard Applicant completed section 7 in accordance with the key.

Cue Notes Results SAT UNSAT

3. Procedure Step: 8 ACCEPTANCE CRITERIA Standard Applicant completed section 8 in accordance with the key.

Cue Notes Results SAT UNSAT Terminating Cue: Jet Pump Operability Test completed; Items found NOT meeting acceptance criteria are in accordance with the key.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A1 Rev 0 Page 5 of 8

RJPM-NRC-M14-A1 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A1 Rev 0 Page 6 of 8

RJPM-NRC-M14-A1 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

The CRS has directed you to complete STP-053-3001, Jet Pump Operability Test, by using the data provided and determine if acceptance criteria are being met.

Initial Conditions:

The plant is operating at 100% power and STP-053-3001, Jet Pump Operability Test, is due to be performed.

Reactor Thermal Power 100%

Recirculation Pumps Fast Speed FCV-A Position 61%

FCV-B Position 60%

Recirculation Loop Flows are in compliance with TS 3.4.1 See attached Jet Pump Data Sheet RJPM-NRC-M14-A1 Rev 0 Page 7 of 8

RJPM-NRC-M14-A1 Rev. 0 JET PUMPS A FCV POSITION ________ B FCV POSITION ________

A LOOP FLOW ________ B LOOP FLOW ________

(C51-R614A) (C51-R614B)

TOTAL FLOW ________

(B33-R613)

JET PUMP D/P

1. _____ 11. _____

2. _____ 12. _____

3. _____ 13. _____

4. _____ 14. _____

5. _____ 15. _____

6. _____ 16. _____

7. _____ 17. _____

8. _____ 18. _____

9. _____ 19. _____

10. _____ 20. _____

RJPM-NRC-M14-A1 Rev 0 Page 8 of 8

RJPM-NRC-M14-A2 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Perform Surveillances Required for Entry Into Single Loop Operation OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 1, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 9

RJPM-NRC-M14-A2 Rev. 0 EXAMINER INFO SHEET Task Standard: Step 4 of Attachment 1 from GOP-0004, Single Loop Operation, is complete in accordance with the answer key.

Synopsis: This task will have the applicant perform a one hour operability using GOP-0004, Attachment 1, Step 4.

Three pictures/diagrams will be provided for the applicant to obtain readings:

The applicant will also be required to use steam tables to perform the final calculation of the operability.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to complete step 4 of Attachment 1 of GOP-0004, Single Loop Operation.

3) Initial Conditions:

The plant experienced a trip of the B Recirc Pump 30 minutes ago.

Steps 1 - 3 of Attach 1 of GOP-0004, Single Loop Operation, have been completed.

Core Thermal Power is 901 MW A and B Recirc Flow Control Valves are in Loop Manual Loop A Flow recorder on C51-R614 is failed downscale Computer Point B33NA005 reads 2.7 mlbm/hr Loop A temperature reads 515°F

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A2 Rev 0 Page 2 of 9

RJPM-NRC-M14-A2 Rev. 0 REFERENCE USE ATTACHMENT 1 SINGLE LOOP OPERATION PAGE 4 OF 5 Step Initials Date/Time NOTE Steps 4, 5, and 6 should be performed concurrently but completed within their respective time limits.

Record initial reading here, then every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on STP-000-0001, Daily Operating Logs.

4 Within one hour of entering Single Loop Operation, verify the following:

4.1. Thermal Power is less than or equal to 77.6% Rated Thermal Power (2400 MWTH)

__INITIALs_

__901 ___CMWTH=__ 29.15___% 77.6%

3091 (TSR 3.4.1.1.2)

AND 4.2. At H13-P680, B33-HYVF060A and B33-HYVF060B, FLOW CONT __INITIALs_

VALVE, is in LOOP MANUAL. (TSR 3.4.1.1.3) 4.3. Total loop flow in running loop is less than 33 kgpm using one of the following methods (N/A method not used): (TSR 3.4.1.1.1)

1. Obtain flow from C51-R614, LOOP A/B FLOW RECORDER, for the operating loop.

___________ kgpm ___N / A__

2. Use computer point for the operating loop (LOOP A -

B33NA005 or B33NA006; LOOP B - B33NA007 or B33NA008) and convert from mlbm/hr to kgpm using the following formula:

__INITIALs_

__2.7__ x _0.0208__ x (124.68) = __ 7.00__

(flow) (sv) (kgpm) where flow = loop flow from computer point in mlbm/hr.

sv = specific volume from steam tables (Vf) (dependent on loop temp) in ft3/lbm.

RJPM-NRC-M14-A2 Rev 0 Page 3 of 9

RJPM-NRC-M14-A2 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Perform Surveillances Required for 202012001001 G 2.1.7 4.4 Entry Into Single Loop Operation 300130003001

REFERENCES:

APPLICABLE OBJECTIVES GOP-0004, Rev 21 RLP-HLO-0053, Obj 7 RLP-OPS-0503, Obj 2, 3 REQUIRED MATERIALS:

GOP-0004, Rev 21, Attachment 1, page 7 of 26 A copy of steam tables SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup

2. Get pictures for the applicants to use to complete attachment 1.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Step 4 of Attachment 1 from GOP-0004, Single Loop Operation, is complete in accordance with the answer key.

RJPM-NRC-M14-A2 Rev 0 Page 4 of 9

RJPM-NRC-M14-A2 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 4.1 Thermal Power is less than or equal to 77.6% Rated Thermal Power (2400 MWTH)

_________ / 3091 CMWTH = _______% 77.6%

Standard Applicant determined that thermal power was less than or equal to 77.6% in accordance with the answer key.

Cue Notes Results SAT UNSAT

2. Procedure Step: 4.2 At H13-P680, B33-HYVF060A and B33-HYVF060B, FLOW CONT VALVE, is in LOOP MANUAL. (TSR 3.4.1.1.3)

Standard Applicant determined FCVs are in Loop Manual using initial conditions.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-A2 Rev 0 Page 5 of 9

RJPM-NRC-M14-A2 Rev. 0

3. Procedure Step: 4.3 Total loop flow in running loop is less than 33 kgpm using one of the following methods (N/A method not used): (TSR 3.4.1.1.1)

2. Use computer point for the operating loop (LOOP A - B33NA005 or B33NA006; LOOP B - B33NA007 or B33NA008) and convert from mlbm/hr to kgpm using the following formula:

____________ x ______________ x (124.68) = _________

(flow) (sv) (kgpm)

Standard Applicant performed calculation and determined that loop flow was less than 33 kgpm in accordance with the answer key.

Cue Notes Results SAT UNSAT Terminating Cue: Step 4 of Attachment 1 from GOP-0004, Single Loop Operation, is complete in accordance with the answer key.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A2 Rev 0 Page 6 of 9

RJPM-NRC-M14-A2 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A2 Rev 0 Page 7 of 9

RJPM-NRC-M14-A2 Rev. 0 OPERATOR CUE SHEET Initial Conditions:

The plant experienced a trip of the B Recirc Pump 30 minutes ago.

Steps 1-3 of Attach 1 of GOP-0004, Single Loop Operation, have been completed.

Core Thermal Power is 901 MW A and B Recirc Flow Control Valves are in Loop Manual Loop A Flow recorder on C51-R614 is failed downscale Computer Point B33NA005 reads 2.7 mlbm/hr Loop A temperature reads 515°F Initiating Cues:

The CRS has directed you to complete step 4 of Attachment 1 of GOP-0004, Single Loop Operation.

RJPM-NRC-M14-A2 Rev 0 Page 8 of 9

RJPM-NRC-M14-A2 Rev. 0 REFERENCE USE ATTACHMENT 1 SINGLE LOOP OPERATION PAGE 4 OF 5 Step Initials Date/Time NOTE Steps 4, 5, and 6 should be performed concurrently but completed within their respective time limits.

Record initial reading here, then every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on STP-000-0001, Daily Operating Logs.

4 Within one hour of entering Single Loop Operation, verify the following:

4.1. Thermal Power is less than or equal to 77.6% Rated Thermal Power (2400 MWTH)

__________CMWTH=____________% 77.6%

3091 (TSR 3.4.1.1.2)

AND 4.2. At H13-P680, B33-HYVF060A and B33-HYVF060B, FLOW CONT VALVE, is in LOOP MANUAL. (TSR 3.4.1.1.3) ___________

4.3. Total loop flow in running loop is less than 33 kgpm using one of the following methods (N/A method not used): (TSR 3.4.1.1.1)

1. Obtain flow from C51-R614, LOOP A/B FLOW RECORDER, for the operating loop.

___________ kgpm

2. Use computer point for the operating loop (LOOP A -

B33NA005 or B33NA006; LOOP B - B33NA007 or B33NA008) and convert from mlbm/hr to kgpm using the following formula:

__________ x __________ x (124.68) = _________

(flow) (sv) (kgpm) where flow = loop flow from computer point in mlbm/hr.

sv = specific volume from steam tables (Vf) (dependent on loop temp) in ft3/lbm.

RJPM-NRC-M14-A2 Rev 0 Page 9 of 9

RJPM-NRC-M14-A3 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Electrically Bypass a Control Rod in the Rod Gang Drive Cabinet OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform Plant X Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 1, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-A3 Rev. 0 EXAMINER INFO SHEET Task Standard: The answer sheet graphic for the Rod Gang Drive System Cabinet Analyzer Board Command Registry is annotated with (1) the binary code for bypassing Rod 12-33 and (2) the bypass switch.

Synopsis: This task will have the applicant use a procedure to determine the binary address of a control rod which requires bypassing in the Rod Gang Drive System (RGDS). The applicant answer sheet has a diagram of the RGDS Analyzer Board Command Registry which he will mark up to show which switches to reposition.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to electrically bypass control rod 12-33 in the Rod Gang Drive Cabinet.

3) Initial Conditions:

The plant is in the process of starting up.

Maintenance is scheduled to be performed on Rod 12-33 that requires it to be electrically isolated.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A3 Rev 0 Page 2 of 8

RJPM-NRC-M14-A3 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Electrically Bypass a Control Rod in 214004004001 G 2.1.37 4.3 the Rod Gang Drive Cabinet

REFERENCES:

APPLICABLE OBJECTIVES SOP-0071, Rev 26 RLP-STM-0500, Obj 15, 20 REQUIRED MATERIALS:

SOP-0071, Rev 26 Handout for Student copy of switch manipulation SIMULATOR CONDITIONS & SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup but pictures/diagrams must be developed for the applicants use.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: The answer sheet graphic for the Rod Gang Drive System Cabinet Analyzer Board Command Registry is annotated with (1) the binary code for bypassing Rod 12-33 and (2) the bypass switch.

RJPM-NRC-M14-A3 Rev 0 Page 3 of 8

RJPM-NRC-M14-A3 Rev. 0 PERFORMANCE:

START TIME:

PROCEDURE NOTE This method is normally used when bypassing Control Rods during Maintenance or Transponder Faults.

All actions in this Section are performed at panel H13-P653, Rod Gang Drive Cabinet.

In determining the binary numbers corresponding to the column and row plant coordinates, use Attachment 5, Full Core Map.

Key Number 1161, located on the Reactor Operator keyring in the Unit Operator desk, will be needed to access the Rod bypass file.

5.10. Electrically Bypassing a Control Rod in the Rod Gang Drive System Cabinet

1. Procedure Step: 5.10.1 On Attachment 5, Full Core Map, or Attachment 6, Binary Numbers for Bypassing Control Rods, determine the binary numbers corresponding to the column (X) and the row (Y) coordinates for the rod to be bypassed.

Standard NA Cue Notes No action is necessary - the applicant will use the information gathered in this step to correctly perform the next step. The applicant determined the correct binary code for rod 12-33, using attachment 5 or 6.

PROCEDURE NOTE RJPM-NRC-M14-A3 Rev 0 Page 4 of 8

RJPM-NRC-M14-A3 Rev. 0 PROCEDURE NOTE The following switches are positioned to enter the Control Rods Binary Address. When the switch is in the UP position, a 1 is input to the Address Register. When the switch is in the DOWN position, a 0 is input to the Address Register.

2. *Procedure Step: 5.10.2 On the Rod Gang Drive System Analyzer Board under Command Register 2, Bypassed Rod Identity, enter the bypassed Control Rod identity.

Standard Applicant marked the answer sheet diagram of the Bypassed Rod Identity with up arrows through only the X2, X0, Y3, and Y1.

Cue Notes Results SAT UNSAT

3. *Procedure Step: 5.10.3 Place BYPASS SWITCH in the UP Position.

Standard Applicant marked the answer sheet diagram of the Rod Gang Drive Cabinet with an up arrow through the bypass switch.

Cue Notes Results SAT UNSAT Terminating Cue: The answer sheet graphic for the Rod Gang Drive System Cabinet Analyzer Board Command Registry is annotated with (1) the binary code for bypassing Rod 12-33 and (2) the bypass switch.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A3 Rev 0 Page 5 of 8

RJPM-NRC-M14-A3 Rev. 0 ANSWER KEY This is from Attach 6 of SOP-0071; it shows the coordinates that the candidate should determine in step 5.10.1 for the given control rod.

This is how the applicants answer sheet looks:

This is how the applicant should mark up the answer sheet after step 5.10.2:

This is how the applicant should mark up the answer sheet after step 5.10.3:

RJPM-NRC-M14-A3 Rev 0 Page 6 of 8

RJPM-NRC-M14-A3 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A3 Rev 0 Page 7 of 8

RJPM-NRC-M14-A3 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

The CRS has directed you to electrically bypass control rod 12-33 in the Rod Gang Drive Cabinet in accordance with the SOP.

Initial Conditions:

The plant is in the process of starting up.

Maintenance is scheduled to be performed on Rod 12-33 that requires it to be electrically isolated.

Student Answer Sheet Directions:

Below is a diagram of the Command Registry for the Rod Gang Drive System Analyzer Board.

Indicate any switch manipulations by drawing an arrow through the switch/switches (if any) that you would reposition.

RJPM-NRC-M14-A3 Rev 0 Page 8 of 8

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• ROD CONTROL AND INFORMATION SYSTEM (SYS #500)

PROCEDURE NUMBER: *SOP-0071 REVISION NUMBER: *026 Effective Date:

• NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0071 REV - 026 PAGE 1 OF 44

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ...................................................................................4 3 PREREQUISITES ......................................................................................................................5 4 SYSTEM STARTUP..................................................................................................................5 4.1 Startup Following System Deenergization or Loss of Power ...........................................5 4.2 Startup with System Energized .........................................................................................6 5 SYSTEM OPERATION .............................................................................................................7 5.1 Performing Control Rod Notch Movement (Note: This Section is Reference Use.) ..................................................................................................................................7 5.2 Performing Continuous Rod Movement (Note: This Section is Reference Use.) ............8 5.3 Continuous Insert Using IN TIMER SKIP (Note: This Section is Reference Use.) ..................................................................................................................................10 5.4 Performing Control Rod Coupling Check (Note: This Section is Reference Use.) ..................................................................................................................................11 5.5 Determining Control Rod Position With one RACS Cabinet Not Functional..................12 5.6 Determining Control Rod Position With the Full Core Display Out of Service...............12 5.7 Notching Control Rods Out of Position '00' (Note: This Section is Reference Use.) .................................................................................................................................15 5.8 Notching Control Rods Out of Positions Other Than '00' (Note: This Section is Reference Use.) .................................................................................................................18 5.9 Difficulty Notching IN a Control Rod. ............................................................................19 5.10 Electrically Bypassing a Control Rod in the Rod Gang Drive System (RGDS)

Cabinet ..............................................................................................................................20 5.11 Bypassing Control Rod Position Information in the Rod Action Control System (RACS) Cabinets ..................................................................................................21 5.12 Entering Substitute Data for Control Rod Position (Note: This Section is Reference Use.) .................................................................................................................22 5.13 Resetting Rod Withdrawal Limiter Induced Control Rod Blocks ....................................22 6 SYSTEM SHUTDOWN.............................................................................................................24 7 REFERENCES ..........................................................................................................................24 8 RECORDS ..................................................................................................................................24 SOP-0071 REV - 026 PAGE 2 OF 44

CONTINUOUS USE ATTACHMENT 1 - VALVE LINEUP - ROD CONTROL AND INFORMATION SYSTEM ...........................................................................................................................25 ATTACHMENT 2 - INSTRUMENT & VALVE LINEUP ROD CONTROL AND INFORMATION SYSTEM (SAFETY RELATED) ........................................................27 ATTACHMENT 3 - ELECTRICAL LINEUP - ROD CONTROL AND INFORMATION SYSTEM ...........................................................................................................................29 ATTACHMENT 4 - CONTROL BOARD LINEUP - ROD CONTROL AND INFORMATION SYSTEM ..............................................................................................30 ATTACHMENT 5 - FULL CORE MAP ..........................................................................................31 ATTACHMENT 6 - BINARY NUMBERS FOR BYPASSING CONTROL RODS.......................32 ATTACHMENT 7 - OPERATOR CONTROL MODULE - CONTROL DEVICES AND STATUS INDICATORS WITH FUNCTION..................................................................37 ATTACHMENT 8 - CONTROL ROD DEFICIENCY REPORT ....................................................43 SOP-0071 REV - 026 PAGE 3 OF 44

CONTINUOUS USE 1 PURPOSE 1.1 To provide instructions for the operation of the Rod Control and Information System.

2 PRECAUTIONS AND LIMITATIONS 2.1 Technical Specification 3.3.2.1 and TSR 3.3.2.1.10 require that Control Rods not be withdrawn in Modes 1 and 2 with the Main Turbine Bypass Valves open and Thermal Power greater than 20% of Rated Thermal Power (RTP). This condition causes the Rod Withdrawal Limiter (RWL) function to be nonconservative.

2.2 Technical Specifications 3.1.3 and 3.1.6 require that each Control Rod be operable in Modes 1 and 2, and that the operable Control Rods comply with the requirements of the Banked Position Withdrawal Sequence in Modes 1 or 2 with Rated Thermal Power (RTP) less than or equal to 10%.

2.3 Technical Specification SR 3.3.2.1.9 requires that, prior to the movement of Control Rods bypassed in Rod Action Control System (RACS), the verification of the bypassing and movement of Control Rods required to be bypassed in RACS is in conformance with applicable analyses by a second licensed operator or other qualified member of the technical staff.

2.4 Test Switch Positions on H13-P651, 652 and 653 have a profound effect on RC&IS operability and should only be changed with the concurrence and/or assistance of I&C.

2.5 For abnormal occurrences involving Control Rod movement above or below the Low Power Setpoint (LPSP), REP-0051, Reactivity Control and Control Rod Movement or AOP-0061, Mispositioned Control Rods should be referenced.

2.6 A Coupling Check shall be performed when any control rod is withdrawn to position 48.

2.7 Hot control rod drive mechanisms (CRDM) have the potential to be exposed to thermal fatigue if subjected to multiple insert signals and/or coupling checks as directed by Sections 5.4 or 5.7.2 of this procedure when dealing with rod movement issues.

Therefore, they should be performed in rapid succession in order to prevent temperature cycling in the CRDM.

SOP-0071 REV - 026 PAGE 4 OF 44

CONTINUOUS USE 3 PREREQUISITES 3.1 Check that the Control Rod Drive Hydraulics System is in operation per SOP-0002, Control Rod Drive Hydraulics.

3.2 Check that the 120 VAC Instrument Bus is in operation per SOP-0048, 120 VAC System.

3.3 Verify system is lined up for startup.

4 SYSTEM STARTUP 4.1 Startup Following System Deenergization or Loss of Power 4.1.1. Verify power switches in OFF as follows:

1. At H13-P680, inside cabinet, CB1, Power Module

2. At H13-P652, one breaker inside cabinet, Rod Pattern Controller

3. At H13-P653, two breakers inside cabinet, Rod Gang Drive

4. At H13-P651, one breaker inside cabinet, Rod Pattern Controller 4.1.2. Perform the Instrument and Valve Lineup per Attachment 2, Instrument &

Valve Lineup Rod Control and Information System (Safety Related).

4.1.3. Perform the Electrical Lineup per Attachment 3, Electrical Lineup - Rod Control and Information System.

4.1.4. Turn power supply switches to ON as follows:

1. At H13-P680, inside cabinet, CB1, Power Module

2. At H13-P652, one breaker inside cabinet, Rod Pattern Controller

3. At H13-P653, two breakers inside cabinet, Rod Gang Drive

4. At H13-P651, one breaker inside cabinet, Rod Pattern Controller SOP-0071 REV - 026 PAGE 5 OF 44

CONTINUOUS USE ATTACHMENT 5 PAGE 1 OF 1 FULL CORE MAP SOP-0071 REV - 026 PAGE 31 OF 44

CONTINUOUS USE ATTACHMENT 6 PAGE 1 OF 5 BINARY NUMBERS FOR BYPASSING CONTROL RODS 04 04- 04- 04- 04- 04-Rod 17 21 25 29 33 37 41 X4 0 0 0 0 0 0 0 X3 0 0 0 0 0 0 0 X2 0 0 0 0 0 0 0 X1 1 1 1 1 1 1 1 X0 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 Y3 0 0 1 1 1 1 1 Y2 1 1 0 0 0 0 1 Y1 1 1 0 0 1 1 0 Y0 0 1 0 1 0 1 0 08 08- 08- 08- 08- 08 08-Rod 13 17 21 25 29 33 37 41 45 X4 0 0 0 0 0 0 0 0 0 X3 0 0 0 0 0 0 0 0 0 X2 1 1 1 1 1 1 1 1 1 X1 0 0 0 0 0 0 0 0 0 X0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 Y3 0 0 0 1 1 1 1 1 1 Y2 1 1 1 0 0 0 0 1 1 Y1 0 1 1 0 0 1 1 0 0 Y0 1 0 1 0 1 0 1 0 1 12 12- 12- 12- 12- 12 12- 12 Rod 09 13 17 21 25 29 33 37 41 45 49 X4 0 0 0 0 0 0 0 0 0 0 0 X3 0 0 0 0 0 0 0 0 0 0 0 X2 1 1 1 1 1 1 1 1 1 1 1 X1 0 0 0 0 0 0 0 0 0 0 0 X0 1 1 1 1 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 1 1 1 1 1 1 1 Y2 1 1 1 1 0 0 0 0 1 1 1 Y1 0 0 1 1 0 0 1 1 0 0 1 Y0 0 1 0 1 0 1 0 1 0 1 0 SOP-0071 REV - 026 PAGE 32 OF 44

CONTINUOUS USE ATTACHMENT 6 PAGE 2 OF 5 BINARY NUMBERS FOR BYPASSING CONTROL RODS 16 16- 16- 16- 16- 16 16- 16 16- 16-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 0 0 0 0 0 0 0 0 0 0 0 0 0 X2 1 1 1 1 1 1 1 1 1 1 1 1 1 X1 1 1 1 1 1 1 1 1 1 1 1 1 1 X0 0 0 0 0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 20 20- 20- 20- 20- 20 20- 20 20- 20-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 0 0 0 0 0 0 0 0 0 0 0 0 0 X2 1 1 1 1 1 1 1 1 1 1 1 1 1 X1 1 1 1 1 1 1 1 1 1 1 1 1 1 X0 1 1 1 1 1 1 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 24 24- 24- 24- 24- 24 24- 24 24- 24-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 1 1 X2 0 0 0 0 0 0 0 0 0 0 0 0 0 X1 0 0 0 0 0 0 0 0 0 0 0 0 0 X0 0 0 0 0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 SOP-0071 REV - 026 PAGE 33 OF 44

CONTINUOUS USE ATTACHMENT 6 PAGE 3 OF 5 BINARY NUMBERS FOR BYPASSING CONTROL RODS 28 28- 28- 28- 28- 28 28- 28 28- 28-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 1 1 X2 0 0 0 0 0 0 0 0 0 0 0 0 0 X1 0 0 0 0 0 0 0 0 0 0 0 0 0 X0 1 1 1 1 1 1 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 32 32- 32- 32- 32- 32 32- 32 32- 32-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 1 1 X2 0 0 0 0 0 0 0 0 0 0 0 0 0 X1 1 1 1 1 1 1 1 1 1 1 1 1 1 X0 0 0 0 0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 36 36- 36- 36- 36- 36 36- 36 36- 36-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 1 1 X2 0 0 0 0 0 0 0 0 0 0 0 0 0 X1 1 1 1 1 1 1 1 1 1 1 1 1 1 X0 1 1 1 1 1 1 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 SOP-0071 REV - 026 PAGE 34 OF 44

CONTINUOUS USE ATTACHMENT 6 PAGE 4 OF 5 BINARY NUMBERS FOR BYPASSING CONTROL RODS 40 40- 40- 40- 40- 40 40- 40 40- 40-Rod 05 09 13 17 21 25 29 33 37 41 45 49 53 X4 0 0 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 1 1 X2 1 1 1 1 1 1 1 1 1 1 1 1 1 X1 0 0 0 0 0 0 0 0 0 0 0 0 0 X0 0 0 0 0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 0 1 1 1 1 1 1 1 1 Y2 0 1 1 1 1 0 0 0 0 1 1 1 1 Y1 1 0 0 1 1 0 0 1 1 0 0 1 1 Y0 1 0 1 0 1 0 1 0 1 0 1 0 1 44 44- 44- 44- 44- 44 44- 44 Rod 09 13 17 21 25 29 33 37 41 45 49 X4 0 0 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 1 1 X2 1 1 1 1 1 1 1 1 1 1 1 X1 0 0 0 0 0 0 0 0 0 0 0 X0 1 1 1 1 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 0 0 0 0 Y3 0 0 0 0 1 1 1 1 1 1 1 Y2 1 1 1 1 0 0 0 0 1 1 1 Y1 0 0 1 1 0 0 1 1 0 0 1 Y0 0 1 0 1 0 1 0 1 0 1 0 48 48- 48- 48- 48- 48 48-Rod 13 17 21 25 29 33 37 41 45 X4 0 0 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 1 1 X2 1 1 1 1 1 1 1 1 1 X1 1 1 1 1 1 1 1 1 1 X0 0 0 0 0 0 0 0 0 0 Y4 0 0 0 0 0 0 0 0 0 Y3 0 0 0 1 1 1 1 1 1 Y2 1 1 1 0 0 0 0 1 1 Y1 0 1 1 0 0 1 1 0 0 Y0 1 0 1 0 1 0 1 0 1 SOP-0071 REV - 026 PAGE 35 OF 44

CONTINUOUS USE ATTACHMENT 6 PAGE 5 OF 5 BINARY NUMBERS FOR BYPASSING CONTROL RODS 52 52- 52- 52- 52- 52-Rod 17 21 25 29 33 37 41 X4 0 0 0 0 0 0 0 X3 1 1 1 1 1 1 1 X2 1 1 1 1 1 1 1 X1 1 1 1 1 1 1 1 X0 1 1 1 1 1 1 1 Y4 0 0 0 0 0 0 0 Y3 0 0 1 1 1 1 1 Y2 1 1 0 0 0 0 1 Y1 1 1 0 0 1 1 0 Y0 0 1 0 1 0 1 0 SOP-0071 REV - 026 PAGE 36 OF 44

RJPM-NRC-M14-A4 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Determine Radiological Brief and Protective Clothing Requirements OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform Plant X Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 1, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 6

RJPM-NRC-M14-A4 Rev. 0 EXAMINER INFO SHEET Task Standard: Applicant determined that a high radiation area brief by RP will be required.

Applicant also determined that single anti-c clothing will be worn.

Synopsis: This task will have the applicant review a RWP and Survey map to determine both the dress-out requirements and the type of brief needed for an evolution.

The evolution is to perform a GVI and check oil level in the RCIC pump and turbine.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to determine (1) the type of brief required, and (2) the protective clothing requirements for entry into this area.

3) Initial Conditions:

Following a quarterly Rad Survey by RP, it was reported that there was a hissing sound coming from somewhere on the north wall of the 95 elevation RCIC cubicle. RP is unable to give any more information on this. Preparations are being made for OPs to investigate.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A4 Rev 0 Page 2 of 6

RJPM-NRC-M14-A4 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Determine Radiological Brief and 217013001001 G 2.3.7 3.5 Protective Clothing Requirements 217004001004

REFERENCES:

APPLICABLE OBJECTIVES RWP-2014-1057, Rev 0 RLP-_______________

EN-RP-101, Rev 7 (Section 5.4)

Handout of Survey Maps of the RCIC cubicle (Contamination & Rad)

REQUIRED MATERIALS:

RWP-2014-1057, Rev 0 (altered to fit our needs)

EN-RP-101, Rev 7 Survey Maps of the RCIC cubicle (Contam and Rad)

SIMULATOR CONDITIONS & SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Applicant determined that a high radiation area brief by RP will be required. Applicant also determined that single anti-c clothing will be worn.

RJPM-NRC-M14-A4 Rev 0 Page 3 of 6

RJPM-NRC-M14-A4 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1. Determine the briefing requirements for entering RCIC cubicle 95 elev.

Standard Applicant used the provided procedure (EN-RP-101) to determine that a High Rad Brief would be required.

Cue Provide applicant with reference material.

Notes Results SAT UNSAT

2. Procedure Step: 2. Determine the protective clothing requirements for entering the RCIC cubicle 95 elev.

Standard Applicant used the provided maps and RWP to determine that Single Anti-Cs would be required.

Cue Notes Results SAT UNSAT Terminating Cue: Applicant determined that a high radiation area brief by RP will be required. Applicant also determined that single anti-c clothing will be worn.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A4 Rev 0 Page 4 of 6

RJPM-NRC-M14-A4 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A4 Rev 0 Page 5 of 6

RJPM-NRC-M14-A4 Rev. 0 OPERATOR CUE SHEET Initial Conditions:

Following a quarterly Rad Survey by RP, it was reported that there was a hissing sound coming from somewhere on the north wall of the 95 elevation RCIC cubicle. RP is unable to give any more information on this. Preparations are being made for OPs to investigate.

Initiating Cues:

The CRS has directed you to determine (1) the type of brief required, and (2) the protective clothing requirements for entry into this area.

Answer Sheet:

Indicate what type of brief must be performed:

Answer:

Indicate what (if any) protective clothing requirements are necessary:

Answer:

RJPM-NRC-M14-A4 Rev 0 Page 6 of 6

Radiological Work Permit redacted for security purposes per SUNSI checklist 2(a)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS Procedure Contains NMM REFLIB Forms: YES NO Effective Procedure Owner: Steven Brewer Governance Owner: David Moore Date

Title:

Manager, RP

Title:

Manager, Fleet RP 11/6/12 Site: PNPS Site: HQN Exception Site Site Procedure Champion Title Date*

ANO Donnie Marvel Manager, RP N/A BRP Charles Sherman Manager, RP N/A CNS Bob Bielke Manager, RP GGNS Tom Trichell Manager, RP IPEC Reid Tagliamonte Manager, RP JAF Eric Wolf Manager, RP PLP Charles Sherman Manager, RP PNPS Steven Brewer Manager, RP RBS Greg Hackett Manager, RP VY David Tkatch Manager, RP W3 John Gumnick Manager, RP N/A NP N/A N/A HQN David Moore Manager, Fleet RP Site and NMM Procedures Canceled or Superseded By This Revision Process Applicability Exclusion: All Sites:

Specific Sites: ANO BRP GGNS IPEC JAF PLP PNPS RBS VY W3 Change Statement Editorial revision to align RP-101 with revision 12 to EN-RP-105, Radiological Work Permits Step 4.0[2], change: 2.5 to: 1.5 Step 4.0[3], change: 2.5 to: 1.5 Step 5.5[2], add: unauthorized (clarify intent of step, change not related to RP-105, rev. 12)

Step 5.5[10], 6th bullet, change: 2.5 to: 1.5 Step 5.5[10], 7th bullet, change: 2.5 to: 1.5

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 2 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ........................................................................................ 3

2.0 REFERENCES

................................................................................. 3 3.0 DEFINITIONS ................................................................................... 4 4.0 RESPONSIBILITIES ......................................................................... 8 5.0 DETAILS ........................................................................................ 9 5.1 PRECAUTIONS AND LIMITATIONS ..................................................... 9 5.2 RADIOLOGICALLY CONTROLLED AREA (RCA) ACCESS CONTROL ..... 12 5.3 RADIATION AREA ACCESS CONTROL ............................................. 13 5.4 HIGH RADIATION AREA (HRA) ACCESS CONTROL ........................... 13 5.5 LOCKED HIGH RADIATION AREA ACCESS CONTROL ........................ 15 5.6 VERY HIGH RADIATION AREA ACCESS CONTROL ............................ 20 5.7 AIRBORNE RADIOACTIVITY AREA ACCESS CONTROL ....................... 23 5.8 CONTAMINATION / HIGH CONTAMINATION AREA ACCESS CONTROL... 24 5.9 MANUAL ENTRY / EXIT ................................................................. 24 5.10 ACCESS CONTROL GUARD .......................................................... 25 5.11 CONTROL AND INVENTORY OF LHRA KEYS ................................... 27 5.12 CONTROL AND INVENTORY OF VERY HIGH RADIATION AREA (VHRA)

KEYS ...................................................................................... 33 5.13 HIGH, LOCKED HIGH, AND VERY HIGH RADIATION AREA BOUNDARY VERIFICATIONS .................................................................................. 35 6.0 INTERFACES ................................................................................. 37 7.0 RECORDS ..................................................................................... 37 8.0 SITE SPECIFIC COMMITMENTS ................................................... 38 9.0 ATTACHMENTS ............................................................................ 38 ATTACHMENT 9.1 QUESTIONS FOR HIGH NOISE ENTRY .................................. 39 ATTACHMENT 9.2 MANUAL DOSE TRACKING CARD ........................................ 40 ATTACHMENT 9.3 APPROVAL FOR LOCKED HIGH RADIATION AREA DEVIATIONS 41 ATTACHMENT 9.4 VHRA ACCESS APPROVAL FORM ...................................... 42 ATTACHMENT 9.5 RESPONSIBILITIES FOR THE ACCESS CONTROL GUARD ........ 43 ATTACHMENT 9.6 LHRA / VHRA KEY LOG ................................................... 44 ATTACHMENT 9.7 SUPPLEMENTAL AREA ACCESS LOG ................................... 45 ATTACHMENT 9.8 RADIOLOGICAL AREA ACCESS KEY LOG ............................. 46

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 3 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 1.0 PURPOSE To provide detailed guidance for the entry requirements and associated controls for:

Radiologically Controlled Areas (RCA)

Radiation Areas (RA)

High Radiation Areas (HRA)

Locked High Radiation Areas (LHRA)

Very High Radiation Areas (VHRA)

Airborne Radioactivity Areas (ARA)

Contamination Area (CA) and High Contamination Area (HCA)

Access Control Guard responsibilities and guidance Access key control, issuance, transfer and inventory HRA, LHRA and VHRA Boundary Verifications

2.0 REFERENCES

2.1 GENERAL

[1] 10 CFR 20 Standards for Protection Against Radiation

[2] NRC Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants

[3] Technical Specifications for all EN units FSAR, UFSAR or USAR as applicable for each unit

[4] INPO 05-008, Guidelines for Radiological Protection at Nuclear Power Stations

[5] INPO SOER 01-01, Unplanned Radiation Exposures Recommendations 3 and 6 (LO-WTHQN-2009-00605, CA-40)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 4 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 2.0, continued 2.2 DEVELOPMENT DOCUMENTS

[1] NRC Information Notice 86-107, Licensee Alert to Reactor Cavity (Incore Shaft)

Entries

[2] NRC Information Notice 88-79, Misuse of Flashing Lights for High Radiation Area Controls

[3] INPO SOER 85-3, Excessive Personnel Radiation Exposure

[4] WANO SOER 2001-1, Unplanned Radiation Exposures

[5] SER 1-04, Continued Problems with Unplanned External Radiation Exposure 3.0 DEFINITIONS

[1] Accessible Area - Areas that can reasonably be occupied by any portion of an individuals whole body and does not require exceptional measures (e.g. the addition of ladder or scaffolding) to gain access.

[2] Airborne Radioactivity Area - An area accessible to individuals, in which the airborne radioactivity levels are equal to or greater than 30% of the DAC values listed in 10CFR20, Appendix B, Table 1, Column 3, OR, to such a degree that an individual present in the area without respiratory protective equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake (ALI) or 12 DAC-hours. (10CFR20)

[3] ALARA - (acronym for as low as is reasonably achievable) means making every reasonable effort to maintain exposures to radiation as far below the dose limits as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed material. (10CFR20)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 5 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 3.0, continued

[4] Barricade - A conspicuous obstacle such as a firmly secured rope, ribbon or rigid construction (e.g. scaffolding, chain link fence, metal, etc...) by itself or used with physical structures such as existing walls or handrails that surrounds an area to prevent inadvertent entry. (USNRC RG 8.38)

[5] Cocoon, Cocooning, Cocooned - To limit access to an area or equipment by means of erecting a physical barrier that prevents inadvertent access.

[6] Contamination Area - An area where removable surface contamination is greater than or equal to 1000dpm/100cm² beta-gamma or greater than or equal to 20 dpm/100 cm² alpha but is less than 100,000 dpm/100 cm² beta-gamma.

[7] Continuous RP Coverage - Direct radiological surveillance provided by RP with the sole responsibility for providing constant monitoring during the entire period personnel are in the work area. Continuous surveillance may be provided as follows:

(a) Locally by maintaining visual or audible contact; OR (b) Remotely by maintaining audible and telemetry, with visual contact; if available OR (c) Remotely by using stay times, time keeping, and audible contact.

[8] Direct Reading Dosimeter (DRD) - A self-reading quartz fiber, electronic, or other type of radiation measuring device used to measure exposures to x-ray or gamma radiation which can be read directly by the individual.

[9] Dosimeter of Legal Record (DLR) - A device used to determine an individuals accumulated external occupational radiation exposure including DDE, LDE and SDE.

This device is inclusive of, but not limited to, OSLDs (optically stimulated luminescent dosimeters) and TLDs (thermoluminescent dosimeters).

[10] Dose or Radiation Dose - A generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as defined in 10 CFR20.1003.

(10CFR20)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 6 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 3.0, continued

[11] Dose Margin - The difference between an individuals Administrative Dose Limit and their accumulated dose for a specific time period.

[12] Dosimetry - Individual monitoring devices issued to and worn by a single individual for assessment of dose equivalent. Dosimetry may consist of a Dosimeter of Legal Record (DLR) and/or a direct-reading dosimeter (DRD). A DRD may be an electronic alarming dosimeter (EAD) or a pocket ion chamber (PIC).

[13] Exposure Guideline - An exposure criterion established to ensure that an exposure limit is not exceeded. An exposure guideline may be exceeded with proper authorization.

[14] Exposure Limit - Maximum radiation dose permitted under specified circumstances.

[15] High Contamination Area (HCA) - An area where the majority of the area has removable surface contamination equal to or greater than 100,000 dpm/100cm2 beta-gamma, or equal to or greater than 500 dpm/100 cm2 alpha.

[16] High Radiation Area - an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 rem (1 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates.

(10CFR20)

[17] Locked High Radiation Area - An area, accessible to individuals, in which radiation levels from sources external to the body could result in an individual receiving a deep dose equivalent greater than or equal to 1 Rem (10 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm ( 12 inches) from the radiation source or from any surface that the radiation penetrates.

[18] Major Portion of the Whole Body - For purposes of external exposure; either the head, trunk, arm above the elbow, or leg above the knee. If an individuals upper arm can reasonably occupy a space, then that space is considered accessible. If only an extremity (e.g., a hand and lower arm) can be inserted into an area, then the area is not accessible.

[19] Monitoring (radiation monitoring, radiation protection monitoring) - The measurement of radiation levels, concentrations, surface area concentrations, or quantities or radioactive material and the use of the results of these measurements to evaluate potential exposures and doses. (10CFR20)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 7 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 3.0, continued

[20] Personal External Alarm (PEA) - A device attached to a workers EAD which emits a loud noise and / or vibrates to allow the worker to be aware of an EAD alarm in high-noise areas.

[21] Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent greater than or equal to 5 mRem (0.05 mSv) in one hour at 30 cm (~ 12 inches) from the radiation source or from any surface that the radiation penetrates. (10CFR20)

[22] Radiation Protection Personnel - RP Technicians, RP Supervisors, Radiological Support Staff with RPM approval and trained, qualified contractor ANSI 18.1 or ANSI 3.1 RP Technicians.

[23] Radiological Barrier - A person, rope/ribbon, or fixed structure that prevents inadvertent entry, in whole or part, of personnel into a radiological area.

[24] Radiologically Controlled Area (RCA) - An area where full radiological controls (such as, contamination monitoring and controlled access/egress) are in effect for the purpose of providing protection and/or information to the individual. At Fleet Nuclear facilities, the main RCA normally includes parts or all of the Auxiliary, Fuel, Radwaste, Reactor Buildings and the Turbine Buildings at BWRs.

[25] Remote Job Coverage - Continuous RP oversight of work activities using RMT, and consists of all of the following: telemetry dosimetry, audio communications, and if available, visual surveillance (either direct or video).

[26] Restricted Area - An area to which access is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area.

(10CFR20)

[27] Verification - After a worker completes an action requiring confirmation, a second person will independently OR concurrently and by direct observation, ensure that the action has been completed as required by this procedure.

[28] Very High Radiation Area - An area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed dose in excess 500 rads (5 grays) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter ( 3 feet) from the radiation source or 1 meter from any surface that the radiation penetrates. (10CFR20)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 8 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 4.0 RESPONSIBILITIES

[1] RPM Peer Group is responsible for the maintenance of this procedure and approval of any changes for access control to radiologically controlled areas.

[2] Radiation Protection Manager (RPM) - Responsible for:

Overall control and implementation of this procedure.

Authorizing all entries into LHRAs with general area dose rates greater than 1.5 R/hr in the actual work area.

Authorizing flashing lights as alternative LHRA controls.

Authorizing all entries into VHRAs.

Overall responsibility for control of LHRA / VHRA Keys.

[3] RP Supervision (RPS) - Responsible for:

Implementation of this procedure.

Conducting pre-job briefings for LHRA entries where general area dose rates greater than 1.5 R/hr in the actual work area.

Directing RP personnel that perform the provisions of this procedure.

Concurs that the location of LHRA / VHRA Access Control Guards ensures compliance with the LHRA / VHRA access controls in this procedure.

[4] Lead Radiation Protection Technicians (RPTs) (The terms Watch, Shift and Chief are equivalent to Lead) - Are responsible for maintaining control/possession of the key to the key storage box AND ensuring that the key storage box is locked at all times except during periods of use.

[5] RP Personnel - Are responsible for performing the provisions of this procedure.

[6] Radiation Workers - Are responsible for being cognizant of RWP requirements, access requirements, and radiological conditions in their work areas.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 9 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.0 DETAILS 5.1 PRECAUTIONS AND LIMITATIONS

[1] Persons working in areas where the access doors or gates are required to be locked (i.e. LHRA, VHRA), are responsible for ensuring that these doors/ gates are closed and locked upon entering or exiting the area unless an authorized access control guard is in use.

[2] A control device or locking mechanism SHALL NOT prevent the safe emergency egress of personnel or interfere with the safe operation of the access barrier.

Individuals should always be able to exit a posted/controlled HRA/LHRA/ VHRA.

[3] Bolted flanges, hatchways, lids, etc. are considered part of the structure and do not require additional locking mechanisms.

[4] Communication to and from the Control Room should occur prior to access when normal system operation could significantly change radiological conditions (e.g.

changes in Reactor Power, securing a system, by-passing a demineralizer, changing hydrogen water chemistry (HWC) flow rates, etc.)

[5] Due to ALARA considerations Locked High Radiation Areas (LHRA) and Very High Radiation Areas (VHRA) are NOT NORMALLY entered for the sole purpose of obtaining periodic survey data.

[6] IF temporary lead shielding is used to make a HRA/LHRA/VHRA or potential HRA/LHRA/VHRA inaccessible, THEN :

(a) The shielding should be secured in place or made not readily removable by the use of lock-wire, bolts or other fasteners that would require a tool to remove.

(b) A sign or label SHALL be placed on the shielding such as Warning - Do Not Remove - High Radiation Levels May Result OR Danger - Do Not Remove -

Very High radiation Levels May Result.

(c) A local audible and visible radiation monitor SHALL be installed to warn personnel IF temporary shielding, used to control access to fuel transfer tube or other plant areas of greater than 100 rem/hr (1Gy) is removed.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 10 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.1, continued

[7] WHEN removing or relocating installed shielding in an HRA or LHRA, THEN establish appropriate posting, barricades, and other required controls for the anticipated increase in dose rates prior to beginning work.

[8] No ladders, equipment or material shall be stored around or used in a manner that would allow access over the enclosure of a LHRA.

[9] An individual who is using manual dose tracking should not re-enter the RCA once computer access is restored until the electronic dose tracking system is updated with the data from the dose tracking card unless RP Supervision authorizes otherwise.

[10] For entry into an area that is a HRA/LHRA/VHRA and posted Hearing Protection Required area, issuance of additional monitoring such as a PEA is required.

Individuals entering a HRA/LHRA/VHRA who have demonstrated difficulty in hearing electronic dosimetry alarms require additional monitoring such as a PEA.

Use Attachment 9.1, Questions for High Noise Entry as an additional planning/briefing tool to ensure adequate controls.

[11] During an emergency, individuals qualified in radiation protection procedures OR personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant RP procedures for entry to, exit from, and work in such areas.

[12] During an emergency (e.g. medical, security, plant):

Dosimetry should still be worn.

Plant personnel may bypass the normal RCA entry / exit process. In such cases, all exposures should be recorded on an appropriate RWP as soon as possible after the event and a condition report generated to document the event.

[13] Access to "Very High Radiation Areas" SHALL normally be prohibited. In the event it is necessary to enter a "Very High Radiation Area", the entry SHALL be controlled in accordance with this procedure which complies with NRC Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants" with the exception of entries into a VHRA for the purpose of verification surveys to downgrade posting. These entries will be controlled in accordance with step 5.6[14].

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 11 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.1, continued

[14] Locks to Locked High Radiation Areas SHALL NOT be controlled by magnetic key card readers.

[15] Neutron dosimetry is required if exposure to neutron radiation is expected to result in a neutron dose greater than 10 mrem during the task. Refer to EN-RP-204, Special Monitoring Requirements.

[16] The Spent Fuel Pool (SFP) barriers are to be controlled with a locking device to prevent inadvertent removal of highly activated components from the SFP.

(a) LHRA locks or VHRA locks are used to secure items stored in the SFP.

(b) This requirement applies to other underwater storage areas, e.g. cask wash down pits, refueling equipment storage pits, etc. as directed by RP Supervision.

(c) Refer to EN-RP-108, Radiation Protection Posting for posting instructions.

[17] Identify accessible Hot Spots during pre-job briefs.

[18] Locks for LHRA rooms SHALL have an individual/unique key.

[19] Each key to a LHRA SHALL be issued with an encumbrance (large ring or key fob) to prevent loss, misplacement or inadvertent removal from the protected area.

[20] Radiological Trip Tickets are used as a tool to ensure workers are aware of the radiation protection requirements for the area they are working in. Refer to EN-RP-100, Radiation Worker Expectations for instructions on the use of Trip Tickets.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 12 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS NOTES

1. Steps in this procedure may be performed in any order, as long as regulatory and station Technical Specification requirements continue to be met.

2. Equivalent forms may be used in place of procedural attachments, however the form must contain as a minimum the information identified in the attachment.

5.2 RADIOLOGICALLY CONTROLLED AREA (RCA) ACCESS CONTROLS

[1] Personnel dosimetry requirements will be in accordance with the RWP.

[2] Specific monitoring and radiological controls for access to Radiologically Controlled Areas SHALL be listed on the appropriate RWP.

[3] Radiation workers and visitors will log into the RCA using the access control computer OR By the manual dose tracking process using Attachment 9.2, Manual Dose Tracking Card.

[4] Prior to access to any RCA, radiation workers and visitors SHALL obtain radiological information from any of the following:

Radiological Work Permits Radiological Survey Maps Plant Status Boards Direct contact with RP personnel

[5] IF access to a radiological area is necessary and is controlled with the use of a key (NOT applicable for LHRA/VHRA access),

THEN Attachment 9.8, Radiological Area Access Key Log may be used.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 13 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.3 RADIATION AREA ACCESS CONTROL

[1] Specific monitoring and radiological controls for access to Radiation Areas SHALL be listed on the appropriate RWP.

[2] As a minimum, each person entering a "Radiation Area" SHALL have:

Dosimeter of Legal Record (DLR)

Direct reading dosimeter Approved RWP White or Red Trip Ticket 5.4 HIGH RADIATION AREA (HRA) ACCESS CONTROL

[1] High Radiation Area entry points require a barricade to prevent inadvertent access.

[2] IF the barricade for an HRA must be temporarily removed, THEN, an RP Technician may maintain direct line-of-sight surveillance of the access to the HRA until the access/barrier is re-secured and verified.

[3] Specific monitoring and radiological controls for access to High Radiation Areas are listed on the appropriate RWP.

[4] As a minimum, each person entering a "High Radiation Area" SHALL have :

DLR Alarming direct reading dosimeter (Electronic Dosimeter)

Stay Time (IF greater than 500 mrem per entry expected)

Approved RWP Pre-Job briefing on radiological conditions in the area.

Red Trip Ticket

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 14 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.4, continued

[5] RP personnel control access to HRAs for all personnel. Personnel requesting such access will be equipped with one or more of the following:

(a) A radiation monitoring device which continuously indicates the radiation dose rate in the area, OR (b) A radiation-monitoring device, such as an electronic dosimeter, that has the capability to display accumulated dose and which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after dose rate levels in the area have been established and personnel have been made knowledgeable of them, OR (c) A direct reading dosimeter and a qualified representative of the RP Department with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area, and who performs periodic radiation surveillance at the frequency specified for the applicable RWP, OR (d) As specified in site Technical Specifications.

[6] WHEN a new, unanticipated HRA is discovered, PERFORM the following:

(a) Ensure all personnel (if any) are immediately evacuated from the area and direct them to report to RP.

(b) Guard or barricade the area to prohibit unauthorized access.

(c) Maintain control of the area at all times. Do not leave the area unguarded for any reason until proper procedural radiological controls have been established.

NOTE Decisions regarding operation of plant equipment and systems are the sole responsibility of licensed Operations personnel. Due to plant conditions, it may not be possible or advisable for Operations to implement requests for equipment system status change.

(d) IF a request is made to Operations to secure equipment or a system lineup to address HRA conditions, THEN confirm action has occurred and verify HRA condition has been eliminated prior to leaving area unguarded.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 15 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.4[6], continued (e) Initiate a Condition Report to document this occurrence.

5.5 LOCKED HIGH RADIATION AREA ACCESS CONTROL

[1] Barricades and blocking devices SHALL be a minimum of 6 feet in height AND Installed in a manner such that they prevent unauthorized access.

[2] No ladders, equipment or material SHALL be stored around or used in a manner that would allow unauthorized access over the enclosure.

[3] IF a change in plant layout or radiological conditions occurs which results in areas with dose rates in excess of 1000 mRem/hr at 30 cm from the source of radiation or any surface that the radiation penetrates, THEN evaluate the use of locking gates OR "cocooning" in the affected area(s) to enhance access control AND to prevent unauthorized entry.

[4] WHEN using the cocooning method, THEN a sign on the barrier must be used to inform the radiation worker of the purpose of the barrier AND of the hazards if the barrier is removed or altered to gain access to the area.

[5] All entrances or access points to Locked High Radiation Areas SHALL be locked with a distinct LHRA lock for the area or room.

[6] Entrances or access points to LHRAs SHALL remain locked EXCEPT during periods of access by personnel under an approved RWP. The following guidelines SHALL be used:

(a) Lock each access to a LHRA, OR (b) Establish an Access Control Guard to prevent unauthorized entry following the guidelines of section 5.10, OR

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 16 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.5[6], continued (c) Control access to an LHRA through the use of a barricade and red flashing light, subject to the conditions described and with RPM approval, as follows:

(1) IF no enclosure exists for purposes of locking a LHRA located within a large area such as containment, AND An enclosure can not be reasonably constructed, THEN

a. Ensure the provision for the use of flashing light is specified within the sites Technical Specifications for LHRAs.

b. Obtain written approval of the Radiation Protection Manager, or designee, using Attachment 9.3, Approval for Locked High Radiation Area Deviations, for use of a barricade and red flashing lights to control access.

c. Once approved, barricade AND Conspicuously post the area.

d. Activate the red flashing light(s) as a warning device.

e. Instruct personnel working or traversing in the vicinity of these alternative controls as to their meaning and significance.

[7] Use a ladder lock, if appropriate, to control access to an LHRA.

(a) The ladder lock, if used SHALL be a minimum of 6 feet in length as per step 5.5[1].

(b) WHEN ladder locks are used to prevent unauthorized access to LHRAs, THEN ensure that BOTH sides of the ladder are blocked to prevent unauthorized access.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 17 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.5, continued

[8] Control LHRA shielded containers such as floor plugs, rad waste cubicles, filter housings, and outside shielded liner storage containers when the following are met:

(a) Dose rates greater than1R/hr at 30 cm, AND.

(b) Contents can be accessed through the use of local installed lifting devices or readily available mobile cranes, AND.

(c) Bolting is not in place to prevent access without tools.

(d) Controls may include:

De-energize cranes with RP admin control of tag out; Use of RP-controlled locks on chain hoists; Use of RP-controlled locking nuts on plug bolts; Removal of lifting lugs used to remove plug and lugs are controlled by RP.

[9] Specific monitoring and radiological controls for access to Locked High Radiation Areas SHALL be made by RP Personnel and listed on the appropriate RWP.

[10] As a minimum, each person entering a Locked High Radiation Area SHALL have:

DLR Alarming direct reading dosimeter (Electronic Dosimeter)

Approved RWP RP Lead technician or RPS approval IF workers are in a field dose rate greater than or equal to 1R/hr, OR Worker dose is expected to be greater than 500 mrem per entry, THEN continuous RP coverage with the use of EN-RP-141, Attachment 9.1 Radiological Stay Time Verification Sheet is required.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.5[10], continued Radiation Protection Manager's approval for entry into LHRAs with general area dose rates greater than 1.5 Rem/hr in the actual work area.

Documented pre-job brief for entry, given by RP personnel. RPS performs the pre-job brief for entry into LHRAs with general area dose rates greater than 1.5 Rem/hr in the actual work area. This brief SHALL cover:

Radiological conditions in the immediate work areas using most recent survey data available AND The scope of the work to be performed.

Red Trip Ticket

[11] While LHRAs are open, the access to the LHRA SHALL be controlled in accordance with site-specific Technical Specifications.

[12] Turnover of radiological coverage by RP personnel during Locked High Radiation Area work should be avoided.

WHEN transfer of the LHRA key is required, PERFORM in accordance with Section 5.11.

[13] The following verification SHALL follow the initial check by the access control guard or RPT and be documented on Attachment 9.6, LHRA / VHRA Key Log.

(a) Upon re-establishing any LHRA boundary controls following work that required access into these areas, a second person SHALL verify the access point(s) are securely locked.

This verification SHALL consist of ensuring the locking mechanism has been replaced, where removed, and that the access point is shut AND locked.

IF the person who performed the initial check was NOT an RPT, THEN the person performing the verification SHALL be an RPT.

(1) Where keys are required to lock doors, VERIFY that the door is closed and secured/locked with a physical challenge of the door.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 19 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.5[13](a), continued (2) IF the door is locked using a padlock and chain or cable, THEN inspect the lock and chain or cable for defects, AND Physically challenge the lock.

(b) Any deficiencies found during this inspection should be:

Documented on a Condition Report AND Reported to RP Supervision immediately.

[14] WHEN a new, unanticipated LHRA is discovered, PERFORM the following:

(a) Ensure all personnel are immediately evacuated from the area and direct them to report to RP.

(b) Guard the area and prohibit unauthorized access.

(c) Maintain control of the area at all times. Do not leave the area unguarded for any reason until proper procedural radiological controls have been established.

NOTE Decisions regarding operation of plant equipment and systems are the sole responsibility of licensed Operations personnel. Due to plant conditions, it may not be possible or advisable for Operations to implement requests for equipment or system status changes.

(d) IF a request is made to Operations to secure equipment or a system lineup to address LHRA conditions, THEN confirm the action has occurred and verify the LHRA condition has been eliminated prior to leaving the area unguarded.

(e) Initiate a Condition Report to document this occurrence.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 20 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.6 VERY HIGH RADIATION AREA ACCESS CONTROL WARNING TO THE EXTENT POSSIBLE, ENTRY INTO A VHRA SHOULD BE FORBIDDEN UNLESS THERE IS A SOUND OPERATIONAL OR SAFETY REASON FOR ENTERING.

WITHOUT PROPER CONTROLS AND MONITORING, PERSONNEL ENTERING THESE AREAS COULD RECEIVE RADIATION EXPOSURE WITH SEVERE OR LIFE-THREATENING CONSEQUENCES.

[1] Barricades and blocking devices SHALL completely enclose the Very High Radiation Area sufficient to thwart undetected circumvention of the barrier.

Fencing or walls around a Very High Radiation Area should extend to the overhead and preclude anyone from climbing over the barricade.

[2] All entrances or access points to Very High Radiation Areas SHALL be locked with a unique lock AND conspicuously posted.

[3] Entrances or access points to VHRAs SHALL remain locked EXCEPT during periods of access by personnel under an approved RWP.

[4] VHRA keys are maintained under the administrative control of the Radiation Protection Manager (or designee).

[5] Each entry into a Very High Radiation Area requires the completion of EN-RP-141, Attachment 9.1, Radiological Stay Time Verification Sheet.

[6] WHEN a VHRA is not locked, THEN direct surveillance, capable of preventing unauthorized or inadvertent access, in accordance with section 5.10 SHALL be maintained.

[7] Access control to a VHRA may be made utilizing an Access Control Guard.

IF entry into a VHRA is made using an Access Control Guard, THEN the Guard SHALL comply with the requirements of section 5.10.

[8] Specific monitoring and radiological controls for access to Very High Radiation Areas SHALL be listed on the appropriate RWP.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 21 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.6, continued

[9] As a minimum, each person entering a Very High Radiation Area SHALL have a(n):

DLR Alarming direct reading dosimeter (Electronic Dosimeter).

Continuous RP coverage and Stay Time Tracking to be in accordance with EN RP-141 Stay time and dose estimate assigned for each entry.

Documented pre-job brief for entry given by an RPS.

This brief SHALL cover radiological conditions in the immediate work area using the most recent survey data available and the scope of the work to be performed.

Approved Job Specific RWP.

Approval of the Radiation Protection Manager, or designee and the on-watch Shift Manager.

Red Trip Ticket

[10] Radiation Protection Manager and Shift Manager's approval:

May only be granted following a documented evaluation of the risks and alternatives.

The approvals may be obtained by telephone.

WHEN permission is obtained, THEN document approval on Attachment 9.4, VHRA Access Approval Form.

The completed Attachment 9.4 SHALL be filed in the appropriate RWP file.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 22 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.6, continued

[11] Each Very High Radiation Area lock has a unique key which allows access to only that area.

(a) Issuance of a key for entry into a Very High Radiation Area requires the permission of the Shift Manager and Radiation Protection Manager (or designee).

(b) Complete the Key Issue section of Attachment 9.6 to obtain a VHRA key only after completing the pre-job brief.

[12] Turnover of radiological coverage by Radiation Protection personnel during Very High Radiation Area work requires RP Supervisor approval.

[13] The following verification SHALL follow the initial check by the access control guard or entrant AND SHALL be documented on Attachment 9.6.

(a) Upon re-establishing any VHRA boundary controls following work that required access into these areas an RPT SHALL verify the access point is securely locked. This verification SHALL consist of ensuring the locking mechanism has been replaced on the access point.

(b) Where keys are required to lock doors, VERIFY that the door is closed and secured/locked with a physical challenge of the door.

(c) IF door locked using a padlock and chain or cable, Inspect the lock and chain or cable for defects, AND Physically challenge the lock.

(d) Any deficiencies found during this inspection should be:

Documented on a condition report AND Reported to RP Supervision immediately.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 23 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.6, continued

[14] Entry into posted VHRA by RP Technicians for the purposes of surveying to downgrade the posting:

(a) SHALL be approved by the Radiation Protection Manager AND Shift Manager.

(b) A Pre-Job brief SHALL be conducted by RPS.

(c) The entry SHALL be performed with a minimum of two (2) RP technicians.

Each is required to have in their possession:

An electronic dosimeter, AND Two continuously indicating high-range dose rate meters, one of which SHALL have telescoping capabilities (e.g., an RO-2A AND a telescoping high range instrument).

(d) Post the area based on the existing radiological conditions upon exit from the area.

5.7 AIRBORNE RADIOACTIVITY AREA ACCESS CONTROL

[1] Specific monitoring and radiological controls for access to Airborne Radioactivity Areas SHALL be listed on the appropriate RWP.

[2] As a minimum, each person entering an Airborne Radioactivity Area SHALL have:

DLR Direct reading dosimeter Approved RWP White or Red Trip Ticket

[3] Respiratory protection devices for access to Airborne Radioactivity Areas, if required, SHALL be issued in accordance with the Respiratory Protection Program Procedures.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 24 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.8 CONTAMINATION/HIGH CONTAMINATION AREA ACCESS CONTROL

[1] Specific monitoring and radiological controls for access to Contamination Areas SHALL be made by RP Personnel and listed on the appropriate RWP.

[2] As a minimum, each person entering a Contamination Area SHOULD have:

DLR Direct reading dosimeter Protective clothing Approved RWP White or Red Trip Ticket 5.9 MANUAL ENTRY/EXIT

[1] Use Attachment 9.2, Manual Dose Tracking Card, for any of the following conditions:

To track dose when the RCA electronic access system is not available OR Any other situation specified by RP.

[2] WHEN entering the RCA using a Dose Tracking Card, Workers should complete the entry section of the dose tracking card.

Normally the card will remain at the control point while the individual is in the RCA.

[3] WHEN exiting the RCA using a Dose Tracking Card, Workers should complete the exit section of the Manual Dose Tracking Card AND Give the Direct Reading Dosimeter (DRD) to RP personnel.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 25 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.10 ACCESS CONTROL GUARD NOTE Person(s) assigned duties as an Access Control Guard for LHRAs / VHRAs wears a vest or other garment as provided by Radiation Protection to identify him / her to other workers as an Access Control Guard.

[1] RP Supervision concurs that the location for the LHRA/VHRA Access Control Guard ensures unobstructed view of the LHRA/VHRA entrance point to ensure compliance with this procedure.

[2] IF an entry into a LHRA/VHRA is made using the conditions of an Access Control Guard, THEN the Access Control Guard SHALL be stationed, OR May be one of the entrants using the following guidelines:

(a) RP personnel are not required to complete Attachment 9.5, but are expected to comply with all requirements.

(b) Non-RP personnel must read, be briefed by an RPS/RPT AND Sign a copy of Attachment 9.5, Responsibilities for the Access Control Guard.

(c) The Access Control Guard SHALL have Attachment 9.5 present with them.

(d) IF the door controlling access to the LHRA/VHRA is to be left open OR Cannot be secured/locked when entering, THEN the Access Control Guard SHALL remain stationed at the door until:

The door is secured/locked, OR They are relieved by RP personnel, OR They are relieved by another RPS/RPT briefed Access Control Guard.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 26 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.10[2], continued (e) IF the Access Control Guard enters a LHRA/VHRA, THEN the Access Control Guard SHALL:

(1) Ensure that the access is secured / locked by physically challenging the access to ensure closure and proper latching.

(2) Ensure that the access is secured / locked by physically challenging the access to ensure closure and proper latching after any additional authorized individuals enter, OR any one part of the work crew exits.

[3] The Access Control Guard SHALL prevent unauthorized entry into the LHRA / VHRA by:

(a) Obtaining verbal or written acknowledgement from the RP that the prospective entrant has permission from and is covered under the provisions of an RWP authorizing access to the LHRA / VHRA for each individual entry.

(b) IF RP acknowledges permission for entry; THEN permit entry to the area.

ELSE DENY entry and instruct the individual to contact RP.

NOTE This next step even applies when all individuals have left and plan on returning to complete or continue work.

[4] WHEN all individuals have exited the LHRA/VHRA area AND The Access Control Guard is no longer going to be present, THEN:

(a) Ensure that access door is secured / locked by physically challenging the access to ensure closure and proper latching, AND.

(b) Notify the RP Technician that the initial check is complete and door should be verified, AND.

(c) Remain at the door until the door is VERIFIED, AND (d) Document verification on Attachment 9.6.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 27 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.10, continued

[5] WHEN re-establishing any LHRA / VHRA boundary, AND Notified an initial check was completed, THEN RP personnel SHALL verify the access point(s) are secured/locked at the access point(s), AND Document verification on Attachment 9.6.

5.11 CONTROL AND INVENTORY OF LHRA KEYS

[1] LHRA Keys should be number-stamped for identification AND LHRA keys should be of a type that is not easily reproducible.

[2] The ready for issue keys SHALL be maintained in a locked box :

The locked box SHALL be kept locked by the Lead RPT except while in use; The key for the box SHOULD be in the control of the Lead RPT, shift tech, or RPS; Transfer of the key for the locked box SHALL be recorded in the appropriate RP log; IF the box key is lost, THEN control the key storage box AND Notify the RPS.

[3] Keys that are spare and not specifically assigned or ready for issue will be recorded in the RP logbook.

[4] The on-coming and off-going Lead RP Technicians perform a shift inventory:

(a) Inventory ready to issue keys as directed in step 5.11[5], AND (b) Audit the in-use Attachment 9.6, LHRA/VHRA Key Log, AND (c) Documents in shift LHRA key inventory in the turnover log.

[5] PERFORM the following unless the key lock box is secured with tamper tape, tamper seals, etc. AND the tamper closure has not been broken.

(a) Account for all keys assigned to the ready for issue inventory.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 28 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.11[5], continued (b) Ensure that Attachment 9.6 accurately reflects the current status of the keys.

(1) IF any key is not accounted for, THEN attempt to determine location of key AND VERIFY that the door controlled by the key is locked and secure.

(2) IF key cannot be located, THEN notify RP Supervision AND Record the missing key in RP log.

(3) IF it is determined that a LHRA key is lost, THEN immediately notify the Radiation Protection Manager, AND RP Supervision will make a determination as to whether LHRA door(s)/padlock(s) will be re-keyed and new key(s) issued.

(4) Initiate a Condition Report.

[6] Weekly Inventory is performed by RP Supervision or designee, who:

(a) Reviews AND initials any ready for issue Attachment(s) 9.6.

(b) Verifies proper accounting documentation for satellite issued keys.

[7] The emergency entry LHRA key issued to the Operations Shift Manager SHALL be inventoried monthly by the RPM or designee.

[8] Monthly Inventory is performed by the RP Supervision or designee.

(a) ALL LHRA keys SHALL be inventoried.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 29 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.11[8], continued (b) Ensure the current Attachment 9.6 accurately reflects the current status of the keys assigned to the satellite issue areas.

(1) IF any key is not accounted for, THEN attempt to determine location of key, AND VERIFY that the door controlled by the key is locked and secure.

(2) IF key cannot be located, THEN notify RPS AND RPM.

(3) Commence an investigation as to the key whereabouts.

(4) IF it is determined that the key is lost, THEN initiate a Condition Report, AND RPS will make a determination as to whether all LHRA doors/padlocks will be re-keyed and new keys issued.

(c) Monthly inventory is documented as completed:

In the RP Logbook OR Other media as a permanent record.

[9] LHRA Key Issue (a) Keys for LHRAs affected by the movement of irradiated fuel are controlled by AND are issued by RP Supervision or designee.

(b) Keys designated for access to LHRAs may only be issued to:

Currently trained and qualified ANSI 3.1 RP personnel in accordance with the facility qualification matrices.

Those Radiation Protection staff personnel approved in writing by the RPM.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 30 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.11[9], continued (c) The issuance of a LHRA key is recorded on EN-RP-101, Attachment 9.6, LHRA / VHRA Key Log.

(1) The key SHALL be issued with an encumbrance (e.g. bulky key ring or key fob).

(2) The key assignment sheet will record:

The key number, AND Location/Door, AND Date & time of issue, AND Issued by, AND RWP, AND Individual assigned custody of the key.

(3) Key Custodians:

a. SHALL Maintain control of the key at all times, AND

b. SHALL NOT allow unauthorized or unqualified personnel to access LHRAs controlled by the key.

(d) WHEN the RPT is responsible for multiple LHRAs (as during outages),

THEN the Lead RPT OR RPS SHALL issue a LHRA key to the RPT for a specific area, e.g. the Vapor Containment.

(1) Attachment 9.7, Supplemental Area Access Log SHALL be maintained by the RPTs assigned to that area to record:

The key custody, AND The LHRA areas accessed AND LHRA locked verification.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 31 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.11[9](c), continued (2) Transfer of keys SHALL be recorded in the Custody block of the Custody section of Attachment 9.7.

(3) The completed Attachment 9.7 SHALL be maintained with the corresponding Attachment 9.6, LHRA / VHRA Key Log.

(e) RPTs may transfer keys to one another at job sites.

(1) The new key custodian must meet the requirements of step 5.11[9](b).

(2) These transfers SHALL be reported to the Lead RPT AND Recorded on Attachment 9.7, LHRA Supplemental Area Access Log.

(3) IF transfer of LHRA key custodian responsibilities from one individual to another is necessary and Attachment 9.7 is not being utilized, THEN PERFORM the following:

a. VERIFY that the new key custodian meets the requirements of step 5.11[9](b).

b. New key custodian requests that the Lead RPT complete the entries (by phone if necessary) to accept responsibility from the current key custodian.

c. Current key custodian transfers the LHRA key to the new key custodian.

d. Lead RP, upon being notified of key transfer, from the new key custodian,

1. Completes the appropriate entries in the Key Transfer column on Attachment 9.6 AND

2. MARK N/A in the appropriate blank in the columns on Attachment 9.6, if used, for Key Return and Verification.

(f) Keys should be returned to the issue point at end of shift or assignment.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 32 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.11[9], continued (g) WHEN the key is returned, THEN Attachment 9.6 will be used to record:

That the area(s) is / are re-locked including date and time, AND The Initials, date and time of the keys return, AND Verification of key return by Lead RP Technician or shift tech, or by RPS, RPM, etc.

[10] Emergency Entry into Locked High Radiation Areas NOTE This provision allows for the emergency access to LHRA areas.

(a) The Operations Shift Manager has control of an Emergency Use Only LHRA key.

(b) The use of LHRA keys, not in the possession of the RP Department, SHALL be limited to emergency entries into a LHRA AND SHALL be maintained by the Operations Shift Manager.

(c) In the event that use of a LHRA key is needed for access, contact with Lead RPT SHALL be made as soon as possible following the entry to the area.

(d) In the case of emergency entry being required and at the direction of the Operations Shift Manager, Operations personnel may utilize the emergency key to gain access to the LHRA.

(e) IF the LHRA key controlled by the Operations Shift Manager is used, THEN NOTIFY RP Supervision AND NOTIFY the RPM, AND Initiate a Condition Report.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 33 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.12 CONTROL AND INVENTORY OF VERY HIGH RADIATION AREA (VHRA) KEYS

[1] VHRA keys will be uniquely identified (not keyed the same as any other type of locks used in the plant), AND VHRA keys will be of a type that is not easily reproducible.

[2] Keys and padlocks used for control of VHRAs SHALL be administered through the RPM or designee.

[3] VHRA Key Inventory Requirements (a) The RPM or designee performs a VHRA key inventory monthly.

(b) The VHRA key inventory includes an audit of the current Attachment 9.6.

(c) This inventory SHALL account for all keys.

(d) Ensure Attachment 9.6 accurately reflects the current status of the keys.

(1) IF any key is not accounted for, THEN attempt to determine the location of key, AND VERIFY the area controlled by the missing key is locked and secure.

(2) IF key cannot be located, THEN notify all RP Personnel.

(3) Commence an investigation as to the key whereabouts.

(4) IF it is determined that the key is lost, THEN the RPM will make a determination as to whether all VHRA doors/padlocks will be re-keyed and new keys issued.

(5) Initiate a Condition Report.

(e) Monthly inventory is documented as completed:

In the RP Logbook OR Other media as a permanent record.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 34 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.12, continued

[4] VHRA Key Issue (a) VHRA keys will be issued to RP Personnel only, as outlined in step 5.11[9](a),

with RPM approval, for specific use.

(b) The RPM AND the Operations Shift Manager SHALL approve the issuance of VHRA keys by completing Attachment 9.4, VHRA Access Approval Form.

(c) The issuance of VHRA keys SHALL be for specifically briefed activities.

(d) Issuance of VHRA keys SHALL be recorded on the appropriate sections of Attachment 9.6.

(e) VHRA keys should not be transferred from one custodian to another while in use.

IF a transfer of VHRA keys is necessary, PERFORM the following:

(1) New Key Custodian MUST be approved by RPM and briefed by RPS.

a. Ensure the appropriate entries in the Key Issue columns on Attachment 9.6 are made by notifying the RPM or designee that they accept responsibility from the current key custodian.

b. This notification may be made by phone or in person.

(2) Current Key Custodian, upon approval of RPM, transfers the VHRA key to the new key custodian.

(3) RPM or designee,: upon being notified of key transfer from the new key custodian, Completes the appropriate entries in the Key Transfer column on Attachment 9.6 AND MARK N/A in the appropriate blank in the columns on Attachment 9.6 if used, for Key Return and Verification.

(f) Upon the return of the keys the appropriate sections of Attachment 9.6 will be completed following the requirements for verifications of step 5.6[13].

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 35 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.13 HIGH, LOCKED HIGH AND VERY HIGH RADIATION AREA BOUNDARY VERIFICATIONS NOTE Increasing the periodicity of boundary verifications should be considered during outages.

[1] High Radiation Area (HRA) Boundary verifications (a) SHALL be, at a minimum and as applicable, performed weekly, AND.

(b) SHALL consist of the following:

Ensuring the boundary is continuous to avoid unauthorized entry.

Ensuring the swing gate, if used, is functional.

Verifying the swing gate alarm, if used, is operational.

Verifying the postings are in accordance with EN-RP-108, Radiation Protection Posting.

Verifying that unauthorized access is not inadvertently created by the positioning or placement of piping, conduit, tool boxes, cable trays, ladders, scaffolding, etc., that could facilitate access to the area.

Documenting that the check is completed in the RP Logbook or other media.

[2] Locked High Radiation Area (LHRA) and Very High Radiation Area (VHRA) Boundary verifications (a) SHALL be, as a minimum and as applicable, performed weekly AND (b) SHALL consist of:

Ensuring the locking mechanism is in place.

Physically challenging the access entrance.

IF padlock with chain/cable is used, THEN ensure there is no slack in the chain/cable that would allow unauthorized entry.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 36 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.13[2](b), continued IF cocooning or barricade is used, THEN ensure boundary is continuous and secure to avoid unauthorized entry.

WHEN using the cocooning method, THEN a sign on the barrier must be used to inform the radiation worker of the purpose of the barrier AND The hazards present if the barrier is removed or altered to gain access to the area.

VERIFY that the door or area is posted in accordance with EN-RP-108, Radiation Protection Posting.

VERIFY that unauthorized access is not inadvertently created by the positioning or placement of piping, conduit, tool boxes, cable trays, ladders, scaffolding, etc., that could facilitate access to the area.

Document that the check is completed in RP Logbook or other media.

[3] Door and gate integrity inspections and alarm (audio/visual) inspections (a) SHALL be, as a minimum, conducted semi-annually AND (b) SHALL, at a minimum and as applicable, consist of the following:

Verify the locking mechanism is in place.

Verify the latch is in alignment with AND sits well in the frames keyway.

Ensure locking cylinders and striker plates are intact.

Verify the door or gates framework is intact.

Verify handles are in working order.

Verify the auto closure device performs satisfactory.

Documentation that inspection is complete in RP Logbook (or other media).

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 37 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 5.13, continued

[4] IF defects or deficiencies are found, THEN Control the area AND Notify RPS.

6.0 INTERFACES

[1] EN-RP-100, Radiation Worker Expectations

[2] EN-RP-108, Radiation Protection Posting

[3] EN-RP-105, Radiological Work Permits

[4] EN-RP-109, Hot Spot Program

[5] EN-RP-141 Job Coverage

[6] EN-RP-204, Special Monitoring Requirements

[7] EN-RP-501, Respiratory Protection Program

[8] EN-IS-115, Hearing Conservation Program 7.0 RECORDS

[1] Attachment 9.2, Manual Dose Tracking Card

[2] Attachment 9.3, Approval For Locked High Radiation Area Deviations

[3] Attachment 9.4, VHRA Access Approval Form

[4] Attachment 9.6, LHRA / VHRA Key Log

[5] Attachment 9.7, Supplemental Area Access Log

[6] Attachment 9.8, Radiological Area Access Key Log

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 38 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS 8.0 SITE SPECIFIC COMMITMENTS Step Site Document Commitment Number or Reference All W3 10 CFR 20 P-675, P-678, P-705, P-21823, P-22156, P-22157 All W3 IEC 76-03 P-2481 All W3 TS 6.12 P-2808 All W3 FSAR P-6368 All W3 Regulatory Guide 8.8 P-11726 All W3 IEN 86-044 P-12725 All W3 SOER 85-03 P-16214 5.1 [14] W3 LER 90-016 P-17611 9.0 ATTACHMENTS 9.1 Questions for High Noise Entry 9.2 Manual Dose Tracking Card 9.3 Approval for Locked High Radiation Area Deviations 9.4 VHRA Access Approval Form 9.5 Responsibilities for the Access Control Guard 9.6 LHRA / VHRA Key Log 9.7 LHRA Supplemental Area Access Log 9.8 Radiological Area Access Key Log

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 39 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.1 QUESTIONS FOR HIGH NOISE ENTRY Sheet 1 of 1 Consider the following questions to determine if additional measures should be taken prior to entry into HRA/LHRA/VHRA, Hearing Protection Required Areas, or for individuals who have demonstrated the inability to hear electronic dosimeter alarms. Check the box which applies.

If the answer to question 1 or 2 is Yes an additional monitoring device such as a PEA is Required.

If the answer is Yes to any of the other questions consider the need for additional monitoring devices such as PEAs, telemetry dosimetry, or continuous RP coverage.

1. Is the entry into a HRA/LHRA/VHRA and Hearing Protection Required Area? Yes No

2. Is the entry into a HRA/LHRA/VHRA and has the individual demonstrated the inability to hear an electronic dosimeter alarm? Yes No

3. Are noise levels in the work area or along the transit path such that they could prevent an individual from hearing the dosimeter alarm? Yes No

4. Will electric or air powered tools be used during the course of the job that will generate noise levels high enough to prevent someone from hearing the electronic dosimeter alarm? Yes No

5. Would it be possible during the course of the entry for the individual to receive an accumulated dose of greater than 30 mrem based on RWP EAD setpoints? Yes No

6. Will hearing protection be worn by the individual which might impair the ability to hear the electronic dosimeter alarm? Yes No

7. Will the position of the body, position of the electronic dosimeter, or the amount of protective clothing have an impact on the ability of the individual to hear the dosimeter alarm? Yes No

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 40 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.2 MANUAL DOSE T RACKING CARD Sheet 1 of 1 Name / Badge Number: DLR # Issue Date: Entry Dose Margin:

Entry Section RWP # Date Workers Time In DRD Number DRD Dose In Initials*

• Initials indicate that the individual has read, understands and will comply with the RWP and/or entry/work requirements; they also indicate that the individual currently meets all training requirements for access to the RWP area.

Exit Section Time Out Dose Out New Margin Computer Updated (RP initial)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 41 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.3 APPROVAL FOR LOCKED HIGH RADIATION AREA DEVIATIONS Sheet 1 of 1 This form is to be used for approval of Locked High Radiation Area deviations. Exceptions must comply with the controls required by 10CFR20, and station Technical Specifications.

Description of Deviation Requested: ___________________________________________

Approval for the use of barricade and red flashing light for controlling access to a LHRA.

Other (specify) _________________________________________________________

Location of Requested Deviation: ___________________________________________

Justification and/or comments (including why physical barrier cant be erected):_________

Deviation Requested by: ____________________________ Date: _______________

Deviation Approved by: _____________________________ Date: _______________

Radiation Protection Manager (or designee)

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 42 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.4 VHRA ACCESS APPROVAL FORM Sheet 1 of 1 Location/Description of Area: ________________________________________________

Date of Request: __________________ Applicable RWP #_________________________

Requestor: __________________________________ Department __________________

Access Review and Approval (all items must be completed and checked off):

The RWP instructions are adequate to address radiological conditions expected for this work.

The ALARA Review (if required by procedure) for this work is complete and dose estimate calculated if necessary.

A briefing has been performed for all personnel involved with this work.

A documented evaluation of the risks and alternatives associated with this entry has been performed and is adequate.

Approval: _____________________________ Date: ___________ Time: ___________

Radiation Protection Manager or designee Approval: _____________________________ Date: ___________ Time: ___________

Operations Shift Manager The RP Manager, or designee, granted approval for access to VHRA, by telephone.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 43 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.5 RESPONSIBILITIES FOR THE ACCESS CONTROL G UARD Sheet 1 of 1 (To be completed by all Personnel except qualified RPTs)

Date: ________________ Time: __________________

Area to be attended (Unit, Bldg, Elevation, Room #, etc.):

Individual signing as the Access Control Guard reads, understands and accepts the following responsibilities:

1. Have the signed Attachment 9.5 present with them while performing duties as an Access Control Guard.

2. Wears the vest / garment as provided by RP, to identify the individual as a LHRA / VHRA Guard.

3. Has received a briefing on the LHRA/VHRA boundary that he/she is guarding.

4. If the door controlling access to the LHRA/VHRA is to be left open OR cannot be secured/locked when entering, the Access Control Guard SHALL remain stationed with a direct line-of-sight and control at the door until:

Access/barrier is secured/locked and verified by RP Relieved by RP qualified personnel Relieved by another RPS/RPT briefed Access Control Guard

5. If the Access Control Guard enters the LHRA/VHRA, the Access Control Guard SHALL:

Ensure that the access is secured / locked by physically challenging the access to ensure closure and proper latching.

Ensure that the access is secured/locked by physically challenging the access to ensure closure and proper latching after any additional authorized individuals enter or any part of the work crew exits.

6. Prevent unauthorized entry into the LHRA / VHRA by performing the following actions for any individual requesting access to the area:

Obtain verbal or written acknowledgement from the RP that the prospective entrant has permission from and is covered under the provisions of an RWP authorizing access to the LHRA / VHRA for each individual entry.

IF RP acknowledges permission for entry; THEN permit entry to the area. Otherwise DENY entry and instruct the individual to contact RP.

7. Ensure that personnel are able to exit the LHRA / VHRA at any time, and are not prevented from leaving the area by a locked or obstructed access.

8. If, at any point, you do not believe that access to the LHRA / VHRA is being adequately controlled then contact RP personnel immediately

9. When all individuals have exited the LHRA/VHRA area AND the Access Control Guard is no longer going to be present:

(a) Ensure that access door is secured / locked by physically challenging the access/barrier and ensuring proper latching (b) Notify the RP Technician that the initial check is complete and door(s) should be verified.

(c) Remain at the door until the door is VERIFIED secured/locked by RP Personnel

10. Indicate completion of the initial check on Attachment 9.6.

Access Control Guard Name: ____________________________________

Access Control Guard Signature ____________________________________ Date: ________________

RP Technician Performing Brief: ____________________________________ Date: ________________

NUCLEAR MANAGEMENT NON-QUALITY RELATED EN-RP-101 REV. 7 MANUAL INFORMATIONAL USE Page 44 of 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.6 LHRA / VHRA KEY LOG Sheet 1 of 1 KEY ISSUE KEY RETURN AREA LOCKED KEY TRANSFER Key # Location/Door Key

• Key Custodian Date Key Returned By: ** Area Locked By: Transfer to:

Accessed Issued (Print Name) Time Init/Date/Time Init/Date/Time (Print Name) by: (Signature) Issued Key Verified Returned Area Verified Locked by: Init/Date/Time RWP Init/Date/Time Init/Date/Time RP Supervisor (or designee) Review: __________________________________ Date: ________________

• If key is issued to an area and Attach. 9.7, LHRA Supplemental Access Area Access Log is being utilized, then indicate Attachment 9.7 for Area Locked column.

• • If an RPT is down posting an area and a lock is no longer required, then the appropriate blank in this column Area Locked By should be marked Down posted then Initials/Date/Time placed in the Key Returned By block.

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 45 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.7 SUPPLEMENTAL AREA ACCESS LOG Sheet 1 of 1 Key #_____________ Location_____________________________

Custody Area Entry Verification Key Custodian Date Door / Area Area Locked by: Verified Locked Print Name Time Accessed (Print Name) by:

Signature Init/Date/Time (Print Name)

Init/Date/Time RP Supervisor (or designee) Review: ______________________________ Date: _______________

NUCLEAR NON-QUALITY RELATED EN-RP-101 REV. 7 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 46 OF 46 ACCESS CONTROL FOR RADIOLOGICALLY CONTROLLED AREAS ATTACHMENT 9.8 RADIOLOGICAL AREA ACCESS KEY LOG Sheet 1 of 1

• Key Custodian Date Key Returned By:

Key # Location/Door Accessed (Print Name) Time Print/Date/Time (Signature) Issued

• FORM IS NOT APPLIED OR ALLOWED FOR KEY ISSUING ACCESS TO LHRA OR VHRA

RJPM-NRC-M14-A5 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Calculate Heatup Rate and Time to Uncover Core During Loss of SDC Event OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 2, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 6

RJPM-NRC-M14-A5 Rev. 0 EXAMINER INFO SHEET Task Standard: Heatup Rate is determined to be between 9 and 9.2°F, and Time to TAF is determined to be between 104.5 and 106.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Synopsis: The plant was preparing for refueling when a loss of shutdown cooling occurs.

This task will have the applicant choose and use Thermal Hydraulic Curves from OSP-0037, Shutdown Operations Protection Plan to determine (1) the Heat Up Rate, and (2) the Time to Top of Active Fuel for given conditions.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

Determine (1) the Heatup Rate, and (2) the Time to Top of Active Fuel.

3) Initial Conditions:

The plant is in Mode 5; it has been shutdown for 6 days, Preparations for refueling have just been completed.

The Reactor cavity is currently flooded, with water temperature at 130°F.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A5 Rev 0 Page 2 of 6

RJPM-NRC-M14-A5 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Calculate Heatup Rate and Time to 400077004001 G 2.1.25 4.2 Uncover Core During Loss of SDC Event

REFERENCES:

APPLICABLE OBJECTIVES OSP-0037, Rev 30 RLP-STM-0053, Obj 3, 11 REQUIRED MATERIALS:

OSP-0037, Rev 30 SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup 2.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Heatup Rate is determined to be between 9 and 9.2°F, and Time to TAF is determined to be between 104.5 and 106.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

RJPM-NRC-M14-A5 Rev 0 Page 3 of 6

RJPM-NRC-M14-A5 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1. Determine Heat Up Rate Standard Applicant determined that the Heatup rate is 9°F/hr (acceptable range is between 9 and 9.2), using the correct graph on Attachment 9, page 14 of 32.

Cue Notes Results SAT UNSAT

2. Procedure Step: 2. Determine Time to TAF Standard Applicant determined that the time to TAF is between 104.5 and 106.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> using the correct graph on Attachment 9, page 15 of 32.

Cue Notes The applicant must use the multiplier (0.97) to achieve correct answer.

Results SAT UNSAT Terminating Cue: Heatup Rate is determined to be between 9 and 9.2°F, and Time to TAF is determined to be between 104.5 and 106.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A5 Rev 0 Page 4 of 6

RJPM-NRC-M14-A5 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A5 Rev 0 Page 5 of 6

RJPM-NRC-M14-A5 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

Determine (1) the Heatup Rate, and (2) the Time to Top of Active Fuel.

Initial Conditions:

The plant is in Mode 5; it has been shutdown for 6 days, Preparations for refueling have just been completed.

The Reactor cavity is currently flooded, with water temperature at 130°F.

Answer Sheet:

1) Heatup Rate = ______________

2) Time to Top of Active Fuel = ________________

RJPM-NRC-M14-A5 Rev 0 Page 6 of 6

REFERENCE USE G12.1.28 RIVER BEND STATION STATION OPERATING MANUAL

• OPERATION SECTION PROCEDURE

• SHUTDOWN OPERATIONS PROTECTION PLAN (SOPP)

PROCEDURE NUMBER: *OSP-0037 REVISION NUMBER: *030 Effective Date:

• NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

REFERENCE USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER OSP-0037 REV - 030 PAGE 1 OF 62

REFERENCE USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE.................................................................................................................................3 2 REFERENCES .........................................................................................................................3 3 DEFINITIONS .........................................................................................................................5 4 PROCEDURE...........................................................................................................................9 5 CONTINGENCY PLANS......................................................................................................16 6 RECORDS..............................................................................................................................17 ATTACHMENT 1 - SHUTDOWN COOLING FUNCTION COLOR STATES ........................18 ATTACHMENT 2 - INVENTORY CONTROL FUNCTION COLOR STATES .......................20 ATTACHMENT 3 - AC POWER CONTROL FUNCTION COLOR STATES..........................21 ATTACHMENT 4 - FUEL POOL COOLING FUNCTION COLOR STATES..........................22 ATTACHMENT 5 - CONTAINMENT CONTROL FUNCTION COLOR STATES .................23 ATTACHMENT 6 - FUEL BUILDING VENTILATION FUNCTION COLOR STATES ........24 ATTACHMENT 7 - REACTIVITY CONTROL FUNCTION COLOR STATES ......................25 ATTACHMENT 8 - FIRE FUNCTION COLOR STATES..........................................................26 ATTACHMENT 9 - THERMAL HYDRAULIC CURVES .........................................................27 ATTACHMENT 10 - EOI CORPORATE OUTAGE MANAGEMENT NUCLEAR SAFETY PHILOSOPHY..............................................................................................59 ATTACHMENT 11 - APPROVAL FOR DEPARTURE FROM THE REQUIREMENTS OF THE SHUTDOWN OPERATIONS PROTECTION PLAN..................................62 OSP-0037 REV - 030 PAGE 2 OF 62

REFERENCE USE 1 PURPOSE This procedure provides guidelines for Operations and Outage Management personnel to evaluate the availability of plant equipment required to meet the EOI Corporate Outage Management Nuclear Safety Philosophy (Attachment 10).

This procedure is intended for use when the plant is in Mode 4 or Mode 5 during scheduled, forced (unscheduled), and refueling outages.

This procedure reinforces the expectation that the Operations Shift Managers-Outage maintains overall responsibility for control of the key shutdown safety functions. Activities with the potential to challenge decay heat removal, lower reactor coolant system inventory, result in a loss of electrical power, or affect reactivity such as fuel or control rod movement are overseen by the shift manager. In addition, the Operations Shift Managers concur with the release and closure of outage and system work windows that have an impact on the shutdown safety functions.

(SOER 09-01, Recommendation 3, SOER 94-01 Recommendation 3a).

SOER 09-1 should be referred to prior to making any changes to this procedure to ensure the requirements of the SOER continue to be met.

2 REFERENCES 2.1 ADM-0096, Risk Management Program Implementation and On-Line Maintenance Risk Assessment 2.2 AOP-0004, Loss of Offsite Power 2.3 AOP-0027, Fuel Handling Mishaps 2.4 AOP-0050, Station Blackout 2.5 AOP-0051, Loss of Decay Heat Removal 2.6 GOP-0002, Power Decrease/Plant Shutdown 2.7 OSP-0034, Control of Obstructions for Primary Containment/Fuel Building Operability 2.8 OSP-0041, Alternate Decay Heat Removal 2.9 SOP-0003, Reactor Recirculation System 2.10 SOP-0031, Residual Heat Removal System operating procedure.

2.11 SOP-0091, Fuel Pool Cooling and Cleanup system operating procedure OSP-0037 REV - 030 PAGE 3 OF 62

REFERENCE USE 2.12 SOP-0140, Suppression Pool Cleanup And Alternate Decay Heat Removal 2.13 EN-OU-108, Shutdown Safety Management Program (SSMP) 2.14 NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

2.15 INPO, INPO Outage Management Guidelines.

2.16 NSAC-175L, Safety Assessment of BWR Risk During Operations (Grand Gulf).

2.17 EPRI Draft Report, BWR Generic Risk Management Guidelines.

2.18 A-15693, There will be a Standing Team for Outage Risk Oversight during future refueling outages.

2.19 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants.

2.20 EN-DC-401, Configuration Risk Management Program.

2.21 G13.18.14.0*189 Time to Boil, Heat Up Rate, and Time to Top of Active Fuel Curves Accounting for Power Uprate 2.22 G13.18.12.3*171, Shutdown Safety Function Defense in Depth Color Codes 2.23 ER-2005-0168-009, RF13 Decay Heat Evaluations 2.24 NUREG 0612 Control of Heavy Loads at Nuclear Power Plants 2.25 INPO SOER 09-1, Shutdown Safety 2.26 CR-RBS-2009-3895 2.27 SOER 94-1 Recommendation 3 Non-Conservative Decision Making OSP-0037 REV - 030 PAGE 4 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 1 OF 32 THERMAL HYDRAULIC CURVES Before fuel shuffle TIME TO 200F (Mins)

OSP-0037 REV - 030 PAGE 27 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 2 OF 32 THERMAL HYDRAULIC CURVES TAF (HRS)

OSP-0037 REV - 030 PAGE 28 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 3 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE (DEGREES/HR)

OSP-0037 REV - 030 PAGE 29 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 4 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200F (HRS)

OSP-0037 REV - 030 PAGE 30 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 5 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT 85 INCHES 120 110 100 90 80 70 60 Before Fuel Shuffle 50 40 HEAT UP RATE (DEGREES/HR) 30 20 10 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 31 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 6 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT 85 INCHES AND RX WATER TEMPERATURE AT 110 DEGREES 18 17 16 15 14 13 12 11 10 9

TAF (HRS) 8 Before Fuel Shuffle 7

6 Mulitiplier Temp 5 0.86 170 0.90 150 4 0.95 130 3 1.00 110 1.05 90 2

1 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 32 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 7 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F CURVE BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT MAIN STEAM LINES AND RX WATER TEMPERATURE AT 110 DEGREES 4

3 2 Before Fuel Shuffle TIME TO 200F (HRS)

Multiplier Temp 0.33 170 1

0.55 150 0.77 130 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 33 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 8 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE BEFORE FUEL SHUFFLE FOR RX WATER AT THE MAIN STEAM LINES 120 110 100 90 80 70 60 50 Before Fuel Shuffle 40 HEAT UP RATE (DEGREES PER HOUR) 30 20 10 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 34 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 9 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT MAIN STEAM LINES AND RX WATER TEMPERATURE AT 110 DEGREES 20 15 Before Fuel Shuffle TAF (HRS) 10 Multiplier Temp 0.86 170 0.91 150 5 0.95 130 1.00 110 1.05 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 35 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 10 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F CURVE BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT FLANGE AND RX WATER TEMPERATURE AT 110 DEGREES 5

4 3

Before Fuel Shuffle 2

TIME TO 200F (HRS)

Multiplier Temp 0.33 170 0.55 150 1 0.77 130 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 36 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 11 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT FLANGE 100 90 80 70 60 50 40 Before Fuel Shuffle HEAT UP RATE (DEGREES/HR) 30 20 10 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 37 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 12 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL BEFORE FUEL SHUFFLE FOR RX WATER LEVEL AT FLANGE AND RX WATER TEMPERATURE AT 110 DEGREES 30 25 20 Before fuel Shuffle 15 TAF (HRS)

Multiplier Temp 10 0.88 170 0.92 150 0.96 130 1.00 110 5 1.04 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 38 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 13 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F CURVE BEFORE FUEL SHUFFLE FOR FLOODED CONDITIONS AND RX WATER TEMPERATURE AT 110 DEGREES 25 20 15 Before Fuel Shuffle 10 TIME TO 200F (HRS)

Multiplier Temp 0.33 170 0.55 150 5 0.77 130 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 39 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 14 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE BEFORE FUEL SHUFFLE FOR FLOODED CONDITIONS 20 15 10 Before Fuel Shuffle HEAT UP RATE (DEGREES/HR) 5 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 40 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 15 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL BEFORE FUEL SHUFFLE FOR FLOODED CONDITIONS AND RX WATER TEMPERATURE AT 110 DEGREES 300 250 200 Before Fuel Shuffle 150 TAF (HRS)

Multiplier Temp 100 0.92 170 0.95 150 0.97 130 1.00 110 50 1.02 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 41 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 16 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200F (HRS)

OSP-0037 REV - 030 PAGE 42 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 17 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE (DEGREES/HR)

OSP-0037 REV - 030 PAGE 43 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 18 OF 32 THERMAL HYDRAULIC CURVES TAF (HRS)

OSP-0037 REV - 030 PAGE 44 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 19 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200F (HRS)

OSP-0037 REV - 030 PAGE 45 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 20 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE AFTER FUEL SHUFFLE WITH RX WATER LEVEL AT 85 INCHES 60 50 40 30 After Fuel Shuffle 20 HEAT UP RATE (DEGREES/HR) 10 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 46 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 21 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL AFTER FUEL SHUFFLE FOR RX WATER LEVEL AT 85 INCHES AND RX WATER TEMPERATURE AT 110 DEGREES 25 20 15 TAF (HRS)

After Fuel Shuffle 10 Mulitiplier Temp 0.86 170 5 0.90 150 0.95 130 1.00 110 1.05 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 47 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 22 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F AFTER FUEL SHUFFLE FOR RX WATER LEVEL AT THE MAIN STEAM LINES AND RX WATER TEMPERATURE AT 110 DEGREES 5

4 3

After Fuel Shuffle 2

TIME TO 200F (HRS)

Multiplier Temp 0.33 170 0.55 150 1 0.77 130 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 48 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 23 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE FOR RX WATER AT THE MAIN STEAM LINES AFTER FUEL SHUFFLE 60 50 40 After Fuel Shuffle 30 20 HEAT UP RATE (DEGREES/HR) 10 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 49 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 24 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL AFTER FUEL SHUFFLE AT MAIN STEAM LINES AND RX WATER TEMPERATURE AT 110 DEGREES 30 25 20 After Fuel Shuffle 15 TAF (HRS) 10 Mulitiplier Temp 0.86 170 0.91 150 0.95 130 5

1.00 110 1.05 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 50 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 25 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F AFTER FUEL SHUFFLE FOR RX WATER LEVEL AT THE FLANGE AND RX WATER TEMPERATURE AT 110 DEGREES 6

5 4

3 After Fuel Shuffle TIME TO 200F (HRS) 2 Multiplier Temp 0.33 170 0.55 150 0.77 130 1 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 51 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 26 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE FOR WATER LEVEL AT FLANGE AFTER FUEL SHUFFLE 45 40 35 30 25 After Fuel Shuffle 20 15 HEAT UP RATE (DEGREES/HR) 10 5

0 0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 52 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 27 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL AFTER FUEL SHUFFLE FOR RX WATER LEVEL AT FLANGE AND RX WATER TEMPERATURE AT 110 DEGREES 40 35 30 25 After Fuel Shuffle 20 TAF (HRS) 15 Multiplier Temp 0.88 170 10 0.92 150 0.96 130 1.00 110 1.04 90 5

0 0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 53 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 28 OF 32 THERMAL HYDRAULIC CURVES TIME TO 200 F AFTER FUEL SHUFFLE FOR FLOODED CONDITIONS AND RX WATER TEMPERATURE AT 110 DEGREES 35 30 25 20 After Fuel Shuffle 15 TIME TO 200F (HRS)

Multiplier Temp 10 0.33 170 0.55 150 0.77 130 5 1.00 110 1.23 90 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 54 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 29 OF 32 THERMAL HYDRAULIC CURVES HEAT UP RATE FOR FLOODED CONDITIONS AFTER FUEL SHUFFLE 8

7 6

5 After Fuel Shuffle 4

3 HEAT UP RATE (DEGREES/ HR) 2 1

0 0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 55 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 30 OF 32 THERMAL HYDRAULIC CURVES TIME TO TOP OF ACTIVE FUEL AFTER FUEL SHUFFLE FOR FLOODED CONDITIONS AND RX WATER TEMPERATURE AT 110 DEGREES 400 350 300 250 After Fuel Shuffle 200 TAF (HRS) 150 Multiplier Temp 0.92 170 0.95 150 100 0.97 130 1.00 110 1.02 90 50 0

0 5 10 15 20 25 30 35 40 TIME AFTER SHUTDOWN (DAYS)

OSP-0037 REV - 030 PAGE 56 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 31 OF 32 THERMAL HYDRAULIC CURVES BEFORE FUEL SHUFFLE Days After Decay Heat Up Heat Up Heat Heat Time To Time To Time To Time To Time To Time To Time To Time To Shutdown Heat Rate Rate Up Up Mode Mode Mode Mode Top of Top of Top of Top of Flooded Flange Rate Rate Change Change Change Change Active Active Active Active MSL 85 in. Flooded Flange MSL 85 in. Fuel Fuel Fuel Fuel 85 Flooded Flange MSL in.

Days MBtu/hr (F/HR) (F/HR) (F/HR) (F/HR) hrs hrs hrs hrs hrs hrs hrs hrs 1 67.4 16.19 87.59 109.54 110.61 5.56 1.03 .82 .81 60.78 6.79 4.34 4.24 2 55.1 13.24 71.61 89.55 90.43 6.8 1.26 1.01 1.00 74.35 8.30 5.30 5.19 3 48.3 11.6 62.77 78.5 79.27 7.76 1.43 1.15 1.14 84.81 9.47 6.05 5.92 4 43.7 10.5 56.79 71.02 71.72 8.57 1.58 1.27 1.25 93.74 10.47 6.69 6.54 5 40.2 9.66 52.24 65.33 65.98 9.32 1.72 1.38 1.36 101.9 11.38 7.27 7.11 6 37.8 9.08 49.12 61.43 62.04 9.91 1.83 1.47 1.45 108.37 12.1 7.73 7.56 7 35.4 8.5 46.01 57.53 58.1 10.58 1.96 1.56 1.55 115.72 12.92 8.26 8.07 8 33.7 8.1 43.8 54.77 55.31 11.12 2.05 1.64 1.63 121.56 13.58 8.67 8.48 10 30.7 7.37 39.9 49.89 50.38 12.2 2.26 1.8 1.79 133.44 14.9 9.52 9.31 20 23.2 5.57 30.15 37.7 38.08 16.15 2.99 2.39 2.36 176.57 19.72 12.6 12.32 40 17.2 4.13 22.35 27.95 28.23 21.78 4.03 3.22 3.19 238.17 26.6 16.99 16.62 OSP-0037 REV - 030 PAGE 57 OF 62

REFERENCE USE ATTACHMENT 9 PAGE 32 OF 32 THERMAL HYDRAULIC CURVES AFTER FUEL SHUFFLE Time To Time To Time To Time To Time To Time To Time To Time To Heat Up Heat Up Heat Up Heat Up Top of Top of Top of Top of Days After Decay Mode Mode Mode Mode Rate Rate Rate Rate Active Active Active Active Shutdown Heat Change Change Change Change Flooded Flange MSL 85 in. Fuel Fuel Fuel Fuel Flooded Flange MSL 85 in.

Flooded Flange MSL 85 in.

Days MBtu/hr (F/hr) (F/hr) (F/hr) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) 4 31.09 7.47 40.41 50.53 51.03 12.05 2.23 1.78 1.76 131.74 14.71 9.40 9.19 5 28.60 6.87 37.17 46.49 46.94 13.10 2.42 1.94 1.92 143.21 16.00 10.22 9.99 6 26.90 6.46 34.95 43.71 44.14 13.93 2.57 2.06 2.04 152.31 17.01 10.87 10.63 7 25.19 6.05 32.73 40.94 41.34 14.87 2.75 2.20 2.18 162.63 18.16 11.60 11.35 8 23.98 5.76 31.16 38.97 39.35 15.63 2.89 2.31 2.29 170.84 19.08 12.19 11.92 10 21.84 5.25 28.39 35.50 35.85 17.15 3.17 2.54 2.51 187.53 20.95 13.38 13.08 20 16.51 3.97 21.45 26.83 27.09 22.70 4.20 3.35 3.32 248.15 27.72 17.70 17.31 40 12.24 2.94 15.90 19.89 20.09 30.61 5.66 4.52 4.48 334.72 37.38 23.88 23.35 OSP-0037 REV - 030 PAGE 58 OF 62

RJPM-NRC-M14-A6 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Determine Personnel Call-Out Availability OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 20 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 3, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 7

RJPM-NRC-M14-A6 Rev. 0 EXAMINER INFO SHEET Task Standard: Operators who are not available to work have been annotated on applicant cue sheet and match the answer key.

Synopsis: This task will require an SRO to recognize which operators for a given shift are available to work after one crew member calls in sick. The SRO will use EN-OM-123, Fatigue Management Program, Section 5.2, Work Hour Limits for Covered Individuals to determine who is eligible to be called in.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

As the off-going CRS, determine any operators from the crew C schedule that are not eligible to be called in without violating fatigue rule. Annotate any ineligible operators on the cue sheet and include the reason he/she is not eligible.

3) Initial Conditions:

The plant is operating at 100% power.

A non-licensed operator has called in sick for the next day shift on March 23, 2014.

The shift will be below minimum staffing requirements.

All operators from crew C worked only the hours they were scheduled and no vacation is scheduled.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A6 Rev 0 Page 2 of 7

RJPM-NRC-M14-A6 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Determine Personnel Call-Out 300013003003 G 2.1.5 3.9 Availability

REFERENCES:

APPLICABLE OBJECTIVES EN-OM-123, Rev 5 FRR-GET-OM123, Obj 3 REQUIRED MATERIALS:

EN-OM-123, Rev 5 SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup 2.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Operators who are not available to work have been annotated on applicant cue sheet and match the answer key.

RJPM-NRC-M14-A6 Rev 0 Page 3 of 7

RJPM-NRC-M14-A6 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1. Determine ineligible operators and reasons using EN-OM-123, Sect 5.2.

Standard Applicant determined that Justin Lawrence is ineligible; max 16 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Applicant determined that Mike Melancon is ineligible; max 72 in 7 days.

Applicant determined that Lane Watts is ineligible; min 34 hr break in 9 days Cue Provide the applicant the Crew C schedule and a copy of EN-OM-123.

Notes Results SAT UNSAT Terminating Cue: Operators who are not available to work have been annotated on applicant cue sheet and match the answer key.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A6 Rev 0 Page 4 of 7

RJPM-NRC-M14-A6 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A6 Rev 0 Page 5 of 7

RJPM-NRC-M14-A6 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

As the off-going CRS, determine any operators from the crew C schedule that are not eligible to be called in without violating fatigue rule. Annotate any ineligible operators on the cue sheet and include the reason he/she is not eligible.

Initial Conditions:

The plant is operating at 100% power.

A non-licensed operator has called in sick for the next day shift on March 23, 2014.

The shift will be below minimum staffing requirements.

All operators from crew C worked only the hours they were scheduled and no vacation is scheduled.

The following are not eligible to work the day shift on Sunday, March 23, 2014:

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

If _____________________ worked this shift, he would violate the _______________ rule.

RJPM-NRC-M14-A6 Rev 0 Page 6 of 7

RJPM-NRC-M14-A6 Rev 0 Note all times represent work hours (shift turnover time has been removed).

For this exercise, disregard the 54-hour rolling average limit.

Schedule Report for Operations / Team C March 3/08 3/09 3/10 3/11 3/12 3/13 3/14 3/15 3/16 3/17 3/18 3/19 3/20 3/21 3/22 3/23 3/24 3/25 3/26 3/27 2014 SAT SUN MON TUE W ED THU FRI SAT SUN MON TUE W ED THU FRI SAT SUN MON TUE W ED THU Duplessis, 1800 1800 0600 0600 0600 0600 0600 0600 0600 0600 1800 1800 1800 1800 John 0600 0600 1800 1830 1800 1900 1600 1600 1600 1600 0600 0600 0600 0600 Frost, 1800 1800 1800 0600 0600 0600 0600 0600 0600 0600 0600 0600 1800 1800 1800 Justin 0600 0600 0600 1800 1800 1800 1800 1600 1600 1600 1600 2000 0600 0600 0600 Lawrence, 1800 1800 0600 0600 0600 0600 0600 0600 0600 0600 1800 1800 1800 1800 Justin 0600 0600 1800 1800 1800 1800 1600 1600 1600 1600 0600 0600 0600 0600 Melancon, 1800 1800 0600 0600 0600 0600 0600 0600 0600 0600 0600 0600 0600 0600 0600 Mike 0600 0600 1800 1800 1800 1800 1800 1800 1800 1800 1800 1600 1600 1600 1600 Thames, 1800 1800 1800 0600 0600 0600 0600 0600 0600 0600 0600 1800 1800 1800 Matt 0600 0600 0600 1830 1800 1800 1800 1600 1600 1600 1800 0600 0600 0600 Umberger, 1800 1800 0700 0600 0600 0600 0600 0600 0600 0600 0600 1800 1800 1800 1800 Dave 0600 0600 1800 1800 1800 1800 1900 1600 1600 1600 1600 0600 0600 0600 0600 Watts, 1800 1800 1800 1800 1800 1800 1800 1800 0600 0600 0600 0600 0600 0600 Lane 0600 0600 0600 0600 0600 0600 0600 0700 1800 1800 1800 1800 1800 1800 River Bend Station Initial License Exam Page 7 of 7

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF 58 Fatigue Management Program Procedure Contains NMM eB REFLIB Forms: YES NO Effective Procedure John McCann Governance John McCann Date Owner: VP, Nuclear Safety, EP, Owner: VP, Nuclear Safety,

Title:

and Licensing

Title:

EP, and Licensing 8/4/2013 Site: HQN / WPO Site: HQN / WPO Exception Site Site Procedure Champion Title Date*

12/29/2013 ANO Michael Chisum Gen Mgr, Plant Ops N/A BRP N/A CNS 9/29/2013 GGNS Ogden J Miller Gen Mgr, Plant Ops 9/22/2013 IPEC John Dinelli Gen Mgr, Plant Ops 11/3/2013 JAF Brian Sullivan Gen Mgr, Plant Ops 10/27/2013 PLP Anthony L. Williams Gen Mgr, Plant Ops 9/22/2013 PNPS Steve Verrochi Gen Mgr, Plant Ops 9/1/2013 RBS Richard Gadbois Gen Mgr, Plant Ops 12/29/2013 VY Vincent Fallacara Gen Mgr, Plant Ops 8/4/2013 W3 Kimberly Cook Gen Mgr, Plant Ops N/A NP N/A HQN Karl Jones GMPO, Fleet Ops Support CNS has a site-specific implementing procedure for the Fatigue Management Program Site and NMM Procedures Canceled or Superseded By This Revision:

EN-OM-123-02, Working Hour Limits eSOMS Users Guide Process Applicability Exclusion: All Sites:

Specific Sites: ANO BRP GGNS IPEC JAF PLP PNPS RBS VY W3 Change Statement Major rewrite including changes to incorporate the following:

Implement 10 CFR 26 rule change regarding 54-hour limit in lieu of (on-line) Minimum Days Off.

Replace Opt-In / Opt-Out steps and forms with a simplified administrative process.

Condition Report corrective actions per CR-HQN-2012-253.

Enhancements per WT-HQN-2012-271.

Enhancements identified by benchmark LO-HQNLO-2011-0176.

Editorial changes for step numbering and attachment numbering to reflect above changes.

• Requires justification for the exception: The effective date is based on the pilot plant (WF3) transition to a new work hour limit being implemented by this procedure revision. Remaining sites are adopting the new limit at selected dates after pilot plant implementation, based on site-specific work schedules and planned activities.

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 2 OF 58 Fatigue Management Program TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ........................................................................................ 3

2.0 REFERENCES

.................................................................................. 4 3.0 DEFINITIONS.................................................................................... 5 4.0 RESPONSIBILITIES ......................................................................... 9 5.0 DETAILS ...................................................................................... 14 5.1 IDENTIFICATION OF COVERED INDIVIDUALS .......................................... 14 5.2 W ORK HOUR LIMITS FOR COVERED INDIVIDUALS ................................. 16 5.3 EXCEPTIONS TO W ORK HOUR CONTROLS ........................................... 18 5.4 SCHEDULES ..................................................................................... 19 5.5 TRANSITIONS.................................................................................... 22 5.6 CALCULATING HOURS W ORKED ........................................................244 5.7 FATIGUE ASSESSMENTS .................................................................... 27 5.8 CONFLICT RESOLUTION ....................................................................311 5.9 W AIVER PROCESS ...........................................................................322 5.10 AUDITS AND REVIEWS ......................................................................366 5.11 DOCUMENT STORAGE ....................................................................... 37 5.12 REPORTS ........................................................................................377 6.0 INTERFACES ................................................................................388 7.0 RECORDS ...................................................................................... 39 8.0 SITE SPECIFIC COMMITMENTS .................................................... 39 9.0 ATTACHMENTS ............................................................................. 39 Attachment 9.1 Fatigue Assessment..40 Attachment 9.2 Fatigue Assessor Guidelines...44 Attachment 9.3 Covered Individual PQ&S Control Form - Addition.............45 Attachment 9.4 Covered Individual PQ&S Control Form - Removal............47 Attachment 9.5 Typical Annual Review Report ....48 Attachment 9.6 Overview of Response to Emergent Condition ....52 Attachment 9.7 Waiver Basis and Approval..... 55 Attachment 9.8 Covered Worker Capability Assessment ............ 56

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 3 OF 58 Fatigue Management Program 1.0 PURPOSE

[1] This procedure establishes the administrative controls for fatigue management, as required by 10 CFR 26 Subpart I, and includes the following:

Entergys policy for the management of fatigue for all individuals who are subject to the Fitness for Duty Program described in EN-NS-102; Method for implementing work hour controls for Covered Individuals; Limitations and processes for granting waivers and exceptions of work hour controls; Processes and requirements regarding fatigue assessments, and; Program provisions for training, recordkeeping, reporting, periodic reviews and audits.

[2] Fatigue management is one of several aspects of the Fitness-for-Duty program (Reference 2.0 [4]). The Fitness-for-Duty program, including fatigue management concepts apply to:

All persons who are granted unescorted access to an Entergy protected area, and, All persons who are required to physically report to an Entergy Technical Support Center (TSC) or Emergency Operations Facility (EOF) by Entergy site-specific Emergency Plan.

[3] Fatigue management concepts include:

Knowledge of contributors to worker fatigue, Ability to identify symptoms of worker fatigue, and Responsibility of individuals to report to work well rested, mentally alert, and fit for duty with respect to fatigue consistent with Entergy Policy EN-PL-202, Personnel Expectations Related to Fatigue Management (Reference 2.0 [12]).

[4] Fatigue management concepts AND the following additional requirements apply to those members of the FFD population (including Entergy employees and contractor /

vendor personnel) who are also identified as Covered Individuals:

Work Hour Limits as specified in Section 5.2 of this procedure.

Waivers and exceptions to the Work Hour Limits

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 4 OF 58 Fatigue Management Program 1.0 cont.

[5] The potential for excessive fatigue is not solely based on actual hours worked but can result from other factors, such as:

(a) stressful working conditions (b) sleep disorders (c) accumulation of sleep debt (d) disruptions of circadian rhythms associated with shift work

[6] Behavioral observation is also an aspect of the Fitness-for Duty program and includes observations related to drugs, alcohol, and fatigue. Actions in response to a suspected impaired condition must be in accordance with that program; Reference 2.0 [4].

2.0 REFERENCES

[1] 10 CFR Part 26 Subpart I, Managing Fatigue

[2] NEI 06-11, Managing Personnel Fatigue at Nuclear Power Reactor Sites, Revision 1, October 2008

[3] Regulatory Guide 5.73, Fatigue Management for Nuclear Power Plant Personnel, March 2009

[4] EN-NS-102, Fitness for Duty Program

[5] Ventyx Personnel Qualifications and Scheduling (PQ&S) User Guide

[6] ENN-HR-130, Shared Resource Assignment

[7] ENS-HR-130, Entergy Shared Resources

[8] EN-WM-104, On Line Risk Assessment

[9] EN-OU-108, Shutdown Safety Management Program

[10] EN-QV-109, Audit Process

[11] Entergy System Policy; Attendance and Absenteeism

[12] EN-PL-202, Personnel Expectations Related to Fatigue Management

[13] EN-EP-309, Fatigue Management for Hurricane Response Activities

[14] EN-IS-113, Reporting & Investigating Occupational Injuries / Illnesses and Near Misses.

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 16 OF 58 Fatigue Management Program 5.2 WORK HOUR LIMITS FOR COVERED INDIVIDUALS NOTES Work hour tracking is accomplished using the eSOMS PQ&S software. Use of an alternate compliance tool, such as by contractors / vendors, requires approval of the site SME.

Work hour limits for covered workers may only be exceeded during Exceptions (Section 5.3) or when evaluated and approved using the Waiver Process (Section 5.9).

[1] Work hour limits for individuals performing Covered Work consist of the following:

(a) Maximum of 16 work hours in any 24-hour period.

(b) Maximum of 26 work hours in any 48-hour period.

(c) Maximum of 72 work hours in any 7-day period..

(d) Minimum 10-hour break between successive work periods, except that an 8-hour break is allowed when necessary to accommodate a crews scheduled transition between work schedules or shifts.

(e) Minimum 34-hour break in any 9-day period.

(f) 54-hour rolling average, as described in 5.2[3].

(g) Minimum Days Off (MDO), as described in 5.2[4].

[2] Limits 5.2[1](a) through (e) apply for online and offline plant conditions. Limit 5.2[1](f) must be used when the plant is online and limit 5.2[1](g) is typically applied when the plant is offline, for individuals working on outage activities. However, limit 5.2[1](f) may also be used in lieu of limit 5.2[1](g) when the plant is offline.

[3] The 54-hour rolling average limit (5.2[1](f)) is a maximum average of 54 work hours per week calculated using a rolling average period of up to 6 weeks. The requirements of the averaging calculation are modeled in the PQ&S software and include the following characteristics:

(a) The duration over which the work hour average is calculated is called the Averaging Period (Definition 3.0[3]). The Averaging Period may be set by the Watchbill Coordinator for the work group or for individuals. The Averaging Period must be in 1 week increments up to a maximum duration of 6 weeks.

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 17 OF 58 Fatigue Management Program 5.2[3] cont.

(b) The date and time during the week when a work hour average in calculated is called the Calculation Milestone (Definition 3.0[5]). The Calculation Milestone is established in this procedure and any changes must be approved through the procedure change process.

(c) When a work shift starts during one calendar day and ends during the next calendar day, the work hours for that shift are attributed to the calendar days in which the hours were actually worked.

(d) PQ&S performs forward and backward calculations for the work hour average. The forward calculation considers future scheduled work hours to support schedule adjustments that may be needed to assure future compliance with the limit. The backward calculation is the verification of compliance with the limit based on actual hours worked for the current Averaging Period.

(e) Transitions into and out of an Averaging Period are discussed in Section 5.5.

[4] The MDO limits (5.2[1](g)) available for use during a unit outage depend on the category of covered work (Definition 3.0[11]) being performed. For the purposes of this limit, a Day Off is defined in 3.0[13].

(a) During the first 60 days of a unit outage, the MDO limit for Maintenance covered work is a minimum of 1 day off in any rolling 7-day period, for individuals working on outage activities.

(b) During the first 60 days of a unit outage, the MDO limit for Operations, Radiation Protection, Chemistry, and Fire Brigade covered work is a minimum of 3 days off in each successive, non-rolling 15-day period, for individuals working on outage activities.

(c) During the first 60 days of a unit outage or planned security system outage, the MDO limit for Security covered work is a minimum of 4 days off in each successive, non-rolling 15-day period.

(d) During the first 60 days of an unplanned security system outage or increased threat condition, Security covered work is not subject to the 54-hour rolling average limit or the MDO limit (e) The 60-day periods in (a) through (d) above may be extended for individuals in 7-day increments for each non-overlapping 7-day period that the individual worked not more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during the unit outage, security system outage, or increased threat condition, as applicable. Note that this allowance provides for a maximum extension of 56 days based on a maximum of eight non-overlapping 7-day periods available in a 60-day period.

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 58 Fatigue Management Program 5.2[4] cont.

(f) For dual unit sites, when one unit is in an outage and the other unit is on-line, any Covered Individual who performs outage tasks may work a schedule based on the MDO limits. However, at least 2 Senior Operators and 2 Reactor Operators in the control room for the on-line unit must remain on a schedule using the 54-hour average online limit.

(g) When an Entergy employee or contractor/vendor works during two or more unit outages or security system outages (or a combination thereof) at one or more Entergy sites, and the interval(s) between successive outages is less than 9 days, the supervisor must ensure that the individual has had a 34-hour break period in the previous 9 days and that the maximum ceilings limits (5.2 [1](a) to (c)) are met.

[5] Shift turnover time may be excluded from the calculation of actual hours worked. Only one period of shift turnover, either at the beginning or the end of the shift but not both, may be excluded for purposes of calculating the break period. These allowances for excluding shift turnover time may not include activities such as holdover for late arrivals, holdover for event investigations, and early arrival for required meetings, training, and special evolution briefings.

[6] Within shift rest breaks during which the individual is provided opportunity and accommodations for restorative sleep may be excluded from the calculation of actual hours worked.

[7] When working during the transition from daylight savings time to standard time, the extra hour incurred during that shift may be excluded from the calculation of actual hours worked. The actual hours worked are used in the calculation for the reverse transition.

[8] Incidental time worked by a covered individual, when the cumulative time during a single break period exceeds a nominal 30 minutes, must be included in the work hours calculation.

5.3 EXCEPTIONS TO WORK HOUR CONTROLS

[1] Force-on-force tactical exercises During the conduct of NRC-evaluated force-on-force tactical exercises, hours worked in excess of 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> during the week of the exercise may be excluded from the calculation of the 54-hour average limit. This allowance applies to all Security personnel whose work schedules are modified to support the exercise. In the event that the exercise is conducted while the plant is offline and MDO limits are being used, the work shifts may be excluded from the MDO compliance calculation. The PQ&S code available for recording the hours over 54 and the MDOs eligible for exclusion under this provision is FTR ID EX-FF.

NUCLEAR NON-QUALITY RELATED EN-OM-123 REV 5 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 19 OF 58 Fatigue Management Program 5.3 cont.

[2] Common defense and security Work Hour Control requirements need not be met when informed in writing by the NRC that these requirements, or any subset thereof, are waived for security personnel in order to assure the common defense and security, for the duration of the period defined by the NRC. The applicable FTR ID is EX-CDS.

[3] Plant emergencies Work Hour Control requirements need not be met during declared emergencies, as defined in the Site Emergency Plan. However, information regarding hours worked needs to be available in PQ&S to support a transition to post-emergency work schedules. The PQ&S code available for recording eligible hours under this provision is FTR ID EX-PE.

[4] Hurricane Response Situations A site-specific rule exemption approved by NRC for Waterford-3 applies for Work Hour Control requirements during hurricane response situations as implemented in Reference 2.0 [13]. The waiver process (Section 5.9) is available for other sites to manage work force requirements in this situation.

5.4 SCHEDULES

[1] For personnel defined as covered workers in this procedure work hour schedules are developed and maintained in eSOMS PQ&S software. Details on developing a work schedule can be found in the Ventyx PQ&S User Guide (Reference 2.0[5]).

[2] When designing schedules, the following factors should be considered with the performance objective of preventing impairment from fatigue due to the duration, frequency, or sequencing of successive shifts:

Duration of scheduled work period (typically <12 hour shifts)

Duration of break period Consistent start times for work periods (e.g. 6 or 7 a.m.)

Considerations of start times that are consistent with circadian factors Consistent stop times for work periods Consistent rotation (e.g., if working a 5-week shift rotation, the scheduled work days and days off are repeated every five weeks)

Stable 24-hour shift rotation (e.g., 3x8s, 2x12s, 2x10s with four hours un-staffed)

RJPM-NRC-M14-A7 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Review Data to Determine if Control Rod Withdrawal is Supported OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: January 15, 2014 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 6

RJPM-NRC-M14-A7 Rev. 0 EXAMINER INFO SHEET Task Standard: Applicant determines that control rod withdrawal is supported.

Synopsis: The plant is performing a post-scram startup. Section C, Step 10 of GOP-0001, Plant Startup, requires verification of temperature/pressure limits within 15 minutes of pulling control rods. This task will have the applicant review a completed copy of STP-050-0700, RCS Pressure/Temperature Limits Verification, to determine if the plant can support withdrawing control rods.

The applicant will arrive at the conclusion that conditions do meet requirements.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

You are the CRS. STP-050-0700 has just been completed. Determine if current plant conditions support withdrawing control rods.

3) Initial Conditions:

The plant experienced and is recovering from a Reactor scram. All actions of GOP-0003, Scram Recovery have been performed. The plant is performing a startup in accordance with GOP-0001, Plant Startup.

Step 10 of Section C, Approach to Critical states, Within 15 minutes prior to control rod withdrawal to bring the Reactor critical, verify Reactor Coolant System temperature and pressure are to the right of the criticality limit line of Technical Specifications Figure 3.4.11-1 curve C (STP-050-0700).

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A7 Rev 0 Page 2 of 6

RJPM-NRC-M14-A7 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Review Data to Determine if Control 300066003001 202001 G 2.2.1 4.4 Rod Withdrawal is Supported K5.05

3.3 REFERENCES

APPLICABLE OBJECTIVES GOP-0001, Rev 81 RLP-HLO-0405, Obj 1 STP-050-0700, Rev 306 Tech Spec 3.4.11 Condition C REQUIRED MATERIALS:

Marked up copy of STP-050-0700, Rev 306 SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup 2.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Applicant determines that control rod withdrawal is supported.

RJPM-NRC-M14-A7 Rev 0 Page 3 of 6

RJPM-NRC-M14-A7 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1 Determine if Control Rod Withdrawal is supported.

Standard Applicant used data provided from step 7.1.2 to plot a point on Attachment 3 of STP-050-0700.

Applicant recognized the point is to the right of Curve C.

Cue Notes The applicant will use the marked up copy of STP-050-0700.

Results SAT UNSAT Terminating Cue: Applicant determines that control rod withdrawal is supported.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A7 Rev 0 Page 4 of 6

RJPM-NRC-M14-A7 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A7 Rev 0 Page 5 of 6

RJPM-NRC-M14-A7 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

You are the CRS. STP-050-0700 has just been completed. Determine if current plant conditions support withdrawing control rods.

Initial Conditions:

The plant experienced and is recovering from a Reactor scram. All actions of GOP-0003, Scram Recovery have been performed. The plant is performing a startup in accordance with GOP-0001, Plant Startup.

Step 10 of Section C, Approach to Critical states, Within 15 minutes prior to control rod withdrawal to bring the Reactor critical, verify Reactor Coolant System temperature and pressure are to the right of the criticality limit line of Technical Specifications Figure 3.4.11-1 curve C (STP-050-0700).

Answer Below:

Control Rod Withdrawal is supported by current plant conditions.

YES NO RJPM-NRC-M14-A7 Rev 0 Page 6 of 6

RJPM-NRC-M14-A8 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Determine if Normal Dose Limits Will Be Exceeded OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 9, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 6

RJPM-NRC-M14-A8 Rev. 0 EXAMINER INFO SHEET Task Standard: Applicant determines that emergency dose limits would not be exceeded.

Applicant determines that only the Emergency Director or Emergency Plant Manager has the authority to permit exceeding dose limits in excess of 10CFR20 limits during an emergency. 75 rem is the highest authorized dose under the condition that the dose is voluntary for an action to save a life or protect large populations.

Synopsis: This task will have the applicant review plant conditions and operator dose records to determine if the operator will exceed dose limits for an emergency evolution. The applicant will further determine whose authority is needed to exceed dose limits for a given emergency condition.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

1) Using the initial conditions below, determine if allowed dose limits will be exceeded when performing an evolution during emergency conditions.

2) Assuming a condition where normal limits would need to be exceeded, who has the authority to permit exceeding dose limits?

3) What is the highest dose that can be authorized by the person named in #2 above?

4) and under what conditions could this authorization be made?

3) Initial Conditions:

The plant has declared a general emergency due to offsite dose. To restore cooling to the reactor, it is necessary to enter the aux building where the dose rate is 1840 mr/hr.

A qualified operator has a year-to-date dose of 1568 mr and can do the job in 20 minutes

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A8 Rev 0 Page 2 of 6

RJPM-NRC-M14-A8 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Determine if Normal Dose Limits Will 200076005002 G 2.3.4 3.7 Be Exceeded 300052003002 G 2.3.14

3.8 REFERENCES

APPLICABLE OBJECTIVES EIP-2-012, Rev 21 REQUIRED MATERIALS:

EIP-2-012, Rev 21 SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup 2.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Applicant determines that normal dose limits would not be exceeded.

Applicant determines that only the Emergency Director or Emergency Plant Manager has the authority to permit exceeding dose limits in excess of 10CFR20 limits during an emergency. 75 rem is the highest authorized dose under the condition that the dose is voluntary for an action to save a life or protect large populations.

RJPM-NRC-M14-A8 Rev 0 Page 3 of 6

RJPM-NRC-M14-A8 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1 Determine if allowed dose limits will be exceeded when performing an evolution during emergency conditions.

Standard Applicant checked the box indicating that limit would not be exceeded during emergency conditions.

Cue Notes Calculated Operator dose = 613 mr for a ytd dose of 2181 mr.

Results SAT UNSAT

2. Procedure Step: 2 Determine who has authority to permit exceeding the normal limit.

Standard Applicant indicated the emergency director could authorize exceeding limits.

Cue Notes Results SAT UNSAT Terminating Cue: Applicant determines that emergency dose limits would not be exceeded.

Applicant determines that only the Emergency Director or Emergency Plant Manager has the authority to permit exceeding dose limits in excess of 10CFR20 limits during an emergency.

75 rem is the highest authorized dose under the condition that the dose is voluntary for an action to save a life or protect large populations.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A8 Rev 0 Page 4 of 6

RJPM-NRC-M14-A8 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A8 Rev 0 Page 5 of 6

RJPM-NRC-M14-A8 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

1) Using the initial conditions below, determine if the allowed dose limits will be exceeded when performing an evolution during emergency conditions.

2) Assuming a condition where normal limits would need to be exceeded, who has the authority to permit exceeding dose limits?

3) What is the highest dose that can be authorized by the person named in #2 above?

4) and, under what conditions could this authorization be made?

Initial Conditions:

The plant has declared a general emergency due to offsite dose. To restore cooling to the reactor, it is necessary to enter the aux building where the dose rate is 1840 mr/hr.

A qualified operator has a year-to-date dose of 1568 mr and can do the job in 20 minutes 1)Given current plant conditions, the operator would exceed allowed limits.

YES NO

2) In a condition where normal limits would need to be exceeded, who has the authority to permit exceeding limits?

3) The highest dose allowed that can be authorized by #2 is _________________.

4) The condition(s) under which the authorization can be made in #3 is:

RJPM-NRC-M14-A8 Rev 0 Page 6 of 6

REFERENCE USE RIVER BEND STATION STATION SUPPORT MANUAL

• EMERGENCY IMPLEMENTING PROCEDURE

• RADIATION EXPOSURE CONTROLS PROCEDURE NUMBER: *EIP-2-012 REVISION NUMBER: *21 Effective Date:

• 11/15/12 NOTE : SIGNATURES ARE ON FILE.

TemRev 1 AddCounter 12 Att Enc DS MSet REGULAR KWN OFF

• INDEXING INFORMATION

TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE .................................................................................................................................2 2 REFERENCES .........................................................................................................................2 3 DEFINITIONS .........................................................................................................................2 4 RESPONSIBILITIES ...............................................................................................................2 5 GENERAL ................................................................................................................................3 6 PROCEDURE...........................................................................................................................4 7 DOCUMENTATION .............................................................................................................11 ATTACHMENT 1 - RADIATION EXPOSURE LIMITS AND GUIDELINES .........................12 ATTACHMENT 2 THYROID COMMITTED DOSE EQUIVALENT GRAPH.. ...................13 ATTACHMENT 3 POTASSIUM IODIDE ADMINISTRATION FORM.. ..............................14 EIP-2-012 REV - 21 PAGE 1 OF 14

1 PURPOSE This procedure provides instructions for establishing special radiation exposure controls during an emergency.

2 REFERENCES 2.1 Title 10, Code of Federal Regulations, Part 20, (10 CFR 20) "Standards for Protection Against Radiation" 2.2 RBNP-024, River Bend Station Radiation Protection Plan 2.3 Company Procedure RP-205, Prenatal Monitoring 2.4 FDA Guidance, Potassium Iodide as a Thyroid Blocking Agent in Radiation Emergencies, December 2001 2.5 Company Procedure RP-202, Personnel Monitoring 2.6 EN-EP-FAP-009, Use of KI for the Emergency Response Organization 3 DEFINITIONS 3.1 Committed Dose Equivalent (CDE) - The dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material during the 50 year period following the intake.

3.2 Total Effective Dose Equivalent (TEDE) - The sum of the Deep Dose Equivalent (DDE) (from external exposure) and the Committed Effective Dose Equivalent (CEDE) (from internal exposure).

4 RESPONSIBILITIES 4.1 Emergency Director (ED) or Emergency Plant Manager (EPM) - The ED or EPM is responsible for authorizing individuals assigned to their responsible areas/facilities to receive exposures in excess of 10 CFR 20 limits and approving the issuance of potassium iodide (KI). Normally the Emergency Plant Manager will perform the authorizations for EIP-2-012 REV - 21 PAGE 2 OF 14

personnel assigned to the OSC, TSC and Security and the Emergency Director will perform the authorizations for personnel assigned to the JIC and EOF.

4.2 Radiological Coordinator (RC) or Radiological Assessment Coordinator (RAC) - The RC or RAC is responsible for advising the ED or EPM, tracking the dose history for those individuals authorized to receive exposures in excess of 10 CFR 20 limits, and notifying the Nuclear Regulatory Commission of any overexposures.

4.3 Onshift Senior Radiation Protection Technician (SRPT) - The Onshift SRPT is responsible for performing the duties of the Radiological Coordinator per this procedure until his arrival at the Technical Support Center.

5 GENERAL 5.1 During a classified emergency, the administrative exposure controls of the River Bend Station Radiation Protection Plan RBNP-024 and RP-202, Personnel Monitoring, are suspended; however, efforts shall be made to maintain personnel exposures within the limits established by 10 CFR 20.

5.2 Due to rapidly changing conditions during an emergency, administrative approvals for exceeding established exposure limits are suspended. Only the Emergency Director or Emergency Plant Manager shall have the authority for authorizing exposures in excess of 10 CFR 20 limits (included as Attachment 1 for reference).

5.3 During the emergency phase of an accident, the Radiation Work Permit (RWP) provisions of the River Bend Radiation Protection Plan are suspended, but shall be re-implemented at the termination of an emergency when the recovery phase is initiated.

5.4 Potassium Iodide (KI) (thyroid blocking agent) is available in the Control Room, Technical Support Center, Decontamination Room (Second Floor of the Services Building), Emergency Operations Facility (EOF) and the Offsite Monitoring Team Emergency Kits.

5.5 A declared pregnant female shall not be assigned any functions during a declared emergency which may cause her to exceed the dose limits of 10CFR20.1208 (See Attachment 1); however, a female who declares herself pregnant after an emergency is declared will be expected to continue to fulfill her assignment until a qualified relief can be found. In this case every effort will be made to limit the female s TEDE to the EIP-2-012 REV - 21 PAGE 3 OF 14

limits specified in Attachment 1, consistent with the needs of the Emergency Response Organization.

5.6 Responsibility for authorizing federal, state, and local emergency workers (Ex. EMS, fire, law enforcement, National Guard, etc.)

responding to RBS to incur exposures in excess of the EPA Protective Action Guides for the general population rests with the unit of government for whom the emergency worker is employed. This also applies to the issuance of potassium iodide (KI). LDEQ is the State of Louisiana s lead agency on radiological matters to include technical assessment and emergency worker s protection during events at nuclear power plants.

6 PROCEDURE NOTE The actions of this procedure may be completed in any sequence, however, the sequence presented is recommended.

NOTE During a declared emergency, the exposure limits for federal, state, and local emergency workers responding to RBS are the EPA Protective Action Guides. The responsibility for authorizing extensions above these limits or issuing KI to these emergency workers resides with the respective federal, state, or local governmental agency for which the emergency worker is employed.

6.1 The Emergency Director or Emergency Plant Manager should:

6.1.1. Use 10CFR20 exposure limits contained in Attachment 1.

These limits apply to all members of the Emergency Response Organization, whether or not every person has completed Radiation Worker Training.

6.1.2. When assigning members of the emergency organization to perform tasks which may result in exposures in excess of the 10 CFR 20 limits (see Attachment 1, Section A):

EIP-2-012 REV - 21 PAGE 4 OF 14

1. Consult with the Radiological Coordinator or Radiological Assessment Coordinator to determine the person's current exposure history to verify the amount of exposure the individual may receive without exceeding the 10 CFR 20 limit.

2. Authorize each individual a maximum exposure limit, not to exceed the limits in Attachment 1, Section B.

NOTE The Emergency Director or Emergency Plant Manager shall initiate a log for the documentation of emergency information. The Operations Shift Manager shall use the Control Room log.

3. Document the authorization of each individual in the log.

4. In accordance with the Entergy Operations Inc. policy concerning exposures to females who may be pregnant, no female who suspects she is pregnant should be assigned any responsibilities during an emergency which could result in exposures in excess of the 10 CFR 20 limits.

6.1.3. Ensure that any individual believed to have received greater than 25 rem (250 mSv) TEDE is promptly relieved from the Emergency Response Organization.

NOTE SCBA's and other masks do not preclude the consideration of the dissemination of KI.

6.1.4. Authorize the use of KI, as necessary.

6.2 The Radiological Coordinator or Radiological Assessment Coordinator should:

6.2.1. Ensure that current exposure margins are readily available for the emergency organization.

6.2.2. When time permits, consult with the Emergency Director or Emergency Plant Manager on the methods available to prevent excessive exposures during the emergency.

EIP-2-012 REV - 21 PAGE 5 OF 14

NOTE SCBA's and other masks do not preclude the consideration of the dissemination of KI.

6.2.3. Consult with the Emergency Director or Emergency Plant Manager regarding the use of KI by emergency response personnel involved in actions to save a life of another individual, mitigate accident consequences, or prevent major releases of radioactivity to the environment I.A.W. Section 6.4 6.2.4. Inform emergency workers who are authorized emergency exposure in excess of 10 CFR 20 limits regarding the relative risks involved with excessive radiation exposure.

6.2.5. Determine the need to process emergency worker DLRs.

6.2.6. Initiate efforts to obtain a medical evaluation of any individual who receives greater than 25 rem (250 mSv) TEDE, during emergency operations by a physician who is familiar with acute effects of radiation exposure. These individuals shall not be subjected to any further radiation exposure until approved by the Radiation Protection Manager and the General Manager of Plant Operations.

NOTE The following notification will be made in accordance with the reporting requirements of 10 CFR 20.2202 and 20.2203.

6.2.7. As soon as practical during an emergency, make oral reports of radiation overexposures to the Nuclear Regulatory Commission followed by a written report. Written reports should be provided within 30 days as provided by 10 CFR 20.2203 except when the emergency continues for more than 30 days, then the written report shall be provided within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after termination of the emergency.

6.2.8. Ensure that TEDE dose received during an emergency is recorded on each individual's dose history file. All occupational doses, including emergency doses, are required to be included as part of an individual's accumulated dose EIP-2-012 REV - 21 PAGE 6 OF 14

history and can affect the individual's allowable exposure during the current and subsequent years.

6.2.9. Ensure that declared pregnant females do not exceed the dose limits specified in Attachment 1, and that radiation doses to females, who declare themselves pregnant after the declaration of an emergency, are limited to the extent practical to the limits specified in Attachment 1, consistent with the needs of the Emergency Response Organization.

6.3 The Onshift Senior Radiation Protection Technician should:

6.3.1. Assume duties of the Radiological Coordinator per this procedure until position is filled.

6.3.2. Assist in evaluating radiation exposure levels likely to be encountered during emergency operations.

EIP-2-012 REV - 21 PAGE 7 OF 14

NOTE Completion of Attachment 3 is not required unless a worker s thyroid CDE is expected to be 5 Rem.

Attachment 2 may be used as a reference without completion of Attachment 3.

6.4 Administration of Iodine Blocking Agents.

6.4.1. Assessing the Need to Issue KI

1. If there is a potential need to issue KI, potential recipients of KI may fill out their portion of Attachment 3 in advance.

2. If a worker s thyroid CDE is expected to approach 5 Rem, obtain a copy of Attachment 2, Thyroid Committed Dose Equivalent Graph, and estimate the dose commitment for the thryoid.

3. Verify your calculations/measurements/estimates and record the results on Attachment 3, Potassium Iodide Administration Form.

4. Report the results to the Emergency Director or Emergency Plant Manager and advise him as to the need to issue KI in accordance with this procedure.

5. The Emergency Director or Emergency Plant Manager may approve the issuance of KI via telecon/radio.

6.4.2. KI Issuance Requirements

1. When thyroid CDE is estimated to be 5 rem or greater the following are required:

The Emergency Director or Emergency Plant Manager shall designate the individuals who will receive KI.

The individual to receive KI shall voluntarily elect to take KI.

The individual to receive KI shall read Potassium Iodide precaution information provided by the drug company. The individual shall then complete the appropriate sections of Attachment 3 - Potassium Iodide Administration Form.

EIP-2-012 REV - 21 PAGE 8 OF 14

6.4.3. Distribution of KI NOTE KI is stored in the following locations: Control Room, Technical Support Center, Decontamination Room (second floor of the Services Building), Emergency Operations Facility and Offsite Monitoring Team Emergency Kits.

1. Assemble the individuals who were designated to receive KI and the individuals to administer the KI.

2. Provide the individuals designated to receive KI with copies of:

1. Potassium Iodide precaution information provided by the drug company.

2. Attachment 3 Pottassium Iodide (KI)

Administration Form

3. Ensure personnel read and/or complete the appropriate sections of the above.

EIP-2-012 REV - 21 PAGE 9 OF 14

6.4.4. Guidelines for the Administration of KI NOTE The Emergency Director or Emergency Plant Manager can authorize the administration of KI in the field after the Field Monitoring Team members have complied with the guidelines of this procedure. Completion of the KI documentation may be accomplished at the convenience of the Emergency Director or Emergency Plant Manager.

1. If possible, KI should be administered approximately one-half hour before exposure for maximum blockage.

2. Final uptake is halved if KI is administered within 3-4 hours after exposure.

3. Little benefit is gained with KI administration 10-12 hours after exposure.

4. Once the KI is taken and the Iodine concentration is verified or the calculated dose determined, the tablets should be issued for a minimum of six (6) to a maximum of ten (10) consecutive days. One tablet is issued each day.

5. Verify that each individual receiving KI has completed and signed Attachment 3.

6. Verify that there are no YES blocks marked for allergies or iodine sensitivity on Attachment 3, Potassium Iodide Administration Form.

7. Individuals who have answered YES for allergies or iodine sensitivity to those questions on Attachment 3, will initially be considered to be iodine sensitive and must be treated as follows:

1. The individuals will be relocated or replaced to eliminate or minimize the uptake of radioiodine in the thyroid gland, or

2. The individuals WILL NOT receive KI without the Radiological Coordinator s or Radiological Assessment Coordinator s authorization (after evaluation of the YES answer and the Emergency EIP-2-012 REV - 21 PAGE 10 OF 14

Director s or Emergency Plant Manager s concurrence).

8. Issue each individual designated to receive KI one (1) 130 mg KI tablet.

9. Forward all completed paperwork to the Radiological Coordinator or Radiological Assessment Coordinator.

6.4.5. Final Conditions

1. Ensure that each individual whose estimated exposure to radioiodine exceeded 5 rem has been identified and administered KI, as appropriate.

2. Ensure all necessary forms are completed and reviewed by the Radiological Coordinator or Radiological Assessment Coordinator and the Emergency Director or Emergency Plant Manager.

3. Ensure that each individual who was exposed to radioiodine with a calculated thyroid CDE 5 Rem has been scheduled for bioassay analysis.

7 DOCUMENTATION 7.1 Attachment 3 of this procedure, completed during actual events shall be submitted to permanent plant files (PPF) per EPP-2-100. Attachments from exercises/drills will be used to critique and evaluate exercises/drills performance. This documentation will not be sent to PPF and may be discarded.

EIP-2-012 REV - 21 PAGE 11 OF 14

ATTACHMENT 1 PAGE 1 OF 1 RADIATION EXPOSURE LIMITS AND GUIDELINES A. 10CFR20 RADIATION EXPOSURE LIMITS 5 rem/yr. (50 mSv/yr.) Total Effective Dose Equivalent (TEDE) to the whole body.

50 rem/yr. (500 mSv/yr.) sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE) to an individual organ or tissue other than the lens of the eye.

15 rem/yr. (150 mSv/yr.) Eye Dose Equivalent (LDE) to the eye.

50 rem/yr. (500 mSv/yr.) Shallow Dose Equivalent (SDE) to the skin or an extremity.

50 mrem (0.5 mSv) in a one month period for a declared pregnant female, not to exceed 500 mrem (5 mSv) for the entire pregnancy period (10CFR20.1208).

B. GUIDELINES FOR EMERGENCY EXPOSURES

1. Emergency Total Effective Dose Equivalent (TEDE) limits are:

a. 5 rem (50 mSv) for preplanned emergency actions.

b. 10 rem (100 mSv) for immediate actions taken to prevent major damage to equipment, prevent the release of radioactive materials, or control fires.

c. 25 rem (250 mSv) without consent and 75 rem (750 mSv) on a voluntary basis for action to save a life or to protect large populations.

2. Committed Dose Equivalent to the Thyroid NOTE Although RBS Emergency Plan Table 13.3-10 establishes thyroid exposure guidelines, the difficulty in monitoring thyroid exposure over a short period of time prevents use of these numbers as absolute limits.

Therefore, when radioiodine airborne concentrations are known, projected thyroid doses will be calculated to prevent exceeding these guidelines.

a. To save the life of another individual there is no specified limit. Although respirators should be used where effective to control the dose to emergency workers, thyroid dose should not be a limiting factor for lifesaving missions.

b. To mitigate accident consequences and prevent major releases of radioactivity to the environment or control fires- 100 rem (1 Sv) Committed Dose Equivalent (CDE).

c. Emergency duties including decontamination and first aid, but not related to protecting equipment, the public or for lifesaving - 50 rem (500 mSv) CDE.

3. Shallow Dose Equivalent to the Extremities

a. To save the life of another individual, extremity exposure should not be a factor.

b. To mitigate accident consequences and prevent major release of radioactivity to the environment - 100 rem (1 Sv) Shallow Dose Equivalent (SDE).

c. When preplanned emergency actions are possible -50 rem (500 mSv) SDE.

EIP-2-012 REV - 21 PAGE 12 OF 14

ATTACHMENT 2 PAGE 1 OF 1 THYROID COMMITTED DOSE EQUIVALENT GRAPH Time to 5 REM CDE verses I-131 concentration guideline Duration of Exposure (minutes) resulting in 5 REM CDE 1.00E+05 1.00E+04 1.00E+03 minutes 1.00E+02 1.00E+01 1.00E+00 1.00E-01 1.00E-08 1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 I-131 concentration (uCi/cc)

Instructions for Use:

1. Determine the estimated or actual I-131 airborne concentration in the area(s) of interest. Divide this by the protection factor of the equipment used (if unknown, use 1). Locate this number on the Horizontal Axis.

2. Locate the duration of exposure in minutes on the Vertical Axis. Find the point at which this value intersects with the number from step 1.

3. If this point of intersection is located to the left (below) the line, the thyroid CDE is less than 5 rem.

4. If this point of intersection is located to the right (above) the line, the thyroid CDE is greater than 5 rem.

5. If this point of intersection is located on the line, the thyroid CDE is 5 rem.

EIP-2-012 REV - 21 PAGE 13 OF 14

ATTACHMENT 3 PAGE 1 OF 1 POTASSIUM IODIDE (KI) ADMINISTRATION FORM Name:____________________/_______________/_____________

Last First Middle SSN Yes No Have you any known allergies? If so, please describe major severity of allergy and medications taken if any.

Yes No When eating seafood or shellfish, do you suffer from symptoms of stomach or bowel upset or skin eruption? If so, explain.

Yes No Has any physician told you that you have a sensivity to iodine?

Yes No If you have ever had a gallbladder dye test, kidney x-ray requiring dye injection, thyroid isotope scan, did you have any reactions?

Please explain any Yes answers:

• Known Iodide Allergy/Previous Allergic Reaction: (Mark One) Yes No I verify that I have read and understand the precaution leaflet. I understand that taking thyroid blocking agent (KI) is strictly voluntary.

I (Mark One) Do Do Not choose to take KI when approved.

Signature of Individual Date Duration of Exposure: (minutes) I-131 Concentration: ( Ci/cc in air)

Estimated Thyroid Dose Commitment: (Mark one) < 5 Rem > 5 Rem Respiratory Protection Worn During Exposure: (Mark One) Yes No Respiratory Protection Factor:______________ Date of Exposure:___________________

CAUTION If the above allergic reaction statement

• is marked Yes, then do not administer KI.

Approved: Mark if telecon/radio approval ED/EPM Date/Time Individual notified KI is approved for use: (Date/Time) /

KI taken (Date/Time) /

Notes:

EIP-2-012 REV - 21 PAGE 14 OF 14

RJPM-NRC-M14-A9 Rev 0 NUCLEAR PLANT OPERATOR ADMINISTRATIVE JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Determine Reportability Requirements Following Component Failure OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 15 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate Simulator Control Room X Classroom Prepared: Dave Bergstrom Date: October 9, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-A9 Rev. 0 EXAMINER INFO SHEET Task Standard: Applicant completes the NRC Form 361 in accordance with the highlighted portions of the key.

Synopsis: This task will have the applicant determine a Reportability based on an operator initiated scram.

NOTE: This JPM is Administrative and will be performed in a classroom.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

You are the CRS. Complete the attached Reactor Plant Event Notification Worksheet.

(NRC Form 361)

3) Initial Conditions:

The plant was operating at 100% power when the Feedwater Level Control System causes reactor water level to lower. The ATC Operator anticipates a level 3 and takes the Mode switch to SHUTDOWN. The plant is now stable with FWS-P1A, Feedpump A, feeding the reactor through the Startup Feed Reg Valve. The lowest reactor water level recorded was minus 20 inches.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-A9 Rev 0 Page 2 of 8

RJPM-NRC-M14-A9 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Determine Reportability Requirements 300134003002 G 2.4.30 4.1 Following Component Failure 300015003002

REFERENCES:

APPLICABLE OBJECTIVES EN-LI-108, Rev 8 RLP-HLO-219, Obj 4 EN-LI-108-01, Rev 3 RLP-HLO-0205, Obj 7 10 CFR 50.72/73 NUREG 1022 REQUIRED MATERIALS:

10 CFR 50.72/73 NUREG 1022 SIMULATOR CONDITIONS &/or SETUP:

1. This is a classroom/Admin JPM - There is no simulator setup 2.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD:

Applicant completes the NRC Form 361 in accordance with the highlighted portions of the key.

RJPM-NRC-M14-A9 Rev 0 Page 3 of 8

RJPM-NRC-M14-A9 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1 Determine Reportability of the event. and complete NRC Form 361.

Standard Applicant completed the Form 361 in accordance with the key.

Cue Notes The applicant will use the pictures provided to gather readings.

Results SAT UNSAT Terminating Cue: Applicant completes the NRC Form 361 in accordance with the highlighted portions of the key.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-A9 Rev 0 Page 4 of 8

RJPM-NRC-M14-A9 Rev. 0 NRC FORM 361 U.S. NUCLEAR REGULATORY COMMISSION (12-2000) REACTOR PLANT OPERATIONS CENTER EVENT NOTIFICATION WORKSHEET EN # 1234 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 #

(Current Time) RBS 1 (Applicant Name) 225-378-2297 EVENT TIME & ZONE EVENT DATE POWER/MODE BEFORE POWER/MODE AFTER 0800 CST 3/24/2014 100% / Mode 1 0% / Mode 2 EVENT CLASSIFICATIONS 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 UNUSUAL 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 ACOM 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 HFIT (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)

Applicant fills in info about the manual SCRAM here.

The following is the justification for why 10CFR50.72(b)(2)(iv)(B) should be selected:

From page 31: § 50.72(b)(2)(iv)

(B) Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

From page 35:

Note that, if an operator were to manually scram the reactor in anticipation of receiving an automatic reactor scram, this would be reportable just as the automatic scram would be reportable.

ANSWER KEY NOTIFICATIONS YES NO WILL ANYTHING UNUSUAL OR YES (Explain above) NO BE NOT UNDERSTOOD?

NRC RESIDENT STATE DID ALL SYSTEMS YES NO (Explain above)

LOCAL FUNCTION AS REQUIRED?

OTHER GOV AGENCIES ESTIMATED MODE OF OPERATION ADDITIONAL INFO ON BACK UNTIL CORRECTED: 3/4 RESTART DATE: 03/25/2014 MEDIA/PRESS RELEASE (MM/DD/YYYY) YES NO NRC FORM 361 (12-2000) PAGE 1 OF 2 RJPM-NRC-M14-A9 Rev 0 Page 5 of 8

RJPM-NRC-M14-A9 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-A9 Rev 0 Page 6 of 8

RJPM-NRC-M14-A9 Rev. 0 OPERATOR CUE SHEET Initiating Cues:

You are the CRS. Complete the attached Reactor Plant Event Notification Worksheet.

(NRC Form 361)

Initial Conditions:

The plant was operating at 100% power when the Feedwater Level Control System causes reactor water level to lower. The ATC Operator anticipates a level 3 and takes the Mode switch to SHUTDOWN. The plant is now stable with FWS-P1A, Feedpump A, feeding the reactor through the Startup Feed Reg Valve. The lowest reactor water level recorded was minus 20 inches.

RJPM-NRC-M14-A9 Rev 0 Page 7 of 8

RJPM-NRC-M14-A9 Rev. 0 NRC FORM 361 U.S. NUCLEAR REGULATORY COMMISSION (12-2000) REACTOR PLANT OPERATIONS CENTER EVENT NOTIFICATION WORKSHEET EN # 1234 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 #

225-378-2297 EVENT TIME & ZONE EVENT DATE POWER/MODE BEFORE POWER/MODE AFTER 0800 CST 3/24/2014 EVENT CLASSIFICATIONS 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 UNUSUAL 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 ACOM 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 HFIT (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 ANYTHING UNUSUAL OR YES (Explain above) NO BE NOT UNDERSTOOD?

NRC RESIDENT STATE DID ALL SYSTEMS YES NO (Explain above)

LOCAL FUNCTION AS REQUIRED?

OTHER GOV AGENCIES ESTIMATED MODE OF OPERATION ADDITIONAL INFO ON BACK UNTIL CORRECTED: 3/4 RESTART DATE: 03/25/2014 MEDIA/PRESS RELEASE (MM/DD/YYYY) YES NO NRC FORM 361 (12-2000) PAGE 1 OF 2 RJPM-NRC-M14-A9 Rev 0 Page 8 of 8

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF 33 Event Notification and Reporting Procedure Contains NMM REFLIB Forms: YES NO Effective Procedure Owner: Charlene Faison Governance Owner: John McCann Date

Title:

Mgr, Licensing

Title:

VP, Nuc Safety, EP &

Programs Licensing 7/1/2013 HQN HQN Site: Site:

Exception Site Site Procedure Champion Title Date*

ANO Steve Coffman Sr. Licensing Specialist BRP Otto Gustafson Manager, Licensing N/A CNS David Vanderkamp Manager, Licensing GGNS Chris Robinson Manager, Licensing IPEC Steve Prussman Sr. Lead Engineer JAF Jorge OFarrill Licensing Specialist PLP Otto Gustafson Manager, Licensing PNPS Frank McGinnis Licensing Specialist IV RBS Danny Williamson Sr. Licensing Specialist VY Robert Wanczyk Manager, Licensing W3 Michael Mason Sr. Licensing Specialist N/A NP NA NA HQN T.R. Jones Sr. Staff Engineer Site and NMM Procedures Canceled or Superseded By This Revision Process Applicability Exclusion: All Sites:

Specific Sites: ANO BRP GGNS IPEC JAF PLP PNPS RBS VY W3 NP Change Statement Revision 8 contains editorial changes including the following:

- Update References based on NRC issuing Revision 3 of NUREG 1022. There are no changes to the reporting regulations cited in this procedure. Changes in reporting guidance are implemented by a non-editorial revision of REAP which is referenced in this procedure.

- Delete References NRC Information Notices 88-64, 85-27, and 85-89; NUREG 1022 contains applicable information.

- Add NERC Standard NUC-001 per LO-WTHQN-2011-084, CA-230

- Add section continuation headers per Writers Manual formatting requirements

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 2 OF 33 Event Notification and Reporting TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ........................................................................................ 3

2.0 REFERENCES

................................................................................. 3 3.0 DEFINITIONS ................................................................................... 7 4.0 RESPONSIBILITY ............................................................................ 9 5.0 DETAILS ......................................................................................... 10 6.0 INTERFACES ................................................................................. 16 7.0 RECORDS ...................................................................................... 16 8.0 SITE SPECIFIC COMMITMENTS .................................................. 17 9.0 ATTACHMENTS ............................................................................. 17 ATTACHMENT 9.1 REPORTING REQUIREMENTS ..................................................... 18 ATTACHMENT 9.2 PAST OPERABILITY EVALUATIONS .............................................. 31

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 3 OF 33 Event Notification and Reporting 1.0 PURPOSE Various Nuclear Regulatory Commission (NRC) and federal rules and regulations require the submittal of reports. The purpose of this procedure is to identify responsibilities and specify a process for the identification and development of these reports.

2.0 REFERENCES

[1] Federal Regulatory References (a) U. S. Code of Federal Regulations, title 10 (10 CFR).

(b) U. S. Code of Federal Regulations, title 29 (29 CFR).

(c) U. S. Code of Federal Regulations, title 40 (40 CFR).

(d) NRC Regulatory Guide 1.16, "Reporting of Operating Information - Appendix A Technical Specifications" (e) NRC Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants" (f) NRC Regulatory Guide 5.62, "Reporting of Safeguards Events" (g) NRC Regulatory Guide 10.1, "Compilation of Reporting Requirements for Persons Subject to NRC Regulations NUREG-1460" (h) NUREG-1022, "Event Reporting Guidelines: 10 CFR 50.72 and 50.73" (i) NUREG-1304, Reporting of Safeguards Events (j) NRC Generic Letter 91-02, "Reporting Mishaps Involving LLW (Low Level Radwaste) Forms Prepared for Disposal" (k) NRC Generic Letter 91-03, "Reporting of Safeguards Events" (l) NRC RIS 01-14, Position on Reportability Requirements for Reactor Core Isolation Cooling System Failure.

(m) NRC RIS 05-06, Reporting Requirements for Gauges Damaged at Temporary Job Sites. (Refer to 10 CFR 30.50(c)(3) regarding applicability.)

(n) North American Electric Reliability Corporation (NERC) Standard EOP-004-1, Disturbance Reporting.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 4 OF 33 Event Notification and Reporting 2.0 [1] cont.

(o) NERC Standard CIP-001-1, Sabotage Reporting.

(p) NERC Standard NUC-001, Nuclear Plant Interface Coordination (q) U.S. Department of Energy (DOE) Electric Delivery and Energy Reliability Form OE-417, Electric Emergency Incident and Disturbance Report.

(r) Institute of Nuclear Power Operations (INPO)12-009, ICES Reporting Requirements and Standards.

[2] Corporate References (a) EN-LI-102, "Corrective Action Process" (b) EN-LI-106, "NRC Correspondence" (c) EN-LI-108-01, 10 CFR 21 Evaluations And Reporting (d) EN-NS-200, Security Reporting Requirements (e) EN-OP-104, Operability Determinations (f) EN-RP-143, Source Control (g) EN-FAP-OM-012, Notifications of Off-Normal Situations / Corporate Duty Manager Responsibilities (h) ENN-DC-154, NEIL and ANI Standardized Process (i) EN-NS-102, Fitness For Duty Program (j) EN-DC-202, NEI 03-08 Materials Initiative (k) EN-OM-119, "On-Site Safety Review Committee (l) EN-OM-123, Fatigue Management Program (m) EN-RP-113, Response to Contaminated Spills/Leaks (n) NEI 03-08, Guideline for the Management of Materials Issues.

(o) NEI 07-07, Industry Ground Water Protection Initiative for Proper Tritium Leaks or Spills Reporting.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 5 OF 33 Event Notification and Reporting 2.0 cont.

[3] ANO References (a) National Pollutant Discharge Elimination System (NPDES) Permit No.

AR0001392

[4] FitzPatrick References (a) AP-03.04, Information Reporting Requirements (b) AP-09.03, Oil Spill Prevention Control and Countermeasure Plan (c) State Pollutant Discharge Elimination System (SPDES) Permit No. NY-0020109

[5] GGNS References (a) Administrative Procedure 01-S-15-1, GGNS Plant Reporting Requirements (b) Section Procedure 10-S-01-6, Notification of Offsite Agencies and Plant On-Call Emergency Personnel (c) Emergency Plan Procedure 10-S-01-1, Activation of the Emergency Plan (d) Operating Procedure 01-S-06-5, Reportable Events or Conditions (e) NPDES Permit No. MS0029521

[6] Indian Point Energy Center References (a) IP-SMM-LI-108, Event Notification and Reporting (b) Nuclear Security Procedure Number 12, Security and Fitness for Duty Event Reporting (c) SMM-EV-101, IPEC Spill/Release Reporting (d) Appendix B to Technical Specifications - Environmental Protection Plan (e) SPDES Permit Nos.

NY 000 4472 (IP1, 2 and 3)

NY 025 0414 (Simulator Transformer Vault)

NY 023 4826 (Buchanan Gas Turbine Site)

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 6 OF 33 Event Notification and Reporting 2.0 cont.

[7] Pilgrim References (a) PNPS 1.3.9, Reports (b) PNPS 1.3.12, Notification and Recall of Personnel (c) FAA AC 70/7460-IJ, Obstruction Marking and Lighting (1-1-96)

(d) RRG 97-112, Stack Navigation Lights (e) PNPS Offsite Dose Calculation Manual (f) PNPS 1.3.22, Oil Spill Prevention and Countermeasure Plan (g) PNPS 5.5.1, General Fire Procedure (h) PNPS 5.5.2, Special Fire Response Procedure (i) PNPS 5.5.3, Medical Emergency Response Procedure (j) PNPS 5.4.4, Response to Hazardous Material Incidents (k) EP-IP-100, Emergency Classification and Notification (l) NPDES Permit (U.S. EPA #MA0003557/Mass. DPH #359)

[8] RBS References (a) Federal National Pollutant Discharge Elimination System (NPDES Permit #

LA0042731)

(b) NPM Procedure RBNP-026, Reporting of Defects and Noncompliances (c) SSM Procedure EIP-2-006, Notifications

[9] Vermont Yankee References (a) AP-0010, Situational Reporting Requirements (b) AP-0069, Routine Reports due to State and Federal Agencies (c) AP-0156, Notifications and Reports Due (d) AP-0138, State Regulatory Commitment Tracking

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 7 OF 33 Event Notification and Reporting 2.0 cont.

[10] W3 References (a) UNT-006-010, Event Notification and Reporting (b) W2.301, Identification, Evaluation, and Reporting Process for 10 CFR 21 Compliance (c) NPDES Permit No. LA0007374 3.0 DEFINITIONS

[1] Condition Report (CR) - A computer generated or paper form used to identify and document issues as defined in and governed by procedure EN-LI-102.

[2] Immediate Notification - Verbal report to the NRC or other outside agencies as required by regulations, and/or the operating license. Immediate notifications are those required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less from the time of discovery or occurrence of the reportable condition or event.

[3] INPO Consolidated Event System (ICES) - Web-based data entry software that integrates the Nuclear Network Operating Experience (OE) and Construction Experience (CE) with Equipment Performance and Information Exchange (EPIX) failure reporting into one consolidated event reporting and analysis platform.

[4] Licensee Event Report (LER) - A report required to be submitted to the NRC in accordance with 10 CFR 50.73. Initiating conditions and the required format, distribution, and timeliness of LERs are described in detail in that regulation and in the NRC document NUREG-1022, Revision 3, "Event Reporting Guidelines 10 CFR 50.72 and 50.73", dated January 2013.

[5] NERC Standards - Electric reliability standards developed by the North American Electric Reliability Corporation (NERC) which define obligations or requirements of utilities and other entities that operate, plan, and use the bulk power system in North America. The authority to develop and enforce mandatory standards is delegated to NERC by the U. S. Federal Energy Regulatory Commission.

[6] Paperless Condition Reporting System (PCRS) - A computer program used for documenting CRs and tracking actions resulting from the processes described in EN-LI-102.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 8 OF 33 Event Notification and Reporting 3.0 cont.

[7] Past Operability Evaluation - An evaluation performed to determine whether a degraded or non-conforming condition affected the Operability of a structure, system, or component (SSC) prior to the time that the condition was discovered. This evaluation does not have to be prepared in accordance with EN-OP-104 but the basic concepts supplied in EN-OP-104 are applicable to reaching a determination on past operability. The evaluation is used primarily to assist in determining reportability of the condition.

[8] Reportability review - A review of an identified condition to determine whether it is required to be reported.

[9] Reportability Evaluation Assistance Program (REAP) - A computer based compilation of regulatory reporting requirements including discussions, clarifications, and interpretations of reporting requirements. REAP also contains a logic process for determining the potentially applicable regulations regarding specific plant events and/or conditions as well as required reporting timeframes and recipients.

[10] Reportable Condition - Any situation (for example, equipment inoperability, design issue, declaration of an emergency class) that is reportable to the NRC or other outside agencies.

[11] Reportable Event - Any occurrence (for example, Engineered Safety Feature actuation, declaration of an emergency class) that is reportable to the NRC or other outside agencies.

[12] Reporting - As used in this procedure, means providing information in accordance with governing rules and regulations. This reporting may be written or verbal. Reports may be generally categorized as follows:

Periodic - reports required to be submitted in accordance with some specified periodicity; for example, Personnel Occupational Exposure Data Reports, etc.

Condition-Related - reports required to be submitted as a result of some identified adverse condition or event (for example, LERs). These reports may be further categorized as immediate or non-immediate.

Process Driven - reports based on regulatory requirements that do not specify any calendar periodicity and are not the result of an identified adverse condition (for example, reports of changes to the Security or Emergency Plans that do not reduce the effectiveness of those Plans, Reactor Startup Reports).

10 CFR Part 21 - reports driven by 10 CFR Part 21 requirements.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 9 OF 33 Event Notification and Reporting 4.0 RESPONSIBILITY

[1] The VP, Nuclear Safety, EP & Licensing is responsible for maintaining this procedure.

[2] Each sites Department Heads are responsible for implementing this procedure within their departments.

[3] Manager, Licensing is responsible for:

(a) Implementing this procedure at the site.

(b) Assigning responsibility for preparing and processing NRC reports.

(c) Assisting the Operations Department in identifying immediate NRC reporting requirements, if requested.

(d) Reviewing the Operations Departments immediate reportability determinations.

(e) Performing reportability determinations to identify NRC notification requirements other than immediate reporting requirements.

(f) Assisting other departments in evaluating conditions not typically documented in Condition Reports (for example, Fitness for Duty events) for reportability.

(g) Coordinating the development of information necessary to retract notifications made to agencies if it is determined that the initial notification was not required.

(h) Ensuring that periodic and process driven reports are submitted to the appropriate agencies, as required.

[4] Operations Department is responsible for evaluating Condition Reports initiated under procedure EN-LI-102 to identify conditions that must be reported immediately to the NRC or other agencies.

[5] Manager/Superintendent, Plant Security is responsible for assisting the Shift Manager in determining the reportability of safeguards events and in making such reports, as required.

[6] Manager, Environmental/Chemistry/Technical Support is responsible for assisting the Shift Manager in determining the reportability of environmental events and in making such reports, as required.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 10 OF 33 Event Notification and Reporting 4.0 cont.

[7] Licensing Staff Members are responsible for:

(a) Reviewing conditions and events for reportability as assigned by the Manager, Licensing.

(b) Issuing Condition Report / Corrective Actions (CAs) per EN-LI-102 to have Past Operability Evaluations performed on identified conditions when needed for a reportability determination.

(c) Promptly informing the Manager, Licensing if a reportability review indicates a report is required and coordinating the development of information to ensure that the report is made.

(d) Promptly informing the Manager, Licensing in the event that an immediate report made to an outside agency is determined to be not required and coordinating the development of information to ensure that the retraction is made.

(e) Preparing and processing assigned written reports to outside agencies in accordance with site and corporate procedures and in accordance with appropriate regulatory requirements and guidance documents.

5.0 DETAILS

[1] PRECAUTIONS AND LIMITATIONS None

[2] GENERAL REQUIREMENTS (a) Reports required by state and local authorities are addressed in site-specific procedures.

(b) Emergency plan notifications are specified in site emergency planning procedures.

(c) The Reportability Evaluation Assistance Program (REAP), Attachment 9.1 (Reporting Requirements), governing regulations, applicable site procedures, NUREG-1022, or other NRC or industry information sources may be used in determining the reportability of conditions or events. Licensing may also be contacted for assistance in determining reportability.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 11 OF 33 Event Notification and Reporting 5.0 [2] cont.

(d) Written correspondence (for example, reports and letters) to the NRC should be prepared, reviewed and submitted in accordance with EN-LI-106.

(e) Reporting for multiple sites is permissible; however for submittals that are security related, consideration must be given to the reporting timeframe. The requirements for handling safeguards information can make communications between sites difficult and time consuming. Only a report with a long reporting time frame should be considered for multiple site reporting.

[3] IMMEDIATE REPORTS NOTE The Reportability Evaluation Assistance Program (REAP) contains guidance to assist in determining the reportability of plant conditions and events and includes the required reporting timeframes and recipients for each criterion. It may be used in performing reportability determinations, as needed.

(a) Reporting of conditions or events that are not documented in Condition Reports (PCRS) (for example, fitness-for-duty events and logable safeguards events) is the responsibility of the department administering the applicable process.

(b) The Operations Department is responsible for performing an operability determination for conditions and events documented in Condition Reports.

Operability determinations are conducted in accordance with EN-OP-104 or other applicable site procedures.

(c) Upon completion of the operability determination, the Operations Department shall evaluate the condition for immediate reportability.

(d) If the stated condition does not meet any immediate reporting criterion, the Operations Department shall enter the appropriate code in the Condition Report. The basis for the conclusion should be documented in the operability description field of the Condition Report.

(e) If the condition is immediately reportable, the Operations Department informs:

(1) Site Management [for example, the Site Duty Manager; General Manager, Plant Operations; Manager, Operations; Director, Nuclear Safety Assurance].

a. The Site Duty Manager is to notify the Corporate Duty Manager of any offsite agency notifications in accordance with EN-FAP-OM-012 (Reference [2](g)).

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 12 OF 33 Event Notification and Reporting 5.0 [3](e) cont.

(2) Appropriate offsite agencies.

(3) NRC Resident Inspector (4) The Site Vice President and the On-Site Safety Review Committee (OSRC) Chairperson shall be notified immediately if the reportable condition is a technical specifications safety limit violation.

(f) If time allows, NRC Form 361 Reactor Plant Event Notification Worksheet, (available on the NRC website) should be used to record information used in making NRC notifications. The form should be telefaxed, or emailed, to the NRC Operations Center at the time the notification is made. The completed form should also receive a peer/management/Licensing review prior to submittal if time allows.

(g) The reportability determination and notifications made shall be documented in the Condition Report.

[4] INPO Document Reporting Requirements (a) ICES reports are required for each of the conditions or events leading to issuance of the following documents within the time frames specified.

LERs (10 CFR 50.73) - two weeks from issuance of an LER or a revision to an LER (Records of retracted LERs may be retired.)

Plant Event Notices (PENs) (10 CFR 50.72) - five business days from issuance of a PEN (Records of retracted PENs may be retired.)

[5] FOLLOW-UP REPORTS (a) The Manager, Licensing shall assign to appropriate staff personnel the responsibility for:

Reviewing immediate reportability determinations made by Operations.

Reviewing and/or evaluating the conditions and events documented in Condition Reports for reportability.

NOTE The assigned staff member may use the REAP, applicable site procedures, or the guidance referenced in Attachment 9.1 (Reporting Requirements) to assist in the reportability evaluation.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 13 OF 33 Event Notification and Reporting 5.0 [5] cont.

(b) In the event the reportability review indicates that an immediate report is required, but has not been made, the staff member shall promptly inform the Manager, Licensing for action to ensure that the required notification is made.

(1) If it is determined that the immediate notification has not been made, the staff member shall coordinate development of the information necessary to ensure that the notification is made. The circumstances associated with this delayed notification shall be documented in the Condition Report (PCRS) or documented by initiating a new Condition Report.

(c) In the event that the reportability review indicates that an immediate report made to an outside agency was not required, the staff member shall promptly inform the Manager, Licensing for action to determine if the notification should be retracted.

(1) If it is determined that the immediate notification is to be retracted, the staff member shall coordinate development of the information necessary to retract the notification and to ensure that the retraction is made. The revised reportability status shall be documented in the Condition Report (PCRS).

(d) In cases where the reportability review indicates that a written report to an outside agency is required, the staff member shall inform the Manager Licensing. As applicable, the Manager, Licensing will assign a staff member to prepare and process the submittal in accordance with site and corporate procedures and in accordance with appropriate regulatory requirements and guidance documents.

(e) In cases where the Condition Report identifies a potentially reportable condition (for example, past operability concern) but does not contain sufficient information to conclusively determine reportability, the reportability determination shall be classified as Indeterminate until the additional information required to determine reportability (for example, engineering evaluation, past operability) is received and reviewed.

To obtain the information necessary to determine reportability, the Licensing staff may issue a corrective action to the appropriate department in accordance with EN-LI-102. Attachment 9.2 provides additional guidance for past operability determinations.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 14 OF 33 Event Notification and Reporting 5.0 [5](e) cont.

The due date for the action item should be commensurate with the safety significance of the condition. It should be limited to 30 days from the date of initiation of the Condition Report unless there is reasonable assurance that the condition will ultimately be determined non-reportable.

(f) EN-LI-102 provides guidance to the Condition Review Group (CRG) for categorizing Condition Reports. The categorization in part is based upon whether the condition or event is reportable.

If a condition that was initially determined to be not reportable or indeterminate is later determined to be reportable, the Condition Report should be reviewed again by CRG for potential reclassification.

In this case, Licensing should either initiate a new Condition Report to document the reportable condition or issue an action to the Corrective Action Group (CAA) in the Condition Report to have the CRG review the Condition Report for potential categorization change.

(g) The time clock for reportable events shall be in accordance with the guidance in NUREG-1022 and generally starts at the time of occurrence or discovery of the event.

If a Licensee Event Report (LER) is not submitted within 60 days from the event date, the relationship between the event date, discovery date, and report date should be explained in the LER narrative.

(h) LERs made in accordance with 10 CFR 50.73 shall be reviewed by the OSRC on a periodic basis (either before or after submittal to the NRC) in accordance with EN-OM-119.

[6] PERIODIC AND PROCESS DRIVEN REPORTS AND OTHER REPORTS (a) The applicable Department is responsible for assuring that periodic and process driven reports are initiated and for coordinating with the Licensing Department to assure their timely submittal.

(b) Independently, the Licensing Department maintains cognizance of the status of pending periodic and process driven reports required by NRC or Federal regulation using the site commitment management system (or other proceduralized process) and initiates actions to assure that these reports are submitted as required.

5.0 cont.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 15 OF 33 Event Notification and Reporting

[7] 10 CFR Part 21 REPORTS (a) If a Part 21 evaluation concludes that a potential defect or noncompliance could create a substantial safety hazard, the defect or non compliance is reported to the NRC. The following actions are required:

(1) Notify the NRC by facsimile (preferred) or by telephone to the NRC Operations Center within 2 days following receipt of information by the director or responsible corporate officer of a reportable defect or noncompliance.

(2) Provide a follow-up written report to the NRC (Document Control Desk) within 30 days following receipt of information by the director or responsible officer of a reportable defect or noncompliance.

(3) Provide an interim report to the NRC (Document Control Desk) within 60 days of discovery of the deviation or failure to comply if an evaluation cannot be completed within 60 days of discovery. The interim report should describe the deviation or failure to comply that is being evaluated and should also state when the evaluation will be completed.

(b) 10 CFR Part 21 (Part 21), Reporting of Defects and Noncompliance, allows licensees of operating nuclear power plants to reduce duplication by evaluating and reporting Part 21 defects under 10 CFR 50.72, 50.73 or 73.71. This reporting satisfies the requirements of Part 21. However, if a part is on the shelf then a defect would not be reportable under Parts 50 but may be reportable under Part 21.

(c) The Licensing Department should use the following guidance for reporting:

For an LER, if a defect meets one of the criteria of 50.73, Item 11 of NRC Form 366 (LER Form) should be checked. The Other block should be checked and Part 21 should be indicated in the space immediately below when applicable. Item 16 (Abstract) and 17 (Text) should state that the report constitutes a Part 21 notification. The other involved facilities in Item 8 on the LER form may be used if one LER is written for multiple sites.

For conditions reportable under 10 CFR 73.71, Item 11 of the LER form should be checked. If safeguards information is included, LER forms may still be used, but should be appropriately marked. The text should clearly indicate any information that is security or safeguards information.

When transmitting safeguards information, the requirements of 10 CFR 73.71 must be met.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 16 OF 33 Event Notification and Reporting 5.0 [7](c) cont.

If proprietary information is included, Item 17 (text) should be marked appropriately. Include proprietary information only in the narrative and not in the abstract. Clearly indicate what information is proprietary. The requirements of 10 CFR 2.390(b) must be met when transmitting proprietary information.

[8] NEI 03-08 MATERIALS ISSUES REPORTS (a) EN-DC-202 provides Entergy Nuclears program plan for managing nuclear industry materials program issues, including the implementation of the Nuclear Energy Institutes (NEI) Materials Initiative referred to as NEI 03-08. NEI 03-08 guidelines require NRC notifications under certain circumstances. Notifications shall be made in accordance with EN-DC-202.

6.0 INTERFACES

[1] EN-AD-103, Document Control and Records Management Programs

[2] EN-LI-102, "Corrective Action Process"

[3] EN-LI-106, NRC Correspondence

[4] EN-NS-200, Security Reporting Requirements

[5] EN-OP-104, Operability Determinations

[6] EN-DC-202, NEI 03-08 Materials Initiative

[7] EN-OM-119, "On-Site Safety Review Committee

[8] EN-RP-113, Response to Contaminated Spills/Leaks

[9] NEI 07-07, Industry Ground Water Protection Initiative for Proper Tritium Leaks or Spills Reporting.

7.0 RECORDS

[1] Documents prepared in accordance with this procedure are to be stored, annotated, and retained as specified in EN-AD-103 and/or applicable procedures.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 17 OF 33 Event Notification and Reporting 8.0 SITE SPECIFIC COMMITMENTS Document Document Section Procedure Section Site Applicability None N/A N/A N/A 9.0 ATTACHMENTS

[1] Attachment 9.1 - Reporting Requirements

[2] Attachment 9.2 - Past Operability Evaluations

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 1 of 13 10 CFR PART 19 NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS, INSPECTIONS 10 CFR 19.13(b) Annual report of radiation exposure 10 CFR 19.13(c) Notify of radiation exposure at former worker's request 10 CFR 19.13(d) Events involving byproduct, source, or Special Nuclear Material (SNM) causing significant exposure, release, loss of facility, or property damage.

Events involving licensed material causing exposure, release, loss of facility, or property damage Individual exposure and radiation levels 10 CFR PART 20 STANDARDS FOR PROTECTION AGAINST RADIATION 10 CFR 20.1206(g) Individual exposure and radiation levels 10 CFR 20.1906(d)(1) Removable contamination on package surface 10 CFR 20.1906(d)(2) Excessive radiation levels on or near package surface 10 CFR 20.2201(a)(1) Loss or theft of licensed material which could expose persons in unrestricted areas 10 CFR 20.2201(b) Loss or theft of licensed material which could expose persons in unrestricted areas 10 CFR 20.2201(d) Loss or theft of licensed material which could expose persons in unrestricted areas 10 CFR 20.2202(a) Events involving loss of control of byproduct, source, or SNM causing significant exposure or release 10 CFR 20.2202(b) Events involving licensed material causing exposure or release 10 CFR 20.2203(a)(1) Events involving byproduct, source, or SNM causing significant exposure or release Events involving loss of control of licensed material causing exposure or release Individual exposure and excessive radiation levels 10 CFR 20.2203(a)(2) Individual exposure and excessive radiation levels 10 CFR 20.2203(a)(3) Individual exposure and excessive radiation levels 10 CFR 20.2203(a)(4) Individual exposure and excessive radiation levels 10 CFR 20.2204 Individual exposure and excessive radiation levels 10 CFR 20.2205 Events involving byproduct, source, or SNM causing significant exposure or release Events involving loss of control of licensed material causing exposure or release Individual exposure and excessive radiation levels Annual report of radiation exposure 10 CFR 20.2206 Annual report of radiation exposure 10 CFR 20.2207 Reports of transactions involving nationally tracked sealed sources (involves manufacture, transfer, receipt, disassembly or disposal of nationally tracked sealed sources).

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 19 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 2 of 13 10 CFR PART 21 REPORTING OF DEFECTS AND NONCOMPLIANCES 10 CFR 21.21(c) Existence of a defect in a procured component 10 CFR PART 26 FITNESS FOR DUTY PROGRAMS 10 CFR 26.719 Significant FFD policy violations or programmatic failures, and drug and alcohol testing errors. (Note that FFD policy includes the Fatigue Management Program) 10 CFR PART 30 RULES OF GENERAL APPLICABILITY TO DOMESTIC LICENSING OF BYPRODUCT MATERIAL (NOTE: Not applicable to events reported under 10 CFR 50.72) 10 CFR 30.9 Inaccurate or incomplete information submitted to the NRC 10 CFR 30.50 (a) Event that prevents immediate protective actions necessary to avoid exposures in excess of regulatory limits 10 CFR 30.50 (b) An unplanned contamination event; or equipment is disabled or fails to function as designed; or an unplanned fire or explosion damaging licensed material 10 CFR PART 31 GENERAL DOMESTIC LICENSES FOR BYPRODUCT MATERIAL 10 CFR 31.5(c)(5) Failure of shielding or detection of removable radioactive material associated with byproduct material 10 CFR 31.5(c)(8) Transfer of device containing by-product material to a specific licensee 10 CFR 31.5(c)(9)(i) Transfer of a device containing byproduct material to a general licensee 10 CFR PART 34 LICENSES FOR RADIOGRAPHY AND RADIATION SAFETY REQUIREMENTS FOR RADIOGRAPHIC OPERATIONS 10 CFR 34.25(d) Sealed source leakage 10 CFR 34.30(a) Events involving radiographic equipment 10 CFR PART 40 DOMESTIC LICENSING OF SOURCE MATERIAL 10 CFR 40.9 Inaccurate or incomplete information submitted to the NRC 10 CFR 40.25(c)(1) Receipt of depleted uranium 10 CFR 40.25(c)(2) Changes in information on Form NRC 244 10 CFR 40.25(d)(3) & Transfer of depleted uranium when the transferor has a general license (d)(4) 10 CFR 40.35(d) Transfer of depleted uranium when the transferor has a specific license 10 CFR 40.64(a) Inventory of uranium or thorium 10 CFR 40.64(c) Attempt or perceived attempt to commit a theft or unlawful diversion of uranium or thorium 10 CFR 40.67(a) Importation of natural uranium from countries not party to the Convention on the Physical Protection of Nuclear Material

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 20 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 3 of 13 10 CFR PART 50 DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES 10 CFR 50.9 Inaccurate or incomplete information submitted to the NRC 10 CFR 50.9(b) Information with significant implication 10 CFR A safety limit is exceeded 50.36(c)(1)(i)(A) 10 CFR Automatic safety system does not function as required 50.36(c)(1)(ii)(A) 10 CFR 50.36(c)(2) A limiting condition for operation is not met 10 CFR 50.46(a)(3)(ii) Change or error in approved ECCS evaluation model 10 CFR 50.54(p) Changes to security, guard qualification, and safeguards contingency plans made without prior approval 10 CFR 50.54(q) Changes in emergency plans without prior approval 10 CFR 50.54(w)(4) Nuclear emergency where financial assistance may be required 10 CFR 50.54(x) Deviation from Technical Specifications authorized by 10 CFR 50.54(x) 10 CFR 50.54(bb) Decommissioning of a power reactor 10 CFR 50.54(cc)(1) Filing for bankruptcy 10 CFR 50.55(e)(3) Construction/design deficiencies for a construction permit holder 10 CFR 50.55(e)(4) Construction/design deficiencies for a construction permit holder 10 CFR 50.72(a)(1) Declaration of an Emergency Class 10 CFR 50.72(a)(4) Declaration of an Emergency Class 10 CFR 50.72(b)(1) Deviation from Technical Specifications authorized by 10 CFR 50.54(x) 10 CFR 50.72(b)(2)(i) Initiation of shutdown required by the Technical Specifications 10 CFR 50.72(b)(2)(iv) RPS or ESF Actuation 10 CFR 50.72(b)(2)(xi) News Release or Government Agency Notification 10 CFR 50.72(b)(3)(ii) Degraded or Unanalyzed Condition 10 CFR 50.72(b)(3)(iv) RPS or ESF Actuation 10 CFR 50.72(b)(3)(v) Prevented Safety Function 10 CFR 50.72(b)(3)(xii) Transport of Contaminated Person to Medical Facility

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 21 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 4 of 13 10 CFR 50.72(b)(3)(xiii) Major Loss of Assessment, Response, or Communications Capability 10 CFR Completion of shutdown required by Technical Specifications 50.73(a)(2)(i)(A) 10 CFR Operation prohibited by Technical Specifications 50.73(a)(2)(i)(B) 10 CFR Deviation from Technical Specifications authorized by 10 CFR 50.54(X) 50.73(a)(2)(i)(C) 10 CFR 50.73(a)(2)(ii) Degraded or Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) External condition that poses an actual threat 10 CFR 50.73(a)(2)(iv) Reactor Protection System (RPS) or Engineered Safety Feature (ESF)

Actuation 10 CFR 50.73(a)(2)(v) Prevented Safety Function 10 CFR 50.73(a)(2)(vii) Safety System Interaction 10 CFR Airborne Radioactive Release 50.73(a)(2)(viii)(A) 10 CFR Liquid Effluent Release 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix) Safety System Interaction 10 CFR 50.73(a)(2)(x) Internal condition that poses an actual threat 10 CFR 50.74(a) Permanent reassignment of licensed operator 10 CFR 50.74(b) Termination of licensed operator 10 CFR 50.82 Decommissioning of a power reactor 10 CFR 50, App. E, Declaration of an Emergency Class Sec. IV.D.3 10 CFR 50, App. E, Exercise scenarios before use in a full participation exercise Sec. IV.F.2.a 10 CFR 50, App. E, Evacuation Time Estimate analyses using the decennial census data, Sec. IV.4 and 6 10 CFR 50, App. E, Emergency Response Data System (ERDS) Changes Sec. VI.3.a 10 CFR 50, App. E, ERDS Changes Sec. VI.3.b 10 CFR 50, App. G, Programs on beltline material fracture toughness Sec. V.E 10 CFR 50, App. H, Results of beltline material fracture toughness tests Sec. III 10 CFR 50, App. I, Effluent release Sec. IV.A

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 22 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 5 of 13 10 CFR PART 55 OPERATORS' LICENSES 10 CFR 55.25 Operator disability 10 CFR 55.53(g) Felony conviction of licensed operator 10 CFR PART 70 DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL 10 CFR 70.9 Inaccurate or incomplete information submitted to the NRC 10 CFR 70.32(a)(9)(i) Filing for bankruptcy 10 CFR 70.32(c)(2) Changes to the material control and accounting program 10 CFR 70.32(d) Changes to the plan for physical protection of SNM in transmit made without prior approval 10 CFR 70.32(e) Changes to a security plan made without prior approval 10 CFR 70.32(g) Changes to the safeguards contingency plan procedures made without prior approval 10 CFR 70.52(a) Accidental criticality or loss of SNM 10 CFR PART 71 PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL 10 CFR 71.7 Completeness and accuracy of information 10 CFR 71.93(c) Fabrication of shipment package with a decay heat load 10 CFR 71.95 Reduced effectiveness or defects in packaging 10 CFR PART 72 LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE 10 CFR 72.11 Inaccurate or incomplete information submitted to the NRC 10 CFR 72.44(b)(6) Notification of filing for bankruptcy 10 CFR 72.44(f) Changes to the emergency plan of an Independent Spent Fuel Storage Installation (ISFSI) or Monitored Retrievable Storage (MRS).

10 CFR 72.74 Accidental criticality or loss of SNM associated with an ISFSI or an MRS 10 CFR 72.75(a) Declaration of an Emergency Class 10 CFR 72.75(b)(1) Deviation from Technical Specifications authorized by 10 CFR 50.54(x) 10 CFR 72.75(b)(2) Any event or situation related to the health and safety of the public, onsite personnel, or protection of the environment for which a news release or other agency notification is planned or made.

10 CFR 72.75(c)(1) A defect in any spent fuel, High-level radioactive waste (HLW), or reactor-related Greater Than Class C (GTCC) waste storage SSC important to safety.

10 CFR 72.75 (c)(2) A significant reduction in the effectiveness of spent fuel, HLW, or reactor-related GTCC waste storage confinement system during use.

10 CFR 72.75(c)(3) Transport of contaminated person to medical facility

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 23 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 6 of 13 10 CFR 72.75(d)(1) Important safety equipment is disabled or fails to function and no redundant equipment was available.

10 CFR 72.78(a) Nuclear Material Transfer Reports for an ISFSI or MRS 10 CFR 72.212(b)(1) Storage of spent fuel in cask 10 CFR PART 73 PHYSICAL PROTECTION OF PLANTS AND MATERIALS 10 CFR 73.26(i)(6) Failure to receive periodic call during shipment of SNM 10 CFR 73.27 Notification requirements 10 CFR 73.27(a)(1) Delivery of strategic SNM 10 CFR 73.27(b) Receipt of shipment of strategic SNM Failure of shipment of strategic SNM to arrive at destination 10 CFR 73.27(b) & (c) Action taken to trace shipment of strategic SNM 10 CFR 73.67(e)(3)(vii) Loss of any shipment of SNM or spent fuel Recovery of, or accounting for, the lost shipment of SNM or spent fuel 10 CFR 73.67(g)(3)(iii) Loss of any shipment of SNM or spent fuel Recovery of, or accounting for, the lost shipment of SNM or spent fuel 10 CFR 73.71(a) Recovery of, or accounting for, the lost shipment of SNM or spent fuel

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 24 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 7 of 13 10 CFR 73.71(b) Complete loss of offsite communications Loss of security alarm monitoring or assessment capability Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Fire or explosion of suspicious or unknown origin Demonstrations or civil disturbances near site, assault on plant Threatened or actual tampering with or damage to safety or security equipment Actual/attempted theft or unlawful diversion of SNM Below minimum security force or actual/imminent strike by security force Loss of security weapon at site Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information 10 CFR 73.71(c) Loss of security alarm monitoring or assessment capability Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Threatened or actual tampering with or damage to safety or security equipment Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information 10 CFR 73.72 Plans to deliver or take delivery of SNM or irradiated reactor fuel

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 25 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 8 of 13 10 CFR 73, App. G, Complete loss of offsite communications para. I(a)(2) & I(a)(3) 10 CFR 73, App. G, Loss of security weapon at site para. I(a)(3) 10 CFR 73, App. G, Intrusion by unauthorized person into controlled area para. I(b) 10 CFR 73, App. G, Loss of security alarm monitoring or assessment capability para. I(c) Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Below minimum security force or actual/imminent strike by security force Inattentive Security Officer Falsified, lost or stolen badges or key cards Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency 10 CFR 73, App. G, Actual/attempted introduction of contraband para. I(d) Fire or explosion of suspicious or unknown origin 10 CFR 73, App. G, Loss of security alarm monitoring or assessment capability para. II(a) Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Fire or explosion of suspicious or unknown origin Threatened or actual tampering with or damage to safety or security equipment Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 26 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 9 of 13 10 CFR 73, App. G, Loss of security alarm monitoring or assessment capability para. II(b) Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Threatened or actual tampering with or damage to safety or security equipment Fire or explosion of suspicious or unknown origin Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 27 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 10 of 13 10 CFR 73.71(b) Complete loss of offsite communications Loss of security alarm monitoring or assessment capability Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Fire or explosion of suspicious or unknown origin Demonstrations or civil disturbances near site, assault on plant Threatened or actual tampering with or damage to safety or security equipment Actual/attempted theft or unlawful diversion of SNM Below minimum security force or actual/imminent strike by security force Loss of security weapon at site Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information 10 CFR 73.71(c) Loss of security alarm monitoring or assessment capability Loss of AC power to security systems Loss of an alarm or lock on a vital portal or material access area Safeguard system failures or vulnerabilities Actual/attempted introduction of contraband Onsite criminal act Offsite criminal act Bomb or extortion threat Threatened or actual tampering with or damage to safety or security equipment Inattentive Security Officer Falsified, lost or stolen badges or key cards Intrusion by unauthorized person into controlled area Loss of intrusion detection capability within a single zone Suspension of safeguards controls during emergency Potential compromise of safeguards information 10 CFR 73.72 Plans to deliver or take delivery of SNM or irradiated reactor fuel

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 28 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 11 of 13 10 CFR PART 74 MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL 10 CFR 74.11 Reports of loss or theft or attempted theft or unauthorized production of special nuclear material 10 CFR 74.13(b) Inventoried SNM difference 10 CFR 74.15(a) Transfer and receipt of SNM of 1 gram or more of contained uranium-235, uranium-233, or plutonium 10 CFR PART 75 SAFEGUARDS ON NUCLEAR MATERIAL - IMPLEMENTATION OF USA/IAEA AGREEMENT 10 CFR 75.32 Initial inventory report (IAEA Safeguards) 10 CFR 75.34 Changes to inventory of nuclear material (IAEA Safeguards) 10 CFR 75.35 Material status report for physical inventory (IAEA Safeguards) 10 CFR 75.36(b)&(c) Safeguards situations described in licensed conditions (IAEA Safeguards) 10 CFR 75.44(a)(1) Advance notifications of exports and domestic transfers (IAEA Safeguards) 10 CFR 75.44(a)(2) Advance notification of unpackaging of nuclear material imports (IAEA Safeguards) 10 CFR 75.44(c) Delay in receipt or shipment of nuclear material for which advance notification is required (IAEA Safeguards) 10 CFR PART 95 SECURITY FACILITY APPROVAL AND SAFEGUARDING OF NATIONAL SECURITY INFORMATION AND RESTRICTED DATA 10 CFR 95.25(h) Unattended security container (with NSI or RD) found opened 10 CFR 95.57(a)&(b) Violation related to NSI or RD 10 CFR 95.57(c) Document containing NSI or RD is generated, declassified or its classification changed 10 CFR PART 110 EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL 10 CFR 110.50(a)(7) Knowledge of potential packaging deficiencies prior to exporting or importing

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 29 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 12 of 13 10 CFR PART 140 FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS 10 CFR 140.15(a)(1) Financial protection maintained through liability insurance 10 CFR 140.15(b)(1) Financial protection maintained as specified in 140.14(a)(2) 10 CFR 140.15(e) Change in proof of financial protection 10 CFR 140.17(b) Renewal of liability insurance 10 CFR 140.6(a) Bodily injury or property damage from radioactive material 29 CFR OCCUPATIONAL SAFETY & HEALTH ADMINISTRATION 29 CFR 1904.39 Reporting fatalities and multiple hospitalization incidents to OSHA 40 CFR PART 61 NATIONAL EMISSION STANDARDS FOR HAZARDOUS AIR POLLUTANTS 40 CFR 61.145(a), (b), Demolishment of a facility containing friable asbestos materials

& (c) 40 CFR 61.145(d) Friable asbestos materials in a facility being renovated 40 CFR 61.146(a), Demolishment of a facility containing friable asbestos materials (b)(1), (b)(2), and (b)(3) 40 CFR 61.146(a) & Friable asbestos materials in a facility being renovated (b)(4) 40 CFR PART 112 OIL POLLUTION PREVENTION 40 CFR 112.4 Discharge of oil into or upon navigable waters or adjoining shorelines 40 CFR PART 302 DESIGNATION, REPORTABLE QUANTITIES AND NOTIFICATION 40 CFR 302.6 Release of hazardous substance 40 CFR PART 355 EMERGENCY PLANNING AND NOTIFICATION 40 CFR 355.40(b) Release of hazardous substance Release of extremely hazardous substance 40 CFR PART 370 HAZARDOUS CHEMICAL REPORTING: COMMUNITY RIGHT-TO-KNOW 40 CFR 370.21(c)(1) New information concerning hazardous chemical and (c)(2) 40 CFR PART 761 POLYCHLORINATED BIPHENYLS (PCBs) MANUFACTURING, PROCESSING, DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS 40 CFR 761.125(a)(1) PCB spill

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 30 OF 33 Event Notification and Reporting ATTACHMENT 9.1 REPORTING REQUIREMENTS Sheet 13 of 13 NPDES PERMIT NPDES Permit Noncompliance with National Pollutant Discharge Elimination System (NPDES) permit which may endanger health or the environment Unanticipated bypass which exceeds any effluent limitation Planned changes to the permitted activity or facility which may result in noncompliance Planned bypass which will exceed effluent limitations Noncompliance with NPDES permit not reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Activity resulting in discharge of toxic pollutant not limited in NPDES permit, if that discharge will exceed certain notification levels Begin use of toxic pollutant not reported in NPDES application Failure to submit NPDES relevant facts In the event of a pending change in control or ownership of facilities from which authorized discharge emanate Planned physical alterations to NPDES permitted facility NERC/ DOE NERC Standards Actual physical attack that causes major interruptions or impacts to critical CIP-001-1 and infrastructure facilities or to operations.

EOP-004 (DOE Form Actual cyber or communications attack that causes major interruptions of OE-417) electrical systems operations.

Suspected physical attacks that could impact electric power system adequacy or reliability or vandalism which targets components of any security system.

Suspected cyber or communications attacks that could impact electric power system adequacy or vulnerability.

Fuel supply emergencies that could impact electric power system adequacy or reliability.

Loss of major system component that significantly affects integrity of interconnected system operations (applicable to facilities registered by the North American Electric Reliability Corporation (NERC) as transmission owners -

Fitzpatrick and Pilgrim).

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 31 OF 33 Event Notification and Reporting ATTACHMENT 9.2 PAST OPERABILITY EVALUATIONS Page 1 of 3 PURPOSE:

A past operability evaluation is performed to determine whether a degraded or non-conforming condition affected the operability of a structure, system, or component (SSC) within 3 years prior to the time of discovery.

The evaluation is used primarily to assist in determining reportability of the condition. Potentially reportable conditions are discussed below. Additional guidance is contained in NUREG-1022 "Event Reporting Guidelines:

10 CFR 50.72 and 50.73" REPORTABLE CONDITIONS:

Condition Prohibited by Technical Specifications: 10 CFR 50.73(a)(2)(i)(B) requires a Licensee Event Report (LER) to be submitted when firm evidence shows that the Technical Specification (TS) requirements were not met for INOPERABLE equipment. Below are two situations in which the past operability review may be needed to determine reportability under this regulation:

1. A condition which causes a SSC to be inoperable may be identified during a MODE when the TS requirements for the SSC do not apply (refueling outage, for example). However, the condition may have previously existed when the TS requirements did apply. An evaluation may be needed to determine if there is firm evidence that the condition existed during a time when the SSC was required to be operable within the last 3 years. The condition may be reportable if it existed longer than allowed by the TS or TS Action requirements were not met within the required completion times when the INOPERABILITY first occurred.

2. A condition which causes a SSC to be inoperable may be identified when the TS requirements for the SSC apply and the TS requirements are met at the time of discovery. However, an evaluation may be needed to determine if the condition existed within 3 years of discovery. The condition may be reportable if there is firm evidence that the condition of INOPERABILITY existed longer than allowed by the TS or TS Action requirements were not met within the required completion times when the INOPERABILITY first occurred.

Degraded or Unanalyzed Condition: A degraded or unanalyzed condition as defined in 10CFR 50.73(a)(2)(ii) is reportable if the condition existed anytime within 3 years of the discovery of the condition.

Prevented a Safety Function: 10CFR 50.73(a)(2)(v) further requires that an LER be submitted whenever there is a reasonable expectation that the event or condition could have prevented the fulfillment of certain safety functions (firm evidence is not required for this criterion). If a reasonable expectation exists that the condition could have prevented the fulfillment of a safety function during a period of 3 years prior the discovery, an LER is required. Note that an 8-hour notification is also required per 50.72(b)(3)(v) if the prevented safety function existed at the time of discovery.

ISSUANCE OF PAST OPERABILITY ACTIONS:

If the discovered condition results in an INOPERABLE status, then Licensing may need to issue a Corrective Action (CA) per EN-LI-102 (typically called a "past operability action") to obtain the necessary information to determine if a condition is reportable to the NRC. Licensing may use judgment based upon the available information to determine whether there is a potential for past OPERABILITY to be a reportability concern.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 32 OF 33 Event Notification and Reporting ATTACHMENT 9.2 PAST OPERABILITY EVALUATIONS Page 2 of 3 The following is an example of a past operability CA:

Determine if the as-found condition affected OPERABILITY of the [enter the TS required SSC] prior to the time of discovery (within the past 3 years). If there is firm evidence that this condition would have previously prevented [the TS SSC] from performing its specified safety function (i.e., was INOPERABLE) and the Tech Spec Actions were not met, then notify Site Licensing as soon as possible. If [the TS SSC] was INOPERABLE, then also determine the condition of redundant TS required equipment to determine whether a complete loss of safety function could have occurred due to concurrent INOPERABILITY of redundant equipment. Note that firm evidence is not necessary to determine whether a complete loss of safety function could have occurred, only a reasonable expectation that the condition could have prevented the fulfillment of a safety function. Contact Licensing for additional guidance, if needed. When the evaluation is complete, notify Licensing to update the CR reportability tab.

EVALUATION GUIDANCE Conditions Found By Surveillances NUREG-1022 gives specific guidance for evaluating the reportability of a discrepancy found during TS surveillance testing. The guidance states that for routine surveillance testing, it should be assumed that the discrepancy occurred at the time of its discovery unless there is firm evidence, based on a review of relevant information such as the equipment history and the cause of failure, to indicate that the discrepancy existed previously. If a known event occurred between the two surveillances that likely caused the SSC failure (e.g.,

improper maintenance, testing, or component mispositioning), then the SSC is considered to have been INOPERABLE since the event and is potentially reportable under 10CFR50.73(a)(2)(i)(B). The TS requirements would need to be reviewed to determine if they were met during the period of INOPERABILITY.

Similar surveillance failures in multiple SSCs, may indicate that the conditions arose over a period of time and the failure mode should be evaluated to make this determination. If so, the condition likely existed prior to the time of discovery by the surveillance test and may be reportable.

Firm Evidence:

For conditions that would have rendered TS equipment INOPERABLE within 3 years of the time of discovery, an LER is required if firm evidence exists that a condition existed for a time longer than permitted by the TS, even if the condition was not discovered until after the allowable time had elapsed and the condition was rectified immediately upon discovery. A cause determination and/or failure mode analysis may be required to make the determination. The focus of the determination is whether the SSC would have performed its safety function if it had been called upon to do so. If an analysis is needed to determine whether a SSC would have performed its safety function, the evaluation may use actual plant data (i.e., actual stroke times, pump flows, actuation setpoints, etc.). The limiting, conservative safety analysis inputs do not necessarily need to be used for this purpose.

Examples and guidance pertaining to "firm evidence" are contained in NUREG-1022 Event Reporting Guidelines. One example is:

During a surveillance test, a component with a 7-day TS LCO allowed outage time was found to be inoperable. Subsequent review indicated that the component was assembled improperly during maintenance conducted 30 days previously and the post-maintenance test was not adequate to identify the error. Thus, there is firm evidence that the standby component was inoperable for the entire 30 days and an LER is required because the condition existed longer than allowed by the TS.

NUCLEAR NON-QUALITY RELATED EN-LI-108 REV 8 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 33 OF 33 Event Notification and Reporting ATTACHMENT 9.2 PAST OPERABILITY EVALUATIONS Page 3 of 3 ADDITIONAL GUIDANCE Additional guidance may be found in the following

References:

1. NRC Inspection Manual Part 9900: Technical Guidance: Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions adverse to Quality or Safety

2. EN-OP-104, Operability Determinations

3. NUREG-1022, Event Reporting Guidelines, 10 CR 50.72 and 50.73.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF 24 10 CFR 21 Evaluations And Reporting Procedure Contains NMM ECH eB REFLIB Forms: YES NO HQN Effective Procedure Owner: Charlene Faison Governance Owner: John McCann Date

Title:

Mgr, Licensing

Title:

VP, Nuclear Safety, Site: Programs Site: EP & Licensing 09/25/2013 Site Site Procedure Champion Title ANO Robert Clark Sr. Licensing Specialist BRP Barbara Dotson Sr. Licensing Specialist CNS David Vanderkamp Manager, Licensing GGNS Cassandre Justiss Technical Specialist IPEC Steve Prussman Sr. Lead Engineer JAF Mark Hawes Licensing Specialist PLP Barbara Dotson Sr. Licensing Specialist PNPS Frank McGinnis Licensing Specialist RBS Danny Williamson Sr. Licensing Specialist VY Philip Couture Sr. Engineer W3 Joe Williams Sr. Licensing Specialist HQN T. R. Jones Sr. Staff Engineer For site implementation dates see ECH eB REFLIB using site tree view (Navigation panel).

Site and NMM Procedures Canceled or Superseded By This Revision None Process Applicability Exclusion: All Sites:

Specific Sites: ANO BRP CNS GGNS IPEC JAF PLP PNPS RBS VY W3 Change Statement

1. Editorial revision to enhance the wording in Attachment 9.1 Part A

2. This procedure revision addresses an issue documented in CR-HQN-2013-00700

3. Updated Site Procedure Champions Associated PRHQN #: 2013-00013

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 2 OF 24 10 CFR 21 Evaluations And Reporting TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ................................................................................................................................. 3

2.0 REFERENCES

.......................................................................................................................... 3 3.0 DEFINITIONS ............................................................................................................................ 5 4.0 RESPONSIBILITIES.................................................................................................................. 7 5.0 DETAILS .................................................................................................................................... 8 5.1 PRELIMINARY 10 CFR 21 SCREENING REVIEW .................................................................. 9 5.2 10 CFR 21 DISCOVERY DETERMINATION........................................................................... 10 5.3 10 CFR 21 EVALUATION AND INTERNAL REPORTING ...................................................... 11 5.4 NRC NOTIFICATION ............................................................................................................... 12 6.0 INTERFACES .......................................................................................................................... 13 7.0 RECORDS ............................................................................................................................... 13 8.0 SITE SPECIFIC COMMITMENTS ........................................................................................... 14 9.0 ATTACHMENTS ...................................................................................................................... 14 ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION ATTACHMENT 9.2 10 CFR 21 WRITTEN REPORT CONTENTS ATTACHMENT 9.3 REGULATORY DISCUSSION AND APPLICATION OF SINGLE FAILURE CRITERION IN 10 CFR 21 EVALUATIONS

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 3 OF 24 10 CFR 21 Evaluations And Reporting 1.0 PURPOSE To identify administrative controls for reviewing, discovering, evaluating, documenting and reporting to the Nuclear Regulatory Commission (NRC) existence of a defect or failure to comply in a facility, activity or basic components of licensed facilities pursuant to 10 CFR 21.

2.0 REFERENCES

[1] Federal Regulatory References (a) 10 CFR 21, Reporting of Defects and Noncompliance (b) 10 CFR 50, Appendix A, General Design Criteria, Definitions and Explanation (c) 10 CFR 50.36, Technical Specifications (d) 10 CFR 50.72, Immediate Notification Requirements For Operating Nuclear Power Plants (e) 10 CFR 50.73, License Event Report System (f) NUREG 302, Remarks Presented (Questions/Answers Discussed) At Public Regional Meetings to Discuss Regulations (10 CFR Part 21) For Reporting of Defects and Noncompliance July, 1977 (ADAMS Accession No. ML062080399)

(g) NRC-IN-2011-19 Licensee Event Reports Containing Information Pertaining to Defects in Basic Components (h) NUREG-1022, "Event Reporting Guidelines: 10 CFR 50.72 and 50.73"

[4] EOI Corporate References (a) EN-LI-102, "Corrective Action Process" (b) EN-LI-108, Event Notification and Reporting (c) EN-OE-100, Operating Experience Program

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 4 OF 24 10 CFR 21 Evaluations And Reporting

[5] ANO References (a) None

[6] FitzPatrick References (a) Technical Specifications (b) AP-03.04, Information Reporting Requirements (c) [[::JAF-LI-102|JAF-LI-102]], JAF Corrective Action Process

[7] GGNS References (a) None

[8] Indian Point Energy Center References (a) None

[9] Palisades References (a) None

[10] Pilgrim References (a) None

[11] RBS References (a) None

[12] Vermont Yankee References (a) AP 0010 "Situational Reporting Requirements"

[13] W3 References (a) UNT-006.010 Event Reporting and Notification

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 5 OF 24 10 CFR 21 Evaluations And Reporting 3.0 DEFINITIONS Note The term basic component is synonymous with safety-related structure, system and component as defined in 10 CFR 50.2.

In all cases, the term basic component includes safety-related design, analysis, inspection, testing, fabrication, replacement parts, or consulting services that are associated with the component. A commercial grade item is not a basic component part or a basic component until after dedication.

[1] Basic Component for Nuclear Power Reactors - A plant structure, system, component, or part thereof that affects its safety function necessary to assure any of the following:

(a) The integrity of the reactor coolant pressure boundary.

(b) The capability to shutdown the reactor and maintain it in a safe shutdown condition.

(c) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in 10 CFR 100.11 or 10 CFR 50.67.

[2] Commercial Grade Item - A plant structure, system, component, or part thereof that affects its safety-related function that was not designed and manufactured as a basic component.

[3] Defect -

(a) A deviation in a basic component delivered to a purchaser for use in a facility or an activity subject to the provisions of 10 CFR 21 if, on the basis of an evaluation, the deviation could create a substantial safety hazard, or (b) The installation, use or operation of a basic component containing a defect, as defined in Items (a), (c) or (d), herein, or (c) A deviation in a portion of a facility subject to the manufacturing requirements of 10 CFR 50 provided the deviation could, on the basis of an evaluation, create a substantial safety hazard and the portion of the facility containing the deviation has been offered to the purchaser for acceptance, or

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 6 OF 24 10 CFR 21 Evaluations And Reporting (d) A condition or circumstance involving a basic component that could contribute to the exceedance of a safety limit (SL), as defined by the plant Technical Specifications (TS).

[4] Dedication - An acceptance process undertaken to provide reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety-related function and, in this respect, is deemed equivalent to an item designed and manufactured under a 10 CFR 50, Appendix B, quality assurance (QA) program.

[5] Deviation - A departure from the technical requirements included in a procurement document.

[6] Discovery - The completion of the documentation first identifying the existence of a deviation or failure to comply potentially associated with a substantial safety hazard.

[7] Evaluation - The process of determining whether a particular deviation could create a substantial safety hazard, or determining whether a failure to comply is associated with a substantial safety hazard.

[8] Failure to Comply - A condition considered to be in non-compliance with the Atomic Energy Act of 1954 as amended, or the Energy Reorganization Act of 1974 as amended, or any applicable NRC rule, regulation, order, or license.

[9] Paperless Condition Reporting System (PCRS) - A computer program used for documenting a Condition Report (CR) and tracking actions resulting from the processes described in EN-LI-102 and addressed in EN-LI-108.

[10] Procurement Document - A contract that defines the requirements which facilities or basic components must meet in order to be considered acceptable by the purchaser.

This includes specifications, purchase orders, engineering service work scopes and other procurement documents that establish the requirements for purchaser acceptance code requirements, drawings, and procedures referenced as part of the procurement document.

[11] Responsible Officer - The individual vested with executive authority over activities subject to 10 CFR 21. The authority to perform reporting of defects and failures to comply may be delegated.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 7 OF 24 10 CFR 21 Evaluations And Reporting Note For a more detailed discussion of consideration of single failure see Attachment 9.3

[12] Single Failure Criterion - As defined in 10 CFR 50 Appendix A, a single failure means an occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither (a) a single failure of any active component (assuming passive components function properly) or (b) a single failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions.

Note A major reduction in the degree of protection is the loss of a safety-related function to the extent that compliance with the 10 CFR 50 Appendix A single failure criterion is not maintained.

[13] Substantial Safety Hazard - A loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety.

[14] Technical Specification (TS) Safety Limit - As defined in 10 CFR 50.36(c)(1) Safety Limits (SLs) for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity.

4.0 RESPONSIBILITIES

[1] Site Vice President (a) Act as Responsible Officer for reporting of 10 CFR 21 defects and failures to comply to the NRC.

(b) Review 10 CFR 21 Evaluations for which the conclusion supports reportability.

[2] Director Nuclear Safety Assurance reviews/approves 10 CFR 21 Evaluations for which the conclusion supports reportability.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 8 OF 24 10 CFR 21 Evaluations And Reporting

[3] Manager, Licensing reviews/approves 10 CFR 21 Evaluations.

[4] Responsible Department Manager assigns personnel to prepare 10 CFR 21 Evaluations, and reviews the 10 CFR 21 Evaluations.

[5] Site Personnel (a) Promptly initiate CRs that identify deficiencies or conditions that may constitute defect or failure to comply per 10 CFR 21 if not already in PCRS.

(b) Prepare 10 CFR 21 Evaluations, as assigned.

(c) Perform independent reviews of completed 10 CFR 21 Evaluations.

Note 10 CFR 21 evaluations are required to address the following conditions:

1. Failure to Comply - A failure to comply is evaluated for reportability in accordance with 10 CFR 21 and this procedure.

2. Received But Not Installed - A supplier identified deviation in a structure, system, component or part thereof that has been offered to and accepted by a nuclear power plant licensee, but which has not been installed for use, is evaluated for reportability under the provisions of 10 CFR 21. If a licensee sells or transfers a basic component to another entity for use in a nuclear power plant, the utility becomes a supplier as defined in 10 CFR 21 and the supplier notification requirements of 10 CFR 21.21(b),

21.51(a)(2), and 21.51(a)(3) apply.

3. Supplier Unable to Make Reportability Determination - If the supplier of a structure, system, component or part thereof identifies a deviation in the structure, system, component or part thereof supplied to a nuclear power reactor and informs the licensee of the nuclear power reactor that the supplier is unable to perform the evaluation to determine reportability under 10 CFR 21, then the licensee evaluates the deviation for reportability.

If the structure, system, component or part thereof is installed in an operating nuclear power reactor, evaluation and reporting is in accordance with 10 CFR 50.72 and 50.73. Security events are handled in accordance with 10 CFR 73.71.

5.0 DETAILS

[1] Degraded or non-conforming conditions reported in PCRS are initially screened to determine if a potential 10 CFR 21 condition exists and if further evaluation is required. Conditions requiring further 10 CFR 21 evaluation are reviewed for deficiencies that potentially could meet the 10 CFR 21 reporting criteria, provided in

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 9 OF 24 10 CFR 21 Evaluations And Reporting Attachment 9.1. The evaluation process first determines if a deficiency/condition qualifies as a deviation or failure to comply, and second, (if applicable) evaluates the discovered deviation or failure to comply to determine if it constitutes a defect or a failure to comply that is related to a substantial safety hazard, and thus is reportable to the NRC.

5.1 PRELIMINARY 10 CFR 21 SCREENING REVIEW

[1] Many instances of defects and failures to comply reportable under 10 CFR 21 are also reportable per 10 CFR 50.72, 10 CFR 50.73, or 10 CFR 73.71.

(a) Reporting defects which manifest themselves as degraded conditions or failures to comply which manifest themselves as non-conforming conditions under 10 CFR 50.72, 10 CFR 50.73, or 10 CFR 73.71 will satisfy the 10 CFR 21 requirements if the information required to be submitted by 10 CFR 21 is included in the submitted report.

(b) Defects in basic components, which are not installed in a plant system (i.e.,

parts on the shelf), are subject only to the reporting requirements and criteria of 10 CFR 21.

(c) Generally, when several reporting regulations apply, the use of 10 CFR 50.72, 10 CFR 50.73, or 10 CFR 73.71 are the preferred methods of reporting.

(d) The most restrictive reporting time requirement must be adhered to.

[2] When reviewing Condition Reports (CRs) for reportability per EN-LI-102 Section 5.4 the following screening criteria should be considered to determine if further 10 CFR 21 evaluation is required:

(a) Does the CR identify a defect with a basic component as defined in Section 3.0?

(b) Could the identified condition potentially result in a substantial safety hazard as defined in Section 3.0?

(c) Will any potential 10 CFR 21 aspects of the identified condition be evaluated for reportability under 10 CFR 50.72, 50.73 or 73.71?

(d) Has the identified 10 CFR 21 condition been evaluated and documented for 10 CFR 21 reportability by the vendor?

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 10 OF 24 10 CFR 21 Evaluations And Reporting (e) Has the identified 10 CFR 21 condition already been reported to the NRC?

If the initial screening review concludes that no 10 CFR 21 reporting is required, then document the basis for not performing a 10 CFR 21 Evaluation in the CR Reportability section.

[3] If the initial screening review is inconclusive, or if further 10 CFR 21 reportability evaluation is required, assign a corrective action (CA) to perform the evaluation, and then proceed to Step 5.2.

Note The Reportability Evaluation Assistance Program (REAP) contains guidance to assist in determining the reportability of plant conditions and events and includes the required reporting timeframes and recipients for each criterion.

5.2 10 CFR 21 DISCOVERY DETERMINATION Note .1 provides a logical series of questions to be answered to determine 10 CFR 21 reportability. The answers may be documented via the Attachment 9.1 or equivalently included in the corrective action response to the associated Condition Report.

PCRS reviews are considered equivalent to the reviews herein.

[1] Answer Parts A.1 and A.2 of Attachment 9.1 to determine if the deficiency (e.g., a reported condition documented in the PCRS) is being or has been addressed elsewhere.

(a) If the deficiency/condition has been previously reported to the NRC (e.g., LER or vendor notice), initiate a CR or CA and indicate the method used for NRC notification.

(b) If the deficiency/condition has been previously or will be reported in a plant LER, review the LER to ensure that all information required for a 10 CFR 21 report was included, and a supplement issued, if necessary. Assure in LER Form 366 the Other box is checked and include 10 CFR 21 in the description field and within the text of the LER.

[2] If needed, assign a corrective action to the cognizant engineering department to answer Parts A.3 and A.4 of Attachment 9.1 to determine if the deficiency/condition is within the scope of 10 CFR 21.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 11 OF 24 10 CFR 21 Evaluations And Reporting

[3] If needed, answer Part A.5 of Attachment 9.1 to determine if the deficiency/condition does or does not constitute a new 10 CFR 21 Discovery (commonly referred to as a potentially reportable condition).

[4] Provide a clear and complete justification of the final Step A.5 answer.

[5] If it is determined that the deficiency/condition does constitute 10 CFR 21 Discovery of a TS SL related issue, deviation or failure to comply, then (a) document the determination as a QA record (as directed in Section 7), and (b) answer Parts B and C of Attachment 9.1. Continue to Section 5.3.

If it is determined that the deficiency/condition does not constitute 10 CFR 21 Discovery of a TS SL related issue, deviation or failure to comply, then sign Part C of Attachment 9.1 and attach a copy of the determination as directed in Section 7.

5.3 10 CFR 21 EVALUATION AND INTERNAL REPORTING The following only applies if the deficiency/condition does constitute discovery of a deviation or failure to comply.

Note For reports due in 60 days, an allowance for reviews, approvals and evaluation report generation of 14 days is suggested.

[1] As soon as practicable, and in all cases within 60 days of determining that the deficiency/condition does constitute discovery of a deviation or failure to comply (i.e.,

the determination of answers to Attachment 9.1 Part A.5), assign a corrective action to the cognizant engineering department to answer Part B of Attachment 9.1 to determine if the deviation or failure to comply is a defect and/or failure to comply relating to a substantial safety hazard, as defined in 10 CFR 21.

Provide a complete justification of the final evaluation determination. The justification should provide a clear statement regarding why there is (or is not) a substantial safety hazard. Assure the potential defect or failure to comply is accurately and thoroughly discussed. See Attachment 9.3.

OR If the evaluation cannot be completed within 60 days of discovery:

(a) Within one (1) working day of making that determination, notify the Licensing Manager that submittal of an interim report to the NRC is necessary.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 12 OF 24 10 CFR 21 Evaluations And Reporting (b) Provide an interim evaluation report within 60 days of discovery.

The interim evaluation report shall describe the deviation or failure to comply that is being evaluated, and shall state (i) the date that the 10 CFR 21 evaluation will be completed internally, and if applicable, (ii) the date that the defect and/or failure to comply (relating to a substantial safety hazard) would be reported to the NRC.

Note If upon conclusion of the 10 CFR 21 Evaluation the condition is determined to NOT be a reportable condition, retract the interim evaluation report from the NRC.

[2] Have the answers to Attachment 9.1 independently reviewed, resolve any reviewer comments and update Attachment 9.1 answers (if needed).

[3] Obtain reporting manager approval of the Attachment 9.1 answers.

[4] Send the completed evaluation to the following for review:

(a) Licensing Manager (b) Nuclear Safety Assurance Director

[5] If the conclusion supports NRC reportability,(i.e., Attachment 9.1 Part B.6. a or b is Yes) then send the completed evaluation to the Site Vice President within 5 working days of completing the evaluation (i.e., obtaining the Licensing Managers signature).

5.4 NRC NOTIFICATION Note Notifications are made by the Site Vice President or his designee.

[1] Notification to the NRC of a defect or failure to comply is not required if the Site Vice President has actual knowledge that the NRC has been notified in writing of the defect or the failure to comply. If applicable, document this written notification in the corrective action system.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 13 OF 24 10 CFR 21 Evaluations And Reporting

[2] If the conclusion is that a defect and/or failure to comply relating to a substantial safety hazard exists, notify the NRC within 2 days of informing the Site Vice President.

(a) Make initial notification to the NRC Operations Center:

• facsimile - 301-816-5151 (preferred method)

• telephone - 301-816-5100 (b) Verify the facsimile has been received by calling the NRC Operations Center.

[3] Provide a written report to the NRC Document Control Desk within 30 days of the initial notification in accordance with EN-LI-106.

(a) Include information detailed in Attachment 9.2 in the report.

(b) Provide a copy to the following:

• Regional Administrator

• NRC Resident Inspector 6.0 INTERFACES

[1] EN-LI-102 Corrective Action Process is used to report conditions.

[2] EN-LI-108 Event Notification and Reporting is used to screen documented conditions for potential reportability.

[3] EN-LI-106 NRC Correspondence is used for outgoing correspondence preparation.

[4] EN-AD-103 Document Control and Records Management Programs is used to retain records not maintained in PCRS.

7.0 RECORDS

[1] Documents prepared in accordance with this procedure are to be stored, annotated, and retained as specified in EN-AD-103 and/or applicable procedures. If a 10 CFR 21 Discovery Determination and/or Reportability Evaluation originated from a CR in PCRS, then a copy of the completed Attachment 9.1 form should be attached within the associated CR topic location that initiated the 10 CFR 21 process. Each record shall be retained for a minimum of five years.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 14 OF 24 10 CFR 21 Evaluations And Reporting 8.0 SITE SPECIFIC COMMITMENTS Step Site Document Commitment Number or Reference

[1] None None identified 9.0 ATTACHMENTS .1 - 10 CFR 21 Discovery Determination & Reportability Evaluation .2 - 10 CFR 21 Written Reports .3 - Regulatory Discussion and Application of Single Failure Criterion in 10 CFR 21 Evaluations

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 15 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION SHEET Sheet 1 of 5 Consider the answers to the following questions until the determination/evaluation is complete.

Document the answers on this form or in PCRS.

Part A - 10 CFR 21 Discovery Determination

1. Review the deficiency or condition (usually documented in a Condition Report [CR] using the Paperless Condition Report System [PCRS]).

2. Has/Is the deficiency/condition

a. been reported as a 10 CFR 21 condition to the NRC by another source (e.g., a vendor or other plant)?

Yes ___ No/Unknown ___

b. been reported under another process (10 CFR 50.72, 50.73 (the LER process), or 73.71) that will include information required to be submitted by 10 CFR 21?

Yes ___ No ___

c. being evaluated for reportability under another process (10 CFR 50.72, 50.73 (the LER process), or 73.71) that will include information required to be submitted by 10 CFR 21?

Yes ___ No ___

If 2.a is Yes, complete the instruction in paragraph 5.2[1](a).

If 2.b or 2.c is Yes, complete the instruction in paragraph 5.2[1](b).

If 2.a or 2.b or 2.c is Yes, then the determination stops; otherwise the determination continues.

3. Does the deficiency/condition involve

a. a basic component (defined in Section 3)

Yes ___ No ___

b. a nonsafety-related SSC or function associated with compliance to a Technical Specification (TS) safety limit (SL)

Yes ___ No ___

c. compliance with the Atomic Energy Act of 1954 as amended, or the Energy Reorganization Act of 1974 as amended, or any applicable NRC rule, regulation, order, or license?

Yes ___ No ___

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 16 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION SHEET Sheet 2 of 5 If 3.a, 3.b and 3.c are all No, then 10 CFR 21 reporting is not required and the determination stops.

If 3.a is Yes, then complete Step 4, regardless of answers to 3.b and 3.c.

If 3.a is Yes, and 3.b or 3.c is Yes, then complete Steps 4 and 5.

If only 3.b or 3.c is Yes; then skip to Step 4 and complete Step 5.

4. Has the basic component or commercial grade (non safety-related) SSC been

a. accepted for use as safety-related Yes ___ No ___

b. installed in the plant?

Yes ___ No ___

If 4.a is No, then 10 CFR 21 reporting is not required and the determination stops.

If 4.b is Yes, then assure that the issue has been or is being evaluated under 10 CFR 50.72 and/or 50.73, and includes the requirement of 10 CFR 21. Issue a new CR if needed.

If 4.a is Yes and 4.b is No, OR 3.b or 3.c is Yes, then complete Step 5.

If 4.a, 4.b, 3.b and 3.c are all No, then 10 CFR 21reporting is not required and the determination stops.

5. Does the deficiency/condition involve a

a. non-safety-related SSC with a functional failure, deviation or other condition that could challenge a TS SL; or

b. deviation (i.e., a departure from a technical requirement, see Section 3) associated with the basic component or nonsafety-related SSC; or

c. failure to comply with the Atomic Energy Act of 1954 as amended, or the Energy Reorganization Act of 1974 as amended, or any applicable NRC rule, regulation, order, or license?

Yes ___ No ___

Justification of the final Step A.5 answer:

(Continue on Sheet 5, if needed.)

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 17 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION SHEET Sheet 3 of 5 If all 5.a, 5.b and 5.c are No, then, 10 CFR 21 reporting is not required and the determination is finished (i.e., skip Part B), complete Part C, and document this determination as a QA record (see Section 7.0 Records).

If either 5.a, 5.b or 5.c is Yes, a 10 CFR 21 Discovery exists, a formal 10 CFR 21 Reportability Evaluation is required (i.e. go to Part B), and document this determination as a QA record (see Section 7.0 Records).

Part B - 10 CFR 21 Reportability Evaluation

6. Could the discovered TS SL related issue, deviation or failure to comply

a. create a Substantial Safety Hazard (i.e., the loss of a safety-related function to the extent that compliance with the 10 CFR 50 Appendix A single failure criterion is not maintained), or

b. contribute to the exceedance of a TS SL (see Section 3)?

Yes ___ No ___

Justification of the final Step B.6 answer:

(Continue on Sheet 5, if needed.)

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION SHEET Sheet 4 of 5 If either 6.a or 6.b is Yes, then a defect and/or failure to comply relating to a substantial safety hazard exists, the condition must be reported to the NRC per 10 CFR 21, as specified in Section 5.4.

If both 6.a and 6.b are No, then a defect and/or failure to comply relating to a substantial safety hazard does not exist, 10 CFR 21 is not applicable, and the condition is not required to be reported to the NRC under 10 CFR 21.

7. Have the completed answers independently reviewed, resolve review comments, obtain approvals and document this evaluation as a QA record, as specified in Section 7.0 Records.

Part C - Approvals Prepared by: ______________________________ Date: __________________

Reviewed by: ______________________________ Date: __________________

Approved by: ______________________________ Date: __________________

Approved by: ______________________________ Date: __________________

Licensing Manager Approved by: ______________________________ Date: __________________

Nuclear Safety Assurance Director Site VP Notification Date and Time _____________________________________________

NRC Notification Date and Time _______________________________________________

Person making NRC notification _______________________________________________

Person receiving NRC notification ______________________________________________

NRC Notification Number _____________________________________________________

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 19 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.1 10 CFR 21 DISCOVERY DETERMINATION & REPORTABILITY EVALUATION SHEET Sheet 5 of 5 10 CFR 21 Determination/Evaluation Justification Continuation:

Continuation page __ of __

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 20 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.2 10 CFR 21 WRITTEN REPORT CONTENTS Sheet 1 of 1 To the extent that the NRC has not been adequately informed, this provides the minimum content for written reports made to the NRC:

I. Name and Address Name and address of the individual (or individuals) informing the NRC.

II. Facility, Activity or Component Identification of the facility, the activity, or the basic component supplied for such facility or such activity which fails to comply or contains a defect.

Ill. Constructor or Supplier Identification of the firm constructing the facility or supplying the basic component which fails to comply or contains a defect.

IV. Defect and Safety Hazard Nature of the defect or failure to comply and the safety hazard which is created or could be created.

V. Date The date on which the information of such defect or failure to comply was obtained.

VI. Location and Number of Defective Components In the case of a basic component which contains a defect or fails to comply, the number and location of all such components in use at, supplied for, or being supplied for one or more facilities or activities.

VII. Corrective Action The corrective action which has been, is being or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

VIII. Advice Any advice related to the defect or failure to comply about the facility, activity or basic component that has been, is being or will be given to purchasers or licensees.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 21 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.3 REGULATORY DISCUSSION AND APPLICATION OF SINGLE FAILURE CRITERION IN 10 CFR 21 EVALUATIONS Sheet 1 of 4 Regulatory Discussion Changes to 10 CFR 21 that became effective on October 29, 1991 transferred many of the requirements for reporting defects in basic components at operating nuclear power plants from 10 CFR 21 to 10 CFR 50.72, 10 CFR 50.73, and 10 CFR 73.71. The transfer of these reporting requirements is contained in 10 CFR 21.2(c) and states:

"For persons licensed to operate a nuclear power plant under part 50 or part 52 of this chapter, evaluation of potential defects and appropriate reporting of defects under Paragraphs 50.72, 50.73 or Paragraph 73.71 of this chapter satisfies each person's evaluation, notification, and reporting obligation to report defects under this part and the responsibility of individual directors and responsible officers of these licensees to report defects under section 206 of the Energy Reorganization Act of 1974.

Hence, there are only two conditions under which nuclear power plants perform reportability evaluations per 10 CFR 21. There is also a third condition where nuclear power plant licensees may choose to perform an evaluation under 10 CFR 21. The process described in this procedure applies to these three conditions as follows:

Failures to Comply - The rule changes that became effective on October 29, 1991 did not affect the reporting requirements previously contained in 10 CFR 21 for failures to comply with the Atomic Energy Act of 1954, as amended, or any applicable rule, regulation, order, or license of the Commission associated with the facility, activity, or basic components purchased by nuclear power plants that relate to substantial safety hazards. Therefore, failures to comply are still evaluated for reportability in accordance with 10 CFR 21 and this procedure.

Received But Not Installed - If a supplier identifies a deviation in a structure, system, component or part thereof that has been offered to and accepted by a nuclear power plant licensee, but which has not been installed for use, the licensee evaluates the deviation for reportability under the provisions of 10 CFR 21. If the structure, system, component or part thereof is installed in an operating nuclear power reactor, reporting is in accordance with 10 CFR 50.72 and 50.73. Security events are handled in accordance with 10 CFR 73.71.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 22 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.3 REGULATORY DISCUSSION AND APPLICATION OF SINGLE FAILURE CRITERION IN 10 CFR 21 EVALUATIONS Sheet 2 of 4 Supplier Unable to Make Reportability Determination - If the supplier of a structure, system, component or part thereof identifies a deviation in the structure, system, component or part thereof supplied to a nuclear power reactor and informs the licensee of the nuclear power reactor that the supplier is unable to perform the evaluation to determine reportability under 10 CFR 21, then the licensee evaluates the deviation for reportability. The evaluation may be performed either under the provisions of 10 CFR 21 or under the applicable requirements of 10 CFR 50.72, 50.73, (or 10 CFR 73.71 for safeguards events), at the discretion of the licensee.

In addition, it has been determined that a utility that sells a basic component to another utility for use in a nuclear power plant is a supplier as defined in 10 CFR 21. Therefore the supplier notification requirements of 10 CFR 21.21(b), 21.51(a)(2), and 21.51(a)(3) apply.

Application of Single Failure Criterion in Part 21 Evaluations NUREG 0302, Revision 1, Response to Question 2, page 21.3(k)-1 can be cited as a justification why a condition may be reportable under 10 CFR 21 depending on how the single failure criterion is applied. The example does apply in the case where a deviation causes or could cause the failure of a redundant basic component, which is not always the case; as shown herein, the example was not intended to mean an automatic single failure must be assumed in all cases.

In its entirety, the paragraph cited above states the following:

Question 2: Are defects in redundant components reportable under Part 21?

Response 2:

A deviation (i.e., a departure from a procurement document specification) which, based on an evaluation, causes or could cause the failure of a redundant basic component is a reportable defect under Part 21. The loss of safety function of a basic component is considered a major reduction in the degree of protection provided to the public health and safety. It is possible that the defect might also exist in the redundant basic component which could result in a loss of safety function. The existence of a defective basic component, considering a single failure of its counterpart redundant basic component, could result in loss of safety function. Actually, the counterpart component need not fail. It could be removed from service for other reasons such as routine maintenance or inspection.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 23 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.3 REGULATORY DISCUSSION AND APPLICATION OF SINGLE FAILURE CRITERION IN 10 CFR 21 EVALUATIONS Sheet 3 of 4 For the reasons discussed below, interpreting this example to indicate that if one basic component in a redundant system configuration has degradation, a single failure must be assumed in the redundant component in evaluating whether there is a loss of safety function is logically inconsistent.

The cited example is in the context of a deviation which causes or could cause the failure of a redundant basic component. Entergy believes the example is simply saying that if an accident analysis of record credits a redundant component in the accident analysis given a single failure, then if the initial 1 of 2 components has a defect, that defect cannot count as the single failure.

To extend the meaning of this example to apply a single failure in all cases would be contrary to the principles in the regulation and the regulatory record. First note the regulatory structure for evaluating and reporting under 10CFR 50.72 and 10CFR 50.73. These regulations establish a single failure on the redundant component is not automatically applied.

Single Failure Requirements of 10CFR 50.72 and 10CFR 50.73 The determination of whether an event or condition is reportable is predicated on whether the event or condition could have prevented fulfillment of a safety function. This predication is substantiated by 10 CFR 50.72 and 10 CFR 50.73 and NUREG 1022, Revision 2.

10 CFR 50.73(a)(2)(vi) states the following:

Events covered in paragraph (a)(2)(v) of this section may include one or more procedural personnel errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (a)(2)(v) of this section if redundant equipment in the same system was operable and available to perform the required safety function.

NUREG 1022, Section 3.2.7, Revision 2, states the following:

In determining the reportability of an event or condition that affects a system, it is not necessary to assume an additional random single failure in that system; however, it is necessary to consider other existing plant conditions. (See example [4] below).

The licensee may also use engineering judgment to decide when personnel actions could have prevented fulfillment of a safety function. For example, when an individual improperly operates or maintains a component, he might conceivably have made the same error for all of the functionally redundant components (e.g., if he incorrectly calibrates one bistable amplifier in the Reactor Protection System, he could conceivably incorrectly calibrate all bistable amplifiers). However, for an event to be reportable it is necessary that the actions actually affect or involve components in more than one train or channel of a system, and the result of the actions must be undesirable from the perspective of protecting public health and safety of the public.

NUCLEAR NON-QUALITY RELATED EN-LI-108(-01) REV. 3 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 24 OF 24 10 CFR 21 Evaluations And Reporting ATTACHMENT 9.3 REGULATORY DISCUSSION AND APPLICATION OF SINGLE FAILURE CRITERION IN 10 CFR 21 EVALUATIONS Sheet 4 of 4 The following types of events or conditions generally are not reportable under these criteria:

• Independent failure of a single component (unless it is indicative of a generic problem, which alone could have caused failure of a redundant safety system As indicated in Paragraph 50.73(a)(2)(vi) individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.

A design or analysis defect or deviation is reportable under this criterion if it could have prevented fulfillment of the safety function of structures or systems defined in the rules. Reportability of a design or analysis defect or deviation under this criterion should be judged on the same basis that is used for other conditions, such as operator errors and equipment failures. That is, the condition is reportable if there is reasonable expectation of preventing fulfillment of the safety function. Alternatively stated, the condition is reportable if there was reasonable doubt that the safety function would have been fulfilled if the structure or system had been called upon to perform it.

The regulatory record clearly establishes a condition is not reportable if the safety function could be fulfilled, redundant equipment was available, and that an additional single failure need not be postulated (there was no common cause).

Single Failure Requirements of 10CFR 21 Assuming that 10 CFR 21 requires an automatic application of a single failure to the redundant component or system would set a threshold for reporting under 10 CFR 21 lower than 10 CFR 50.72 and 10CFR 50.73. Such an interpretation would be contrary to the principles in 10 CFR 21 which establish that reporting under Part 21 applies when there is a loss of safety function and when there is a major reduction in the degree of protection provided to the public health and safety. Also, such an interpretation, would eliminate the parity between 10 CFR 21 and 10 CFR 50.72 / 10 CFR 50.73 and set a new threshold for reporting under 10 CFR 21.

RJPM-NRC-M14-S1 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Initiate Reactor Core Isolation Cooling for Level Control OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: September 4, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 10

RJPM-NRC-M14-S1 Rev. 0 EXAMINER INFO SHEET Task Standard: Reactor Core Isolation Cooling is running, after an inadvertent trip, lined up for RPV Level Control.

Synopsis: The reactor is shutdown, pressurized and MSIVs are closed following an automatic isolation. This task will align RCIC for level control of the RPV using hard card (OSP-53, Attach. 6). As an alternate path, there is an inadvertent trip of RCIC which the operator will reset per SOP-0035, Section 5.2.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to initiate Reactor Core Isolation Cooling (RCIC); and to restore and maintain RPV level from -10 to 51 inches.

3) Initial Conditions:

The reactor is shutdown and pressurized; MSIVs are closed.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-S1 Rev 0 Page 2 of 10

RJPM-NRC-M14-S1 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Initiate RCIC for Level Control 217007001001 A4.04 3.6 / 3.6 with Inadvertent Trip 217010001001 A4.05 4.1 /

4.1 REFERENCES

APPLICABLE OBJECTIVES OSP-0053, Rev 17, Attach 6 RLP-STM-0209, Obj 8 SOP-0035, Rev 45 REQUIRED MATERIALS: SAFETY FUNCTION:

OSP-0053, Rev 17 Attach 6 (Sim Copy-Hard Card) __4__

SOP-0035, Rev 45 (Sim. Copy)

SIMULATOR CONDITIONS & SETUP:

1. IC # 212

2. Required Power: Shutdown, pressurized, MSIVs closed.

3. HPCS racked out/ tagged out.

4. Feed Pumps tripped.

5. RHR-A in Suppression Pool Cooling.

6. Event T1: Malfunction: RCIC Turbine trip to activate when turbine speed reaches 90%

delete in 10 seconds; this will simulate a spurious trip and allow resetting the trip.

7.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Reactor Core Isolation Cooling is running, being used for RPV Level Control RJPM-NRC-M14-S1 Rev 0 Page 3 of 10

RJPM-NRC-M14-S1 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 1. Arm and depress, RCIC MANUAL INITIATION Pushbutton.

Standard Applicant located/identified and depressed the RCIC Manual Initiation Pushbutton.

Cue Notes The applicant will use the Hard Card for the initiation.

Results SAT UNSAT

2. Procedure Step: 2. Verify the following:

• E51-F045, RCIC STEAM SUPPLY TURBINE STOP VALVE Opens.

Standard Applicant located/identified and verified that E51-F045 Opened.

Cue Notes The RCIC turbine will trip at 90% speed.

Results SAT UNSAT

3. Procedure Step: 2. Verify the following:

• RCIC STEAM SUPPLY and EXHAUST DRAIN POT ISOLATION VALVES Close.

Standard Applicant located/identified and verified that the RCIC Steam Supply and Exhaust Drain Pot Isolation Valves closed.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S1 Rev 0 Page 4 of 10

RJPM-NRC-M14-S1 Rev. 0

4. Procedure Step: 2. Verify the following:

• E51-C002C, GLAND SEAL COMPRESSOR Starts.

Standard Applicant located/identified and verified that the RCIC Gland Seal Compressor started.

Cue Notes Results SAT UNSAT

5. Procedure Step: 2. Verify the following:

• E51-F013, RCIC INJECT ISOL VALVE Opens.

Standard Applicant located/identified and verified that E51-F013, RCIC Injection Isolation Valve Opened.

Cue Notes The injection valve will go closed again as the RCIC turbine trips.

Results SAT UNSAT

6. Procedure Step: 3. Verify RCIC Turbine comes up to speed and stabilizes at 2300-4600 rpm.

Standard Applicant recognized the RCIC turbine tripped by turbine full speed not being achieved and a slowing of turbine speed as well as a status light.

Cue As CRS accept report that RCIC has tripped.

Direct the applicant to, Restore/Inject with RCIC using the SOP.

Notes Applicant should transition from Hard Card to SOP-0035, Reactor Core Isolation Cooling, Section 5.2. (Recovery from RCIC Turbine Trip During Auto Initiation).

Results SAT UNSAT RJPM-NRC-M14-S1 Rev 0 Page 5 of 10

RJPM-NRC-M14-S1 Rev. 0 ALTERNATE PATH:

SOP-0035 Section 5.2, Recovery from RCIC Turbine Trip During AUTO Initiation PROCEDURE NOTE This Section is intended to allow restart of RCIC after turbine trip for other reasons than system isolation.

7. 5.2.1 Close E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR.

• Procedure Step:

Standard Applicant located and closed the Trip/Throttle Valve.

Cue Notes Results SAT UNSAT Procedure Step: 5.2.2 IF a RCIC Turbine Overspeed or local manual trip has occurred, THEN reset the turbine throttle valve locally.

Standard NA Cue Notes No applicant action is necessary - the turbine does not require being reset locally.

8. 5.2.3 Throttle E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR

• Procedure Step:

open to obtain 2500 to 3500 rpm on E51-C002-1, RCIC TURBINE SPEED.

Standard Applicant throttled the RCIC Trip & Throttle Valve and obtained 2500-3500 rpm on the RCIC turbine speed meter.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S1 Rev 0 Page 6 of 10

RJPM-NRC-M14-S1 Rev. 0

9. *Procedure Step: 5.2.4 Open E51-F013, RCIC INJECT ISOL VALVE.

Standard Applicant opened the RCIC injection isolation valve.

Cue Notes Results SAT UNSAT PROCEDURE NOTE Open E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR in very small increments and allow time for the turbine to respond between adjustments to prevent an overspeed.

10. Procedure Step: 5.2.5 Slowly throttle open E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR until HVY-C002, RCIC TURBINE GOV VLV indicates that the governor has control.

Standard Applicant slowly throttled open the RCIC Trip & Throttle Valve and observed the HVY-C002, RCIC Pump/Flow Controller until the governor took control.

Cue Notes There will be deviation/response of the meter to flow exceeding 600 gpm.

Results SAT UNSAT PROCEDURE NOTE Governor control is verified by lowering the turbine rpm slightly using RCIC PUMP/FLOW CONTROLLER, HVY-C002. If the rpm drops and is held steady, the governor is in full control.

11. Procedure Step: 5.2.6 WHEN the governor has control, THEN open E51-C002, RCIC TRIP &

THROTTLE VALVE OPERATOR.

Standard Applicant verified governor control per procedure note above then opened the Trip & Throttle valve completely.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S1 Rev 0 Page 7 of 10

RJPM-NRC-M14-S1 Rev. 0

12. Procedure Step: 5.2.7 IF RCIC operation is desired, THEN refer to one of the following sections:

• Refer to section 5.7 for level control.

• Refer to section 5.8 for pressure control.

Standard Applicant chose and referred to Section 5.7 of SOP-0035.

Cue Notes Results SAT UNSAT SOP-0035 Section 5.7, RPV Level Control Procedure Step: 5.7.1 IF RCIC operating AND aligned to the CST, THEN shift to inject to the RPV as follows:

Standard NA Cue Notes No applicant action is necessary

13. Procedure Step: 5.7.2 With E51-R600, RCIC PUMP FLOW CONTROLLER HVYC002, in AUTO OR MANUAL, adjust RCIC flow setpoint as required to maintain RPV water level.

Standard Applicant verifies that the RCIC pump flow controller is set for 600 gallons per minute.

Cue Notes Results SAT UNSAT Terminating Cue: Reactor Core Isolation Cooling is running, being used for RPV Level Control This completes this JPM.

STOP TIME:

RJPM-NRC-M14-S1 Rev 0 Page 8 of 10

RJPM-NRC-M14-S1 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S1 Rev 0 Page 9 of 10

RJPM-NRC-M14-S1 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The reactor is shutdown and pressurized; MSIVs are closed.

INITIATING CUE:

The CRS has directed you to initiate Reactor Core Isolation Cooling (RCIC); and to restore and maintain RPV level from -10 to 51 inches.

RJPM-NRC-M14-S1 Rev 0 Page 10 of 10

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• REACTOR CORE ISOLATION COOLING SYSTEM (SYS #209)

PROCEDURE NUMBER: *SOP-0035 REVISION NUMBER: *045 Effective Date:

• 06/13/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0035R044EC-A Correct typographical error made during incorporation of SOP-0035R042CM-1. The Gland Seal Compressor must be started prior to opening the E51-F045. Moved step for starting Gland Seal Compressor to Step 4.2.5.

SOP-0035 REV - 045 PAGE 1 OF 77

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES FOR STARTUP AND OPERATION .........................................................6 4 SYSTEM STARTUP..................................................................................................................7 4.1 Placing the RCIC System in Standby ...............................................................................7 4.2 Manual RCIC Startup........................................................................................................13 4.3 RCIC Slow Roll Startup....................................................................................................16 4.4 Manual Initiation...............................................................................................................21 4.5 Placing the RCIC/RHR Backfill into Service ...................................................................21 5 SYSTEM OPERATION .............................................................................................................24 5.1 Operation from Automatic/Manual Initiation ...................................................................24 5.2 Recovery from RCIC Turbine Trip During AUTO Initiation...........................................26 5.3 Resetting a RCIC overspeed trip ......................................................................................27 5.4 Alternate Suppression Pool Makeup.................................................................................28 5.5 Operation of the RCIC Gland Seal Compressor ...............................................................29 5.6 Manual swap of Suppression Pool and CST suction valves .............................................29 5.7 RPV Level Control ...........................................................................................................30 5.8 RPV Pressure Control .......................................................................................................30 5.9 High Velocity Flush of RCIC/RHR Backfill during Plant Shutdown ..............................32 6 SYSTEM SHUTDOWN.............................................................................................................34 6.1 RCIC System Shutdown ...................................................................................................34 7 REFERENCES ...........................................................................................................................36 8 RECORDS..................................................................................................................................37 ATTACHMENT 1 - VALVE LINEUP - REACTOR CORE ISOLATION COOLING (SAFETY RELATED)......................................................................................................38 ATTACHMENT 2 - INSTRUMENT AND VALVE LINEUP - REACTOR CORE ISOLATION COOLING SYSTEM (SAFETY RELATED)............................................45 ATTACHMENT 3 - ELECTRICAL LINEUP - REACTOR CORE ISOLATION COOLING (SAFETY RELATED)...................................................................................58 SOP-0035 REV - 045 PAGE 2 OF 77

CONTINUOUS USE ATTACHMENT 4 - CONTROL BOARD LINEUP - REACTOR CORE ISOLATION COOLING (SAFETY RELATED)...................................................................................60 ATTACHMENT 5 - RCIC TRIP THROTTLE VALVE DIAGRAM ..............................................63 ATTACHMENT 6 - RCIC OPERATION WITH A LOSS OF AC AND DC POWER ...................65 ATTACHMENT 7 - RCIC ALTERNATE FLOW INDICATION AND RPV WATER LEVEL INDICATION......................................................................................................70 SOP-0035 REV - 045 PAGE 3 OF 77

CONTINUOUS USE 1 PURPOSE 1.1 To provide instructions for the operation of the Reactor Core Isolation Cooling (RCIC)

System.

2 PRECAUTIONS AND LIMITATIONS 2.1 Starting the RCIC Pump with an injection line low pressure condition may cause damage due to water hammer. If the RCIC System Line Fill Pump is to be shutdown for an extended period, the RCIC Pump should not be allowed to start.

C 2.2 When performing tests or system lineups, the potential exists for water hammer to occur if the system has not been properly filled and vented.

2.3 If narrow range RPV water level reaches Level 8 during RCIC operation, E51-F045, RCIC STEAM SUPPLY TURBINE STOP VALVE and E51-F013, RCIC INJECT ISOL VALVE will automatically close which shuts down the RCIC Turbine. The RCIC Turbine will restart automatically at RPV Level 2.

C 2.4 The Min Flow valve flow transmitter may go into saturation during pump operation.

This condition may delay min. flow valve opening by as long as 21 seconds. The pump vendor has specified that the RCIC pump can be run for 30 seconds in a dead headed condition before experiencing pump degradation.

2.5 Operation of the RCIC Turbine below 2300 rpm may result in turbine exhaust check valve damage due to chattering.

2.6 Operation of the RCIC Turbine below 1700 rpm may result in insufficient pump flow for cooling of pump internals.

2.7 In the event of a RCIC Turbine overspeed or local manual trip, the Turbine Trip and Throttle Valve must be reset locally. All other trips can be reset remotely from Control Room.

C 2.8 If an overspeed event occurs during the operation of the RCIC Turbine, aligned CST to CST, with a pressure greater than 2000 psig, then the CST piping may have been over-pressurized. The following shall be checked prior to declaring the RCIC system operable:

• IF E51-PIR001 was valved in during the overspeed event, THEN calibrate or replace it.

• Recalibrate transmitter loops associated with flow control transmitters E51-FTN051 and E51-FTN003.

• Perform inspection of piping and equipment as directed by System Engineering.

• Calibration check of E51-PCVF015, RCIC Turb Lube Oil Clr Inlet Press Cntrl Vlv SOP-0035 REV - 045 PAGE 4 OF 77

CONTINUOUS USE 2.9 Cooling for the Gland Seal Compressor is provided by the airflow through the compressor. Do not allow the compressor to operate at discharge air temperature of 370°F or greater as indicated by Annunciator P601-21A-H01, RCIC GS COMP TRIP AIR TEMP HIGH-HIGH.

2.10 The oil temperature leaving the Turbine Lube Oil Cooler should be maintained at 40°F to 160°F. The RCIC System should be shutdown if oil temperature or pump bearing temperatures reach 180°F.

2.11 While warming up the RCIC system, steam flashing in the RCIC line dP transmitters will cause an invalid actuation of the isolation logic for E51-MOVF063, F064, and F076.

This ESF action is not reportable when it occurs in this manner.

C 2.12 When the RCIC System is undergoing maintenance, testing, or any condition which could bottle up the steam supply line (i.e., close the steam line drain valves), then E51-F063, RCIC STEAM SUPPLY INBD ISOL VALVE and E51-F064, RCIC STEAM SUPPLY OUTBD ISOL VALVE should be closed to prevent an unexpected ESF actuation.

C 2.13 If maintenance has been performed on the RCIC Turbine, the slow roll, manual startup Section 4.3 of this procedure should be performed to ensure operability of RCIC.

2.13.1. Following any maintenance activity that opens the RCIC turbine lube oil system, including oil changes, oil samples, and if needed, troubleshooting, three 5-minute slow rolls and a final 30 minute slow roll, each followed by a 15-30 minute shutdown period and oil level check, should be completed to ensure all air has been vented from the oil system. This does not include adding small amounts of oil to maintain oil level in the sightglass.

C 2.14 Anytime RCIC is to be run, Suppression Pool Cooling should be placed in service per SOP-0031, Residual Heat Removal to ensure adequate mixing of the Suppression Pool for temperature monitoring.

C 2.15 There is a potential for early realignment of the HPCS and RCIC suctions from the CST to the Suppression Pool. When the suction swap occurs, flow in the CST suction line drops to zero. The trip signal may clear at this time. If needed the RCIC system may be realigned to the CST if local CST level instruments indicate 3.5 or greater and HPCS suction remains on the Suppression Pool.

C 2.16 The outlet of E51-RVF090 is not routed to a floor drain. If this relief were to lift then personnel in the RCIC area may become contaminated.

2.17 If RCIC is running in support of testing the Technical Specifications require average suppression pool temperature monitoring per STP-057-0700, Suppression Pool Average Water Temperature Verification During Testing That Adds Heat to the Suppression Pool.

SOP-0035 REV - 045 PAGE 5 OF 77

CONTINUOUS USE 2.18 High velocity flushes of the RCIC/RHR backfill system shall not be performed with the reactor temperature greater than or equal to 200ºF until additional detailed analysis on the RCIC/RHR steam supply line flow elbow has been completed by Engineering.

2.19 RCIC/RHR Backfill should not be left in service longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the RCIC Isolation valves are closed. Engineering should be notified before exceeding the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.20 The RCIC pump should not be operated below 50% of rated flow for sustained periods of time. Operation at a 10 to 20% minimum flow condition is intended for startup and shutdown transients only. This is not an intended normal operating condition for the pump because severe internal cavitation at the high-head condition can result in pump damage.

3 PREREQUISITES FOR STARTUP AND OPERATION 3.1 Check Suppression Pool level is operable per Technical Specification 3.6.2.2.

3.2 Check the Instrument Air System is in operation per SOP-0022, Instrument Air System.

3.3 Check the below electric systems are in service and aligned to RCIC:

3.3.1. 125VDC per SOP-0049, 125 VDC System 3.3.2. 120VAC per SOP-0048, 120 VAC System 3.3.3. 480VAC per SOP-0047, 480 VAC System 3.4 Check CST level is greater than 11 ft 1 in.

3.5 Check the Remote Shutdown System is in standby per SOP-0027, Remote Shutdown System.

3.6 Check the Reactor Feed Water Line A downstream of FWS-MOV7A to the Reactor Vessel is lined up per SOP-0009, Reactor Feedwater System.

3.7 Verify the system is lined up for startup.

SOP-0035 REV - 045 PAGE 6 OF 77

CONTINUOUS USE 5.2 Recovery from RCIC Turbine Trip During AUTO Initiation NOTE This section is intended to allow restart of RCIC after turbine trip for other reasons than system isolation.

5.2.1. Close E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR.

5.2.2. IF a RCIC Turbine overspeed or local manual trip has occurred, THEN reset the turbine throttle valve locally.

5.2.3. Throttle E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR open to obtain 2500 to 3500 rpm on E51-C002-1, RCIC TURBINE SPEED.

5.2.4. Open E51-F013, RCIC INJECT ISOL VALVE.

NOTE Open E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR in very small increments and allow time for the turbine to respond between adjustments to prevent an overspeed.

5.2.5. Slowly throttle open E51-C002, RCIC TRIP & THROTTLE VALVE OPERATOR until HVY-C002, RCIC TURBINE GOV VLV indicates that the governor has control.

NOTE Governor control is verified by lowering the turbine rpm slightly using RCIC PUMP FLOW FLOW CONTROLLER HYVC002. If the rpm drops and is held steady, the governor is in full control.

CRITICAL STEP 5.2.6. WHEN the governor has control, THEN open E51-C002, RCIC TRIP &

THROTTLE VALVE OPERATOR.

5.2.7. IF RCIC operation is desired, THEN Refer to one of the following Sections.

• Refer to Section 5.7 for level control.

• Refer to Section 5.8 for pressure control.

SOP-0035 REV - 045 PAGE 26 OF 77

CONTINUOUS USE 5.2.8. WHEN RCIC is no longer required, THEN shut down RCIC per Section 6.

5.3 Resetting a RCIC overspeed trip NOTE After a local manual trip or mechanical overspeed, the Turbine Trip Throttle Valve must be manually reset at the turbine.

Refer to Attachment 5, RCIC Trip Throttle valve diagram 5.3.1. Locally verify E51-MOVC002, RCIC TRIP & THROTTLE VALVE is closed using the handwheel.

NOTE The movement of the emergency trip rod must be in a path parallel with its normal motion, there should be no twisting action associated with the movement of the emergency trip rod.

5.3.2. At the turbine, PUSH and HOLD the emergency trip rod, against spring pressure towards the RCIC trip & throttle valve.

5.3.3. IF required, THEN lift up on the manual hand trip lever.

5.3.4. WHEN 5 seconds have elapsed, THEN SLOWLY RELEASE the emergency trip rod.

NOTE Slight manual manipulation of the tappet nut assembly may be necessary if trouble is incurred while attempting to manually reset the RCIC turbine.

5.3.5. Verify the square edge (flat) of the emergency head lever mates with the square edge (flat) of the tappet nut and that the head lever is engaged with the tappet nut.

5.3.6. Verify the trip lever and tappet nut freely resets and mates with the underside of the tappet nut (metal to metal contact). Push down on the top of the tappet nut to assure it is fully seated against the head bracket.

NOTE If any binding of the mechanism occurs while resetting the trip mechanism, then notify the Shift Manager and the System Engineer.

5.3.7. Verify the trip hook lever is engaged on the latch lever of the trip & throttle valve.

SOP-0035 REV - 045 PAGE 27 OF 77

CONTINUOUS USE 5.3.8. Inform the Control Room the RCIC trip mechanism is reset.

5.4 Alternate Suppression Pool Makeup NOTE At the direction of the CRS/OSM, Steps 5.4.1 through 5.4.3 may be performed after RCIC is placed in service.

5.4.1. Notify Radiation Protection prior to running RCIC.

5.4.2. Place RHR into Suppression Pool Cooling mode per SOP-0031, Residual Heat Removal.

5.4.3. Place Containment Purge in service per SOP-0059, Containment HVAC System.

5.4.4. IF RCIC is running in support of testing, THEN begin monitoring the suppression pool average temperature per STP-057-0700, Suppression Pool Average Water Temperature Verification During Testing That Adds Heat to the Suppression Pool.

5.4.5. Start E51-C002C, GLAND SEAL COMPRESSOR.

CRITICAL STEP 5.4.6. Open E51-F045, RCIC STEAM SUPPLY TURBINE STOP VALVE.

5.4.7. Verify the following valves are closed:

• E51-F025, RCIC STM SPLY DR POT UP STREAM ISOL VALVE

• E51-F026, RCIC STM SUPLY DR POT DN STREAM ISOL VALVE

• E51-F004, RCIC TURB EXH DR POT UP STREAM ISOL VALVE

• E51-F005, RCIC TURB EXH DR POT DN STREAM ISOL VALVE 5.4.8. Verify E51-F019, RCIC MIN FLOW VLV TO SUPPRESSION POOL is open.

5.4.9. WHEN RCIC is no longer required, THEN shut down RCIC per Section 6.

SOP-0035 REV - 045 PAGE 28 OF 77

CONTINUOUS USE 5.5 Operation of the RCIC Gland Seal Compressor 5.5.1. IF the compressor is being run to lower the RCIC Exhaust Drain Trap level, THEN, verify E51-F004, RCIC TURB EXH DR POT UP STREAM ISOL VALVE is open.

5.5.2. Start E51-C002C, GLAND SEAL COMPRESSOR.

NOTE If Annunciator, P601-21A-A02, RCIC TURBINE EXHAUST DRAIN TRAP LEVEL HIGH was alarming, then run the compressor at least 20 minutes after the alarm has cleared.

5.5.3. WHEN the GLAND SEAL COMPRESSOR is no longer needed, THEN stop E51-C002, GLAND SEAL COMPRESSOR.

5.6 Manual swap of Suppression Pool and CST suction valves 5.6.1. Swap from CST to Suppression Pool

1. Close E51-F010, RCIC PUMP CST SUCTION VALVE.

2. WHEN E51-F010 indicates dual indication, THEN open E51-F031, RCIC PUMP SUP PL SUCTION VALVE.

3. Verify both valves fully stoke to prevent adding excessive amounts of water to the Suppression Pool.

5.6.2. Swap from Suppression Pool to CST

1. Close E51-F031, RCIC PUMP SUP PL SUCTION VALVE.

2. WHEN E51-F031 indicates dual indication, THEN open E51-F010, RCIC PUMP CST SUCTION VALVE.

3. Verify both valves fully stroke to prevent adding excessive amounts of water to the Suppression Pool.

SOP-0035 REV - 045 PAGE 29 OF 77

CONTINUOUS USE 5.7 RPV Level Control 5.7.1. IF RCIC operating AND aligned to the CST, THEN shift to inject to the RPV as follows:

1. IF desired swap RCIC Suction to the CST per Section 5.6.2.

2. Open E51-F013, RCIC INJECT ISOL VALVE

3. Close the following valves

1) E51-F022, RCIC TEST BYPASS VLV TO CST

2) E51-F059, RCIC TEST RETURN VLV TO CST CAUTION Operation below 2300 rpm may cause the turbine exhaust check valve damage due to valve chattering. Therefore minimize operation below 2300 rpm. Ref 7.13 5.7.2. With E51-R600, RCIC PUMP FLOW FLOW CONTROLLER HVYC002, in AUTO OR MANUAL, adjust RCIC flow setpoint as required to maintain RPV water level.

5.7.3. IF desired to swap to pressure control (CST to CST), THEN Go To Section 5.8 prior to reaching Level 8.

NOTE Section 5.8 is used to augment pressure control when other sources of level control are available and RPV level is less than Level 8 and greater than Level 2.

5.8 RPV Pressure Control 5.8.1. Place E51-C002C, GLAND SEAL COMPRESSOR to START.

5.8.2. IF necessary reset any RCIC Initiation signals as follows:

1. Verify Reactor Water Level is greater than Level 2.

2. Depress the RCIC INITIATION RESET pushbutton.

3. Verify RCIC INITIATION RESET white light is off.

SOP-0035 REV - 045 PAGE 30 OF 77

CONTINUOUS USE 5.8.3. IF necessary, THEN swap RCIC suction to the CST per Section 5.6.2 or EOP-0005, Enclosure 3 Defeating RCIC High S/P Water Level Suction Transfer Interlock as directed by the OSM/CRS.

5.8.4. Open E51-F059, RCIC TEST RETURN VLV TO CST.

CAUTION Operation below 2300 rpm may cause the turbine exhaust check valve damage due to valve chattering. Therefore minimize operation below 2300 rpm. Ref 7.13 5.8.5. Throttle open E51-F022, RCIC TEST BYPASS VLV TO CST to raise RCIC turbine rpm to greater than 2300 rpm and flow to greater than 150 gpm.

5.8.6. When RCIC pump discharge pressure is less than RCIC steam turbine pressure, Close E51-F013, RCIC INJECT ISOL VALVE.

5.8.7. Control pressure as follows:

• Throttle E51-F022, RCIC TEST BYPASS VLV TO CST open or closed to change RCIC pump Discharge pressure.

• Adjust E51-R600, RCIC PUMP FLOW CONTROLLER HYVC002 in MAN or AUTO to raise and lower pressure AND raise and lower rpm.

5.8.8. IF desired, THEN Maximize the RCIC system as follows:

1. Using E51-R600, RCIC PUMP FLOW FLOW CONTROLLER HYVC002 raise RCIC flow to 600 gpm.

2. Throttle E51-F022, RCIC TEST BYPASS VLV TO CST to raise RCIC pump discharge pressure to 1000 to 1100 psig

3. WHEN Maximizing is no longer required, THEN Go To Step 5.8.7 to control pressure.

SOP-0035 REV - 045 PAGE 31 OF 77

RJPM-NRC-M14-S2 Rev 1 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Reopen MSIVs Following Automatic Isolation OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 20 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: January 21, 2014 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 17

RJPM-NRC-M14-S2 Rev. 1 EXAMINER INFO SHEET Task Standard: All MSIVs are open using SOP-0011, Section 4.2.

Synopsis: The reactor is shutdown, pressurized with the MSIVs closed following an automatic isolation. This task will re-open all MSIVs using SOP-0011, Main Steam System, Section 4.2, Opening MSIVs During Hot Startup/Recovery From Automatic Isolation.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to open the Main Steam Isolation Valves (MSIVs) in accordance with the SOP.

3) Initial Conditions:

The reactor is shutdown and pressurized.

All MSIVs are closed.

Main Condenser vacuum is established.

The Circulating Water System is lined up in accordance with SOP-0006.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-S2 Rev 1 Page 2 of 17

RJPM-NRC-M14-S2 Rev. 1 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Open MSIVs During Recovery from 239007001001 239001 A4.01 4.2 / 4.0 Automatic Isolation 295020 AA1.01 3.6 / 3.6 Reason for Revision Revision 1 created to reflect procedure change within revision 29 of SOP-0011.

REFERENCES:

APPLICABLE OBJECTIVES SOP-0011, Rev 29 RLP-STM-0109, Obj 7, 12 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0011, Main Steam System (Sim Copy) __3__

SIMULATOR CONDITIONS & SETUP:

1. IC # 212

2. Rx Power: Shutdown. pressurized

3. MSIVs closed on an initiation signal; (MSS021A&B) signal has been cleared.

4. Condenser Vacuum established being maintained by Condenser Air Removal Pump A.

5. HPCS racked out/ tagged out; Feed Pumps tripped

6. RHR-A in Suppression Pool Cooling

7. Event T1: Malfunction: RCIC Turbine trip to activate when turbine speed reaches 90%

delete in 10 seconds; this will simulate a spurious trip and allow resetting the trip.

8. Several Alarms are forced on (they continue to come in and out, causing distraction)

P680_4-A5, A11, _5-C6 P870_52-E4, F1, G3, H1, H3 CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: All MSIVs are open using SOP-0011, Section 4.2.

RJPM-NRC-M14-S2 Rev 1 Page 3 of 17

RJPM-NRC-M14-S2 Rev. 1 PERFORMANCE:

START TIME:

SOP-0011 Section 4.2, Opening MSIVs During Hot Startup/Recovery From Automatic Isolation PROCEDURE CAUTION Bypassing the Condenser Low Vacuum trip and opening the MSIVs following an Automatic Isolation could result in a Condenser overpressurization condition. The Condenser Low Vacuum Bypass switches should only be used during startup.

Procedure Step: 4.2.1 IF Main Condenser vacuum is not established, THEN place the following switches in BYPASS:

Standard NA Cue Notes No actions necessary due to initial conditions.

Procedure Step: 4.2.2 Verify the current alignment of the Circulating Water Pumps, Cooling Towers and Condenser Waterboxes is in accordance with SOP-0006, Circulating Water System Standard NA Cue Notes No actions necessary due to initial conditions RJPM-NRC-M14-S2 Rev 1 Page 4 of 17

RJPM-NRC-M14-S2 Rev. 1

1. Procedure Step: 4.2.3 At H13-P601, place the following control switches in the CLOSE position:

B21-AOVF022A MSL A INBD MSIV B21-AOVF022B MSL B INBD MSIV B21-AOVF022C MSL C INBD MSIV B21-AOVF022D MSL D INBD MSIV Standard Applicant placed all four Inboard MSIV control switches to CLOSE.

Cue Notes Results SAT UNSAT

2. Procedure Step: 4.2.3 At H13-P601, place the following control switches in the CLOSE position:

B21-AOVF028A MSL A OUTBD MSIV B21-AOVF028B MSL B OUTBD MSIV B21-AOVF028C MSL C OUTBD MSIV B21-AOVF028D MSL D OUTBD MSIV Standard Applicant placed all four Outboard MSIV control switches to CLOSE.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 5 of 17

RJPM-NRC-M14-S2 Rev. 1 PROCEDURE NOTE Step 2.11 contains the MSIV isolation signals.

3. Procedure Step: 4.2.4 Check that an MSIV isolation signal is not present.

Standard Applicant verified that all MSIV isolation signals listed in Step 2.11 are clear.

Cue As backpanel operator, report that Main Steam Line Flow is zero psid.

As backpanel operator, report that Steam Tunnel Temperature is 126°F Notes There are six isolation signals listed in step 2.11:

RPV Level 1 (-143)

Main Steam Line High Flow (185/189 psid) {backpanel}

MSL Low Pressure, {bypassed when mode switch not in RUN}

Steam Tunnel High Temperature (173°F) {backpanel}

Low Condenser Vacuum, (8.5 inches Hg)

(4) Manual Isolation Pushbuttons (B21H-S25A, B, C, D)

Results SAT UNSAT

4. Procedure Step: 4.2.5 At H13-P680, check the Turbine Pressure Regulator setpoint is greater than RPV pressure.

Standard Applicant compared RPV Pressure to the Turbine Pressure Regulator Setpoint.

Cue Notes Press Regulator pressure is 950 psig; RPV pressure is cycling on Low-Low-Set Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 6 of 17

RJPM-NRC-M14-S2 Rev. 1

5. Procedure Step: 4.2.6 Depress the following pushbuttons to reset the inboard and outboard MSIV isolation logic:

• B21H-S33, INBD ISOLATION SEAL-IN RESET Standard Applicant depressed the B21H-S33, Inboard Isolation Seal-In Reset Pushbutton Cue Notes The pushbuttons S33 and S32 need not be pressed in any particular order, but both must be pushed.

Indications after action MSL Drains white light is on Results SAT UNSAT

6. Procedure Step: 4.2.6 Depress the following pushbuttons to reset the inboard and outboard MSIV isolation logic:

• B21H-S32, OUTBD ISOLATION SEAL-IN RESET Standard Applicant depressed the B21H-S32, Outboard Isolation Seal-In Reset Pushbutton Cue Notes Indications after action MSL Drains white light is on Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 7 of 17

RJPM-NRC-M14-S2 Rev. 1

7. Procedure Step: 4.2.7 Verify open the following to equalize pressure across the outboard MSIVs:

B21-MOVF067A, MSL A DRAIN VALVE B21-MOVF067B, MSL B DRAIN VALVE B21-MOVF067C, MSL C DRAIN VALVE B21-MOVF067D, MSL D DRAIN VALVE Standard Applicant opened the 67A, B, C, and D drain valves, by moving the respective control switches from CLOSE to AUTO.

Cue Notes The 6 valves in step 4.2.7 may be verified in any order.

Valve is open when red light is ON and green light is OFF.

Results SAT UNSAT

8. Procedure Step: 4.2.7 Verify open the following to equalize pressure across the outboard MSIVs:

B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE Standard Applicant opened B21-MOVF068 drain valve by turning the switch to open and holding it until the red light is on and the green light is off.

Cue Notes The 6 valves in step 4.2.7 may be verified in any order.

Results SAT UNSAT

9. Procedure Step: 4.2.7 Verify open the following to equalize pressure across the outboard MSIVs:

• B21-MOVF086, MSL DRAIN HDR SHUTOFF VALVE Standard Applicant verified the keylock B21-MOVF086 drain valve open, by verifying red light is ON and green light is OFF.

Cue Notes The 6 valves in step 4.2.7 may be verified in any order.

Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 8 of 17

RJPM-NRC-M14-S2 Rev. 1

10. Procedure Step: 4.2.8 Verify B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE is closed.

Standard Applicant verified the warmup header supply valve closed by checking the red light off and the green light on.

Cue Notes Results SAT UNSAT Procedure Step: 4.2.9 IF the differential pressure across the outboard MSIVs is greater than 50 psid as indicated by the difference between MSS-PI101, MAIN STEAM HEADER PRESSURE and Main Condenser Vacuum, THEN perform the following to reduce the differential pressure:

Standard Cue Notes No applicant action is necessary; this is a placekeeper.

11. Procedure Step: 4.2.9

1. Open at least one of the following valves:

• DTM-AOV12A MSL EQUALIZING HDR DRAIN BYPASS VALVE

• DTM-AOV12B MSL EQUALIZING HDR DRAIN BYPASS VALVE

• DTM-AOV5A BYP VLV BEFORE SEAT DR

• DTM-AOV5B BYP VLV BEFORE SEAT DR Standard Applicant opened one or more of the above valves to equalize pressure (Red lights on/Green lights off)

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 9 of 17

RJPM-NRC-M14-S2 Rev. 1

12. Procedure Step: 4.2.9

2. WHEN the differential pressure across the outboard MSIVs is less than or equal to 50 psid, THEN close the following valves that were opened in the previous step:

• DTM-AOV12A MSL EQUALIZING HDR DRAIN BYPASS VALVE

• DTM-AOV12B MSL EQUALIZING HDR DRAIN BYPASS VALVE

• DTM-AOV5A BYP VLV BEFORE SEAT DR

• DTM-AOV5B BYP VLV BEFORE SEAT DR Standard Applicant opened one or more of the above valves to equalize pressure (Red lights on/Green lights off)

Cue Notes Results SAT UNSAT

13. 4.2.10 Place the following control switches in AUTO:

• Procedure Step:

B21-AOVF028A, MSL A OUTBD MSIV B21-AOVF028B, MSL B OUTBD MSIV B21-AOVF028C, MSL C OUTBD MSIV B21-AOVF028D, MSL D OUTBD MSIV Standard Applicant placed control switches for Outboard MSIVs to AUTO.

Cue Notes The 4 valves in step 4.2.10 may be manipulated in any order.

As each switch is placed in AUTO, the associated MSIV will come open.

Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 10 of 17

RJPM-NRC-M14-S2 Rev. 1

14. Procedure Step: 4.2.11 Check outboard MSIVs are open.

Standard Applicant verified outboard MSIVs open, by verifying red lights ON and green lights OFF.

Cue Notes Results SAT UNSAT Procedure Step: 4.2.12 On H13-P623, Observe the indicating lights and ammeters to confirm the energization of both solenoids on each outboard MSIV.

Standard NA Cue As a backpanel operator, indicate that all readings are acceptable.

Notes No actions are taken by the applicant except to simulate going to the backpanel and check the two pilot solenoid energized lights for each of the four outboard MSIVs and the four ammeters registering some amount of current to show that they are energized.

RJPM-NRC-M14-S2 Rev 1 Page 11 of 17

RJPM-NRC-M14-S2 Rev. 1

15. Procedure Step: 4.2.13 Verify closed the following:

B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE Standard Applicant verified the drain bypass valve closed by checking the red light off and the green light on.

Cue Notes Results SAT UNSAT

16. Procedure Step: 4.2.13 Verify closed the following:

B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE Standard Applicant closed the drain valve by placing the control switch to CLOSE and verified the green light on and the red light off.

Cue Notes Results SAT UNSAT

17. Procedure Step: 4.2.13 Verify closed the following:

B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE Standard Applicant closed the drain bypass valve by placing and holding the control switch to CLOSE until the red light is off and the green light is on.

Cue Notes Results SAT UNSAT

18. Procedure Step: 4.2.13 Verify closed the following:

B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE Standard Applicant closed the drain valve by placing the control switch to CLOSE and verified the green light on and the red light off.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 12 of 17

RJPM-NRC-M14-S2 Rev. 1

19. Procedure Step: 4.2.14 Verify open the following:

B21-MOVF016, MSL WARMUP HDR INBD CONTMT ISOL VLV B21-MOVF019, MSL WARMUP HDR OUTBD CONTMT ISOL VLV B21-MOVF085, MSL WARMUP HDR SHUTOFF VALVE Standard Applicant verified the three warmup header valves OPEN by checking the red lights on and the green lights off.

Cue Notes Results SAT UNSAT

20. Procedure Step: 4.2.15 Open slowly B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE to equalize pressure across the inboard MSIVs.

Standard Applicant throttled opened the warmup header supply valve.

Cue Notes Applicant will move back and forth between throttling the 20 valve and checking pressures per the next step.

How aggressively the applicant is willing to open the 20 valve will determine the time it takes to equalize pressure; It takes about 7 minutes to equalize pressure Results SAT UNSAT

21. Procedure Step: 4.2.16 Monitor differential pressure across the inboard MSIVs using the following:

C33-R605, REACTOR PRESSURE MSS-PI101, MAIN STEAM HEADER PRESSURE.

Standard Applicant monitored differential pressure across the inboard MSIVs using the appropriate meters.

Cue Notes The Reactor Pressure meter is on insert 3 of the 680 panel The Main Steam pressure meter is on insert 52 of the 870 panel.

Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 13 of 17

RJPM-NRC-M14-S2 Rev. 1 PROCEDURE NOTE Opening an MSIV can cause a Reactor water level transient.

Slowly opening the MSIV using the OPEN/SLOW TEST position and intermittent release of the TEST pushbutton can mitigate the level transient.

22. 4.2.17 WHEN differential pressure across the inboard MSIVs is less than or

• Procedure Step:

equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

B21-AOVF022A, MSL A INBD MSIV Standard Applicant placed the A inboard MSIV switch to AUTO (AFTER d/p was equal or less than 50 psid).

Cue Notes Applicant may elect to use the SLOW-TEST button Results SAT UNSAT

23. 4.2.17 WHEN differential pressure across the inboard MSIVs is less than or

• Procedure Step:

equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

B21-AOVF022B, MSL B INBD MSIV Standard Applicant placed the B inboard MSIV switch to AUTO.

Cue Notes Results SAT UNSAT

24. 4.2.17 WHEN differential pressure across the inboard MSIVs is less than or

• Procedure Step:

equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

B21-AOVF022C, MSL C INBD MSIV Standard Applicant placed the C inboard MSIV switch to AUTO.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S2 Rev 1 Page 14 of 17

RJPM-NRC-M14-S2 Rev. 1

25. 4.2.17 WHEN differential pressure across the inboard MSIVs is less than or

• Procedure Step:

equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

B21-AOVF022D, MSL D INBD MSIV Standard Applicant placed the D inboard MSIV switch to AUTO.

Cue Notes Results SAT UNSAT

26. Procedure Step: 4.2.18 Check inboard MSIVs are open.

Standard Applicant checked all four inboard MSIVs indicate open (Red lights on/Green lights off)

Cue Notes Results SAT UNSAT Terminating Cue: All MSIVs are open using SOP-0011, Section 4.2.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-S2 Rev 1 Page 15 of 17

RJPM-NRC-M14-S2 Rev. 1 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S2 Rev 1 Page 16 of 17

RJPM-NRC-M14-S2 Rev. 1 OPERATOR CUE SHEET INITIAL CONDITIONS:

The reactor is shutdown and pressurized.

All MSIVs are closed.

Main Condenser vacuum is established.

The Circulating Water System is lined up in accordance with SOP-0006.

INITIATING CUE:

The CRS has directed you to open the Main Steam Isolation Valves (MSIVs) in accordance with the SOP.

RJPM-NRC-M14-S2 Rev 1 Page 17 of 17

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• MAIN STEAM SYSTEM (SYS #109)

PROCEDURE NUMBER: *SOP-0011 REVISION NUMBER: *028 Effective Date: *01/22/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0011 REV - 028 PAGE 1 OF 91

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES ......................................................................................................................7 4 SYSTEM STARTUP..................................................................................................................8 4.1 Opening MSIVs With Reactor Cold, Depressurized ........................................................8 4.2 Opening MSIVs During Hot Startup/Recovery From Automatic Isolation .....................12 5 SYSTEM OPERATION .............................................................................................................17 5.1 Manual Isolation of a Main Steam Line During Power Operation ...................................17 5.2 Manual Automatic Depressurization System Operation...................................................18 5.3 Returning an Isolated Main Steam Line to Service During Power Operations ................19 5.4 Automatic Depressurization System Operation................................................................20 5.5 Manual Operation of SVV Compressors ..........................................................................22 5.6 Slow-Opening the MSIVs .................................................................................................22 5.7 Temporary Air Supply to SVV Headers From IAS Diesel Air Compressor When PVLCS Is Not Operable (EC-0000000035) ...........................................................23 5.8 SVV Dryer Purge/Depressurization Adjustments.............................................................28 5.9 Alternating SVV Compressors..........................................................................................29 5.10 Starting up the SVV System Following System Depressurization ...................................30 5.11 Setting Dryer Skids A&B Cycle Time..............................................................................31 5.12 Alternating Outboard MSIV Air Strainers........................................................................31 5.13 Opening Outboard MSIVs with Inboard MSIVs Open in Hot Standby ...........................31 5.14 Cycling Safety Relief Valves (SRV) at Low Power .........................................................35 6 SYSTEM SHUTDOWN.............................................................................................................37 6.1 Main Steam System Shutdown .........................................................................................37 6.2 Automatic Depressurization System Shutdown................................................................39 6.3 SVV System Shutdown.....................................................................................................40 7 REFERENCES ...........................................................................................................................41 8 RECORDS..................................................................................................................................41 ATTACHMENT 1A - VALVE LINEUP - MAIN STEAM SYSTEM (SAFETY RELATED) .......................................................................................................................42 SOP-0011 REV - 028 PAGE 2 OF 91

CONTINUOUS USE ATTACHMENT 1B - VALVE LINEUP - MAIN STEAM SYSTEM .............................................50 ATTACHMENT 2A - INSTRUMENT AND VALVE LINEUP - MAIN STEAM SYSTEM (SAFETY RELATED) .....................................................................................62 ATTACHMENT 2B - INSTRUMENT AND VALVE LINEUP - MAIN STEAM SYSTEM ...........................................................................................................................63 ATTACHMENT 3A - ELECTRICAL LINEUP - MAIN STEAM SYSTEM (SAFETY RELATED) .......................................................................................................................72 ATTACHMENT 3B - ELECTRICAL LINEUP - MAIN STEAM SYSTEM ..................................75 ATTACHMENT 4A - CONTROL BOARD LINEUP - MAIN STEAM SYSTEM (SAFETY RELATED)......................................................................................................80 ATTACHMENT 4B - CONTROL BOARD LINEUP - MAIN STEAM SYSTEM.........................87 ATTACHMENT 5 - PURGE RATE CORRECTION FACTOR (C-1) ............................................91 SOP-0011 REV - 028 PAGE 3 OF 91

CONTINUOUS USE 1 PURPOSE 1.1 This procedure is to provide instructions for the operation of the Main Steam and Automatic Depressurization Systems.

2 PRECAUTIONS AND LIMITATIONS 2.1 Technical Specification 3.6.1.3 contains the Limiting Conditions for Operations with an inoperable MSIV.

2.2 Technical Specifications 3.3.5.1, 3.5.1 and Technical Requirement 3.3.5.1.1 contain the Limiting Conditions for Operations pertaining to ADS.

2.3 Opening Main Steam Isolation Valves (MSIVs) or the Main Steam Shutoff Valves (MSSVs) with a differential pressure greater than 200 psid across the valves can result in valve damage. A differential pressure of less than or equal to 50 psid is preferred when opening the valves.

C 2.4 Reactor power shall be limited to less than or equal to 75% RTP with one Main Steam Line out of service.

2.5 During startup and periods of low steam flow operation, the drain system flow path is maintained in order to protect downstream plant equipment from the eroding effects of moisture carryover from condensate in the steam lines.

2.6 The MSIVs isolate on a 1 out of 2 signal taken twice.

2.7 The Main Steam Line Drain Isolation Valves isolate on a 2 out of 2 signal.

2.8 Operation of Logic A components is described. Logic B components are similar and are indicated by parenthesis.

2.9 Placing the ADS A(B) Manual Inhibit Control Switch in the INHIBIT position prevents automatic ADS valve operation.

2.10 Performing a slow open of the MSIVs can prevent RPV pressure/level transients while opening MSIVs.

SOP-0011 REV - 028 PAGE 4 OF 91

CONTINUOUS USE 2.11 The Main Steam System automatically isolates when any of the following conditions exist:

2.11.1. Reactor Low-Low-Low Water Level, Level 1 (-143 in.)

2.11.2. Main Steam Line High Flow

• MSL A, 185 psid

• MSL B, 189 psid

• MSL C, 189 psid

• MSL D, 189 psid 2.11.3. Main Steam Line Low Pressure, 849 psig, bypassed when Reactor Mode Switch not in RUN position.

2.11.4. Steam Tunnel High Temperature, 173°F 2.11.5. Low Condenser Vacuum, 8.5 inches Hg, may be bypassed during startup.

2.11.6. Manual Isolation via B21H-S25(A)(B)(C)(D), MANUAL ISOLATION Pushbuttons.

2.12 SVV Air Compressors supply the normal air supply to the SRV Air System with a capacity of 17 scfm air at 175 psig.

2.12.1. IF SVV Compressors can not operate, THEN the SRV/ADS Valves remain operable if an alternate air system can supply air to the respective SVV Air Headers at a minimum of 131 psig.

2.12.2. Failure to have A&B Skid Dryers cycle within 15 sec of each other can result in excessive compressor cycling.

2.13 Depressurizing/pressurizing the SVV System or isolation/restoration of supply headers to accumulators disables/enables suppression pool level transmitters E51-LTN036A(E) and E22-LTN055C(G). This defeats/enables the RCIC and HPCS Suppression Pool High Level automatic transfer from the CST to the suppression pool. TS 3.3.5.1-1.3.e and TR 3.3.5.1-1.3.e for HPCS, and TS 3.3.5.2-1.4 and TR 3.3.5.2-1.4 for RCIC, contains the specific requirements. Consider actions to preclude unplanned, inadvertent actuation.

2.14 WHEN air header pressure lowers to less than 55 psig, THEN the Main Control Valve and Air Pilot Valve internal positioning springs overcome the air pressure and reposition these valves. This causes air to be vented from under the MSIV piston by both the Main Control Valve and Air Pilot Valve, and provide accumulator air pressure to the top of the piston via the Main Control Valve, causing the MSIV to close.

SOP-0011 REV - 028 PAGE 5 OF 91

CONTINUOUS USE 2.15 Placing the MSIV switch in CLOSE will cause the MSIV to fast close.

2.16 IF alarms at local panel SVV-PNL30A(B) are experienced due to dewpoint, 11°F at 131 psig, THEN a WR should be generated.

2.17 IF the local dewpoint transmitter reads greater than +3°F at 131 psig, for the SVV Dryer to remain in service, THEN the minimum design criteria of +19°F at 131 psig should be verified using a more accurate dewpoint monitor from M&TE, and a CR should be generated for an engineering evaluation.

M 2.18 WHEN dry-stroking a MSIV, defined as the movement of the main poppet from closed to open and back closed, or open to closed and back open without the presence of steam or water, THEN allow 10 minutes between each valve stroke for the valve internals to cool down (Reference 7.7, ER 2001-0398). System Engineering should perform an Engineering Evaluation if a MSIV is going to be dry-stroked more than 5 consecutive times.

2.19 Alignment of the Circulating Waterboxes should be in accordance with the following table during applicable conditions of operation to maintain adequate cooling to the condenser.

Plant Condition Waterbox Requirements Startup and Power Ascension up to 2318 The D Waterbox can remain Out of MWt Service until 2318 MWt.

>2318 MWt All waterboxes are required to be in service Following a Reactor Scram or with the Any 1 water box maybe used to provide Mode switch in Shutdown sufficient cooling (Waterbox configuration should be in alignment with SOP-0006 to prevent runout of the Circulating Water Pumps) 2.20 The Steam Bypass Valve exhaust diffusers are located directly above the B and C waterbox tube bundles. All Main Steam Lines drains are located in the North end of the condenser, directly over the A waterbox tube bundle. If Steam Bypass Valves or Main Steam Line drains are admitted to the condenser during normal operation, with a water box out of service, there could be local overheating of the dry tubes in the vicinity of the drains causing tubes to expand and buckle. The Applicable water box should be in service before admitting steam to the box via the Main Steam Line Drains or Bypass Valves.

2.21 Bypassing the Condenser Low Vacuum trip and opening the MSIVs following an Automatic Isolation could result in a Condenser overpressurization condition. The Condenser Low Vacuum Bypass switches should only be used during startup.

SOP-0011 REV - 028 PAGE 6 OF 91

CONTINUOUS USE 3 PREREQUISITES 3.1 Verify Outboard and Inboard Isolation signals are reset with the following pushbuttons:

• B21H-S32, OUTBD ISOLATION SEAL-IN RESET

• B21H-S33, INBD ISOLATION SEAL-IN RESET 3.2 Check that the Instrument Air System is in operation per SOP-0022, Instrument Air System.

3.3 Check that the 125 VDC System is in operation per SOP-0049, 125 VDC System.

3.4 Check that the 120 VAC System is in operation per SOP-0048, 120 VAC System.

3.5 Check that the 480 VAC System is in operation per SOP-0047, 480 VAC System.

3.6 Check that the EHC Oil System is in operation per SOP-0014, EHC Oil System.

3.7 Check that the Circ Water, Cooling Towers, and Vacuum Priming System is in operation per SOP-0006, Circulating Water, Cooling Tower and Vacuum Priming.

3.8 Check that the MSIV Positive Leakage Control System is in operation per SOP-0034, MSIV Sealing System (Positive Leakage Control System) and Penetration Valve Leakage Control.

3.9 Check that the Auxiliary Building HVAC System is in operation per SOP-0065, HVAC-Auxiliary Building.

3.10 Check that the Containment HVAC System is in operation per SOP-0059, Containment HVAC System.

3.11 Check that the Turbine Building HVAC System is in operation per SOP-0064, Turbine Building HVAC System.

3.12 Check that Nuclear Boiler Instrumentation is in operation per SOP-0001, Nuclear Boiler Instrumentation.

3.13 Check that the Penetration Valve Leakage Control System is in standby per SOP-0034, MSIV Sealing System (Positive Leakage Control) and Penetration Valve Leakage Control.

3.14 Verify that the system is lined up for startup.

SOP-0011 REV - 028 PAGE 7 OF 91

CONTINUOUS USE 4 SYSTEM STARTUP NOTE All switch operations take place at H13-P601 unless otherwise noted.

4.1 Opening MSIVs With Reactor Cold, Depressurized 4.1.1. IF Main Condenser vacuum is not established, THEN place the following keylock switches to BYPASS:

• B21H-S24A, DIV 1 CONDENSER LOW VACUUM BYPASS

• B21H-S24B, DIV 2 CONDENSER LOW VACUUM BYPASS

• B21H-S24C, DIV 3 CONDENSER LOW VACUUM BYPASS

• B21H-S24D, DIV 4 CONDENSER LOW VACUUM BYPASS 4.1.2. Verify the current alignment of the Circulating Water Pumps, Cooling Towers and Condenser Waterboxes is in accordance with SOP-0006, Circulating Water System.

4.1.3. WHEN Reactor water temperature reaches 190°F, THEN perform the following:

NOTE GOP-0001, Plant Startup can be referenced for tag-out of B21-MOVF001, RX DN STREAM HEAD VENT TO DW EQPT DR SUMP and B21-MOVF002, RX UP STREAM HEAD VENT TO DW EQPT DR SUMP.

1. Close B21-MOVF001, RX DN STREAM HEAD VENT TO DW EQPT DR SUMP.

2. Close B21-MOVF002, RX UP STREAM HEAD VENT TO DW EQPT DR SUMP.

3. Open B21-MOVF005, RX HEAD VENT TO MSL A.

SOP-0011 REV - 028 PAGE 8 OF 91

CONTINUOUS USE 4.1.4. WHEN a positive Reactor pressure has been established, THEN equalize around and open the MSIVs as follows:

1. Place the following control switches in AUTO to open the outboard MSIVs:

• B21-AOVF028A, MSL A OUTBD MSIV

• B21-AOVF028B, MSL B OUTBD MSIV

• B21-AOVF028C, MSL C OUTBD MSIV

• B21-AOVF028D, MSL D OUTBD MSIV

2. Check open the outboard MSIVs.

3. On H13-P623, observe the indicating lights and ammeters to confirm the energization of both solenoids on each outboard MSIV.

4. Verify closed the following:

• B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE

5. Verify open the following:

• B21-MOVF016, MSL WARMUP HDR INBD CONTMT ISOL VLV

• B21-MOVF019, MSL WARMUP HDR OUTBD CONTMT ISOL VLV

• B21-MOVF085, MSL WARMUP HDR SHUTOFF VALVE

6. Open slowly B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE to equalize pressure across the inboard MSIVs.

SOP-0011 REV - 028 PAGE 9 OF 91

CONTINUOUS USE

7. Monitor differential pressure across the inboard MSIVs using the following:

• C33-R605, REACTOR PRESSURE

• MSS-PI101, MAIN STEAM HEADER PRESSURE NOTE Opening an MSIV can cause a Reactor water level transient. Slowly opening the MSIV using the OPEN/SLOW TEST position and intermittent release of the TEST Pushbutton can mitigate the level transient.

CRITICAL STEP

8. WHEN differential pressure across the inboard MSIVs is less than or equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

• B21-AOVF022A, MSL A INBD MSIV

• B21-AOVF022B, MSL B INBD MSIV

• B21-AOVF022C, MSL C INBD MSIV

• B21-AOVF022D, MSL D INBD MSIV

9. Check the inboard MSIVs are open.

10. On H13-P622, observe the indicating lights and ammeters to confirm the energization of both solenoids on each inboard MSIV.

11. Close B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE.

12. Open the following:

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE SOP-0011 REV - 028 PAGE 10 OF 91

CONTINUOUS USE 4.1.5. IF bypassed in Step 4.1.1 AND Condenser vacuum is greater than 8.5 inches Hg and stable, THEN place the following keylock switches in NORM:

• B21H-S24A, DIV 1 CONDENSER LOW VACUUM BYPASS

• B21H-S24B, DIV 2 CONDENSER LOW VACUUM BYPASS

• B21H-S24C, DIV 3 CONDENSER LOW VACUUM BYPASS

• B21H-S24D, DIV 4 CONDENSER LOW VACUUM BYPASS 4.1.6. WHEN Reactor power is at 10%, THEN close the following:

1. At H13-P870:

• DTM-MOVSV1, MS STOP BEFORE SEAT DR

• DTM-MOVSV3, MS STOP BEFORE SEAT DR

• DTM-MOVSV5, MS STOP BEFORE SEAT DR

• DTM-MOVSV7, MS STOP BEFORE SEAT DR

• DTM-MOV175, MSL TO SJAE COND DR

2. At H13-P601:

• DTM-AOV12A, MSL EQUALIZING HDR DRAIN BYPASS

• DTM-AOV12B, MSL EQUALIZING HDR DRAIN BYPASS

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE SOP-0011 REV - 028 PAGE 11 OF 91

CONTINUOUS USE

3. At H13-P870:

• DTM-AOV5A, BYP VLV BEFORE SEAT DR

• DTM-AOV5B, BYP VLV BEFORE SEAT DR

• DTM-MOV51A, MSL TO MSR1 COND DR

• DTM-MOV51B, MSL TO MSR2 COND DR

• DTM-MOV142, MSL TO OFG COND DR 4.2 Opening MSIVs During Hot Startup/Recovery From Automatic Isolation CAUTION Bypassing the Condenser Low Vacuum trip and opening the MSIVs following an Automatic Isolation could result in a Condenser overpressurization condition. The Condenser Low Vacuum Bypass switches should only be used during startup.

4.2.1. IF Main Condenser vacuum is not established, THEN place the following switches in BYPASS:

• B21H-S24A, DIV 1 CONDENSER LOW VACUUM BYPASS

• B21H-S24B, DIV 2 CONDENSER LOW VACUUM BYPASS

• B21H-S24C, DIV 3 CONDENSER LOW VACUUM BYPASS

• B21H-S24D, DIV 4 CONDENSER LOW VACUUM BYPASS 4.2.2. Verify the current alignment of the Circulating Water Pumps, Cooling Towers and Condenser Waterboxes is in accordance with SOP-0006, Circulating Water System.

SOP-0011 REV - 028 PAGE 12 OF 91

CONTINUOUS USE 4.2.3. At H13-P601, place the following control switches in the CLOSE position:

• B21-AOVF022A, MSL A INBD MSIV

• B21-AOVF022B, MSL B INBD MSIV

• B21-AOVF022C, MSL C INBD MSIV

• B21-AOVF022D, MSL D INBD MSIV

• B21-AOVF028A, MSL A OUTBD MSIV

• B21-AOVF028B, MSL B OUTBD MSIV

• B21-AOVF028C, MSL C OUTBD MSIV

• B21-AOVF028D, MSL D OUTBD MSIV NOTE Step 2.11 contains the MSIV isolation signals.

4.2.4. Check that an MSIV isolation signal is not present.

4.2.5. At H13-P680, check the Turbine Pressure Regulator setpoint is greater than RPV pressure.

4.2.6. Depress the following pushbuttons to reset the inboard and outboard MSIV isolation logic:

• B21H-S33, INBD ISOLATION SEAL-IN RESET

• B21H-S32, OUTBD ISOLATION SEAL-IN RESET 4.2.7. Verify open the following to equalize pressure across the outboard MSIVs:

• B21-MOVF067A, MSL A DRAIN VALVE

• B21-MOVF067B, MSL B DRAIN VALVE

• B21-MOVF067C, MSL C DRAIN VALVE

• B21-MOVF067D, MSL D DRAIN VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-MOVF086, MSL DRAIN HDR SHUTOFF VALVE SOP-0011 REV - 028 PAGE 13 OF 91

CONTINUOUS USE 4.2.8. Place the following control switches in AUTO:

• B21-AOVF028A, MSL A OUTBD MSIV

• B21-AOVF028B, MSL B OUTBD MSIV

• B21-AOVF028C, MSL C OUTBD MSIV

• B21-AOVF028D, MSL D OUTBD MSIV 4.2.9. Check outboard MSIVs are open.

4.2.10. On H13-P623, observe the indicating lights and ammeters to confirm the energization of both solenoids on each outboard MSIV.

4.2.11. Verify closed the following:

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE 4.2.12. Verify open the following:

• B21-MOVF016, MSL WARMUP HDR INBD CONTMT ISOL VLV

• B21-MOVF019, MSL WARMUP HDR OUTBD CONTMT ISOL VLV

• B21-MOVF085, MSL WARMUP HDR SHUTOFF VALVE 4.2.13. Open slowly B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE to equalize pressure across the inboard MSIVs.

4.2.14. Monitor differential pressure across the inboard MSIVs using the following:

• C33-R605, REACTOR PRESSURE

• MSS-PI101, MAIN STEAM HEADER PRESSURE SOP-0011 REV - 028 PAGE 14 OF 91

CONTINUOUS USE NOTE Opening an MSIV can cause a Reactor water level transient. Slowly opening the MSIV using the OPEN/SLOW TEST position and intermittent release of the TEST Pushbutton can mitigate the level transient.

CRITICAL STEP 4.2.15. WHEN differential pressure across the inboard MSIVs is less than or equal to 50 psid, THEN place the following switches in AUTO to open the inboard MSIVs:

• B21-AOVF022A, MSL A INBD MSIV

• B21-AOVF022B, MSL B INBD MSIV

• B21-AOVF022C, MSL C INBD MSIV

• B21-AOVF022D, MSL D INBD MSIV 4.2.16. Check inboard MSIVs are open.

4.2.17. On H13-P622, observe the indicating lights and ammeters to confirm the energization of both solenoids on each inboard MSIV.

4.2.18. Close B21-MOVF020, MSL WARMUP HDR SUPPLY VALVE.

4.2.19. Open the following:

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE SOP-0011 REV - 028 PAGE 15 OF 91

CONTINUOUS USE 4.2.20. IF bypassed in Step 4.2.1 AND Main Condenser Vacuum is greater than 8.5 inches Hg and stable, THEN place the following keylock switches in NORM:

• B21H-S24A, DIV 1 CONDENSER LOW VACUUM BYPASS

• B21H-S24B, DIV 2 CONDENSER LOW VACUUM BYPASS

• B21H-S24C, DIV 3 CONDENSER LOW VACUUM BYPASS

• B21H-S24D, DIV 4 CONDENSER LOW VACUUM BYPASS 4.2.21. WHEN Reactor power is at 10%, THEN close the following:

• B21-MOVF021, MSL WARMUP HDR COND DRAIN BYP VALVE

• B21-MOVF068, MSL DRAIN HDR COND DRAIN BYP VALVE

• B21-AOVF033, MSL WARMUP HDR COND DRAIN VALVE

• B21-AOVF069, MSL DRAIN HDR COND DRAIN VALVE SOP-0011 REV - 028 PAGE 16 OF 91

RJPM-NRC-M14-S3 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Alternate Feedwater Level Control Channels OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: September 4, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-S3 Rev. 0 EXAMINER INFO SHEET Task Standard: Reactor Feedwater Level Control Channel has been swapped to A; FWLC is in Manual and maintaining level in the green band.

Synopsis: The reactor is at 100% power with FWLC Channel B selected. This task will align FWLC Channel A for level control using SOP-0009, Section 5.1. As an alternate path, there is a gradual upscale failure of the RPV Level Control Signal, which the operator will respond to by taking manual control of the FWLC System per AOP-0006, Condensate/Feedwater Failures.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to Swap Feedwater Level Control Channel from B to A.

3) Initial Conditions:

The reactor is at 100% power.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-S3 Rev 0 Page 2 of 8

RJPM-NRC-M14-S3 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Alternate Feedwater Level Control 259018001001 259001 A2.07 3.7 / 3.8 Channels A4.05 4.0 / 3.9 AA1.02 4.0 /

4.0 REFERENCES

APPLICABLE OBJECTIVES SOP-0009, Rev 6 RLP-STM-0107 Feedwater Level Control, AOP-0006, Rev 19 Obj 14 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0009, Rev 6 (Sim Copy) __4__

AOP-0006, Rev 19 (Sim. Copy)

SIMULATOR CONDITIONS & SETUP:

1. IC # 213

2. Required Power: 100%.

3. FWLC Channel B selected.

4. Event T1: zdi6(69) = 1 (Master FWLC AUTO selected)

5. Malfunction: B21001B (FWLC A transmitter failure) ; delay 5 sec, ramp to 60 in 3 minutes on T1 ; (A level will ramp upscale) 6.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Reactor Feedwater Level Control Channel has been swapped to A; FWLC is in Manual and maintaining level in the green band.

RJPM-NRC-M14-S3 Rev 0 Page 3 of 8

RJPM-NRC-M14-S3 Rev. 0 PERFORMANCE:

START TIME:

Procedure Step: 5.1.1 IF automatically controlling level on the Master Flow controller, THEN alternate the feedwater level control signals as follows:.

Standard NA Cue Notes No applicant action is necessary.

1. Procedure Step: 5.1.1.1 Ensure no deviation on C33-R600, FW REG VALVES MASTER FLOW CONTROLLER and place to MANUAL.

Standard Applicant located/identified and verified no deviation existed on C33-R600 by observing the needle to the left of the tape set; Applicant depressed the left black button on the Flow Controller and observed the yellow light ON and the green light OFF Cue Notes Results SAT UNSAT

2. 5.1.1.2 Swap the level control input by depressing either A or B on the RX

• Procedure Step:

LVL A/B SELECT pushbutton.

Standard Applicant located/identified and depressed the A Select pushbutton.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S3 Rev 0 Page 4 of 8

RJPM-NRC-M14-S3 Rev. 0

3. Procedure Step: 5.1.1.3 Check for proper operation, then return C33-R600, FW REG VALVES MASTER FLOW CONTROLLER to AUTO as follows:

1) Adjust tape set 2 inches above actual vessel level and observe the deviation signal is positive.

Standard Applicant located/identified and adjusted the tape set up two inches.

Applicant noted a positive deviation signal by observing the needle to the left of the tape set moving down.

Cue Notes Results SAT UNSAT

4. Procedure Step: 5.1.1.3 Check for proper operation, then return C33-R600, FW REG VALVES MASTER FLOW CONTROLLER to AUTO as follows:

2) Lower tape set 2 inches below actual vessel level and observe the deviation signal is negative.

Standard Applicant the tape set down two inches and noted a negative deviation signal by observing the needle to the left of the tape set moving up.

Cue Notes Results SAT UNSAT

5. Procedure Step: 5.1.1.3 Check for proper operation, then return C33-R600, FW REG VALVES MASTER FLOW CONTROLLER to AUTO as follows:

3) Match tape set to actual vessel level and observe in order to null the deviation signal.

Standard Applicant adjusted the tape set to null the deviation signal by observing the needle to the left of the tape set moving to the null position.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S3 Rev 0 Page 5 of 8

RJPM-NRC-M14-S3 Rev. 0

6. 5.1.1.3 Check for proper operation, then return C33-R600, FW REG

• Procedure Step:

VALVES MASTER FLOW CONTROLLER to AUTO as follows:

4) WHEN the level signal is nulled, THEN depress the AUTO Pushbutton and check the green light above the pushbutton is on.

Standard Applicant located/identified and observed the deviation signal to be null.

Applicant depressed the right black button on the Flow Controller and noted the yellow light OFF and the green light above the button ON.

Cue Notes When the AUTO pushbutton is depressed, a malfunction will begin ramping in. The A level signal begins rising which will cause a signal to the FWLC system to lower reactor water level.

When the applicant notices this abnormal situation, the applicant will transition to the ALTERNATE PATH.

Results SAT UNSAT ALTERNATE PATH:

AOP-0006 IMMEDIATE OPERATOR ACTIONS

7. 5.1 Manually control the feedwater level control system and/or reduce

• Procedure Step:

reactor power to mitigate any level transient..

Standard Applicant placed the FW REG VALVES MASTER FLOW CONTROLLER to Manual by depressing the left black button on C33-R600.

Applicant adjusted the Master Controller to obtain an actual RPV level in the green band.

Cue Notes ARP-680-3-C08 directs swapping the controller back to B level control.

Results SAT UNSAT Terminating Cue: Reactor Feedwater Level Control Channel has been swapped to A; FWLC is in Manual and maintaining level in the green band.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-S3 Rev 0 Page 6 of 8

RJPM-NRC-M14-S3 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S3 Rev 0 Page 7 of 8

RJPM-NRC-M14-S3 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The reactor is at 100% power.

INITIATING CUE:

The CRS has directed you to Swap Feedwater Level Control Channel from B to A.

RJPM-NRC-M14-S3 Rev 0 Page 8 of 8

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• REACTOR FEEDWATER SYSTEM (SYS #107)

PROCEDURE NUMBER: *SOP-0009 REVISION NUMBER: *060 Effective Date: *08/22/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0009R059EC-A Provide clarification on Startup Feedwater Reg Valve position limitations to ensure adequate margin for valve modulation while maintaining reactor level by adding P&L 2.22 and Cautions at Steps 4.6.1 and 4.7.1. (Ref CR-RBS-2013-4392 CA 11).

SOP-0009 REV - 060 PAGE 1 OF 98

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES FOR STARTUP AND OPERATIONS .......................................................8 4 SYSTEM STARTUP..................................................................................................................9 4.1 Reactor Feed Pump/Motor and Gear Increaser Lube Oil Startup.....................................9 4.2 Filling and Venting a Reactor Feed Pump Using CNS.....................................................11 4.3 Reactor Feed Pump Seal Venting (Cold and Warm Conditions)......................................16 4.4 Warming a Reactor Feed Pump ........................................................................................19 4.5 Reactor Feed Pump Startup...............................................................................................19 4.6 Establishing Flow to the Vessel Through the Startup Level Control Valve.....................25 4.7 Placing Startup Level Control in Automatic.....................................................................28 4.8 Placing a FWREG Valve in Manual with Startup Level Control Valve in Auto .............29 4.9 Transfer from StartUp Level Control to Master Level Control ........................................30 4.10 Single Element to Three Element Control Transfer..........................................................31 4.11 Placing the Second or Third FWREG Valve in Service ...................................................31 5 SYSTEM OPERATION .............................................................................................................33 5.1 Alternating Feedwater Level Control Signals...................................................................33 5.2 Restoring Reactor Water Level to Normal Following a Reactor Scram from High Power .......................................................................................................................34 5.3 Augmenting FWREGs While at High Power with Startup FWREG................................35 5.4 Removing Startup FWREG from FWREG Augmenting Mode........................................36 5.5 Manual Stroking of Start Up FWREG ..............................................................................36 5.6 Mitigating Feedwater Line Thermal Stratification ...........................................................37 5.7 Stroking FWS-MOV27A(B)(C), FWREG VLV 1A(1B)(1C) INLT to Prevent Thermal Binding ...............................................................................................................38 5.8 Isolation/Restoration of FWS-E1A(B), 1st PT HTR ........................................................38 5.9 Adding Feed Pump Gear Increaser Oil .............................................................................40 6 SYSTEM SHUTDOWN.............................................................................................................41 6.1 Reactor Feed Pump Shutdown ..........................................................................................41 6.2 Removing a FWREG Valve from Service ........................................................................49 SOP-0009 REV - 060 PAGE 2 OF 98

CONTINUOUS USE 6.3 Three Element to Single Element Control Transfer..........................................................50 6.4 Transfer from Master Level Control to Startup Level Control.........................................50 7 REFERENCES ...........................................................................................................................51 8 RECORDS..................................................................................................................................51 ATTACHMENT 1 - VALVE LINEUP - FEEDWATER SYSTEM.................................................52 ATTACHMENT 2A - INSTRUMENT AND VALVE LINEUP - FEEDWATER SYSTEM ...........................................................................................................................68 ATTACHMENT 2B - INSTRUMENT AND VALVE LINEUP - FEEDWATER PUMP AND DRIVE LUBE OIL..................................................................................................74 ATTACHMENT 3 - ELECTRICAL LINEUP - FEEDWATER SYSTEM......................................86 ATTACHMENT 4 - CONTROL BOARD LINEUP - FEEDWATER SYSTEM ............................91 ATTACHMENT 5 - HYDROGEN VENT RIG................................................................................97 ATTACHMENT 6 - FWS-MOV27A, B, C STROKE TEMPERATURE ........................................98 SOP-0009 REV - 060 PAGE 3 OF 98

CONTINUOUS USE 1 PURPOSE 1.1 The purpose of this procedure is to provide instructions for operation of the Feedwater and Feedwater Level Control Systems.

2 PRECAUTIONS AND LIMITATIONS 2.1 The RFP should be tripped if any of the following occur:

2.1.1. Either RFP/Motor or Gear Increaser Lube Oil temperature reaches 190°F 2.1.2. Pump bearing temperature reaches 218°F 2.1.3. Motor bearing temperature reaches 195°F 2.1.4. Gear increaser bearing temperature exceeds 190°F 2.2 If seal parameters exceed the following, then perform an orderly down power per GOP-002 Plant Shutdown, or GOP-0005, Power Maneuvering, to within the capacity of the remaining operating RFPs OR start another pump and secure the affected pump.

• Seal cooling temperature with the RFP running is greater than 190°F

• Seal leakage exceeds a solid pencil stream.

2.3 Anytime FWS-V75, S/U RECIRCULATION MANUAL ISOLATION VALVE is closed FWS-MOV103, LONG CYCLE CLEANUP ISOL and its bypass valve FWS-V90, FEEDWATER SYSTEM S/U RECIRC ISOL VLV BYP should be closed and tagged out. This will prevent overpressurization of the low-pressure feedwater piping downstream of FWS-FV104, START UP RECIRC FCV.

2.4 Following initial startup of the Feedwater Pump Lube Oil System per Section 4.1, if the FWL-P5A(B)(C), GEAR INCR AUX OIL PMP Control Switch is moved from AUTO, the auto start feature of the pump will be deactivated. To reactivate the auto start feature, in addition to returning the Control Switch to AUTO, the RX FWP P1A(B)(C) LUBE OIL SYSTEM START Pushbutton must be depressed.

2.5 Operation with feedpump oil level visible below the add mark in the sight glass, is not recommended. This will not result in immediate machine degradation but does decrease the amount of oil that is being cooled and may decrease the life of internal components.

2.6 Gear increaser oil additions are based on use of the tubular sight glasses. Oil level should be in the lower 1/2 inch of the tubular sight glass. Oil addition is required when oil level is at the bottom of the tubular sight glass.

SOP-0009 REV - 060 PAGE 4 OF 98

CONTINUOUS USE 2.7 FWL-P3A(B)(C), RX FWP A(B)(C) AUX OIL PMP 3A(B)(C) DC oil pumps are provided to protect the feed pump bearings on loss of power to the feed pumps. Feed pumps should not be run with FWL-P3A(B)(C) out of service.

2.7.1. The DC Oil Pump has the capacity to allow the RFP to operate but should only be used if the RFP is required to assure adequate core cooling.

2.8 Normal Feedwater Pump Motor current should be greater than 200 amps. Operation of a Feedwater Pump Motor below 200 amps indicates a minimum flow valve malfunction that could lead to pump damage. If the motor amps are less than 200 amps, then pump flow can be checked using the following computer points CNMFY007, CNMFY008, CNMFY009. Flow should be greater than or equal to 3050 gpm.

2.9 RFP/Motor and Gear Increaser temperature limits are as follows:

2.9.1. Normal Lube Oil: 120°F-150°F 2.9.2. Maximum Lube Oil: 190°F 2.9.3. Maximum Motor Bearing: 185°F 2.9.4. Maximum Pump Bearing: 200°F 2.9.5. Maximum Gear Increaser Bearings: 190°F 2.9.6. Maximum Seal Cooling Temperature with RFP running: 190°F 2.10 Lube Oil pressure limits for RFP start are as follows:

2.10.1. RFP Motor: 10 to 20 psig 2.10.2. RFP Gear Increaser: 15 to 30 psig 2.11 Maximum allowable differential pressure across a RFP filter is as follows:

2.11.1. RFP Lube Oil Filter, FWL-FLT28A/B/C: 5 psig 2.11.2. Gear Increaser Lube Oil Filter, FWL-FLT30A/B/C: 25 psig 2.12 Maximum normal seal cooling water temperature with the RFP running is 140°F.

2.13 RFP Full Load Motor Current should be limited to 311 amps; however, operation above 311 amps has been evaluated and pre-approved as long as amps are limited to less than 350 amps. A Condition Report is required to be generated when amps exceed 311 amps for trending and evaluation of motor long term degradation under this condition.

2.13.1. Reactor power should be maintained as high as possible while allowing for adequate control margin for FWREG position and RFP amps.

SOP-0009 REV - 060 PAGE 5 OF 98

CONTINUOUS USE 2.14 RFP Starts are limited as follows:

2.14.1. After a minimum 60-minute idle period, two consecutive starts are allowed.

2.14.2. After pump has reached normal operating temperature due to running for greater than 15 minutes, one start is permissible.

2.14.3. The motor must run for 15 minutes or stand idle for 60 minutes before additional start attempts can be made.

2.14.4. Avoid starting RFPs during pump coastdown. (Approximately 2 minutes) 2.15 Low Flow Conditions Limitations:

2.15.1. If a RFP is to be secured during a planned shutdown, its min flow valve should be operational.

2.15.2. Isolation of RFP min flow valves is acceptable provided one of the three min flow valves remains in automatic.

2.15.3. If a RFP is to be secured during an unplanned shutdown, the RFP(s) with isolated min flow valve(s) should be secured first.

2.15.4. Avoid sustained reactor feed pump operation under low flow conditions.

Sustained operation with low flow and high vibration leads to accelerated wear of bearings, seals, and rotating components. Options to reduce vibration such as opening the reactor Feed Pump Min Flow Valve or the Long Cycle Cleanup Valve to increase Feed Pump flow and/or lowering Rx power as needed to allow shutting down a Reactor Feed Pump should be considered. It is recommended that the min flow valves not be used during operation below the power range until the reactor is shutdown due to increased sensitivity to level and power excursions that can result in a possible reactor scram.

2.15.5. When manually opening the feed pumps min flow valve(s) for low load conditions to lower vibrations and/or to maintain min amps on pumps, priority should be given to using FWR-FV2B or FWR-FV2C. Using FWR-FV2A results in additional vibration due to differences in min flow piping runs.

2.15.6. Whenever FWS-FV104, START UP RECIRC FCV is open with RWCU in service, feedwater line thermal stratification may occur. Monitor feedwater header temperature per SOP-0110, Tamaris Temperature Scanner, display 33.

2.16 If a RFP trip occurs and a restart is attempted, unusually high differential pressure across the pump discharge valve may prevent the valve from opening.

2.17 The maximum permissible feedwater flow through the startup level control valve is 10%

or 1,312,576 1bm/hr.

SOP-0009 REV - 060 PAGE 6 OF 98

CONTINUOUS USE 2.18 If a RFP loses suction pressure and continues to operate, it should be tripped immediately and a thorough maintenance inspection shall be performed.

2.19 Feedwater system may be partially drained during loss of offsite power due to loss of instrument air. This condition allows various feedwater lines to drain to the condenser and results in the Condensate Pumps operating in excess of design flow rate when restarted. Prior to return to service, the feedwater system must be filled and vented.

2.20 When a RFP has been secured, to reduce RFP discharge pressure and vibration, a Condensate Pump may be secured per SOP-0007.

2.21 FWREG position should be limited to less than or equal to 92% open to allow an adequate margin for valve modulation while maintaining reactor level.

2.21.1. Reactor power should be maintained as high as possible while allowing for adequate control margin for FWREG position and RFP amps.

2.21.2. Starting a RFP with feed reg valves near 92% open can adversely impact reactor water level and feed pump suction pressure because the FWLCS may not be able to compensate when the associated min flow valve comes open.

Opening the feed pump discharge valve further challenges the FWLCS.

Lowering power to approx 85% should be consider to provide control margin for the feed reg valves before the RFP is started.

2.22 C33-LVF002, START UP FWREG VALVE position should be limited to less than or equal to 65% open to allow an adequate margin for valve modulation while maintaining reactor level.

2.23 All controls and indications are on H13-P680 unless otherwise noted.

2.24 Excessive venting with RFP Suction Temperature >150 oF offsets normal seal water flow and can overheat internal seal components. Seals are very sensitive to temperatures above 150 oF. If the seal outlet temperature rises above 140oF (FWS-TI12A, B, C or FWS-TI13A, B, C), seal venting should be stopped as soon as possible after a steady stream of water issues from the vent. IF Reactor Feed Pump has not been isolated from hot, pressurized CNM/FWS, THEN venting is not required.

C 2.25 Leakage through HWC valves allows hydrogen to continue to be introduced into the Feedwater System. This includes idle Reactor Feed Pumps. Sampling may be required to ensure hydrogen pockets do not exist prior to performing feedwater system maintenance.

2.26 C33 and B21 Narrow Range Reactor Water Level instrument channels on different reference legs may indicate varying levels due to local temperatures in the drywell, thermal growth of sensing lines, dynamic effects in the reactor downcomer, etc. The STP-000-0001 channel check criteria account for these sensed pressure differences as well as instrument inaccuracies.

SOP-0009 REV - 060 PAGE 7 OF 98

CONTINUOUS USE 3 PREREQUISITES FOR STARTUP AND OPERATIONS 3.1 Verify the below listed Electrical Systems are in operation:

3.1.1. 125 VDC per SOP-0049, 125 VDC System 3.1.2. 120 VAC per SOP-0048, 120 VAC System 3.1.3. 480 VAC per SOP-0047, 480 VAC System 3.1.4. 13.8 kV per SOP-0045, 13.8 kV System 3.2 Verify Condensate System is in operation per SOP-0007, Condensate System.

3.3 Verify Feedwater System is filled and vented.

3.4 Verify Turbine Plant Component Cooling Water System (CCS) is operating and supplying the RFP Lube Oil Coolers, Gear Increaser Coolers, Seal Water Coolers, and Feed Pump Motor Coolers per SOP-0017, Turbine Plant Component Cooling Water System.

3.5 Verify all reactor RFP/Motor and Gear Increaser oil levels normal.

3.6 Verify Instrument Air System is in operation per SOP-0022, Instrument Air System.

3.7 Verify system is lined up for startup.

SOP-0009 REV - 060 PAGE 8 OF 98

CONTINUOUS USE 5 SYSTEM OPERATION 5.1 Alternating Feedwater Level Control Signals 5.1.1. IF automatically controlling level on the Master Flow controller, THEN alternate the feedwater level control signals as follows:

1. Ensure no deviation on C33-R600, FW REG VALVES MASTER FLOW CONTROLLER and place to MANUAL.

2. Swap the level control input by depressing either A or B on the RX LVL A/B SELECT Pushbutton.

3. Check for proper operation, then return C33-R600, FW REG VALVES MASTER FLOW CONTROLLER to AUTO as follows:

C 1) Adjust tape set 2 inches above actual vessel level and observe the deviation signal is positive.

C 2) Lower tape set 2 inches below actual vessel level and observe the deviation signal is negative.

3) Match tape set to actual vessel level in order to null the deviation signal.

4) WHEN the level signal is nulled, THEN depress the AUTO Pushbutton and check green light above the pushbutton is on.

4. Adjust C33-R600, FW REG VALVES MASTER FLOW CONTROLLER Tape Set to maintain the reactor level requested by the OSM/CRS.

5.1.2. IF automatically controlling level on the Startup FWREG Valve Flow Controller, THEN alternate the feedwater level control signals as follows:

1. Check no deviation on C33-R602, START UP FWREG VALVE FLOW CONTROLLER and place to MANUAL.

2. Swap the level control input by depressing either A or B on the RX LVL A/B SELECT Pushbutton.

SOP-0009 REV - 060 PAGE 33 OF 98

CONTINUOUS USE

3. Check for proper operation, then return C33-R602, START UP FWREG VALVE FLOW CONTROLLER to AUTO as follows:

1) Adjust tape set 2 inches above actual vessel level and observe the deviation signal is positive.

2) Lower tape set 2 inches below actual vessel level and observe the deviation signal is negative.

3) Match tape set to actual vessel level in order to null the deviation signal.

4) Depress the AUTO Pushbutton and check green light above the pushbutton is on.

4. Adjust C33-R602, START UP FWREG VALVE FLOW CONTROLLER Tape Set to maintain the level requested by the OSM/CRS.

5.2 Restoring Reactor Water Level to Normal Following a Reactor Scram from High Power 5.2.1. Check reactor water level on the narrow range is greater than 10 inches and rising.

5.2.2. Reduce the number of running Feedwater Pumps to one.

NOTE Normally one FWREG is left in service. However, if necessary for level control, all three FWREGs may be taken out of service.

1. Reduce the number of in service FWREGs by taking manual control and closing the selected FWREGs and associated isolation valve:

• For C33-LVF001A close FWS-MOV27A, FWREG VLV 1A INLT Valve

• For C33-LVF001B close FWS-MOV27B, FWREG VLV 1B INLT Valve

• For C33-LVF001C close FWS-MOV27C, FWREG VLV 1C INLT Valve 5.2.3. Select 1 ELEM on the SINGLE ELEMENT THREE ELEMENT Select Switch.

5.2.4. WHEN feedwater flow requirements are within the capability of the Startup FWREG, place the Startup FWREG in service per Section 6.4.

SOP-0009 REV - 060 PAGE 34 OF 98

CONTINUOUS USE

• G12.1.7 RIVER BEND STATION STATION OPERATING MANUAL

• ABNORMAL OPERATING PROCEDURE

• CONDENSATE/FEEDWATER FAILURES PROCEDURE NUMBER: *AOP-0006 REVISION NUMBER: *019 Effective Date: *03/10/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER AOP-0006 REV - 019 PAGE 1 OF 8

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE/DISCUSSION ..........................................................................................................3 2 PRECAUTIONS AND LIMITATIONS ....................................................................................3 3 SYMPTOMS ..............................................................................................................................4 4 AUTOMATIC ACTIONS ..........................................................................................................5 5 IMMEDIATE OPERATOR ACTIONS .....................................................................................6 6 SUBSEQUENT OPERATOR ACTIONS ..................................................................................6 7 REFERENCES ...........................................................................................................................8 AOP-0006 REV - 019 PAGE 2 OF 8

CONTINUOUS USE 1 PURPOSE/DISCUSSION 1.1 The purpose of this procedure is to provide instructions for the operator in the event of a Condensate/Feedwater System failure resulting in either rising or lowering reactor water level. This procedure also addresses a total loss of feedwater.

1.2 A rising reactor water level can occur from a feedwater flow control failure either in the level controller circuitry or the feedwater regulating valve. A main turbine trip, and if in the RUN mode, a reactor scram occur if reactor water level reaches Level 8.

1.3 A lowering reactor water level can be caused by a partial or total loss of feedwater flow.

A partial loss of feedwater flow can be caused by a feedwater flow control failure or a trip of a condensate pump, heater drain pump, or feedwater pump. A total loss of feedwater can be caused by a trip of all running feed/condensate pumps. If a total loss of feedwater is experienced, a reactor scram occurs at Level 3.

2 PRECAUTIONS AND LIMITATIONS If all reactor feed pumps trip and a minimum flow path for the condensate pumps is not available, a reactor feed pump must be promptly started for protection of the condensate pumps.

If a reactor feed pump can not be promptly started, optional minimum flow paths for the condensate pumps must be considered.

AOP-0006 REV - 019 PAGE 3 OF 8

CONTINUOUS USE 3 SYMPTOMS 3.1 Rising level

• Feed flow/steam flow mismatch

• Reactor water level rising

• Rising neutron flux 3.2 Lowering level

• Feed flow/steam flow mismatch

• Reactor water level lowering

• Condensate pump auto trip

• Heater drain pump auto trip

• Reactor feed pump auto trip AOP-0006 REV - 019 PAGE 4 OF 8

CONTINUOUS USE 4 AUTOMATIC ACTIONS 4.1 The following occur at reactor water Level 3:

• Reactor scram

• Recirculation pump downshift 4.2 The following occurs at reactor water Level 4:

If less than 3 reactor feed pumps are running, as sensed by reactor feed pump suction flow, a Reactor Recirculation flow control valve runback is initiated.

4.3 The following occur at reactor water Level 8:

• Reactor scram when the mode switch is in RUN

• Main turbine trip

• Reactor feed pump trip 4.4 If feed flow is less than 19.9% of rated after a 15 second time delay, Reactor Recirculation Pumps downshift.

AOP-0006 REV - 019 PAGE 5 OF 8

CONTINUOUS USE 5 IMMEDIATE OPERATOR ACTIONS 5.1 Manually control the feedwater level control system and/or reduce reactor power to mitigate any level transient.

NOTE Steps in the following section may be performed concurrently as appropriate.

6 SUBSEQUENT OPERATOR ACTIONS 6.1 IF reactor water level cannot be maintained greater than or equal to +31 inches, THEN start any of the following equipment to maintain reactor vessel level:

• Condensate pumps per SOP-0007, Condensate System

• Heater drain pumps per SOP-0010, MSR & FW Heaters, Extraction Steam and Drains

• Reactor feed pumps per SOP-0009, Feedwater System 6.2 IF any of the following flow control valves have failed, THEN attempt to take manual control or isolate the flow control valve to control reactor vessel level:

• CNM-FV114, CONDENSATE MIN RECIRC VALVE isolated by closing CNM-V95, CNM-FV114 INLET ISOLATION VALVE

• FWS-FV104, START UP RECIRC FCV isolated by closing FWS-MOV103, LONG CYCLE CLEANUP ISOL

• FWR-FV2A, RX FWP A MIN FLOW VALVE isolated by closing FWR-V1, RFP A MIN FLOW MAN ISOL

• FWR-FV2B, RX FWP B MIN FLOW VALVE isolated by closing FWR-V2, RFP B MIN FLOW MAN ISOL

• FWR-FV2C, RX FWP C MIN FLOW VALVE isolated by closing FWR-V3, RFP C MIN FLOW MAN ISOL

• HDL-FV20A, HTR DR PUMPS P1A(B) RECIRC LINE CONTROL VLV

• HDL-FV20B, HTR DR PUMPS P1C(D) RECIRC LINE CONTROL VLV

• CNM-FCV200, CONDENSATE PREFLT VSL BYPASS FLOW CONTROL VALVE AOP-0006 REV - 019 PAGE 6 OF 8

CONTINUOUS USE 6.3 IF any feedwater level control valve has failed, THEN close the associated isolation valve as follows:

• FWS-MOV27A, FWREG VLV A INLT

• FWS-MOV27B, FWREG VLV B INLT

• FWS-MOV27C, FWREG VLV C INLT 6.4 IF all the following conditions exist:

• All reactor feed pumps trip.

• CNM-FV114, MAIN CONDENSATE RECIRC FLOW CONTROL is not available for automatic operation.

• A reactor feed pump cannot be promptly started.

THEN perform any of the following:

• Place at least one Condensate Demineralizer in RECYCLE per SOP-0093, Condensate Demineralizer System.

• At TB, 67 ft el, TZ-8, JTB-RAK17, verify CNS-LV105, HOTWELL REJECT is in AUTO, at H13-P680, open CNS-LV104, CONDENSER HOTWELL LEVEL CONTROL VALVE.

• Take Manual control of CNM-FV114, MAIN CONDENSATE RECIRC FLOW CONTROL and establish minimum flow.

• Establish long cycle condensate water cleanup per SOP-0007, Condensate System.

NOTE The following step limits backflow of RWCU to the CST via the CRD pump minimum flow lines and CNS-LV105.

C 6.5 IF all reactor feed pumps trip AND RWCU is operating, THEN perform one of the following:

• Promptly start a reactor feed pump.

• Close FWS-MOV7B AOP-0006 REV - 019 PAGE 7 OF 8

CONTINUOUS USE 6.6 Refer To AOP-0007, Loss of Feedwater Heating.

6.7 IF CNM-P1C, CONDENSATE PUMP P1C has tripped, THEN shut down Condensate Oxygen injection per SOP-0123, Hydrogen Water Chemistry H2 and O2 System.

6.8 IF power is lost to the Condensate Pumps AND the possibility of Condensate draining to the Hotwell exists, THEN perform the following:

6.8.1. Close CNM-MOV3A(B)(C), CNDS PUMP 1A(1B)(1C) DISCH.

6.8.2. Monitor Hotwell level.

1. Take actions to isolate Hotwell as required to prevent over filling.

6.8.3. WHEN power is restored to Condensate Pumps, THEN restart as necessary.

6.9 IF all Condensate Pumps are off, THEN perform the following:

6.9.1. Close all the following valves:

• CNM-V3105A, CNM-FLT1A BACKWASH AIR SUPPLY

• CNM-V3105B, CNM-FLT1B BACKWASH AIR SUPPLY

• CNM-V3105C, CNM-FLT1C BACKWASH AIR SUPPLY

• CNM-V3105D, CNM-FLT1D BACKWASH AIR SUPPLY

• CNM-V3105E, CNM-FLT1E BACKWASH AIR SUPPLY 6.9.2. Manually line up CRD pump suction to the CST per SOP-0002.

7 REFERENCES 7.1 USAR Sections

• 5.4.9

• 7.7.1.3

• 10.4.7 7.2 System Design Criteria System Number 107 AOP-0006 REV - 019 PAGE 8 OF 8

RJPM-NRC-M14-S4 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Parallel Offsite Power With Div 1 EDG Supplying ENS-SWG1A OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: September 4, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-S4 Rev. 0 EXAMINER INFO SHEET Task Standard: Offsite power is supplying ENS-SWG1A in parallel with the Div I EDG.

Synopsis: The reactor is at 100% power. Div I EDG is the sole source of power to ENS-SWG1A. This task will parallel offsite power from RSS 1 onto the bus using SOP-0053, Standby Diesel Generator and Auxiliaries, Section 5.1, Paralleling an Offsite Power Source to the Standby Diesel from the Control Room.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to parallel Offsite Power to ENS-SWG1A in accordance with SOP-0053 using the normal supply breaker ACB06.

3) Initial Conditions:

The reactor is at 100% power.

Div I Standby Diesel Generator is supplying ENS-SWG1A.

RSS 1 is energized and ENS-ACB06, Normal Supply Breaker is Open.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-S4 Rev 0 Page 2 of 8

RJPM-NRC-M14-S4 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Parallel Offsite Power With Div 1 EDG 264009001001 264000 A2.01 3.5 / 3.6 Supplying ENS-SWG1A 295003 AA1.02 4.2 /

4.3 REFERENCES

APPLICABLE OBJECTIVES SOP-0053, Rev 325 RLP-STM-0309S, Obj 8 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0053, Standby Diesel Generator __6__

and Auxiliaries (Sim Copy)

SIMULATOR CONDITIONS & SETUP:

1. IC # 213

2. Rx Power: 100%

3. Div 1 EDG supplying ENS-SWG1A.

4. ENS-ACB06 is open.

5.

6.

7.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Offsite power is supplying ENS-SWG1A in parallel with the Div I EDG.

RJPM-NRC-M14-S4 Rev 0 Page 3 of 8

RJPM-NRC-M14-S4 Rev. 0 PERFORMANCE:

START TIME:

SOP-0053 Section 5.1, Paralleling an Offsite Power Source to the Standby Diesel from the Control Room PROCEDURE NOTE The following controls and indications are located on Panel H13-P877 unless otherwise stated.

1. *Procedure Step: 5.1.1 IF ENS-ACB06(26), NORMAL SUPPLY BRKR is to be closed, THEN place the REMOTE SYNCH SW to NORM:

Standard Applicant placed the remote synch switch to NORM.

Cue Notes Results SAT UNSAT Procedure Step: 5.1.2 IF ENS-ACB04(24), ALTERNATE SUPPLY BRKR is to be closed, THEN place the REMOTE SYNCH SW to ALTN:

Standard NA Cue Notes No actions necessary due to initial conditions.

RJPM-NRC-M14-S4 Rev 0 Page 4 of 8

RJPM-NRC-M14-S4 Rev. 0

2. Procedure Step: 5.1.3 Adjust diesel voltage, as observed on V-1RUN-1SYDA(B)01, RUNNING VOLTAGE to approximately 1-2 volts above V-1IN-1SYDA(B)01, INCOMING VOLTAGE using the STBY DIESEL GENERATOR A(B) VOLTAGE REGULATOR CONT.

Standard Applicant located/identified and used the named A voltage meters to raise the Running (EDG) Voltage 1-2 volts higher than the incoming (grid) voltage by adjusting the Div 1 EDG Voltage Regulator Controller.

Cue Notes Results SAT UNSAT

3. *Procedure Step: 5.1.4 Adjust diesel speed, using the STBY DIESEL GENERATOR A(B)

GOVERNOR CONTROL, to bring the frequency within the range of grid frequency. Adjust speed so the SY-1-SYDA(B)01, STBY BUS A(B)

SYNCHROSCOPE indicator is rotating slowly in the SLOW direction (counterclockwise) at a rate of one revolution in greater than or equal to 4 seconds and less than or equal to 6 seconds.

Standard Applicant diesel speed using the governor control to match grid frequency and caused the synchroscope to rotate counterclockwise one revolution every four to six seconds.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S4 Rev 0 Page 5 of 8

RJPM-NRC-M14-S4 Rev. 0 PROCEDURE CAUTION Do not close the Normal or Alternate bus feeder breaker with the synchroscope indicator standing still if the bus is being supplied by the EGS-EG1A(B), STANDBY DIESEL GENERATOR.

When synchronizing the D/G and its connected loads back to offsite power, it is possible for the D/G to unload at a rapid rate as soon as the preferred source breaker is closed. This is due to the governor changing to droop mode. If this occurs, immediately raise the load back to the desired value using the governor control switch.

4. *Procedure Step: 5.1.5 When the synchroscope indicator is moving slowly in the SLOW direction AND the synchroscope indicator is 5 minutes to 2 minutes before the 12 oclock position, THEN close the desired feeder breaker, ENS-ACB06, NORMAL SUPPLY BRKR.

Verify the red breaker closed light comes ON.

If not, return the breaker handswitch to TRIP.

Standard Applicant closed ENS-ACB06 with the proper synchroscope indications.

Applicant verified the red breaker closed light illuminated.

Cue Notes Results SAT UNSAT

5. Procedure Step: 5.1.6 As soon as diesel load has stabilized, return the REMOTE SYNC SW to OFF.

Standard Applicant positioned the Remote Synch Switch to OFF.

Cue Notes Results SAT UNSAT Terminating Cue: Offsite power is supplying ENS-SWG1A in parallel with the Div I EDG.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-S4 Rev 0 Page 6 of 8

RJPM-NRC-M14-S4 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S4 Rev 0 Page 7 of 8

RJPM-NRC-M14-S4 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The reactor is at 100% power.

Div I Standby Diesel Generator is supplying ENS-SWG1A.

RSS 1 is energized and ENS-ACB06, Normal Supply Breaker is Open.

INITIATING CUE:

The CRS has directed you to parallel Offsite Power to ENS-SWG1A in accordance with SOP-0053 using the normal supply breaker ACB06.

RJPM-NRC-M14-S4 Rev 0 Page 8 of 8

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• STANDBY DIESEL GENERATOR AND AUXILIARIES (SYS#309)

PROCEDURE NUMBER: *SOP-0053 REVISION NUMBER: *325 Effective Date: *03/13/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION TRACKING NUMBER DETAILED DESCRIPTION OF CHANGES SOP-0053 REV - 325 PAGE 1 OF 115

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES ......................................................................................................................14 4 SYSTEM STARTUP..................................................................................................................15 4.1 Placing EGS-EG1A(B), STBY DIESEL GENERATOR A(B) in Standby .....................15 4.2 Placing EGS-EG1A(B), STBY DIESEL GENERATOR A(B) in Maintenance Mode .................................................................................................................................19 4.3 Barring/Air Rolling Standby Diesel .................................................................................21 4.4 Warming Up the Diesel Post Maintenance.......................................................................23 4.5 Non-Emergency Starting, Loading and Paralleling the Standby Diesel from the Control Room....................................................................................................................24 4.6 Non-Emergency Starting, Loading and Paralleling the Standby Diesel from the Local Control Panel ..........................................................................................................29 4.7 Manual Start of Standby Diesel with an Automatic Start Signal Present.........................34 5 SYSTEM OPERATION.............................................................................................................38 5.1 Paralleling an Offsite Power Source to the Standby Diesel from the Control Room.......38 5.2 Paralleling an Offsite Power Source to the Standby Diesel from the Local Control Panel..................................................................................................................................40 5.3 Operation from Automatic Start .......................................................................................43 5.4 Unloading Diesel Fuel Oil ................................................................................................44 5.5 Cross-Connecting Air Receivers Within a Single Division..............................................48 5.6 Cross Connecting Air Receivers Across Divisions ..........................................................50 5.7 Operation of Diesel Fuel Offload Berme Drain System Following Oily Contamination of Berme...................................................................................................53 5.8 Verifying Proper Operation of EGF-P1A(B), FUEL OIL TRANSFER PUMP...............54 5.9 Swapping Fuel Oil Transfer Pump EGF-P1A(B) Discharge Strainers.............................54 5.10 Swapping the Fuel Oil Duplex Filters. .............................................................................55 5.11 Swapping EGF-STR3A(B)/ STR3D(E) FUEL OIL STRAINER ....................................56 5.12 Swapping Lube Oil Duplex Filters EGO-FLT1A/1D (1B/1E).........................................56 5.13 Swapping Lube Oil Duplex Strainers ...............................................................................57 5.14 Restoration of the Standby Diesel from a Tripped Condition ..........................................58 SOP-0053 REV - 325 PAGE 2 OF 115

CONTINUOUS USE 5.15 Makeup Water Addition to Standby Diesel Jacket Water ................................................61 5.16 Operation of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) as the only power supply to the bus ....................................................................................................62 5.17 Manual Air Compressor Operation...................................................................................63 5.18 Draining Division I(II) Jacket Water Standpipe for Level Control ..................................64 6 SYSTEM SHUTDOWN.............................................................................................................65 6.1 Shutdown of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) from the Control Room....................................................................................................................65 6.2 Shutdown of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) from the Local Control Panel ..........................................................................................................68 6.3 Emergency Shutdown of the Diesel - Loss of Control Air...............................................71 6.4 Emergency Shutdown of the Diesel - Loss of DC Control Power ...................................72 6.5 Shutdown and Resetting of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) with LOCA/LOP Signal Present.......................................................................................73 7 REFERENCES ...........................................................................................................................75 8 RECORDS..................................................................................................................................77 ATTACHMENT 1A - VALVE LINEUP - STANDBY DIESEL GENERATOR EG1A.................78 ATTACHMENT 1B - VALVE LINEUP - STANDBY DIESEL GENERATOR EG1B .................84 ATTACHMENT 2A - INSTRUMENT & VALVE LINEUP - STANDBY DIESEL GENERATOR EG1A .......................................................................................................90 ATTACHMENT 2B - INSTRUMENT & VALVE LINEUP - STANDBY DIESEL GENERATOR EG1B .......................................................................................................94 ATTACHMENT 3A - ELECTRICAL LINEUP - STANDBY DIESEL GENERATOR EG1A ................................................................................................................................97 ATTACHMENT 3B - ELECTRICAL LINEUP - STANDBY DIESEL GENERATOR EG1B.................................................................................................................................100 ATTACHMENT 4A - CONTROL BOARD LINEUP - STANDBY DIESEL GENERATOR EG1A ................................................................................................................................103 ATTACHMENT 4B - CONTROL BOARD LINEUP - STANDBY DIESEL GENERATOR EG1B.................................................................................................................................107 ATTACHMENT 5 - ENGINE PARAMETERS ...............................................................................111 ATTACHMENT 6 - KW VS KVAR (.8PF) .....................................................................................112 ATTACHMENT 7 - FUEL OIL DAY TANK LEVEL INDICATION ............................................113 ATTACHMENT 8 - FUEL OIL STORAGE TANK LEVEL INDICATION ..................................114 ATTACHMENT 9 - FITTING LOCATION FOR SECTION 6.5.2.................................................115 SOP-0053 REV - 325 PAGE 3 OF 115

CONTINUOUS USE 1 PURPOSE 1.1 The purpose of this procedure is to outline the steps necessary to startup, operate and shutdown EGS-EG1A(B), STBY DIESEL GENERATOR A(B).

2 PRECAUTIONS AND LIMITATIONS 2.1 All instructions are written for EGS-EG1A, STBY DIESEL GENERATOR A with nomenclature for EGS-EG1B, STBY DIESEL GENERATOR B in parenthesis.

2.2 High crankcase pressure indicates the possible existence of an explosive gas mixture. Allow the engine to cool for 15 minutes to allow fumes and vapors to dissipate before removing any engine covers. With the exhaust fan running the atmospheric pressure in the room will be lower and the manometers will read higher than specified.

2.3 Placing the diesel in MAINTENANCE mode requires simultaneous operation of the STBY DIESEL ENGINE MODE switch on H13-P877 and the MAINT MODE SELECT switch on EGS-PNL3A(B). Similarly, to place the diesel back in OPERATIONAL mode requires simultaneous operation of the STBY DIESEL ENGINE MODE switch and the RETURN TO OPERATIONAL pushbutton on EGS-PNL3A(B).

2.4 When the diesel is returned to the OPERATIONAL mode the FIELD FLASHING RELAY READY light located on EGE-CAB01A(B) should be lit. If not, the Exciter Shutdown Relay may have failed to reset.

2.5 EGS-EG1A(B), STBY DIESEL GENERATOR A(B) has a continuous rating of 3130 Kw at 0.8 power factor. Do not operate the diesel generator with a power factor of less than 0.8 when operating in parallel with other sources, and do not exceed 3130 Kw load.

2.6 ERIS computer points will be used to ensure voltage and watt limits are not exceeded.

Frequency will be recorded using the MCR or the Local control room meter; however voltage and watt readings will be obtained from the ERIS points. Voltage and watt meters can be used for adjustments when not at the control band limits.

2.7 Diesel Generator Governor oil level shall be checked Prior to, During and Following any diesel run. Acceptable oil levels are greater than the fill mark during standby conditions and visible in the sightglass while operating.

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CONTINUOUS USE 2.8 Prior to manually starting the diesel for other than emergency conditions, the engine should be barred over 2 revolutions and air rolled with the cylinder cocks open to insure the cylinders are clear unless the start is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the last engine shutdown. Roll the engine per Section 4.3 if required. The origin of any water detected in a cylinder must be determined and any cylinder head that leaks due to a crack shall be replaced.

(Ref. 7.21, 7.23) 2.8.1. Air rolls do not apply to any engine that has been removed from service after a run. However, the engine shall be rolled over with the airstart system at the time it is returned to service.

2.8.2. The OPERABLE engines are not required to be air-rolled if the plant is already in an Action Statement of Technical Specification 3.8.1 and 3.8.2.

2.9 Ensure that the rear air system is available prior to attempting any Barring / Air Rolling of a Diesel generator. The barring device and Air Roll function is supplied from rear air only.

2.10 Parallel the Diesel Generator to the Standby Bus with the synchroscope rotating slowly in the "fast" (clockwise) direction. Do not attempt to close a diesel generator output breaker with the synchroscope indicator standing still, if there is power available to the bus from another source.

2.11 If the diesel is run for one hour or greater, check and drain from the day tank any accumulated water via EGF-V11(V41), DAY TANK TK2A(2B) DRAIN.

2.12 Lube oil must be added only through the fill connection on the sump. Do not overfill the sump.

2.13 If EGS-EG1A(B), STBY DIESEL GENERATOR A(B) is declared inoperable, refer to Technical Specification 3.8.1 and 3.8.2.

2.14 Never have 2 synchroscopes in the same division on at the same time.

2.15 If the diesel generator is paralleled with the standby bus normal or alternate breaker and a LOCA signal occurs, the diesel generator output breaker will open. The diesel generator breaker can not be closed as long as bus voltage is being supplied by the normal or alternate supply and the LOCA signal still exists.

2.16 Sustained operation of the engine at critical speeds of 190, 285, 350 and 415 RPM should be avoided. (Ref. 7.19) 2.17 If a diesel start signal is activated while the diesel is not available, the signal will remain sealed in. If the diesel is then made available, the diesel engine will auto start. To prevent this, if Control air pressure is greater than 45 psig then Section 6.5.2 must be performed. If pressure is less than 45 psig, then the EMERGENCY START RESET switch on H13-P877 must be depressed before returning the diesel to Operational.

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CONTINUOUS USE 2.18 During a Station Blackout with the Div. 1 or 2 Diesel Generator failing to deliver power to their respective buses due to a malfunction of the Excitation System, (when diesel engine has attained rated speed) the Field Flashing of the failed D/G should be secured to conserve the battery, and to prevent heating the excitation cabinet.

2.19 Short duration runs and light load (less than 40%, or 1200 kW) operation should be avoided.

After a period of light load or no-load run, the diesel should be loaded to greater than or equal to 2700 kW, for a time period as specified below:

2.19.1. At least one hour, if the engine was run at less than 1200 kW for greater than 30 minutes but less than one hour, OR 2.19.2. At least two hours, if the engine was run at less than 1200 kW for equal to or greater than one hour but less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, OR 2.19.3. At least four hours, if the engine was run at less than 1200 kW for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or longer. (Ref. 7.22; CR-RBS-2004-3156) 2.20 To minimize crankshaft torsional stresses, continuous engine operation at critical speeds shall not be allowed. Minimize the time the engine is operated between 453 and 457 RPM (60.4 to 60.9 Hz). (Ref. 7.19) 2.21 Engine cylinder exhaust gas temperature should be within 75°F of the average for all cylinders. Any cylinder temperature exceeding this limit should be investigated by maintenance.

2.22 Prelube of the engine should be performed before all non-emergency starts.

2.23 Per System Engineering the following conditions should be used to determine if a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is required to allow the engine mass and crankshaft temperatures to equalize, prior to any normal Diesel Generator start:

2.23.1. When after re-energizing the Lube Oil or Jacket Water Heaters from a de-energized state, the Lube Oil and Jacket Water outlet temperatures are greater than 140°F and 115°F respectively with less than or equal to a 40°F differential a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is not required prior to any normal start. This temperature criteria should be used for Diesel Generator outages of 3 days or less.

2.23.2. If the engine block has been allowed to cool to ambient temperature, such as for maintenance with the heaters de-energized for more than 3 days, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is required with the Lube Oil Circulating Pump, Jacket Water Circulating Pump, and associated heaters operating prior to any normal start.

2.23.3. If necessary contact the System Engineer for guidance in determination of temperature criteria.

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CONTINUOUS USE 2.23.4. Use of Section 4.4, Warming Up the Diesel Post Maintenance should be minimized. Warming up by slow start and controlled slow loading is intended for situations where insufficient time is available to allow the heater to warm the system, and an expedited return to service is critical due to plant conditions or shutdown LCO status. Although EC 5759 was written for schedule preservation, and the method poses very little risk (maybe slightly more wear and tear on the Diesel over the long term), this method should not be made a matter of routine.

Operations management or Duty Manager should approve use of Section 4.4.

2.24 Anytime work is done on the fuel oil day tank level instrumentation, the control switch for the fuel oil transfer pump must be placed in "OFF" to preclude pumping fuel oil to the roof.

2.25 If the forward starting subsystem DC control power is lost, the diesel engine will still be able to start, but there will be NO tripping capability.

2.26 If the diesel is started automatically on a LOCA, all automatic shutdowns are bypassed except overspeed and generator differential. The reinstatement of all trips following an automatic start requires the following:

2.26.1. DIV 2 -Depress the RHR DIV 2 INITIATION RESET pushbutton (H13-P601 INSERT 17B).

DIV 1 - Depress the LPCS/RHR DIV 1 INITIATION RESET pushbutton (H13-P601 INSERT 21B) 2.27 If the diesel is started automatically on a Loss of Power (LOP) or manually started using either of the STBY DIESEL ENGINE EMERGENCY START pushbuttons, all automatic shutdowns are bypassed except overspeed, generator differential, jacket water out high temperature, and lube oil out high temperature. The reinstatement of all trips following an above described start requires the following:

2.27.1. LOP - Depress STBY DIESEL ENGINE A(B) EMERGENCY START RESET pushbutton.

2.27.2. REMOTE MANUAL EMERGENCY START(CR) - Depress STBY DIESEL ENGINE A(B) EMERGENCY START RESET.

2.27.3. LOCAL PANEL MANUAL EMERGENCY PB - Diesel must be shutdown.

2.27.4. (LOCAL, EGS-PNL4A) - for LOP 2.28 The EXCITATION SHUTDOWN RESET pushbutton should only be used on a loss of excitation with the engine still running. The excitation shutdown will normally auto-reset when the engine is stopped.

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CONTINUOUS USE 2.29 The FIELD FLASHING RELAY READY white light will be energized when the K1 relay is in the reset (closed) position and the DG is not at voltage. It will deenergize when K1 is in the exciter shutdown position, the voltage relay contacts open (DG at voltage), or when the pressure switches are closed (DG at speed).

2.30 The FIELD FLASHING RELAY READY white light should be verified to be energized after any Exciter Shutdown Reset operation or any time the DG is placed in a Standby lineup.

2.31 Operating data pertaining to all diesel generator start attempts shall be obtained per PEP-0026, Diesel Generator Operating Logs.

2.32 Visual daily inspection between adjacent cylinder heads and the general block top are required during any period of continuous operation following automatic diesel generator startup. (L/C 3.3).

2.33 Whenever the Diesel is being shutdown, adjust Generator frequency to 59.7 Hz (Div I DG) or 60 Hz (Div II DG) after the Generator output breaker has been opened, prior to stopping the engine. (Ref. 7.19) 2.34 Before the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> air roll after an engine run, drain any liquid from the Turbo Charger Casing Drain per PEP-0026.

2.35 When the diesel is running in parallel with the grid, a fault on the grid could cause a loss of the bus associated with the diesel concurrent with a trip/lockout of the diesel. To reduce the chances of this occurring, time spent with the diesel paralleled to the grid should be minimized. (Ref. 7.11) 2.36 Duplex lube oil and fuel oil filters and strainers should be swapped while the engine is running, if at all possible. It may be loaded or unloaded, isochronous or synchronous with the grid. System pressures should be checked after the swap. If it is necessary to swap a duplex filter or strainer while shut down, the engine should be started in test mode (normal start) and run long enough to check that pressures are normal.

2.37 During diesel fuel oil unloading, a fire watch shall be stationed at the unloading area and two (2) 150 pound dry chemical extinguishers placed near unloading area. (Ref. 7.17) 2.38 EGS-EG1A(B), STBY DIESEL GENERATOR A(B) shall not be run in parallel with the Main Generator through STX-XNS1C, NORM STA XFMR. (Ref. 7.15) 2.39 Failing to de-energize control power to the K1 relay prior to depressurizing D/G control air will result in the inability of the K1 relay to auto reset when control air is restored. The K1 relay must be manually reset if this condition occurs. Control air should be restored prior to reenergizing control power to the K1 relay. (Ref. 7.36) 2.40 If desired, EGS-EG1A(B), STBY DIESEL GENERATOR A(B) may be started and run using only one air receiver tank. (Ref. TSI-015)

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CONTINUOUS USE 2.41 EGA-C4A (C5A)(C4B)(C5B) will not operate in AUTO if its associated start air receiver pressure is less than 30 psig. Placing the control switch to RUN will allow the compressor to start. This should only be used for initial startups and repressurizing. DO NOT repressurize the air receiver should pressure fall to less than 25 psig while the diesel generator is loaded, this could cause a trip or uncontrolled loading of the diesel generator.

2.42 When operating in the NORMAL mode, if a TRIP annunciator(s) should come in, and the diesel does not trip, immediately check the amber UNIT TRIPPED light. If the UNIT TRIPPED light is ON, STOP the diesel. If the light is OFF, evaluate the annunciator(s) via other instrumentation. If the trip condition does NOT exist or can not be verified, attempt to RESET the annunciators(s). If the alarm can not be reset and the diesel has run for 2 minutes, manually stop the diesel. (Ref. 7.12) 2.43 The diesel generator will continue to run without control air pressure in emergency conditions. Upon complete loss of Control air pressure, it should not be restored until the diesel is shutdown.

2.44 If the diesel is paralleled with the grid and a Ground Fault Trip/Lockout occurs, the diesel will not Auto start on an Emergency Auto signal. Refer to Section 4.7. (Ref. 7.13) 2.45 When operating the diesel generator at reduced loads, care should be exercised to avoid reverse power trips.

2.46 While in MODES 1, 2 & 3 placing an OPERABLE Emergency Diesel Generator into

'MAINTENANCE' mode causes the diesel to be INOPERABLE. STP-000-0102, Power Distribution Alignment Check shall be performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> unless the diesel is restored to OPERABLE status in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (Ref. 7.29) 2.47 During movement of recently irradiated fuel assemblies in the Primary Containment or Fuel Building AND while in Modes 4 and 5, placing an operable Diesel Generator into MAINTENANCE mode causes the Diesel Generator to be inoperable. Tech Spec 3.8.2, AC Sources - Shutdown, shall be immediately referred to and the action requirements complied with. (Ref. 7.29) 2.48 Non-essential 125 VDC controls are fed from the BYS batteries. The BYS battery chargers are lost on a LOCA or LOP. Use AOP-0014, Loss of 125 VDC for what is lost if the BYS batteries are not available. Safety-related control functions are not affected.

2.49 Non-essential 120 VAC controls are fed from SCA-PNL15A1(B1). In the event of a LOP, EGS-TI64A(B), multi-point (Doric) temperature indicator will be lost. Therefore upon loss of this indicator the operator should use local thermometer readings. Refer to Attachment 5, Engine Parameters for alternate indications. Safety-related control functions are not affected.

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CONTINUOUS USE 2.50 Loss of FORWARD DC power prevents the following indicator lights from coming on:

• UNIT AVAIL EMERGENCY STATUS

• DC CONTROL POWER ON

• UNIT TRIPPED

• READY TO LOAD 2.51 The electric signal that actuates the Maintenance mode is momentary. MAINTENANCE mode is retained by a self-sealing pilot on pneumatic control valve EGS-PNL3A(B)-P2.

There is no electrical seal-in. The RETURN TO OPERATIONAL signal energizes a solenoid valve, which opens a vent path to break this pneumatic seal-in, and P-2 defaults to the OPERATIONAL position. If while in MAINTENANCE mode, the control panel is depressurized, the P-2 self seal-in is lost, and the control system will come up in the OPERATIONAL mode when pressure is restored.

2.52 While running in the test mode, any manual EMERGENCY START signal will activate the governor and voltage regulator pre-position circuits which return the frequency to near 59.7 Hz (Div I DG) or 60 Hz (Div II DG) and voltage to near 4160 v. The output breaker is not signaled to trip. If the DG happens to be synchronous and loaded, the net affect will be a loss of kw and kvar, over a 6 to 10 second period.

2.53 The Diesel Engine oil sump level dip stick indications are as follows:

STANDBY (Diesel Generator not running with keep warm oil pumps running):

• Maintain the oil sump level greater than or equal to the T7 Mark per Tech Spec 3.8.3 and less than or equal to the FULL Mark.

• As long as the oil level is maintained greater than the LOW STBY mark, the Diesel is capable of safely starting and running if required for an emergency.

RUNNING:

• Maintain oil sump level greater than or equal to the LOW RUN Mark.

• The T6 and T7 Marks are not used for oil sump level when the diesel is running.

C 2.54 Special requirements for restoring from Diesel Engine Maintenance/Tagouts:

• If the engine fuel oil system has been tagged out and drained for maintenance, the fuel oil lines shall be refilled by manually operating the DC Fuel Oil Pump for 1 to 2 minutes. This should be done promptly after releasing the tagout to ensure the lines are full before any diesel starts.

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CONTINUOUS USE 2.55 When ENS-SWG1A or B is deenergized such as during a bus outage, the undervoltage relays generate a Loss of Power (LOP) signal to start the associated Standby Diesel Generator. If the diesel is tagged out, the LOP signal will seal in and can go undetected until the tagout is cleared. Therefore, to prevent an auto start of the diesel upon diesel restoration, prior to restoring the diesel, relays R3A and R3B in the side panels at EGS-PNL3A(B) should be checked to ensure that the LOP start is not present. The relay is tripped if the red button in the middle of the relay is not flush (recessed) with the case.

Refer to Section 6.5 of this procedure to reset the relays. (Ref. 7.37) 2.56 When reviewing this procedure for pending operations or system configuration realignments, ensure vulnerabilities to common cause and common mode failures are evaluated for current plant conditions to protect safety sources and safety trains. (SOER 03-1 Recommendation 2 Emergency Power Reliability)(Ref. 7.41) 2.57 When the Diesel is operating synchronized to the grid, the diesel generator shall be declared inoperable. This is because if a Loss Offsite Power were to occur during operations when synchronized to the grid the resultant operations with the diesel powering the Div I(II) bus will cause the diesel frequency to be outside TS 3.8.1.2 and 3.8.1.7 frequency limits.

(Ref 7.39)

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CONTINUOUS USE 2.58 Starting 4.16kV and certain 480VAC loads while the DG is parallel to off-site power can result in the diesel output breaker tripping on overload condition (Ref. 7.42).

2.58.1. To prevent exceeding the maximum load rating of 3130kW when EGS-EG1A is paralleled to the off-site power supply, manual start of equipment on the following switchgears should not be permitted:

• ENS-SWG1A

• EJS-SWG1A

• EJS-SWG2A Additionally if NNS-SWG1A is being powered from RTX-XSR1C, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1A

• NNS-SWG4A (and NNS-SWG4B if cross-tied)

Additionally if NNS-SWG1C is being powered from NNS-SWG1A AND NNS-SWG1A is being powered from RTX-XSR1C, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1C

• E22-S004 Auto start of 4.16kV loads during parallel load testing may result in overload of EGS-EG1A or trip of ENS-ACB07, STBY D/G A OUTPUT BRKR.

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CONTINUOUS USE 2.58.2. To prevent exceeding the maximum load rating of 3130kW when EGS-EG1B is paralleled to the off-site power supply, manual start of equipment on the following switchgears should not be permitted:

• ENS-SWG1B

• EJS-SWG1B

• EJS-SWG2B Additionally if NNS-SWG1B is being powered from RTX-XSR1D, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1B

• NNS-SWG4B (and NNS-SWG4A if cross-tied)

Additionally if NNS-SWG1C is being powered from NNS-SWG1B AND NNS-SWG1B is being powered from RTX-XSR1D, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1C

• E22-S004 Auto start of 4.16kV loads during parallel load testing may result in overload of EGS-EG1B or trip of ENS-ACB27, STBY D/G B OUTPUT BRKR.

2.59 The GERB viscous damper is not required for operability of the Div I or Div II Diesel Generator. Issues with the GERB should be identified and reported via the Condition Report process. (Ref. 7.43) 2.60 Annunciator, P877-31A(32A)-C03, ENS*SWG1A(B) SPLY OR DIST BRKR INOPERATIVE may alarm momentarily whenever ENS-ACB07(27), STBY D/G A(B)

OUTPUT BREAKER is manipulated.

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CONTINUOUS USE 3 PREREQUISITES 3.1 The Fire Protection Water System to the Standby Diesel Generator EGS-EG1A(B) Room is in service per SOP-0037, Fire Protection Water System Operating Procedure.

3.2 The Makeup Water System is available for makeup to the Jacket Water Standpipe per SOP-0099, Makeup Water System.

3.3 The Normal Service Water System is operating per SOP-0018, Normal Service Water.

3.4 The Standby Service Water System is operable per SOP-0042, Standby Service Water System.

3.5 The following electrical systems are operable:

3.5.1. 4160VAC per SOP-0046, 4.16 KV System (except on loss of power start) 3.5.2. 480VAC per SOP-0047, 480 VAC System 3.5.3. 120VAC per SOP-0048, 120 VAC System 3.5.4. 125VDC per SOP-0049, 125 VDC System 3.6 Diesel Generator Building HVAC in operation per SOP-0061, Diesel Generator Building Ventilation.

3.7 Obtain copy of PEP-0026, Diesel Generator Operating Logs for use in all start attempts.

3.8 The Instrument Air System is operable per SOP-0022, Instrument Air System.

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CONTINUOUS USE 5 SYSTEM OPERATION 5.1 Paralleling an Offsite Power Source to the Standby Diesel from the Control Room NOTE The following controls and indications are located on Panel H13-P877 unless otherwise stated.

5.1.1. IF ENS-ACB06(26), NORMAL SUPPLY BRKR is to be closed, THEN place the REMOTE SYNC SW to NORM.

5.1.2. IF ENS-ACB04(24), ALTERNATE SUPPLY BRKR is to be closed, THEN place the REMOTE SYNC SW to ALTN.

5.1.3. Adjust diesel voltage, as observed on V-1RUN-1SYDA(B)01, RUNNING VOLTAGE to approximately 1- 2 volts above V-1IN-1SYDA(B)01, INCOMING VOLTAGE using the STBY DIESEL GENERATOR A(B) VOLTAGE REGULATOR CONT.

5.1.4. Adjust diesel speed, using the STBY DIESEL GENERATOR A(B) GOVERNOR CONTROL, to bring the frequency within the range of grid frequency. Adjust speed so the SY-1-SYDA(B)01, STBY BUS A(B) SYNCHROSCOPE indicator is rotating slowly in the SLOW direction (counterclockwise) at a rate of one revolution in greater than or equal to 4 seconds and less than or equal to 6 seconds.

CAUTION Do not close the Normal or Alternate bus feeder breaker with the synchroscope indicator standing still if the bus is being supplied by the EGS-EG1A(B), STANDBY DIESEL GENERATOR.

When synchronizing the D/G and its connected loads back to offsite power, it is possible for the D/G to unload at a rapid rate as soon as the preferred source breaker is closed. This is due to the governor changing to the droop mode. If this occurs, immediately raise the load back to the desired value using the governor control switch.

CRITICAL STEP 5.1.5. WHEN the synchroscope indicator is moving slowly in the SLOW direction AND the synchroscope indicator is 5 minutes to 2 minutes before the 12 o'clock position, THEN close the desired feeder breaker, ENS-ACB06(26), NORMAL SUPPLY BRKR or ENS-ACB04(24), ALTERNATE SUPPLY BRKR. Verify the red breaker closed light comes ON. If not, return the breaker handswitch to TRIP.

SOP-0053 REV - 325 PAGE 38 OF 115

CONTINUOUS USE 5.1.6. As soon as diesel load has stabilized, return the REMOTE SYNC SW to OFF.

5.1.7. WHEN the diesel is operating synchronized to the grid, THEN declare the Diesel Generator inoperable. (Ref 7.39) 5.1.8. IF desired to load the Diesel Generator, THEN perform the following:

NOTES When raising load, lead with load and follow with VARS.

Generator loading should be done in greater than 150 seconds to minimize mechanical stress and wear on the diesel generator.

The following table is provided as a recommendation for loading limitation. KW and KVAR values listed are approximate values only.

While D/G load is at 2000 KW, PEP-0026 data collection is required.

STBY DIESEL GEN OUTPUT HIGH may alarm briefly while maintaining KW in the proper band due to differences in meter and instrument calibration.

ERIS point EGSEY007(EGSEY005) should be used to verify 3100 KW is not exceeded.

LOADING TIME 0 KW to 1000 KW in 60 to 90 Seconds Then < 600 KVAR Operate at 1000 KW and < 600 KVAR for 60 to 90 Seconds 1000 KW to 2000 KW in 60 to 90 Seconds Then < 1200 KVAR Operate at 2000 KW and < 1200 KVAR for 60 to 90 Seconds or until required PEP-0026 data is collected 2000 KW to 3100 KW in 60 to 90 Seconds Then < 1800 KVAR

1. Raise diesel generator load with the GOVERNOR CONTROL and adjust VARS using the VOLTAGE REGULATOR CONTROL. Use Attachment 6, KW vs KVAR (.8PF) as a guide to verify the generator is not operated at less than a 0.8 power factor.

SOP-0053 REV - 325 PAGE 39 OF 115

CONTINUOUS USE

2. WHEN diesel generator load is 2000 KW, THEN record required PEP-0026 data.

5.1.9. WHEN desired, THEN shutdown the Standby Diesel Generator per Section 6.1 or 6.2 of this procedure.

5.2 Paralleling an Offsite Power Source to the Standby Diesel from the Local Control Panel NOTE The following controls and indications are on Panels EGS-PNL01A(B) and EGS-PNL03A(B) unless otherwise stated.

5.2.1. Verify the blue REMOTE SYNCHRONIZING SELECTOR SWITCH OFF light is on.

5.2.2. IF the NORMAL SPLY TO STBY BUS ENS-SWG1A(B) is to be closed, THEN place the SYNCHRONIZING CONTROL to NORM.

5.2.3. IF the ALTERNATE SPLY TO STBY BUS ENS-SWG1A(B) is to be closed, THEN place the SYNCHRONIZING CONTROL to ALTN.

5.2.4. Select the phase of bus voltage to be monitor on the BUS VOLT voltmeter.

5.2.5. Adjust diesel generator voltage, as observed on RUNNING VOLTS, to approximately 1-2 volts above INCOMING VOLTS using the VOLTAGE REGULATOR CONTROL.

5.2.6. Adjust diesel generator speed to bring the frequency within the range of grid frequency using the GOVERNOR CONTROL. Adjust speed so the SYNCHROSCOPE indicator is rotating slowly in the SLOW direction (counterclockwise) at a rate of one revolution in greater than or equal to 4 seconds and less than or equal to 6 seconds.

SOP-0053 REV - 325 PAGE 40 OF 115

RJPM-NRC-M14-S5 Rev. 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Initiate Standby Liquid Control OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 5 Actual Time (min):

JPM RESULTS*: (Circle one) SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: September 4, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-S5 Rev. 0 EXAMINER INFO SHEET TASK Standby Liquid Control (SLC) Pump B injecting to the RPV and SLC pump A STANDARD: secured.

SYNOPSIS: The plant has experienced an ATWS and requires injection with SLC. This task will have the applicant attempt to align SLC pump A injecting to the RPV using the OSP-0053 Hard Card, but due to a failure of the A Suction Valve to Open, the Applicant will be required to inject with SLC B instead.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to inject with SLC A.

3) Initial Conditions:

A plant transient has occurred resulting in an Anticipated Transient Without Scram (ATWS) and entry into the Emergency Operating Procedures. EOP-0001a directs injecting with Standby Liquid Control.

RJPM-NRC-M14-S5 Rev 0 Page 2 of 8

RJPM-NRC-M14-S5 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Initiate Standby Liquid Control 222007001001 223001 A2.01 4.3/4.4 223001 A2.13 3.3/

3.4 REFERENCES

APPLICABLE OBJECTIVES:

OSP-0053, Rev 034, Attachment 13 RLP-STM-0403 Objective 3, 4, 6 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0059, Rev 034, Section 5.14 __5__

(Simulator copy)

SIMULATOR CONDITIONS & SETUP:

1. IC # 215

2. Required Power: Post-ATWS,with RPV level stable between Level 1 and 2

3. Enclosure 16 inserted to allow applicant to open IAS-MOV106

4. Enclosure 24 inserted

5. Malfunction: SLC A pump suction valve fails to open.

6.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Standby Liquid Control (SLC) Pump B injecting to the RPV and SLC pump A secured.

RJPM-NRC-M14-S5 Rev 0 Page 3 of 8

RJPM-NRC-M14-S5 Rev. 0 PERFORMANCE:

START TIME:

OSP-0053, Attachment 13 Hard Card.

1. *Procedure Step: 1. Place SLC PUMP A(B) (NOT BOTH), control switch to RUN.

Standard Applicant obtained a key, located/identified and manipulated the SLC Pump A keylock switch to RUN.

Cue Notes Results SAT UNSAT

2. *Procedure Step: 2. Perform the following:

Verify the following:

1. SQUIB CONTINUITYA(B), light goes Off

2. C41-F001A(B), SLC PUMP A(B) SUCT VLV, Opens.

3. C41-C001A(B) SLC PUMP A(B), Starts Standard Applicant located and verified that the A Squib Continuity light is OFF Applicant recognized that the A Pump Suction Valve did not open.

Cue As the CRS, acknowledge the report of the SLC Pump A failure.

Notes The second bullet of Step 2 will direct the applicant to the Alternate Path.

Results SAT UNSAT RJPM-NRC-M14-S5 Rev 0 Page 4 of 8

RJPM-NRC-M14-S5 Rev. 0 ALTERNATE PATH:

3. Procedure Step: 2. Perform the following:

IF any required actions do not occur, THEN perform the following:

1. Place SLC PUMP A control switch to STOP.

2. Repeat steps 1 and 2 for the Alternate pump.

Standard Applicant manipulated the SLC Pump A keylock switch to STOP.

Cue Notes Results SAT UNSAT

4. *Procedure Step: 1. Place SLC PUMP B, control switch to RUN.

Standard Applicant obtained a key, located/identified and manipulated the SLC Pump B keylock switch to RUN.

Cue Notes Results SAT UNSAT

5. 2. Perform the following:

• Procedure Step:

Verify the following:

1. SQUIB CONTINUITY B, light goes Off

2. C41-F001B, SLC PUMP B SUCT VLV, Opens.

3. C41-C001B SLC PUMP B, Starts Standard Applicant located and verified that the B Squib Continuity light is OFF Applicant located and verified that the B Pump Suction Valve opened Applicant located and verified that SLC Pump B started..

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S5 Rev 0 Page 5 of 8

RJPM-NRC-M14-S5 Rev. 0

6. Procedure Step: 3. Notify CRS of SLC injection status.

Standard NA Cue As the CRS, acknowledge the report of the SLC Pump B running.

Notes

7. 4. Verify IAS-MOV106 is Open. (Enclosure 16 may be required)

• Procedure Step:

Standard Applicant located/identified and opened IAS-MOV106 by placing the control switch in the OPEN position and by checking the RED indicating light ON and GREEN indicating light OFF.

Cue As CRS, inform the applicant that Enclosure 16 is installed Notes Results SAT UNSAT

8. Procedure Step: 5. Record SLC Tank Level in gallons.

Standard Applicant located/identified the meter for SLC Tank level and wrote the number down (or circled the gallons on the back of the Hard Card).

Cue .

Notes Results SAT UNSAT Terminating Cue Standby Liquid Control (SLC) Pump B injecting to the RPV and SLC pump A secured STOP TIME:

RJPM-NRC-M14-S5 Rev 0 Page 6 of 8

RJPM-NRC-M14-S5 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S5 Rev 0 Page 7 of 8

RJPM-NRC-M14-S5 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

A plant transient has occurred resulting in an Anticipated Transient Without Scram (ATWS) and entry into the Emergency Operating Procedures. EOP-0001a directs injecting with Standby Liquid Control.

INITIATING CUE:

The CRS has directed you to inject with SLC A.

RJPM-NRC-M14-S5 Rev 0 Page 8 of 8

CONTINUOUS USE ATTACHMENT 13 PAGE 1 OF 2 INITIATING STANDBY LIQUID CONTROL

1. Place SLC PUMP A(B) (NOT BOTH), control switch to RUN.

2. Perform the following:

Verify the following:

1. SQUIB CONTINUITY A(B), light goes Off.

2. C41-F001A(B), SLC PUMP A(B) SUCT VLV, Opens.

3. C41-C001A(B), SLC PUMP A(B), Starts.

IF any required actions do not occur, THEN perform the following:

1. Place SLC PUMP A(B), control switch to STOP.

2. Repeat steps 1 and 2 for the Alternate pump.

3. Notify CRS of SLC injection status.

4. Verify IAS-MOV106 is Open. (Enclosure 16 may be required).

5. Record SLC Tank Level. gallons _____________

6. Monitor SLC Tank Level and report to CRS when Hot Shutdown Boron Weight is achieved.

7. WHEN SLC injection is no longer required, THEN place SLC PUMP A (B) control switch to STOP.

OSP-0053 REV - 017 PAGE 44 OF 62

CONTINUOUS USE ATTACHMENT 13 PAGE 2 OF 2 INITIATING STANDBY LIQUID CONTROL STANDBY LIQUID CONTROL INJECTION REQUIREMENTS TANK LEVEL PRIOR TANK LEVEL AFTER TANK LEVEL AFTER TO INJECTION INJECTION OF 69 lb Boron INJECTION OF 166 lb Boron (approximately 16 min inj time) (approximately 38 min inj time)

GAL GAL GAL NOTE WHEN tank level falls between values, THEN the smaller value should be used.

1531 905 0 1550 924 19 1600 974 69 1700 1074 169 1800 1174 269 1900 1274 369 2000 1374 469 2100 1474 569 2200 1574 669 2300 1674 769 2400 1774 869 2500 1874 969 2600 1974 1069 2700 2074 1169 2800 2174 1269 2900 2274 1369 3000 2374 1469 3100 2474 1569 3200 2574 1669 3300 2674 1769 3400 2774 1869 3500 2874 1969 3600 2974 2069 3700 3074 2169 3800 3174 2269 3900 3274 2369 4000 3374 2469 4100 3474 2569 4200 3574 2669 OSP-0053 REV - 017 PAGE 45 OF 62

RJPM-NRC-M14-S6 Rev. 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Emergency Operation of Containment Coolers with Service Water OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 10 Actual Time (min):

JPM RESULTS*: (Circle one) SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

X Perform Plant Simulate X Simulator Control Room Prepared: Dave Bergstrom Date: August 29, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 10

RJPM-NRC-M14-S6 Rev. 0 EXAMINER INFO SHEET TASK Both HVR-UC1A and UC1B running with service water providing cooling and all STANDARD: HVN valves isolated SYNOPSIS: This task will align service water as the cooling medium to the containment unit coolers and start a second safety related unit cooler.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM. When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to conduct Emergency Operation of Containment Unit Coolers with Service Water in accordance with SOP-0059 Section 5.14 for BOTH Containment Unit Coolers

3) Initial Conditions:

A plant transient has occurred resulting in entry into the Emergency Operating Procedures. EOP-0002 directs maximizing containment cooling. The containment unit coolers have no cooling water flow due to the isolation of HVN.

RPV Level is stable between Level 1 & Level 2.

RJPM-NRC-M14-S6 Rev 0 Page 2 of 10

RJPM-NRC-M14-S6 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Emergency Operation of Containment 222007001001 223001 A2.01 4.3/4.4 Coolers with Service Water 223001 A2.13 3.3/

3.4 REFERENCES

APPLICABLE OBJECTIVES:

SOP-0059, Rev 034, Section 5.14 RLP-STM-0403 Objective 3, 4, 6 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0059, Rev 034, Section 5.14 __5__

(Simulator copy)

SIMULATOR CONDITIONS & SETUP:

1. IC # 215

2. Required Power: ATWS, with RPV level stable between Level 1 and Level 2

3. Reactor water level should be stable such that HVN has isolated (except for one failed valve to support alternate path) but service water has not automatically valved in requiring manual action per SOP-0059 Section 5.14

4. HVN-MOV129 should have failed to isolate on Level 2 but be capable of manual isolation for alternate path portion of JPM.

5. Override: HVN-MOV129P 100%

6. T1 LO_HVN-MOV129-G override ON

7. T1 LO_HVN-MOV129-R override OFF delay 30 seconds

8. Event T1 - zdi2(308)

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Both HVR-UC1A and UC1B running with service water providing cooling and all HVN valves isolated.

RJPM-NRC-M14-S6 Rev 0 Page 3 of 10

RJPM-NRC-M14-S6 Rev. 0 PERFORMANCE:

START TIME:

PROCEDURE NOTE:

This section is only to be used in emergencies. If it is desired to use Service Water at other times, refer to SOP-0116, Turbine and Radwaste Building HVAC Chilled Water System.

1. Procedure Step: 5.14.1 Verify the following valves are closed

• HVN-MOV127, CHW SPLY OUTBD ISOL Standard Applicant located/identified and verified HVN-MOV127 closed by checking the GREEN indicating light ON and RED indicating light OFF Cue Notes Results SAT UNSAT

2. Procedure Step: 5.14.1 Verify the following valves are closed

• HVN-MOV128, CHW RTN OUTBD ISOL Standard Applicant located/identified and verified HVN-MOV128 closed by checking the GREEN indicating light ON and RED indicating light OFF.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-S6 Rev 0 Page 4 of 10

RJPM-NRC-M14-S6 Rev. 0

3. *Procedure Step: 5.14.1 Verify the following valves are closed

• HVN-MOV129, CHW SPLY SHUTOFF VLV Standard ALTERNATE PATH:

Applicant located/identified and recognized HVN-MOV129 failed to isolate on Level 2 by checking the RED indicating light ON and GREEN indicating light OFF.

Applicant took action to manually isolate HVN-MOV129 by placing the control switch in the CLOSE position and verifying the GREEN indicating light ON and RED indicating light OFF.

Cue As CRS accept report that HVN-MOV129 has failed to isolate.

Notes Results SAT UNSAT

4. Procedure Step: 5.14.1 Verify the following valves are closed

• HVN-MOV130, CHW RTN SHUTOFF VLV Standard Applicant located/identified and verified HVN-MOV130 closed by checking the GREEN indicating light ON and RED indicating light OFF Cue Notes Results SAT UNSAT RJPM-NRC-M14-S6 Rev 0 Page 5 of 10

RJPM-NRC-M14-S6 Rev. 0

5. Procedure Step: 5.14.1 Verify the following valves are closed

• HVN-MOV102, CHW RTN INBD ISOL Standard Applicant located/identified and verified HVN-MOV102 closed by checking the GREEN indicating light ON and RED indicating light OFF.

Cue Notes Results SAT UNSAT

6. 5.14.1 Verify the following valves are closed

• Procedure Step:

• HVN-MOV22A, CONTMT UC1A DISCH Standard Applicant located/identified and closed HVN-MOV22A placing the control switch in the CLOSE position and by checking the GREEN indicating light ON and RED indicating light OFF.

Cue Notes HVN-MOV22A(B) isolate on Level 1 and will therefore still be open requiring the applicant to close them. This is not a failure / alternate path.

Results SAT UNSAT

7. 5.14.1 Verify the following valves are closed

• Procedure Step:

• HVN-MOV22B, CONTMT UC1B DISCH Standard Applicant located/identified and closed HVN-MOV22B placing the control switch in the CLOSE position and by checking the GREEN indicating light ON and RED indicating light OFF.

Cue Notes HVN-MOV22A(B) isolate on Level 1 and will therefore still be open requiring the applicant to close them. This is not a failure / alternate path.

Step 7 is a continuation of the step 6 bullet. Both the 22A and B valves should be closed for this condition.

Results SAT UNSAT RJPM-NRC-M14-S6 Rev 0 Page 6 of 10

RJPM-NRC-M14-S6 Rev. 0

8. 5.14.2 At H13-P870, open the following valves:

• Procedure Step:

1. SWP-MOV502A, CONTAINMENT UC SUPPLY Standard Applicant located/identified and opened SWP-MOV502A by placing the control switch in the OPEN position and checking the GREEN indicating light OFF and the RED indicating light ON.

Cue Notes The opening of SWP-MOV502A & B are NOT sequence dependent.

Since BOTH Unit Coolers require service water, all 4 valves will be opened.

Results SAT UNSAT

9. 5.14.2 At H13-P870, open the following valves:

• Procedure Step:

1. SWP-MOV502B, CONTAINMENT UC SUPPLY Standard Applicant located/identified and opened SWP-MOV502B by placing the control switch in the OPEN position and checking the GREEN indicating light OFF and the RED indicating light ON.

Cue Notes The opening of SWP-MOV502A & B are NOT sequence dependent.

Step 9 is a continuation of the step 8 bullet. Both the 502A and B valves should be opened for this condition.

Results SAT UNSAT RJPM-NRC-M14-S6 Rev 0 Page 7 of 10

RJPM-NRC-M14-S6 Rev. 0

10. 5.14.2 At H13-P870, open the following valves:

• Procedure Step:

2) SWP-MOV503A, CONTAINMENT UC RETURN Standard Applicant located/identified and opened SWP-MOV503A by placing the control switch in the OPEN position and checking the GREEN indicating light OFF and the RED indicating light ON.

Cue Notes The opening of SWP-MOV503A & B are NOT sequence dependent.

Results SAT UNSAT

11. 5.14.2 At H13-P870, open the following valves:

• Procedure Step:

2) SWP-MOV503B, CONTAINMENT UC RETURN Standard Applicant located/identified and opened SWP-MOV503B by placing the control switch in the OPEN position and checking the GREEN indicating light OFF and the RED indicating light ON Cue Notes The opening of SWP-MOV503A & B are NOT sequence dependent.

Step 11 is a continuation of the step 10 bullet. Both the 503A and B valves should be opened for this condition.

Results SAT UNSAT

12. 5.14.3 Verify HVR-UC1A(B), CONTMT UNIT CLR A(B) is running

• Procedure Step:

Standard Applicant recognized that only one HVR unit cooler was running and started the second cooler by depressing its START pushbutton and by checking the RED indicating light ON and GREEN indicating light OFF Cue Notes Results SAT UNSAT Terminating Cue: HVR-UC1A and HVR-UC1B are running with service water providing cooling water flow and all HVN valves are isolated This completes this JPM.

STOP TIME:

RJPM-NRC-M14-S6 Rev 0 Page 8 of 10

RJPM-NRC-M14-S6 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-S6 Rev 0 Page 9 of 10

RJPM-NRC-M14-S6 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

A plant transient has occurred resulting in entry into the Emergency Operating Procedures. EOP-0002 directs maximizing containment cooling. The containment unit coolers have no cooling water flow due to the isolation of HVN.

RPV Level is stable between Level 1 & Level 2.

INITIATING CUE:

The CRS has directed you to conduct Emergency Operation of Containment Unit Coolers with Service Water in accordance with SOP-0059 Section 5.14 for BOTH Containment Unit Coolers RJPM-NRC-M14-S6 Rev 0 Page 10 of 10

CONTINUOUS USE G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• CONTAINMENT HVAC SYSTEM (SYS #403)

PROCEDURE NUMBER: *SOP-0059 REVISION NUMBER: *034 Effective Date: *06/14/11 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0059R033EC-A Added Notes for Steps 5.2, 5.4, and 5.6.3 referencing the annunciators that are expected to alarm during this procedure.

SOP-0059 REV - 034 PAGE 1 OF 83

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES FOR STARTUP AND OPERATION .........................................................5 4 SYSTEM STARTUP..................................................................................................................6 4.1 Annulus Pressure Control System (APCS) Startup ..........................................................6 4.2 Containment Cooling System Startup...............................................................................7 4.3 Placing the Reactor Sample Panel Filter Unit in Service .................................................7 5 SYSTEM OPERATION .............................................................................................................8 5.1 Shifting Annulus Pressure Control Fans...........................................................................8 5.2 Standby Gas Treatment Manual Startup ...........................................................................9 5.3 Automatic or Manual Initiation of Standby Gas...............................................................11 5.4 Restoration of Standby Gas...............................................................................................16 5.5 Containment High Volume Purge.....................................................................................20 5.6 Containment Low Volume Purge......................................................................................24 5.7 Drywell High Volume Purge ............................................................................................27 5.8 Drywell Low Volume Purge .............................................................................................31 5.9 Containment Venting Using the Containment Purge Filter Exhaust Fan in Recirculation Mode...........................................................................................................35 5.10 Containment Venting With Standby Gas Treatment Train In Operation .........................38 5.11 Drywell Recirculation Using the Containment Purge Filter Exhaust Fan ........................40 5.12 Alternating Containment Unit Coolers .............................................................................43 5.13 Using the Reactor Sample Panel Filter Unit to support RWCU F/D Backwashes .......................................................................................................................43 5.14 Emergency Operation of Containment Unit Coolers with Service Water ........................44 5.15 Performing a Breaker Operability for HVR-UC1A(B).....................................................44 5.16 Lowering or Maintaining Containment Pressure Without Using the Containment Purge Filter Exhaust Fan .............................................................................45 6 SYSTEM SHUTDOWN.............................................................................................................46 6.1 Annulus Pressure Control System (APCS) Shutdown......................................................46 6.2 Shutdown of Containment Unit Coolers ..........................................................................47 SOP-0059 REV - 034 PAGE 2 OF 83

CONTINUOUS USE 6.3 Shutdown of Containment Dome Recirculation Fans ......................................................47 6.4 Isolation of the Reactor Sample Panel Filter Unit ...........................................................47 7 REFERENCES ...........................................................................................................................48 8 RECORDS..................................................................................................................................48 ATTACHMENT 1A - VALVE LINEUP - ANNULUS PRESSURE CONTROL SYSTEM (SAFETY RELATED) .....................................................................................49 ATTACHMENT 1B - VALVE LINEUP - CONTAINMENT/DRYWELL PURGE SYSTEM (SAFETY RELATED) .....................................................................................50 ATTACHMENT 1C - VALVE LINEUP - CONTAINMENT COOLING SYSTEM (SAFETY RELATED)......................................................................................................52 ATTACHMENT 2A - INSTRUMENT LINEUP - ANNULUS PRESSURE CONTROL SYSTEM (SAFETY RELATED) .....................................................................................54 ATTACHMENT 2B - INSTRUMENT LINEUP - CONTAINMENT/DRYWELL PURGE SYSTEM (SAFETY RELATED) .......................................................................56 ATTACHMENT 2C - INSTRUMENT LINEUP - CONTAINMENT COOLING SYSTEM (SAFETY RELATED) .....................................................................................64 ATTACHMENT 2D - INSTRUMENT LINEUP - ANNULUS MIXING SYSTEM (SAFETY RELATED)......................................................................................................68 ATTACHMENT 3A - ELECTRICAL LINEUP - ANNULUS PRESSURE CONTROL SYSTEM (SAFETY RELATED) .....................................................................................69 ATTACHMENT 3B - ELECTRICAL LINEUP - CONTAINMENT/DRYWELL PURGE SYSTEM (SAFETY RELATED) .....................................................................................70 ATTACHMENT 3C - ELECTRICAL LINEUP - CONTAINMENT COOLING SYSTEM (SAFETY RELATED) .....................................................................................72 ATTACHMENT 3D - ELECTRICAL LINEUP - ANNULUS MIXING SYSTEM (SAFETY RELATED)......................................................................................................74 ATTACHMENT 4A - CONTROL BOARD LINEUP - ANNULUS PRESSURE CONTROL SYSTEM (SAFETY RELATED) .................................................................75 ATTACHMENT 4B - CONTROL BOARD LINEUP - CONTAINMENT/DRYWELL PURGE (SAFETY RELATED)........................................................................................77 ATTACHMENT 4C - CONTROL BOARD LINEUP - CONTAINMENT COOLING SYSTEM (SAFETY RELATED) .....................................................................................80 ATTACHMENT 4D - CONTROL BOARD LINEUP - ANNULUS MIXING SYSTEM (SAFETY RELATED)......................................................................................................82 SOP-0059 REV - 034 PAGE 3 OF 83

CONTINUOUS USE 1 PURPOSE 1.1 The purpose of this procedure is to provide instructions for the operation of:

1.1.1. Annulus Pressure Control System 1.1.2. Containment/Drywell Purge System 1.1.3. Containment Cooling System 2 PRECAUTIONS AND LIMITATIONS 2.1 Altering Containment HVAC alignments or operating configurations may cause a change in radiological conditions. Radiation Protection should be kept abreast of changes.

2.2 If venting/purging of Primary Containment is being performed as directed by the Emergency Operating Procedures, the Emergency Director should be notified in order to evaluate the impact on offsite dose.

2.3 With an Auto Start signal present for GTS-FN1A(B), SGT EXH FAN A(B), the fan which is stopped will automatically restart upon receiving a low flow signal from the running train, whether abnormal conditions still exist or not. When auto start signals are cleared, before stopping the running fan, it is necessary to press the STOP Pushbutton of the non running fan to prevent an unnecessary automatic restart of that fan.

2.4 Containment pressure is maintained greater than or equal to -0.3 psig and less than or equal to +0.3 psig when in Modes 1, 2 and 3.

2.5 Containment temperature is maintained greater than or equal to 70!F to ensure operability of Standby Liquid Control Solution Class 1 supports and compliance with Environmental Design Criteria spec 215.150.

2.6 Containment to Annulus high negative differential pressure will be reached much sooner and could cause an ESF actuation when purging or venting the Containment with the Annulus Pressure Control System secured.

2.7 Containment is not to be vented in accordance with Section 5.9 with Standby Gas Treatment in operation. Refer To Section 5.10. Starting HVR-FN14 may cause automatic initiation of Standby Gas Treatment.

2.8 All controls and indications are located on H13-P863 unless noted otherwise.

SOP-0059 REV - 034 PAGE 4 OF 83

CONTINUOUS USE 2.9 HVR-UC1A(B)(C), CONTMT UNIT CLR A(B)(C) outlet temperatures are controlled by M HVR-TIC26A(B)(C), CONTMT UNIT CLR 1A(B)(C) INTAKE TEMPERATURE INDICATING CONTROLLER. The pre LOCA temperature is to be less than or equal to 90oF the Technical Specification upper limit for average containment temperature during normal operation to ensure the Design Basis Accident temperature of 185oF will not be exceeded. A minimum environmental design criterion of 70oF is specified for the general containment area. The setpoints of controllers HVR-TIC26A(B)(C) may be adjusted to maintain Containment temperature greater than or equal to 70oF and less than or equal to 90oF. (Ref. 7.8)

C 2.10 Simultaneous operation of both divisions of Standby Gas Treatment on the Aux.

Building, can cause excessive vacuum which draws air and potential contamination out of floor drains on the Aux Building 114 elevation. Therefore, prior to starting GTS-FN1A and B, SGT EXH FAN A and B Radiation Protection will be notified to install hardware cloth and filter media per ER-RB-2005-0342 in the Auxiliary Bldg floor drain hubs to prevent the spread of contamination.

3 PREREQUISITES FOR STARTUP AND OPERATION 3.1 Verify Normal Service Water System in operation per SOP-0018, Normal Service Water.

3.2 Verify Instrument Air System in operation per SOP-0022, Instrument Air System.

3.3 Verify the below listed electrical systems are in operation:

3.3.1. 480VAC per SOP-0047, 480 VAC System 3.3.2. 120VAC per SOP-0048, 120 VAC System 3.3.3. 125VDC per SOP-0049, 125 VDC System 3.4 Verify Chilled Water System in operation per SOP-0116, Turbine and Radwaste Building HVAC Chilled Water System.

3.5 Verify Auxiliary Building HVAC System in operation per SOP-0065, HVAC-Auxiliary Building.

3.6 Verify Standby Gas Treatment System in standby per SOP-0043, Standby Gas Treatment System.

3.7 Verify Digital Radiation Monitoring System in operation per SOP-0086, Digital Radiation Monitoring System.

3.8 Verify system is lined up for startup.

SOP-0059 REV - 034 PAGE 5 OF 83

CONTINUOUS USE 5.14 Emergency Operation of Containment Unit Coolers with Service Water NOTE This section is only to be used in emergencies. If it is desired to use Service Water at other times, refer to SOP-0116, Turbine and Radwaste Building HVAC Chilled Water System.

5.14.1. Verify the following valves are closed:

" HVN-MOV127, CHW SPLY OUTBD ISOL

" HVN-MOV128, CHW RTN OUTBD ISOL

" HVN-MOV129, CHW SPLY SHUTOFF VLV

" HVN-MOV130, CHW RTN SHUTOFF VLV

" HVN-MOV102, CHW RTN INBD ISOL

" HVN-MOV22A(B), CONTMT UC1A(B) DISCH 5.14.2. At H13-P870, open the following valves:

1. SWP-MOV502A(B), CONTAINMENT UC SUPPLY

2. SWP-MOV503A(B), CONTAINMENT UC RETURN 5.14.3. Verify HVR-UC1A(B), CONTMT UNIT CLR A(B) is running.

NOTE Chilled Water or Service Water is not required when performing a breaker operability.

5.15 Performing a Breaker Operability for HVR-UC1A(B) 5.15.1. Start HVR-UC1A(B), CONTMT UNIT CLR A(B).

5.15.2. Stop HVR-UC1A(B), CONTMT UNIT CLR A(B).

SOP-0059 REV - 034 PAGE 44 OF 83

RJPM-NRC-M14-C1 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Defeat Offgas High Radiation Isolation Interlock OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 5 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform Plant X Simulate Simulator X Control Room Prepared: Dave Bergstrom Date: September 9, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 8

RJPM-NRC-M14-C1 Rev. 0 EXAMINER INFO SHEET Task Standard: The electrical lead from terminal 19 of TB0019 on H13-P845 Bay C is lifted and covered.

Synopsis: This task will defeat the Offgas high radiation isolation interlocks for N64-F060, Offgas Discharge to Vent Valve using EOP Enclosure 34.

This will maintain the main condenser available during an ATWS.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to install Enclosure 34 (Defeating Offgas High Radiation Isolation Interlocks).

3) Initial Conditions:

The plant has experienced an ATWS, all MSIVs are open.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-C1 Rev 0 Page 2 of 8

RJPM-NRC-M14-C1 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Defeat Offgas High Radiation Isolation 200033005004 295002 AA1.02 2.9 / 2.9 Interlock

REFERENCES:

APPLICABLE OBJECTIVES EOP-0005, Rev 312, Attach 34 RLP-HLO-0516, Obj 1 REQUIRED MATERIALS: SAFETY FUNCTION:

OSP-0053, Rev 17 Attach 34 __9__

SIMULATOR CONDITIONS & SETUP:

1. NA - This is a Control Room JPM.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: The electrical lead from terminal 19 of TB0019 on H13-P845 Bay C is lifted and covered.

RJPM-NRC-M14-C1 Rev 0 Page 3 of 8

RJPM-NRC-M14-C1 Rev. 0 PERFORMANCE:

START TIME:

1. Procedure Step: 3.1 OBTAIN EOP-0005 ENCL 34 tool kit from the Control Room Emergency Locker.

3.1.1 INSPECT kit for:

One (1) regular screwdriver One (1) piece of tygon tubing Standard Applicant obtained the key to the Control Room Emergency Locker Applicant located the Control Room Emergency Locker Applicant obtained the Enclosure 34 tool kit.

Applicant inspected and found 1 screwdriver and 1 piece of tubing.

Cue After inspecting kit, applicant should be told to leave the kit in the locker.

Notes Results SAT UNSAT Procedure Step: 3.2 DEFEAT Offgas high radiation isolation interlock as follows:

Standard NA Cue Notes No applicant action is necessary - this is a placekeeper for the procedure.

2. Procedure Step: 3.2.1 Location H13-P845 Bay C Standard Applicant located/identified Panel H13-P845, Bay C.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-C1 Rev 0 Page 4 of 8

RJPM-NRC-M14-C1 Rev. 0

3. Procedure Step: 3.2.1 Affected Terminal Board: TB0019 (Left side of bay, 1st column of terminal boards from panel door, top terminal board)

Standard Applicant located/identified Terminal Board TB0019.

Cue Notes Results SAT UNSAT

4. Procedure Step: 3.2.1.1 REMOVE the orange lead from Terminal 19 on Terminal Board TB0019 AND COVER with Tygon tubing.

Standard Applicant located/identified and simulated removing the screw in a counterclockwise motion from terminal 19.

Cue Notes Results SAT UNSAT

5. 3.2.1.1 REMOVE the orange lead from Terminal 19 on Terminal Board

• Procedure Step:

TB0019 AND COVER with Tygon tubing.

Standard Applicant located/identified and removed the orange lead from terminal 19 on Terminal Board TB0019 Applicant placed the tygon tubing from the kit on the exposed lug.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-C1 Rev 0 Page 5 of 8

RJPM-NRC-M14-C1 Rev. 0

6. Procedure Step: 3.3 VERIFY N64-F060, OFF GAS DISCH TO VENT VLV is open.

Standard Applicant located and verified the Offgas Disch Valve is open (red light on and green light is off)

Cue When requested, indicate the red light is on and the green light is off.

Notes Results SAT UNSAT Terminating Cue: Enclosure 34, Defeating Offgas High Radiation Isolation Interlocks, has been installed This completes this JPM.

STOP TIME:

RJPM-NRC-M14-C1 Rev 0 Page 6 of 8

RJPM-NRC-M14-C1 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-C1 Rev 0 Page 7 of 8

RJPM-NRC-M14-C1 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The plant has experienced an ATWS, all MSIVs are open.

INITIATING CUE:

The CRS has directed you to install Enclosure 34 (Defeating Offgas High Radiation Isolation Interlocks).

RJPM-NRC-M14-C1 Rev 0 Page 8 of 8

CONTINUOUS USE ENCLOSURE 34 DEFEATING OFFGAS HIGH RADIATION ISOLATION INTERLOCKS 1.0 PURPOSE To provide instructions for defeating the Offgas high radiation isolation interlocks for N64-F060, OFF GAS DISCH TO VENT VLV to maintain the main condenser available during an ATWS.

2.0 REQUIRED TOOLS-EQUIPMENT 2.1 EOP-0005 ENCL 34 tool kit containing one (1) regular screwdriver and one (1) piece of tygon tubing.

3.0 INSTRUCTIONS 3.1 OBTAIN EOP-0005 ENCL 34 tool kit from the Control Room Emergency Locker.

3.1.1 INSPECT kit for:

One (1) regular screwdriver One (1) piece of tygon tubing 3.2 DEFEAT Offgas high radiation isolation interlocks as follows:

3.2.1 Location

H13-P845 Bay C Affected Terminal Board: TB0019 (Left side of bay, 1st column of terminal boards from panel door, top terminal board)

1. REMOVE the orange lead from Terminal 19 on Terminal Board TB0019 AND COVER with Tygon tubing.

3.3 VERIFY N64-F060, OFF GAS DISCH TO VENT VLV is open.

4.0 REFERENCES

4.1 GE Elem Diag 828E243AA, Process Radiation Mon System 4.2 GE Elem Diag 828E257AA, Off-Gas Control System 4.3 PID-31-4F, OFFGAS ENCLOSURE 34 PAGE 1 OF 2 EOP-0005 REV.-313 PAGE 129 OF 139

CONTINUOUS USE ENCLOSURE 34 DEFEATING OFFGAS HIGH RADIATION ISOLATION INTERLOCKS MAIN CONTROL ROOM NSSSS BACKPANELS All panel numbers are prefixed with H13-P878 P879 P637 P821 P634 P619 P612 P607 P618 P631 P654 P652 P610 P653 P651 P655 P628 P621 P629 P670 P692 P622 P613 P642 P632 P623 P691 P669 P671 P693 P625 P630 P694 P672 P845 P614 P604 P600 P877 P601 P680 P808 ENCLOSURE 34 PAGE 2 OF 2 EOP-0005 REV.-313 PAGE 130 OF 139

RJPM-NRC-M14-C2 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Bypass a Local Power Range Monitor (LPRM) Detector OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 15 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform Plant X Simulate Simulator X Control Room Prepared: Dave Bergstrom Date: September 10, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 12

RJPM-NRC-M14-C2 Rev. 0 EXAMINER INFO SHEET Task Standard: LPRM 2A-38-47 has been bypassed using REP-0037.

Synopsis: This task will bypass an LPRM.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

LPRM 2A-38-47 has failed. The CRS has directed you to bypass it in accordance with REP-0037, LPRM Operability, Section 4.1.

3) Initial Conditions:

The plant is in Mode 1.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-C2 Rev 0 Page 2 of 12

RJPM-NRC-M14-C2 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Bypass a Local Power Range Monitor 215011001001 215005 A4.04 3.2 / 3.2 (LPRM) Detector 215005 A4.06 3.6 /

3.8 REFERENCES

APPLICABLE OBJECTIVES REP-0037, Rev 15, RLP-STM-0503, Obj 18, 21 REQUIRED MATERIALS: SAFETY FUNCTION:

REP-0037, Rev 15 - Section 4.1 __7__

SIMULATOR CONDITIONS & SETUP:

1. NA - This is a Control Room JPM.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: LPRM 2A-38-47 has been bypassed using REP-0037.

RJPM-NRC-M14-C2 Rev 0 Page 3 of 12

RJPM-NRC-M14-C2 Rev. 0 PERFORMANCE:

START TIME:

PROCEDURE NOTE Sections of this procedure may be performed as needed to support plant configuration. Performing the sections in sequential order is not required. Steps within a section should be performed in sequential order.

1. Procedure Step: 4.1.1 Record the performance of the applicable steps of section 4.1 on the LPRM BYPASS SHEET (Attachment).

Standard Applicant located Attachment 2 of REP-0037.

Cue Notes Results SAT UNSAT

2. Procedure Step: 4.1.2 Ensure that personnel bypassing an LPRM satisfy ADM-0007 requirements for performing this procedure by being any of the following: Licensed Operator, Qualified STA, or Qualified I&C Technician.

Standard Applicant filled in the appropriate line on Attachment 2.

Cue Applicant qualifies as a Licensed Operator for the simulation of this JPM.

Notes Results SAT UNSAT

3. Procedure Step: 4.1.3 Contact the On-Duty Reactor Engineer to notify them of the intent to bypass an LPRM, so that they can evaluate potential to affect core monitoring functions.

Standard Applicant notified Reactor Engineer of the intent to bypass LPRM 2A-38-47.

Applicant initialed the appropriate line on Attachment 2.

Cue As the Reactor Engineer, acknowledge the report about the intent to bypass LPRM 2A-38-47.

Notes Results SAT UNSAT RJPM-NRC-M14-C2 Rev 0 Page 4 of 12

RJPM-NRC-M14-C2 Rev. 0

4. Procedure Step: 4.1.4 Record the LPRM location and level on the LPRM BYPASS sheet.

Standard Applicant recorded and initialed LPRM 2A-38-47 on Attachment 2.

Cue Notes Results SAT UNSAT

5. Procedure Step: 4.1.5 Using Attachment 1 as a guide, determine the APRM channel associated with the LPRM.

Standard Applicant identified APRM H from Attachment 1.

Applicant records APRM H on Attachment 2, and initialed step.

Cue Notes Results SAT UNSAT

6. Procedure Step: 4.1.6 Obtain permission from the Operations Shift Manager (OSM) or the Control Room Supervisor (CRS) prior to bypassing an LPRM.

Request the OSM/CRS to have the At-the-Controls (ATC) operator bypass the affected APRM on 1H13-P680 prior to manipulating any switches.

Standard Applicant communicated/got permission to bypass LPRM 2A-38-47 Applicant communicated/requested that the ATC Operator bypass APRM H.

Cue As OSM/CRS, give permission to bypass the LPRM.

(Applicant may request the OSM/CRS initial the step on Attachment 2)

As ATC Operator, inform the applicant that APRM H is bypassed.

Notes Results SAT UNSAT RJPM-NRC-M14-C2 Rev 0 Page 5 of 12

RJPM-NRC-M14-C2 Rev. 0

7. Procedure Step: 4.1.7 Have a qualified member of plant staff (Step 4.1.2) provide concurrence with the performance of steps 4.1.8 - 4.1.17 Standard Applicant requested a peer check.

Cue Inform the applicant that you are qualified and will provide concurrence with the required steps.

Notes Results SAT UNSAT

8. Procedure Step: 4.1.8 Verify that no more than 33 LPRM signals are bypassed. Record the number of bypassed LPRMs on the LPRM BYPASS SHEET Standard Applicant verified the number of bypassed/failed LPRM signals.

Applicant records Total Number of LPRM signals bypassed as zero and initials Attachment 2 as the Performer.

Cue Inform the applicant that there are no other bypassed/failed LPRM signals.

Notes Applicant could obtain this data from walking around to each of the LPRM panels or by viewing info from the Plant Computer.

Results SAT UNSAT

9. Procedure Step: 4.1.9 Record the number of LPRMs that are in OPERATE by placing the affected APRM METER FUNCTION SWITCH in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Record the number on the LPRM BYPASS SHEET.

Standard Applicant located/identified and simulated placing APRM H Meter Function Switch to COUNT. (located on Panel P672)

{Cue}

Applicant divided 80/5 and recorded 16 on the appropriate line of Attachment 2 then initialed as Performer.

Cue When applicant simulates placing switch in COUNT, use a pen to indicate 80% on the meter.

Notes Results SAT UNSAT RJPM-NRC-M14-C2 Rev 0 Page 6 of 12

RJPM-NRC-M14-C2 Rev. 0

10. Procedure Step: 4.1.10 Operations will determine if bypassing the LPRM will render its APRM inoperable per step 3.2.

Standard Applicant referenced step 3.2 of REP-0037 and determined that APRM H will not be rendered inoperable by bypassing LPRM 2A-38-47.

Applicant initials as Performer on the appropriate line of Attachment 2.

Cue Notes Results SAT UNSAT

11. Procedure Step: 4.1.11 Operations will determine if bypassing the LPRM will render PBDS channel A or B inoperable per step 3.3.

Standard Applicant referenced step 3.3 of REP-0037 and determined that PBDS will not be rendered inoperable by bypassing LPRM 2A-38-47.

Applicant initials as Performer on the appropriate line of Attachment 2.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-C2 Rev 0 Page 7 of 12

RJPM-NRC-M14-C2 Rev. 0 PROCEDURE NOTE The following step brings in ALARM No. 2168, LPRM Downscale on H13-P680.

12. 4.1.12 Using Attachment 1 as a guide, select the desired LPRM to be

• Procedure Step:

bypassed with the LPRM SELECTOR SWITCH and the METER FUNCTION SWITCH. Observe that the bypass light on the APRM panel meter is not lit Standard Applicant located/identified and selected 2 on the Selector Switch.

Applicant located/identified and selected A on the Function Switch.

Applicant located/identified the bypass light.

Applicant initials as Performer on the appropriate line of Attachment 2.

Cue Inform or represent to the applicant that the light is not lit.

Notes The applicant may inform the ATC Operator of the incoming ALARM Results SAT UNSAT

13. 4.1.13 Bypass the selected LPRM by placing the S1 switch of the LPRM in

• Procedure Step:

the BYPASS position.

Standard Applicant located/identified the card for 2A-38-47 and simulated placing the switch to BYPASS.

Cue Indicate that the switch is in the bypass position.

Notes The card is located on the 2nd row below the meter to the far left Results SAT UNSAT

14. Procedure Step: 4.1.14 Observe that the bypass light on the APRM panel meter is lit.

Record on LPRM BYPASS SHEET.

Standard Applicant located/identified and noted the bypass light lit.

Applicant records the current Date and Time, then initials as Performer on the appropriate line of Attachment 2.

Cue Indicate that the bypass light is lit.

Notes Results SAT UNSAT RJPM-NRC-M14-C2 Rev 0 Page 8 of 12

RJPM-NRC-M14-C2 Rev. 0

15. Procedure Step: 4.1.15 Determine the number of the LPRMs that are in OPERATE by placing the METER FUNCTION SWITCH for the selected APRM on the panel meter in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Check that the number of LPRMs in OPERATE is one less than recorded on the LPRM BYPASS SHEET, step 4.1.9.

4. Record the number on the LPRM BYPASS SHEET.

Standard Applicant located/identified and simulated placing APRM H Meter Function Switch to COUNT.

{Cue}

Applicant divided 75/5 and recorded 15 on the appropriate line of Attachment 2 then initialed as Performer.

Cue When applicant simulates placing switch in COUNT, use a pen to indicate 75% on the meter.

Notes Results SAT UNSAT Procedure Step: 4.1.16 If other LPRMs in the same APRM are required to be bypassed, continue with bypassing those LPRMs before continuing to step 4.1.17.

Standard NA Cue Notes No applicant action is required for this step.

RJPM-NRC-M14-C2 Rev 0 Page 9 of 12

RJPM-NRC-M14-C2 Rev. 0

16. Procedure Step: 4.1.17 Using either an OD-3 edit or a 3D Monicore Core Power and Flow edit, verify that the APRM reading is within +/- 2% of rated thermal power (RTP)

Standard Applicant verified that APRM H is reading within +/- 2% of RTP Applicant initialed as Performer on the appropriate line of Attachment 2.

Cue When requested inform the applicant that APRM H and Core Thermal Power are equal.

Notes The 9 substeps of 4.1.17 are only implemented when there is a mismatch.

Results SAT UNSAT

17. Procedure Step: 4.1.18 Request the OSM/CRS have the ATC return the affected APRM to service Standard Applicant requested that APRM H is returned to service (un-bypassed).

{Cue}

Applicant initialed as Performer on the appropriate line of Attachment 2.

Cue When requested, inform the applicant that the ATC Operator has returned APRM H to service.

Notes Results SAT UNSAT

18. Procedure Step: 4.1.19 Record the date and WR/WO# on the LPRM BYPASS SHEET.

Standard Applicant recorded the current date and time on the appropriate line of Attachment 2.

Cue Inform the Applicant that the WR# is 123456.

Notes Results SAT UNSAT Terminating Cue: LPRM 2A-38-47 has been using REP-0037.

This completes this JPM.

Cue:

STOP TIME:

RJPM-NRC-M14-C2 Rev 0 Page 10 of 12

RJPM-NRC-M14-C2 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-C2 Rev 0 Page 11 of 12

RJPM-NRC-M14-C2 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The plant is in Mode 1.

INITIATING CUE:

LPRM 2A-38-47 has failed. The CRS has directed you to bypass it in accordance with REP-0037, LPRM Operability, Section 4.1.

RJPM-NRC-M14-C2 Rev 0 Page 12 of 12

REFERENCE USE

• G12.1.23 RIVER BEND STATION STATION OPERATING MANUAL

• REACTOR ENGINEERING PROCEDURE

• LPRM OPERABILITY PROCEDURE NUMBER: *REP-0037 REVISION NUMBER: *15 Effective Date: *February 2, 2010 NOTE : SIGNATURES ARE ON FILE.

TemRev 2 AddCounter 84 Att Enc DS MSet REGULAR KWN OFF REFERENCE USE

• INDEXING INFORMATION

REFERENCE USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER REP-0037 REV - 15 PAGE 1 OF 14

REFERENCE USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE............................................................................................................................3 2 REFERENCES ....................................................................................................................3 3 PRECAUTION AND LIMITATIONS .................................................................................4 4 PROCEDURE......................................................................................................................5 4.1 Bypassing an LPRM....................................................................................................5 4.2 Restoring an LPRM to Service ....................................................................................8 5 DOCUMENTATION.........................................................................................................11 ATTACHMENT 1 - LPRM INPUTS TO APRM CHANNELS.................................................12 ATTACHMENT 2 - LPRM BYPASS .......................................................................................13 ATTACHMENT 3 - LPRM RESTORE ...................................................................................14 REP-0037 REV - 15 PAGE 2 OF 14

REFERENCE USE 1 PURPOSE 1.1 To provide procedural guidance for bypassing LPRMs and restoring LPRMs to service in mode 1.

1.2 To provide procedural controls to ensure that the APRM operability requirement as a function of the operable LPRMs is met.

1.3 To provide procedural controls to ensure that the Core Monitor functioning requirement as a function of the operable LPRMs is met.

2 REFERENCES 2.1 RBS Technical Specification Bases B3.3.1.1.2a-d.

2.2 GEK-83360, Power Range Neutron Monitoring System Dec., 1983 2.3 GE Elem. Diagram 851E230AA, Power Range Neutron Mon. Sys. Rev.

11.

2.4 EN-AD-103, Document Control and Records Management Activities 2.5 NEDE-24810 Vol. 2, Process Computer, June, 1981 2.6 Commitment 11825 2.7 ADM-0007, Selection, Training, Qualification and Evaluation of Plant Staff Personnel.

2.8 River Bend CR-1999-1857-015 REP-0037 REV - 15 PAGE 3 OF 14

REFERENCE USE 3 PRECAUTION AND LIMITATIONS 3.1 Bypassing and un-bypassing an LPRM may affect the reading of the APRM channel associated with the LPRM. The gain of the APRM may require adjustment to indicate the Desired Fraction of Rated Thermal Power per Technical Specification 3.3.1.1s Surveillance Requirement 3.3.1.1.2.

3.2 Each APRM must have 2 operable LPRM inputs per level and a minimum of 11 operable LPRMs to be operable as required by Technical Specification Bases B3.3.1.1.2a-d.

3.3 Each PBDS channel (APRM A & B only) must have 8 valid LPRM inputs as required by Technical Specification Bases B3.3.1.3.

3.4 Bypassing and un-bypassing an LPRM may cause LPRM Drift Messages to print on the PPC printer in the Main Control Room.

3.5 This procedure is for bypassing and restoring LPRMs in mode 1 only.

REP-0037 REV - 15 PAGE 4 OF 14

REFERENCE USE 4 PROCEDURE NOTE Sections of this procedure may be performed as needed to support plant configuration. Performing the sections in sequential order is not required. Steps within a section should be performed in sequential order.

4.1 Bypassing an LPRM 4.1.1. Record the performance of the applicable steps of section 4.1 on the LPRM BYPASS SHEET (Attachment 2).

4.1.2. Ensure that personnel bypassing an LPRM satisfy ADM-0007 requirements for performing this procedure by being any of the following: Licensed Operator, Qualified STA, or Qualified I&C Technician.

1. Operations will complete the determination if the APRM remains operable following the bypassing of an LPRM.

4.1.3. Contact the On-Duty Reactor Engineer to notify them of the intent to bypass an LPRM so that they can evaluate potential to affect core monitoring functions.

4.1.4. Record the LPRM location and level on the LPRM BYPASS SHEET.

4.1.5. Using Attachment 1 as a guide determine the APRM channel associated with the LPRM.

4.1.6. Obtain permission from the Operations Shift Manager (OSM) or the Control Room Supervisor (CRS) prior to bypassing an LPRM. Request the OSM/CRS to have the At-the-Controls (ATC) operator bypass the affected APRM on 1H13-P680 prior to manipulating any switches.

4.1.7. Have a qualified member of plant staff (Step 4.1.2 ) provide concurrence with the performance of steps (4.1.8-4.1.17).

4.1.8. Verify that no more than 33 LPRM signals are bypassed.

Record the number of bypassed LPRMs on the LPRM BYPASS SHEET.

REP-0037 REV - 15 PAGE 5 OF 14

REFERENCE USE 4.1.9. Record the number of LPRMs that are in OPERATE by placing the affected APRM METER FUNCTION SWITCH in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Record the number on the LPRM BYPASS SHEET.

(Commitment 11825) 4.1.10. Operations will determine if bypassing the LPRM will render its APRM inoperable per step 3.2.

4.1.11. Operations will determine if bypassing the LPRM will render PBDS channel A or B inoperable per step 3.3.

NOTE The following step brings in ALARM No. 2168 LPRM downscale on H13-P680 4.1.12. Using Attachment 1 as a guide, select the desired LPRM to be bypassed with the LPRM SELECTOR SWITCH and the METER FUNCTION SWITCH. Observe that the bypass light on the APRM panel meter is not lit. (Commitment 11825) 4.1.13. Bypass the selected LPRM by placing the S1 switch of the LPRM in the BYPASS position.

4.1.14. Observe that the bypass light on the APRM panel meter is lit.

Record on LPRM BYPASS SHEET. (Commitment 11825) 4.1.15. Determine the number of the LPRMs that are in OPERATE by placing the METER FUNCTION SWITCH for the selected APRM on the panel meter in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Check that the number of LPRMs in OPERATE is one less than recorded on the LPRM BYPASS SHEET , Step 4.1.9.3.

4. Record the number on the LPRM BYPASS SHEET.

(11825)

REP-0037 REV - 15 PAGE 6 OF 14

REFERENCE USE 4.1.16. IF other LPRMs in the same APRM are required to be bypassed, continue with bypassing those LPRMs before continuing to step 4.1.17 4.1.17. Using either an OD-3 edit or a 3D Monicore Core Power and Flow edit, operations will verify that the APRM reading is within ! 2% of rated thermal power (RTP) or operations will adjust the APRM as follows:

1. Record the As Found reading of the APRM to be adjusted on the LPRM BYPASS SHEET.

2. Record the CTP readings in percent power on the LPRM BYPASS SHEET.

3. Verify or place the APRM mode switch S1 in the OPERATE position.

4. Verify or place METER FUNCTION switch S2 in the AVERAGE position.

5. Adjust the gain of the affected APRM (using feedback control potentiometer R16, located on auxiliary unit Z404) until the APRM panel meter or the PPC display reads as close as possible, the percent power recorded in step 4.1.17.2 .

6. Obtain an OD-3 or 3D Monicore Core Power and Flow edit.

7. Record the As Left reading of the adjusted APRM and the CTP on the LPRM BYPASS SHEET.

TS 8. Verify the absolute difference between the Average Power Range Monitor (APRM) and the calculated power " 2%

RTP. Record Yes or No on the LPRM BYPASS SHEET.

(Tech Spec SR 3.3.1.1.2)

9. If the difference between the APRM and the calculated core thermal power cannot be adjusted within -2% to

+2%, immediately notify the Operations Shift Manager/

Control Room Supervisor (OSM/CRS) and refer to LCO 3.3.1.1.

4.1.18. Request the OSM/CRS to have the ATC return the affected APRM to service.

REP-0037 REV - 15 PAGE 7 OF 14

REFERENCE USE 4.1.19. Record the date and WR/WO# (if applicable) on the LPRM BYPASS SHEET .

4.1.20. Send the LPRM BYPASS SHEET , Attachment 2, to the On-Duty Reactor Engineer for review and update of the LPRM Bypass Log.

4.2 Restoring an LPRM to Service 4.2.1. Record the performance of the applicable steps of section 4.2 on the LPRM RESTORE SHEET (Attachment 3).

4.2.2. Ensure that personnel restoring an LPRM satisfy ADM-0007 requirements for performing this procedure by being any of the following: a Licensed Operator, Qualified STA, or Qualified I&C Technician.

4.2.3. Contact Reactor Engineering to notify them of the intent to restore an LPRM to service so that Reactor Engineering can evaluate the potential to affect core monitoring functions.

4.2.4. Record the LPRM location and level.

4.2.5. Using Attachment 1 as a guide, determine the APRM channel associated with the LPRM.

4.2.6. Obtain permission from the Operations Shift Manager (OSM) or the Control Room Supervisor (CRS) prior to restoring an LPRM to service. Request the OSM/CRS to have the ATC bypass the affected APRM on 1H13-P680 prior to manipulating any switches.

4.2.7. Have a qualified member of plant staff ( Step 4.2.2) provide concurrence with the performance of steps ( 4.2.8.3 - 4.2.17).

4.2.8. Determine the number of LPRMs that are in OPERATE by placing the METER FUNCTION SWITCH for the selected APRM on the panel meter in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Record the number on the LPRM RESTORE SHEET .

(Commitment 11825)

REP-0037 REV - 15 PAGE 8 OF 14

REFERENCE USE 4.2.9. Using Attachment 1 as a guide, select the desired LPRM to be restored with the LPRM SELECTOR SWITCH and the METER FUNCTION SWITCH. Observe that the bypass light on the APRM panel meter is lit. (Commitment 11825) 4.2.10. Place the selected LPRM in service by placing the S1 switch of the LPRM in the OPERATE position.

4.2.11. Observe that the bypass light on the APRM panel meter is not lit. Record on LPRM RESTORE SHEET . (Commitment 11825) 4.2.12. Determine number of LPRMs that are in OPERATE by placing the METER FUNCTION SWITCH for the selected APRM on the panel meter in the COUNT position.

1. There is one LPRM for every 5% division.

2. Divide the meter value by 5.

3. Verify that the number of LPRMs in OPERATE is one greater than recorded in LPRM RESTORE SHEET , step 4.2.8.3.

4. Record the number on the LPRM RESTORE SHEET .

(11825) 4.2.13. With the LPRM SELECTOR SWITCH and the METER FUNCTION SWITCH, select the desired LPRM and display the reading on the front meter.

4.2.14. Reactor engineering will evaluate the restored LPRM indication / reading as follows and provide a desired value to have I&C adjust the reading to, as necessary.

1. If in a symmetric A sequence pattern, compare the LPRM reading to a symmetric LPRM reading.

2. Does the LPRM reading compare favorably to the other LPRM readings in that string?

3. Does the LPRM reading compare favorably with other LPRM readings of a similar level in that part of the core?

REP-0037 REV - 15 PAGE 9 OF 14

REFERENCE USE 4.2.15. IF the reactor engineer provides a desired value for the restored LPRM to be adjusted to, THEN I&C will adjust the LPRM gain using the appropriate LPRM GAIN Control/Switch as shown below until the LPRM indication matches the desired value.

GAIN SW (S2) POSITION GAIN ADJUSTMENT CONTROL L R5 M R3 H R1

1. Record final adjusted value as appropriate.

4.2.16. IF other LPRMs in the same APRM are required to be restored, continue with restoring those LPRMs before continuing to step 4.2.17 4.2.17. Using either an OD-3 edit or a 3D Monicore Core Power and Flow edit, operations will verify that the APRM reading is within ! 2% of rated thermal power (RTP) or operations will adjust the APRM as follows:

1. Record the As Found reading of the APRM to be adjusted on the LPRM RESTORE SHEET .

2. Record the CTP readings in percent power on the LPRM RESTORE SHEET .

3. Verify or place the APRM mode switch S1 in the OPERATE position.

4. Verify or place METER FUNCTION switch S2 in the AVERAGE position.

5. Adjust the gain of the affected APRM (using feedback control potentiometer R16, located on auxiliary unit Z404) until the APRM panel meter or the Core Monitor Edit or PPC display reads as close as possible, the percent power recorded in step 4.2.17.2 .

6. Obtain an OD-3 or 3D Monicore Core Power and Flow edit.

REP-0037 REV - 15 PAGE 10 OF 14

REFERENCE USE

7. Record the As Left reading of the adjusted APRM on the LPRM RESTORE SHEET .

TS 8. Verify the absolute difference between the Average Power Range Monitor (APRM) and the calculated power " 2%

RTP. Record Yes or No on the LPRM RESTORE SHEET . (Tech Spec SR 3.3.1.1.2)

9. If the difference between the APRM and the calculated core thermal power cannot be adjusted within -2% to

+2%, immediately notify the Operations Shift Manager/

Control Room Supervisor (OSM/CRS) and refer to LCO 3.3.1.1.

4.2.18. Request the OSM/CRS to have the ATC operator return the affected APRM to service.

4.2.19. After the LPRM is returned to service, notify Reactor Engineering so they can determine if an OD-1 is warranted or not.

4.2.20. Send the LPRM RESTORE SHEET , Attachment 3, to the On-Duty Reactor Engineer for review and update of the LPRM Bypass Log.

5 DOCUMENTATION 5.1 Reactor Engineering is responsible for:

5.1.1. Submitting records to Permanent Plant File within 30 days of final approval 5.1.2. Ensuring that the records submitted contain the required documentation REP-0037 REV - 15 PAGE 11 OF 14

REFERENCE USE ATTACHMENT 1 PAGE 1 OF 1 LPRM INPUTS TO APRM CHANNELS APRM A APRM B APRM C APRM D 2A-06-39 2A-14-39 1A-14-47 1A-22-47 3A-38-39 3A-46-39 2A-46-47 3A-06-31 4A-22-23 4A-30-23 3A-30-31 4A-38-31 6A-38-07 5A-14-07 4A-14-15 5A-22-15 1B-30-47 2B-38-47 5A-46-15 2B-14-39 2B-14-31 3B-22-31 2B-06-39 3B-46-39 3B-46-31 5B-06-15 3B-38-39 4B-30-23 4B-30-15 6B-38-15 4B-22-23 5B-14-07 2C-22-39 3C-30-39 6B-38-07 2C-38-47 4C-06-23 4C-14-23 1C-30-47 3C-22-31 5C-38-23 5C-46-23 2C-14-31 5C-06-15 6C-22-07 6C-30-07 3C-46-31 6C-38-15 1D-14-47 1D-22-47 4C-30-15 3D-30-39 2D-46-47 3D-06-31 2D-22-39 4D-14-23 3D-30-31 4D-38-31 4D-06-23 5D-46-23 4D-14-15 5D-22-15 5D-38-23 6D-30-07 5D-46-15 6D-22-07 APRM E APRM F APRM G APRM H 2A-22-39 3A-30-39 1A-30-47 2A-38-47 4A-06-23 4A-14-23 2A-14-31 3A-22-31 5A-38-23 5A-46-23 3A-46-31 5A-06-15 6A-22-07 6A-30-07 4A-30-15 6A-38-15 1B-14-47 1B-22-47 2B-22-39 3B-30-39 2B-46-47 3B-06-31 4B-06-23 4B-14-23 3B-30-31 4B-38-31 5B-38-23 5B-46-23 4B-14-15 5B-22-15 6B-22-07 6B-30-07 5B-46-15 2C-14-39 1C-14-47 1C-22-47 2C-06-39 3C-46-39 2C-46-47 3C-06-31 3C-38-39 4C-30-23 3C-30-31 4C-38-31 4C-22-23 5C-14-07 4C-14-15 5C-22-15 6C-38-07 2D-38-47 5C-46-15 2D-14-39 1D-30-47 3D-22-31 2D-06-39 3D-46-39 2D-14-31 5D-06-15 3D-38-39 4D-30-23 3D-46-31 6D-38-15 4D-22-23 5D-14-07 4D-30-15 6D-38-07 REP-0037 REV - 15 PAGE 12 OF 14

REFERENCE USE ATTACHMENT 2 PAGE 1 OF 1 LPRM BYPASS Step Name (print) Signature Initials 4.1.2 Step Instruction Initials 4.1.3 Contact Reactor Eng, RE(contacted):

4.1.4 LPRM 4.1.5 APRM 4.1.6 Obtain permission to bypass LPRM / Bypass APRM (OSM/CRS)

Step Instructions Performer Concurrence 4.1.8 Total number of LPRM signals bypassed. Number:

4.1.9.3 Affected APRM, LPRMs in operate. Number:

4.1.10 Operations - Determine if APRM will remain operable 4.1.11 Operations - Determine if PBDS A or B will remain operable 4.1.12 LPRM in OPERATE 4.1.14 LPRM in BYPASS Date/Time 4.1.15.4 Affected APRM, LPRMs in operate. Number:

Step Instructions Performer Concurrence Operations - Adjust APRM as required As Found As Found As Left As Left Within 4.1.17 APRM CTP APRM CTP  ! 2% RTP 4.1.18 Return APRM to OPERATE (as applicable) 4.1.19 Date WR #

Comments Return to the On-Duty Reactor Engineer when Complete.

Step Instruction On-Duty Reactor Engineer/ Date 4.1.20 Update the LPRM BYPASS LOG REP-0037 REV - 15 PAGE 13 OF 14

REFERENCE USE ATTACHMENT 3 PAGE 1 OF 1 LPRM RESTORE Step Name (print) Signature Initials 4.2.2 Step Instruction Initials 4.2.3 Contact Reactor Eng, RE(contacted):

4.2.4 LPRM 4.2.5 APRM 4.2.6 Obtain permission to restore LPRM / Bypass APRM (OSM/CRS)

Step Instructions Performer Concurrence 4.2.8.3 Number of LPRMs in operate. Number:

4.2.9 LPRM in BYPASS 4.2.11 LPRM in OPERATE Date/Time 4.2.12.4 Number of LPRMs in operate. Number:

Adjust LPRM (yes / no) (Rx Eng) As Left 4.2.14 Desired Value Operations - Adjust APRM as required As Found As Found As Left As Left Within 4.2.17 APRM CTP APRM CTP  ! 2% RTP 4.2.18 Return APRM to OPERATE 4.2.19 Notify RE to determine if OD-1 is required Comments Return to the On-Duty Reactor Engineer when Complete.

Step Instruction On-Duty Reactor Engineer /Date 4.2.20 Update the LPRM BYPASS LOG REP-0037 REV - 15 PAGE 14 OF 14

RJPM-NRC-M14-P1 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Alternate Control Building Chilled Water Pumps within the Standby Division OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 15 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform X Plant X Simulate Simulator Control Room Prepared: Dave Bergstrom Date: September 11, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 9

RJPM-NRC-M14-P1 Rev. 0 EXAMINER INFO SHEET Task Standard: Chilled Water is lined up to HVK Chiller D using SOP-0066, Section 5.3 Synopsis: This task will swap the standby chiller from B to D using SOP-0066, Control Building HVAC Chilled Water System This JPM is written for the field portion of the task which alternates the chilled water system only.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to perform the local lineup for placing Control Building HVK Chiller D, in standby with HVK-P1D Chilled Water Pump. SOP-0066 has been completed through Step 5.3.3.

3) Initial Conditions:

HVK-CHL1A, Control Building Chiller A, is currently in service.

HVK-CHL1B Control Building Chiller B and 1HVK-P1B, Chilled Water Pump B, are lined up for standby operation.

The Unit Operator has placed 1HVK-CHL1B, CONTROL BLDG CHILLER B, in LOCKOUT and 1HVK-P1B, CHILLED WATER PUMP B, in STOP.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-P1 Rev 0 Page 2 of 9

RJPM-NRC-M14-P1 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Alternate Control Building Chilled 291011001001 290003 A4.01 3.2 / 3.2 Water Pumps within the Standby Division

REFERENCES:

APPLICABLE OBJECTIVES SOP-0066, Rev 311 RLP-STM-0402, Obj 4 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0066, Rev 313, Section 5.3 __9__

SIMULATOR CONDITIONS & SETUP:

1. NA - This is an In Plant JPM.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Chilled Water is lined up to HVK Chiller D using SOP-0066, Section 5.3.

RJPM-NRC-M14-P1 Rev 0 Page 3 of 9

RJPM-NRC-M14-P1 Rev. 0 PERFORMANCE:

START TIME:

1. 5.3.4 Locally at the chiller which is currently in standby, unlock and close the

• Procedure Step:

chiller inlet valve.

HVK-V84, HVK CHL1B INLET ISOL Standard Applicant located/identified the Chill Water Inlet Valve for B HVK Chiller Applicant unlocked and closed HVK-V84 by turning the handwheel fully clockwise using the chain.

Cue Inform the applicant that the handwheel is fully clockwise.

Notes Results SAT UNSAT

2. 5.3.5 Locally at the currently out of service chiller, open and lock the chiller

• Procedure Step:

inlet valve.

HVK-V88, HVK CHL1D INLET ISOL Standard Applicant located/identified the Chill Water Inlet Valve for D HVK Chiller Applicant opened and locked HVK-V88 by turning the handwheel fully counter clockwise using the chain.

Cue Inform the applicant that the handwheel is fully counter-clockwise.

Notes Results SAT UNSAT RJPM-NRC-M14-P1 Rev 0 Page 4 of 9

RJPM-NRC-M14-P1 Rev. 0 PROCEDURE NOTE Oil level can be lower than normal if service water temperature is low, greater than or equal to 65°F and less than or equal to 75°F.

For a non-operating chiller an oil level in or above the upper sight glass is normal. When idle, the level may be higher due to the absorption of refrigerant by the oil.

3. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

1. Lube oil level greater than 3/4 of lower sight glass.

Standard Applicant verified lube oil level within specification.

Cue Indicate an oil level in the lower half of the upper sight glass Notes Results SAT UNSAT

4. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

2. Lube oil temperature is greater than or equal to 120°F and less than or equal to 155°F.

Standard Applicant verified lube oil temperature within specification.

Cue Indicate an oil temperature of 140°F.

Notes Results SAT UNSAT

5. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

3. IF the READY Light is off, THEN depress the PUSH TO RESET PRETRIP ANNUNCIATOR Pushbutton.

Standard NA Cue Indicate that the READY light is lit.

Notes No applicant action is necessary for this step.

RJPM-NRC-M14-P1 Rev 0 Page 5 of 9

RJPM-NRC-M14-P1 Rev. 0

6. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

4. READY Light is on.

Standard Applicant verified the READY light is lit.

Cue Indicate the READY light is ON.

Notes The lights are difficult to see - not very bright.

Results SAT UNSAT

7. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

5. SAFETY CIRCUIT Light is on.

Standard Applicant verified the SAFETY CIRCUIT Light is on.

Cue Indicate the safety circuit light is on.

Notes Results SAT UNSAT

8. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

6. LOAD RECYCLE Light is on.

Standard Applicant verified the LOAD RECYCLE Light is on.

Cue Indicate the load recycle light is on.

Notes Results SAT UNSAT RJPM-NRC-M14-P1 Rev 0 Page 6 of 9

RJPM-NRC-M14-P1 Rev. 0

9. Procedure Step: 5.3.6 Locally at the chiller being placed in standby, check the following:

7. Refrigerant visible in evaporator sight glass.

Standard Applicant verified refrigerant level within specification.

Cue Indicate a refrigerant level in the evaporator sight glass.

Notes The sightglass is on the north end of the machine; it is yellow Results SAT UNSAT

10. Procedure Step: 5.3.7 Perform the following for the chiller being placed in standby:

1. Verify SWP-P3D, CHILLER D RECIRC SWP in AUTO.

Standard Applicant informed the Control Room that Steps 5.3.4 through 5.3.6 have been completed.

Cue Accept the report as a Control Room Operator.

Notes The rest of the steps in this section would be performed in the MCR.

Results SAT UNSAT Terminating Cue: Chilled Water is lined up to HVK Chiller D usingSOP-0066, Section 5.3 This completes this JPM.

STOP TIME:

RJPM-NRC-M14-P1 Rev 0 Page 7 of 9

RJPM-NRC-M14-P1 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-P1 Rev 0 Page 8 of 9

RJPM-NRC-M14-P1 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

HVK-CHL1A, Control Building Chiller A, is currently in service.

HVK-CHL1B Control Building Chiller B and 1HVK-P1B, Chilled Water Pump B, are lined up for standby operation.

The Unit Operator has placed 1HVK-CHL1B, CONTROL BLDG CHILLER B, in LOCKOUT and 1HVK-P1B, CHILLED WATER PUMP B, in STOP.

INITIATING CUE:

The CRS has directed you to perform the local lineup for placing Control Building HVK Chiller D, in standby with HVK-P1D Chilled Water Pump. SOP-0066 has been completed through Step 5.3.3.

RJPM-NRC-M14-P1 Rev 0 Page 9 of 9

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• CONTROL BUILDING HVAC CHILLED WATER SYSTEM (SYS #410)

PROCEDURE NUMBER: *SOP-0066 REVISION NUMBER: *313 Effective Date: *10/08/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER SOP-0066 REV - 313 PAGE 1 OF 87

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES ......................................................................................................................5 4 SYSTEM STARTUP..................................................................................................................6 4.1 Control Building Chilled Water Loop Fill and Vent ........................................................6 4.2 Control Building Chilled Water Loop Startup..................................................................10 5 SYSTEM OPERATION .............................................................................................................20 5.1 Alternating Control Building Chilled Water Pump within the Running Division............20 5.2 Alternating Control Building Chilled Water Pump within the Standby Division ............21 5.3 Alternating Control Building Chilled Water Pump and Chiller within the Standby Division...............................................................................................................21 5.4 Alternating Divisions of Control Building Chilled Water ................................................24 5.5 Cooling Control Building Chilled Water Loops with Service Water ...............................27 5.6 Control Building Compression Tank Makeup from Service Water .................................28 5.7 Removing Control Building Chilled Water Loops from Cooling by Service Water .................................................................................................................................29 5.8 Pressurizing HVK-TK1A(1B) ..........................................................................................30 5.9 Transferring Refrigerant to or from Division I Chillers; HVK-CHL1A/C, CONT BLDG CHILLER A/C ..........................................................................................30 5.10 Transferring Refrigerant to or from Division II Chillers; HVK-CHL1B/D, CONT BLDG CHILLER B/D ..........................................................................................34 5.11 Manual Overriding SWP-PVY32A,B,C,D ......................................................................37 5.12 Division I Control Building Chilled Water System Feed and Bleed ................................37 5.13 Division II Control Building Chilled Water System Feed and Bleed...............................39 6 SYSTEM SHUTDOWN.............................................................................................................40 6.1 Control Building Chilled Water Loop Shutdown .............................................................40 7 REFERENCES ...........................................................................................................................41 8 RECORDS..................................................................................................................................42 ATTACHMENT 1A - VALVE LINEUP - CONTROL BUILDING CHILLED WATER (SAFETY RELATED)......................................................................................................43 SOP-0066 REV - 313 PAGE 2 OF 87

CONTINUOUS USE ATTACHMENT 1B - VALVE LINEUP - CONTROL BUILDING CHILLER, HVK-CHL1A (SAFETY RELATED) ........................................................................................56 ATTACHMENT 1C - VALVE LINEUP - CONTROL BUILDING CHILLER, HVK-CHL1B (SAFETY RELATED) ........................................................................................58 ATTACHMENT 1D - VALVE LINEUP - CONTROL BUILDING CHILLER, HVK-CHIL1C (SAFETY RELATED).......................................................................................60 ATTACHMENT 1E - VALVE LINEUP - CONTROL BUILDING CHILLER, HVK-CHL1D (SAFETY RELATED) ........................................................................................62 ATTACHMENT 2 - INSTRUMENT & VALVE LINEUP - CONTROL BUILDING CHILLED WATER (SAFETY RELATED).....................................................................64 ATTACHMENT 3 - ELECTRICAL LINEUP - CONTROL BUILDING CHILLED WATER (SAFETY RELATED).......................................................................................73 ATTACHMENT 4 - CONTROL BOARD LINEUP - CONTROL BUILDING CHILLED WATER (SAFETY RELATED).......................................................................................78 ATTACHMENT 5 - NORMAL OPERATING PARAMETERS - CHILLERS ...............................82 ATTACHMENT 6 - CONTROLLER SETTING CORRELATIONS ..............................................85 ATTACHMENT 7 - METREX VALVE DIAGRAM.......................................................................87 SOP-0066 REV - 313 PAGE 3 OF 87

CONTINUOUS USE 1 PURPOSE 1.1 The purpose of this procedure is to provide the specific steps necessary to startup, operate, and shutdown the Control Building HVAC Chilled Water System.

2 PRECAUTIONS AND LIMITATIONS 2.1 Normally only one Control Building Chilled Water Chiller is in operation. The non-operating division has one chiller in standby. The other two chillers are locked out and valved out of service.

2.2 The HVK-CHL1A(B)(C)(D), CONTROL BUILDING VENT CHILLERS starting is inhibited as follows:

• A chiller is prevented from restarting, by an anti-cycle time delay, until 20 minutes have elapsed from the previous start. This time delay allows for cool down of the chiller motor prior to restart.

• A chiller can not be started until 2 1/2 minutes have elapsed since it was stopped.

This time delay allows sufficient time for the guide vanes to close and ensure no-load starting of the chiller.

2.3 Control Building Ventilation Chilled Water equipment capacities are as follows:

• Chilled Water Pumps - 100% each.

• Water Chillers - 100% each.

2.4 Even though the refrigerant is not toxic at normal temperatures, in heavy concentrations it displaces the air and can cause personnel suffocation.

2.5 When exposed to heater elements or flame, Freon refrigerant becomes highly toxic, and even in low concentrations may cause fatal or serious injury. If a flame or heating element exists in a room when Freon is present, the room or space must be evacuated. IF personnel are required to enter the space, THEN use of breathing apparatus is required until the space is ventilated and the air sampled.

2.6 The following flow transmitters require venting during HVK Loop Fill or if the chiller evaporator has been isolated and drained:

• HVK-FTY5A, B, C, and D

• HVK-FTX5A, B, C, and D SOP-0066 REV - 313 PAGE 4 OF 87

CONTINUOUS USE 2.7 The following flow transmitters require venting during SWP Loop Fill or if the chiller condenser has been isolated and drained:

• SWP-FT69A, B, C, and D 2.8 The following valves require venting per SOP-0018, Normal Service Water during SWP Loop Fill:

• SWP-PVY32A, B, C, and D 2.9 Placing HVK-CHL1B(D) LO FLOW TRIP LOGIC BYPASS key lock switch in the BYPASS position will provide a Flow Normal signal to the Division II chiller control logic allowing associated chiller operation without minimum chilled water or service water flow.

2.10 The following apply for the applicable electro-hydraulic actuated Borg-Warner or Metrexs valves in this system:

• WHEN SWP-PVY32A/B/C/D are required to be physically gagged open or failed open, THEN the applicable HVK-CHL1A/B/C/D is INOPERABLE and UNAVAILABLE. SWP-PVY32A/B/C/D must be capable of modulating to maintain the design criteria for condenser pressure.

• HVK-TV16A/B must be capable of modulating to maintain the design criteria for the Control Room environment and to maintain divisional HVC/HVK administrative OPERABILITY.

2.11 Control Building Chiller normal operating information can be found in Attachment 5, NORMAL OPERATING PARAMETERS - CHILLERS.

2.12 Control Building Chiller and Cable Vault/Switchgear temperature controller setting information can be found in Attachment 6, CONTROLLER SETTING CORRELATIONS.

3 PREREQUISITES 3.1 Check Control Building HVAC System aligned and ready to start per SOP-0058, Control Building HVAC System.

3.2 Check Makeup Water System in operation and aligned to the Chilled Water System per SOP-0099, Makeup Water System.

3.3 Check Normal Service Water System in operation and aligned to the Chilled Water System per SOP-0018, Normal Service Water.

SOP-0066 REV - 313 PAGE 5 OF 87

CONTINUOUS USE 3.4 Check the following electrical systems are in operation and aligned to the Control Building Chilled Water System:

3.4.1. 4.16KV per SOP-0046, 4.16KV System 3.4.2. 480VAC per SOP-0047, 480VAC System 3.4.3. 120VAC per SOP-0048, 120VAC System 3.4.4. 125VDC per SOP-0049, 125VDC System 3.5 Verify the system is lined up for startup.

4 SYSTEM STARTUP NOTE Filling and venting is necessary only when the removal or repair of a system component has allowed the intrusion of air into the system.

Fill/vent of individual components in this section can be performed in any order. Do not backfill from running division.

4.1 Control Building Chilled Water Loop Fill and Vent 4.1.1. At H13-P863, verify STOP Pushbuttons are depressed for the following:

• HVK-CHL1A(B), CONT BLDG CHILLER A(B)

• HVK-CHL1C(D), CONT BLDG CHILLER C(D) 4.1.2. At H13-P863, verify control switches for the following are in STOP:

• HVK-P1A(B), CHILLED WATER PUMP A(B)

• HVK-P1C(D), CHILLED WATER PUMP C(D)

SOP-0066 REV - 313 PAGE 6 OF 87

CONTINUOUS USE 5.2 Alternating Control Building Chilled Water Pump within the Standby Division 5.2.1. IF alternating standby Control Building Chilled Water Pump and Chiller, THEN Go To Section 5.3.

5.2.2. At H13-P863, perform the following:

1. Place the standby HVK-P1A(B)(C)(D), CHILLED WATER PUMP A(B)(C)(D) to STOP.

2. Place the previously out of service HVK-P1A(B)(C)(D) to AUTO.

3. Select the Chilled Water Pump from Step 2 on the CHILLED WATER PUMP SELECTOR for the standby division.

NOTE Alternating Control Building Chilled Water Pump and Chiller within the standby division will cause the following annunciator to alarm:

Division 1: Annunciator, P863-74A-C01, DIV 1 CONT BLDG CHILLED WATER SYS INOPERATIVE Divisiion 2: Annunciator, P863-74A-C06, DIV 2 CONT BLDG CHILLED WATER SYS INOPERATIVE 5.3 Alternating Control Building Chilled Water Pump and Chiller within the Standby Division 5.3.1. Check that the currently operating chiller has been running for at least 20 minutes.

NOTE The controls and indications in this section are located at H13-P863, unless otherwise specified.

The standby chiller is the chiller which is aligned for automatic start but is not operating.

5.3.2. Depress LOCKOUT on HVK-CHL1A(B)(C)(D), CONT BLDG CHILLER (A)(B)(C)(D) Lockout/Reset Pushbutton for the standby chiller.

5.3.3. Place the standby HVK-P1A(B)(C)(D), CHILLED WATER PUMP A(B)(C)(D) to STOP.

SOP-0066 REV - 313 PAGE 21 OF 87

CONTINUOUS USE 5.3.4. Locally at the chiller which is currently in standby, unlock and close chiller inlet valve.

• HVK-V35, HVK-CHL1A INLET ISOL

• HVK-V39, HVK-CHL1C INLET ISOL

• HVK-V84, HVK-CHL1B INLET ISOL

• HVK-V88, HVK-CHL1D INLET ISOL 5.3.5. Locally at the currently out of service chiller, open and lock the chiller inlet valve.

• HVK-V35, HVK-CHL1A INLET ISOL

• HVK-V39, HVK-CHL1C INLET ISOL

• HVK-V84, HVK-CHL1B INLET ISOL

• HVK-V88, HVK-CHL1D INLET ISOL 5.3.6. Locally at the chiller being placed in standby, check the following:

NOTE Oil level can be lower than normal if service water temperature is low, greater than or equal to 65°F and less than or equal to 75°F.

For a non-operating chiller an oil level in or above the upper sightglass is normal. When idle, the level may be higher due to the absorption of refrigerant by the oil.

1. Lube oil level greater than 3/4 of lower sight glass.

2. Lube oil temperature is greater than or equal to 120°F and less than or equal to 155°F.

3. IF the READY Light is off, THEN depress the PUSH TO RESET PRETRIP ANNUNCIATOR Pushbutton.

4. READY Light is on.

5. SAFETY CIRCUIT Light is on.

6. LOAD RECYCLE Light is on.

7. Refrigerant visible in evaporator sight glass.

SOP-0066 REV - 313 PAGE 22 OF 87

CONTINUOUS USE 5.3.7. Perform the following for the chiller being placed in standby:

1. Verify SWP-P3A(B)(C)(D), CHILLER A(B)(C)(D) RECIRC SWP in AUTO.

2. Verify HVK-CHL1A(B)(C)(D)PL, CHLD CPRSR LUBO in AUTO.

5.3.8. Select the desired Chilled Water Pump on the CHILLED WATER PUMP SELECTOR for the standby division.

5.3.9. Place HVK-P1A(B)(C)(D), CHILLED WATER PUMP A(B)(C)(D) selected in Step 5.3.8 to AUTO.

5.3.10. Select the chiller being placed in standby on the CHILLER SUPPLY FLOW ELEMENT SELECTOR Switch.

5.3.11. Depress RESET on HVK-CHL1A(B)(C)(D), CONT BLDG CHILLER A(B)(C)(D) Lockout/Reset Pushbutton for the chiller being placed in standby.

5.3.12. Depress STOP then RESET on HVK-CHL1A(B)(C)(D) Start/Stop/Reset Pushbutton for the chiller being placed in standby.

5.3.13. WHEN alternating Division I Chiller only, THEN on the back of EHS-MCC8A, perform the following:

1. Place HVK-CHL1APL(CPL), CHLD CPRSR LUBO for the standby chiller to ON.

2. Place HVK-CHL1CPL(APL) for the out of service chiller to OFF.

SOP-0066 REV - 313 PAGE 23 OF 87

CONTINUOUS USE NOTE Alternating divisions of Control Building Chilled Water will cause the following annunciator to alarm:

Alternating from Division 1 to Division 2:

H13-P863/74A/A01, CONTROL BLDG CHILLER 1A OR 1C AUTO TRIP H13-P863/74A/A02, CONTROL ROOM AHU 1A AUTO TRIP H13-P863/74A/B01, CONTROL BLDG CHILLER 1A OR 1C PRE-TRIP H13-P863/74A/B04, CONTROL ROOM AHU HIGH DISCHARGE TEMPERATURE H13-P863/74A/C01, DIV 1 CONT BLDG CHILLED WATER SYS INOPERATIVE H13-P863/74A/C02, STBY SWGR AIR HDLG UNIT 2A AUTO TRIP H13-P863/74A/D02, DIV 1 CONTROL BLDG VENT SYS INOP H13-P863/74A/B06, CONTROL BLDG CHILLER 1B OR 1D PRE-TRIP Alternating from Division 2 to Division 1:

H13-P863/74A/A06, CONTROL BLDG CHILLER 1B OR 1D AUTO TRIP H13-P863/74A/A07, CONTROL ROOM AHU 1B AUTO TRIP H13-P863/74A/B04, CONTROL ROOM AHU HIGH DISCHARGE TEMPERATURE H13-P863/74A/B06, CONTROL BLDG CHILLER 1B OR 1D PRE-TRIP H13-P863/74A/C06, DIV 2 CONT BLDG CHILLED WATER SYS INOPERATIVE H13-P863/74A/C07, STBY SWGR AIR HDLG UNIT 2B AUTO TRIP H13-P863/74A/D07, DIV 2 CONTROL BLDG VENT SYSTEM INOP H13-P863/74A/B01, CONTROL BLDG CHILLER 1A OR 1C PRE-TRIP 5.4 Alternating Divisions of Control Building Chilled Water 5.4.1. Check that the operating chiller has been running for at least 20 minutes.

NOTE The controls and indications in this section are located at H13-P863, unless otherwise specified.

CRITICAL STEP 5.4.2. Stop the running Control Building Chilled Water Pump.

SOP-0066 REV - 313 PAGE 24 OF 87

RJPM-NRC-M14-P2 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Cross Connect EDG Air Receivers Within a Single Division OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform X Plant X Simulate Simulator Control Room Prepared: Dave Bergstrom Date: September 11, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 9

RJPM-NRC-M14-P2 Rev. 0 EXAMINER INFO SHEET Task Standard: The Division 2 EDG Forward Air Start System pressurized the Rear Air Start System using SOP-0053, Section 5.5.2.

Synopsis: This task will align the Forward Air Start system compressor to pressurize the Rear Air Start system Air Receivers using SOP-0053, Standby Diesel Generator and Auxiliaries.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to cross-connect the Division 2 Emergency Diesel Generator air receivers in accordance with Section 5.5 of SOP-0053. Pressurize both air receivers to 220 psig.

3) Initial Conditions:

The Div 2 Emergency Diesel Generator rear air compressor, EGA-C4B, is not operable.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-P2 Rev 0 Page 2 of 9

RJPM-NRC-M14-P2 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Cross Connect EDG Air Receivers 400067004001 264000 K1.06 3.2 / 3.2 Within a Single Division 264000 K6.01 3.8 / 3.9 295003 AA1.02 4.2 /

4.3 REFERENCES

APPLICABLE OBJECTIVES SOP-0053, Rev 327 RLP-STM-0309S, Obj 2, 3 REQUIRED MATERIALS: SAFETY FUNCTION:

SOP-0053, Rev 327, Section 5.5 __6__

SIMULATOR CONDITIONS & SETUP:

1. NA - This is an In Plant JPM.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: The Division 2 EDG Forward Air Start System pressurized the Rear Air Start System using SOP-0053, Section 5.5.2.

RJPM-NRC-M14-P2 Rev 0 Page 3 of 9

RJPM-NRC-M14-P2 Rev. 0 PERFORMANCE:

START TIME:

SOP-0053, Emergency Diesel Generator and Auxiliaries Section 5.5.2, Division 2 Air System

1. 5.5.2.1 Close EGA-V3148 (V3140), FORWARD (REAR) AIR START

• Procedure Step:

SUPPLY ISOLATION VALVE for the inoperable compressor.

Standard Applicant identified rear air compressor as being inoperable in accordance with the initial conditions.

Applicant located/identified and closed the rear air start isolation valve (EGA-V3140) by turning the handwheel fully clockwise.

Cue Notes The valve is closed when the handwheel is fully clockwise. Also, this is a rising stem valve. (stem in)

Results SAT UNSAT

2. 5.5.2.2 Open EGA-V3142, REAR AIR START SUPPLY CROSS TIE

• Procedure Step:

ISOLATION VALVE.

Standard Applicant located/identified EGA-V3142, Cross Tie Isolation Valve.

Applicant opened EGA-V3142 by turning the handwheel fully counter-clockwise.

Cue Notes The valve is open when the handwheel is fully counter-clockwise. Also, this is a rising stem valve. (stem out)

Results SAT UNSAT RJPM-NRC-M14-P2 Rev 0 Page 4 of 9

RJPM-NRC-M14-P2 Rev. 0

3. 5.5.2.3 Open EGA-V3170, FORWARD AIR START SYSTEM CROSS-TIE

• Procedure Step:

ISOLATION VALVE.

Standard Applicant located/identified EGA-V3170, Cross Tie Isolation Valve.

Applicant opened EGA-V3170 by turning the handwheel fully counter-clockwise.

Cue Notes The valve is open when the handwheel is fully counter-clockwise. Also, this is a rising stem valve. (stem out)

Results SAT UNSAT

4. 5.5.2.4 Start the operable air compressor by placing EGA-C4B(C5B),

• Procedure Step:

REAR(FORWARD) START AIR COMPRESSOR to RUN.

Standard Applicant identified rear air compressor as being inoperable in accordance with the initial conditions.

Applicant located/identified and started the forward air compressor (EGA-C5B) by turning the switch to RUN.

Cue Notes Results SAT UNSAT

5. Procedure Step: 5.5.2.5 WHEN the desired pressure is reached in the cross tied air receivers, THEN stop the air compressor by placing EGA-C5B, FORWARD START AIR COMPRESSOR to OFF.

Standard Applicant observed the air pressure in the Rear Air Receivers.

When cued that pressure reads 220 psig, applicant stopped the forward air compressor (EGA-C5B) by turning the switch to STOP.

Cue Indicate that the rear and forward air receivers are both reading 220 psig.

Notes Receiver pressure can be read locally in the Diesel Room or at the control panel outside the Diesel Room.

Results SAT UNSAT RJPM-NRC-M14-P2 Rev 0 Page 5 of 9

RJPM-NRC-M14-P2 Rev. 0

6. Procedure Step: 5.5.2.6 Close EGA-V3170, FORWARD AIR START SYSTEM CROSS-TIE ISOLATION VALVE.

Standard Applicant located/identified EGA-V3170, Cross Tie Isolation Valve.

Applicant closed EGA-V3170 by turning the handwheel fully clockwise.

Cue Notes The valve is closed when the handwheel is fully clockwise. Also, this is a rising stem valve.

Results SAT UNSAT

7. Procedure Step: 5.5.2.7 Close EGA-V3142, REAR AIR START SUPPLY CROSS TIE ISOLATION VALVE.

Standard Applicant located/identified EGA-V3142, Cross Tie Isolation Valve.

Applicant closed EGA-V3142 by turning the handwheel fully clockwise.

Cue Notes The valve is closed when the handwheel is fully clockwise. Also, this is a rising stem valve.

Results SAT UNSAT

8. Procedure Step: 5.5.2.8 Open EGA-V3148 (V3140), FORWARD (REAR) AIR START SUPPLY ISOLATION VALVE for the inoperable compressor.

Standard Applicant identified rear air compressor as being inoperable in accordance with the initial conditions.

Applicant located/identified and opened the rear air start isolation valve (EGA-V3140) by turning the handwheel fully counter-clockwise.

Cue Notes The valve is open when the handwheel is fully counter-clockwise. Also, this is a rising stem valve.

Results SAT UNSAT RJPM-NRC-M14-P2 Rev 0 Page 6 of 9

RJPM-NRC-M14-P2 Rev. 0

9. Procedure Step: 5.5.2.9 Place the operable air compressor in AUTO.

Standard Applicant identified forward air compressor as being operable.

Applicant located/identified and manipulated the forward air compressor (EGA-C5B) switch to AUTO.

Cue Notes Results SAT UNSAT

10. Procedure Step: 5.5.2.10 Verify the restorations are independently verified and logged.

Standard Applicant informed the Control Room that the Rear Air Receivers of the Division 2 EDG has been pressurized using the Forward system in accordance with SOP-0053, Section 5.5.2.

Applicant requests and independent verifier.

Cue Accept the report as a Control Room Operator.

Accept the request for an independent verifier.

Notes Results SAT UNSAT Terminating Cue: The Division 2 EDG Forward Air Start System pressurized the Rear Air Start System using SOP-0053, Section 5.5.2.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-P2 Rev 0 Page 7 of 9

RJPM-NRC-M14-P2 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-P2 Rev 0 Page 8 of 9

RJPM-NRC-M14-P2 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

The Div 2 Emergency Diesel Generator rear air compressor, EGA-C4B, is not operable.

INITIATING CUE:

The CRS has directed you to cross-connect the Division 2 Emergency Diesel Generator air receivers in accordance with Section 5.5 of SOP-0053. Pressurize both air receivers to 220 psig.

RJPM-NRC-M14-P2 Rev 0 Page 9 of 9

CONTINUOUS USE

• G12.1.6 RIVER BEND STATION STATION OPERATING MANUAL

• SYSTEM OPERATING PROCEDURE

• STANDBY DIESEL GENERATOR AND AUXILIARIES (SYS#309)

PROCEDURE NUMBER: *SOP-0053 REVISION NUMBER: *327 Effective Date: *09/30/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION TRACKING NUMBER DETAILED DESCRIPTION OF CHANGES SOP-0053R326EC-B Clarified wording of 4th bullet of Step 5.4 Caution. Also relocated Caution to below section 5.4 Title.

SOP-0053 REV - 327 PAGE 1 OF 115

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE...................................................................................................................................4 2 PRECAUTIONS AND LIMITATIONS ....................................................................................4 3 PREREQUISITES ......................................................................................................................14 4 SYSTEM STARTUP..................................................................................................................15 4.1 Placing EGS-EG1A(B), STBY DIESEL GENERATOR A(B) in Standby .....................15 4.2 Placing EGS-EG1A(B), STBY DIESEL GENERATOR A(B) in Maintenance Mode .................................................................................................................................19 4.3 Barring/Air Rolling Standby Diesel .................................................................................21 4.4 Warming Up the Diesel Post Maintenance.......................................................................23 4.5 Non-Emergency Starting, Loading and Paralleling the Standby Diesel from the Control Room....................................................................................................................24 4.6 Non-Emergency Starting, Loading and Paralleling the Standby Diesel from the Local Control Panel ..........................................................................................................29 4.7 Manual Start of Standby Diesel with an Automatic Start Signal Present.........................34 5 SYSTEM OPERATION.............................................................................................................38 5.1 Paralleling an Offsite Power Source to the Standby Diesel from the Control Room.......38 5.2 Paralleling an Offsite Power Source to the Standby Diesel from the Local Control Panel..................................................................................................................................40 5.3 Operation from Automatic Start .......................................................................................43 5.4 Unloading Diesel Fuel Oil ................................................................................................44 5.5 Cross-Connecting Air Receivers Within a Single Division..............................................48 5.6 Cross Connecting Air Receivers Across Divisions ..........................................................50 5.7 Operation of Diesel Fuel Offload Berme Drain System Following Oily Contamination of Berme...................................................................................................53 5.8 Verifying Proper Operation of EGF-P1A(B), FUEL OIL TRANSFER PUMP...............54 5.9 Swapping Fuel Oil Transfer Pump EGF-P1A(B) Discharge Strainers.............................54 5.10 Swapping the Fuel Oil Duplex Filters. .............................................................................55 5.11 Swapping EGF-STR3A(B)/ STR3D(E) FUEL OIL STRAINER ....................................56 5.12 Swapping Lube Oil Duplex Filters EGO-FLT1A/1D (1B/1E).........................................56 5.13 Swapping Lube Oil Duplex Strainers ...............................................................................57 5.14 Restoration of the Standby Diesel from a Tripped Condition ..........................................58 SOP-0053 REV - 327 PAGE 2 OF 115

CONTINUOUS USE 5.15 Makeup Water Addition to Standby Diesel Jacket Water ................................................61 5.16 Operation of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) as the only power supply to the bus ....................................................................................................62 5.17 Manual Air Compressor Operation...................................................................................63 5.18 Draining Division I(II) Jacket Water Standpipe for Level Control ..................................64 6 SYSTEM SHUTDOWN.............................................................................................................65 6.1 Shutdown of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) from the Control Room....................................................................................................................65 6.2 Shutdown of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) from the Local Control Panel ..........................................................................................................68 6.3 Emergency Shutdown of the Diesel - Loss of Control Air...............................................71 6.4 Emergency Shutdown of the Diesel - Loss of DC Control Power ...................................72 6.5 Shutdown and Resetting of EGS-EG1A(B), STBY DIESEL GENERATOR A(B) with LOCA/LOP Signal Present.......................................................................................73 7 REFERENCES ...........................................................................................................................75 8 RECORDS..................................................................................................................................77 ATTACHMENT 1A - VALVE LINEUP - STANDBY DIESEL GENERATOR EG1A.................78 ATTACHMENT 1B - VALVE LINEUP - STANDBY DIESEL GENERATOR EG1B .................84 ATTACHMENT 2A - INSTRUMENT & VALVE LINEUP - STANDBY DIESEL GENERATOR EG1A .......................................................................................................90 ATTACHMENT 2B - INSTRUMENT & VALVE LINEUP - STANDBY DIESEL GENERATOR EG1B .......................................................................................................94 ATTACHMENT 3A - ELECTRICAL LINEUP - STANDBY DIESEL GENERATOR EG1A ................................................................................................................................97 ATTACHMENT 3B - ELECTRICAL LINEUP - STANDBY DIESEL GENERATOR EG1B.................................................................................................................................100 ATTACHMENT 4A - CONTROL BOARD LINEUP - STANDBY DIESEL GENERATOR EG1A ................................................................................................................................103 ATTACHMENT 4B - CONTROL BOARD LINEUP - STANDBY DIESEL GENERATOR EG1B.................................................................................................................................107 ATTACHMENT 5 - ENGINE PARAMETERS ...............................................................................111 ATTACHMENT 6 - KW VS KVAR (.8PF) .....................................................................................112 ATTACHMENT 7 - FUEL OIL DAY TANK LEVEL INDICATION ............................................113 ATTACHMENT 8 - FUEL OIL STORAGE TANK LEVEL INDICATION ..................................114 ATTACHMENT 9 - FITTING LOCATION FOR SECTION 6.5.2.................................................115 SOP-0053 REV - 327 PAGE 3 OF 115

CONTINUOUS USE 1 PURPOSE 1.1 The purpose of this procedure is to outline the steps necessary to startup, operate and shutdown EGS-EG1A(B), STBY DIESEL GENERATOR A(B).

2 PRECAUTIONS AND LIMITATIONS 2.1 All instructions are written for EGS-EG1A, STBY DIESEL GENERATOR A with nomenclature for EGS-EG1B, STBY DIESEL GENERATOR B in parenthesis.

2.2 High crankcase pressure indicates the possible existence of an explosive gas mixture. Allow the engine to cool for 15 minutes to allow fumes and vapors to dissipate before removing any engine covers. With the exhaust fan running the atmospheric pressure in the room will be lower and the manometers will read higher than specified.

2.3 Placing the diesel in MAINTENANCE mode requires simultaneous operation of the STBY DIESEL ENGINE MODE switch on H13-P877 and the MAINT MODE SELECT switch on EGS-PNL3A(B). Similarly, to place the diesel back in OPERATIONAL mode requires simultaneous operation of the STBY DIESEL ENGINE MODE switch and the RETURN TO OPERATIONAL pushbutton on EGS-PNL3A(B).

2.4 When the diesel is returned to the OPERATIONAL mode the FIELD FLASHING RELAY READY light located on EGE-CAB01A(B) should be lit. If not, the Exciter Shutdown Relay may have failed to reset.

2.5 EGS-EG1A(B), STBY DIESEL GENERATOR A(B) has a continuous rating of 3130 Kw at 0.8 power factor. Do not operate the diesel generator with a power factor of less than 0.8 when operating in parallel with other sources, and do not exceed 3130 Kw load.

2.6 ERIS computer points will be used to ensure voltage and watt limits are not exceeded.

Frequency will be recorded using the MCR or the Local control room meter; however voltage and watt readings will be obtained from the ERIS points. Voltage and watt meters can be used for adjustments when not at the control band limits.

2.7 Diesel Generator Governor oil level shall be checked Prior to, During and Following any diesel run. Acceptable oil levels are greater than the fill mark during standby conditions and visible in the sightglass while operating.

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CONTINUOUS USE 2.8 Prior to manually starting the diesel for other than emergency conditions, the engine should be barred over 2 revolutions and air rolled with the cylinder cocks open to insure the cylinders are clear unless the start is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the last engine shutdown. Roll the engine per Section 4.3 if required. The origin of any water detected in a cylinder must be determined and any cylinder head that leaks due to a crack shall be replaced.

(Ref. 7.21, 7.23) 2.8.1. Air rolls do not apply to any engine that has been removed from service after a run. However, the engine shall be rolled over with the airstart system at the time it is returned to service.

2.8.2. The OPERABLE engines are not required to be air-rolled if the plant is already in an Action Statement of Technical Specification 3.8.1 and 3.8.2.

2.9 Ensure that the rear air system is available prior to attempting any Barring / Air Rolling of a Diesel generator. The barring device and Air Roll function is supplied from rear air only.

2.10 Parallel the Diesel Generator to the Standby Bus with the synchroscope rotating slowly in the "fast" (clockwise) direction. Do not attempt to close a diesel generator output breaker with the synchroscope indicator standing still, if there is power available to the bus from another source.

2.11 If the diesel is run for one hour or greater, check and drain from the day tank any accumulated water via EGF-V11(V41), DAY TANK TK2A(2B) DRAIN.

2.12 Lube oil must be added only through the fill connection on the sump. Do not overfill the sump.

2.13 If EGS-EG1A(B), STBY DIESEL GENERATOR A(B) is declared inoperable, refer to Technical Specification 3.8.1 and 3.8.2.

2.14 Never have 2 synchroscopes in the same division on at the same time.

2.15 If the diesel generator is paralleled with the standby bus normal or alternate breaker and a LOCA signal occurs, the diesel generator output breaker will open. The diesel generator breaker can not be closed as long as bus voltage is being supplied by the normal or alternate supply and the LOCA signal still exists.

2.16 Sustained operation of the engine at critical speeds of 190, 285, 350 and 415 RPM should be avoided. (Ref. 7.19) 2.17 If a diesel start signal is activated while the diesel is not available, the signal will remain sealed in. If the diesel is then made available, the diesel engine will auto start. To prevent this, if Control air pressure is greater than 45 psig then Section 6.5.2 must be performed. If pressure is less than 45 psig, then the EMERGENCY START RESET switch on H13-P877 must be depressed before returning the diesel to Operational.

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CONTINUOUS USE 2.18 During a Station Blackout with the Div. 1 or 2 Diesel Generator failing to deliver power to their respective buses due to a malfunction of the Excitation System, (when diesel engine has attained rated speed) the Field Flashing of the failed D/G should be secured to conserve the battery, and to prevent heating the excitation cabinet.

2.19 Short duration runs and light load (less than 40%, or 1200 kW) operation should be avoided.

After a period of light load or no-load run, the diesel should be loaded to greater than or equal to 2700 kW, for a time period as specified below:

2.19.1. At least one hour, if the engine was run at less than 1200 kW for greater than 30 minutes but less than one hour, OR 2.19.2. At least two hours, if the engine was run at less than 1200 kW for equal to or greater than one hour but less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, OR 2.19.3. At least four hours, if the engine was run at less than 1200 kW for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or longer. (Ref. 7.22; CR-RBS-2004-3156) 2.20 To minimize crankshaft torsional stresses, continuous engine operation at critical speeds shall not be allowed. Minimize the time the engine is operated between 453 and 457 RPM (60.4 to 60.9 Hz). (Ref. 7.19) 2.21 Engine cylinder exhaust gas temperature should be within 75°F of the average for all cylinders. Any cylinder temperature exceeding this limit should be investigated by maintenance.

2.22 Prelube of the engine should be performed before all non-emergency starts.

2.23 Per System Engineering the following conditions should be used to determine if a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is required to allow the engine mass and crankshaft temperatures to equalize, prior to any normal Diesel Generator start:

2.23.1. When after re-energizing the Lube Oil or Jacket Water Heaters from a de-energized state, the Lube Oil and Jacket Water outlet temperatures are greater than 140°F and 115°F respectively with less than or equal to a 40°F differential a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is not required prior to any normal start. This temperature criteria should be used for Diesel Generator outages of 3 days or less.

2.23.2. If the engine block has been allowed to cool to ambient temperature, such as for maintenance with the heaters de-energized for more than 3 days, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> warm up is required with the Lube Oil Circulating Pump, Jacket Water Circulating Pump, and associated heaters operating prior to any normal start.

2.23.3. If necessary contact the System Engineer for guidance in determination of temperature criteria.

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CONTINUOUS USE 2.23.4. Use of Section 4.4, Warming Up the Diesel Post Maintenance should be minimized. Warming up by slow start and controlled slow loading is intended for situations where insufficient time is available to allow the heater to warm the system, and an expedited return to service is critical due to plant conditions or shutdown LCO status. Although EC 5759 was written for schedule preservation, and the method poses very little risk (maybe slightly more wear and tear on the Diesel over the long term), this method should not be made a matter of routine.

Operations management or Duty Manager should approve use of Section 4.4.

2.24 Anytime work is done on the fuel oil day tank level instrumentation, the control switch for the fuel oil transfer pump must be placed in "OFF" to preclude pumping fuel oil to the roof.

2.25 If the forward starting subsystem DC control power is lost, the diesel engine will still be able to start, but there will be NO tripping capability.

2.26 If the diesel is started automatically on a LOCA, all automatic shutdowns are bypassed except overspeed and generator differential. The reinstatement of all trips following an automatic start requires the following:

2.26.1. DIV 2 -Depress the RHR DIV 2 INITIATION RESET pushbutton (H13-P601 INSERT 17B).

DIV 1 - Depress the LPCS/RHR DIV 1 INITIATION RESET pushbutton (H13-P601 INSERT 21B) 2.27 If the diesel is started automatically on a Loss of Power (LOP) or manually started using either of the STBY DIESEL ENGINE EMERGENCY START pushbuttons, all automatic shutdowns are bypassed except overspeed, generator differential, jacket water out high temperature, and lube oil out high temperature. The reinstatement of all trips following an above described start requires the following:

2.27.1. LOP - Depress STBY DIESEL ENGINE A(B) EMERGENCY START RESET pushbutton.

2.27.2. REMOTE MANUAL EMERGENCY START(CR) - Depress STBY DIESEL ENGINE A(B) EMERGENCY START RESET.

2.27.3. LOCAL PANEL MANUAL EMERGENCY PB - Diesel must be shutdown.

2.27.4. (LOCAL, EGS-PNL4A) - for LOP 2.28 The EXCITATION SHUTDOWN RESET pushbutton should only be used on a loss of excitation with the engine still running. The excitation shutdown will normally auto-reset when the engine is stopped.

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CONTINUOUS USE 2.29 The FIELD FLASHING RELAY READY white light will be energized when the K1 relay is in the reset (closed) position and the DG is not at voltage. It will deenergize when K1 is in the exciter shutdown position, the voltage relay contacts open (DG at voltage), or when the pressure switches are closed (DG at speed).

2.30 The FIELD FLASHING RELAY READY white light should be verified to be energized after any Exciter Shutdown Reset operation or any time the DG is placed in a Standby lineup.

2.31 Operating data pertaining to all diesel generator start attempts shall be obtained per PEP-0026, Diesel Generator Operating Logs.

2.32 Visual daily inspection between adjacent cylinder heads and the general block top are required during any period of continuous operation following automatic diesel generator startup. (L/C 3.3).

2.33 Whenever the Diesel is being shutdown, adjust Generator frequency to 59.7 Hz (Div I DG) or 60 Hz (Div II DG) after the Generator output breaker has been opened, prior to stopping the engine. (Ref. 7.19) 2.34 Before the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> air roll after an engine run, drain any liquid from the Turbo Charger Casing Drain per PEP-0026.

2.35 When the diesel is running in parallel with the grid, a fault on the grid could cause a loss of the bus associated with the diesel concurrent with a trip/lockout of the diesel. To reduce the chances of this occurring, time spent with the diesel paralleled to the grid should be minimized. (Ref. 7.11) 2.36 Duplex lube oil and fuel oil filters and strainers should be swapped while the engine is running, if at all possible. It may be loaded or unloaded, isochronous or synchronous with the grid. System pressures should be checked after the swap. If it is necessary to swap a duplex filter or strainer while shut down, the engine should be started in test mode (normal start) and run long enough to check that pressures are normal.

2.37 During diesel fuel oil unloading, a fire watch shall be stationed at the unloading area and two (2) 150 pound dry chemical extinguishers placed near unloading area. (Ref. 7.17) 2.38 EGS-EG1A(B), STBY DIESEL GENERATOR A(B) shall not be run in parallel with the Main Generator through STX-XNS1C, NORM STA XFMR. (Ref. 7.15) 2.39 Failing to de-energize control power to the K1 relay prior to depressurizing D/G control air will result in the inability of the K1 relay to auto reset when control air is restored. The K1 relay must be manually reset if this condition occurs. Control air should be restored prior to reenergizing control power to the K1 relay. (Ref. 7.36) 2.40 If desired, EGS-EG1A(B), STBY DIESEL GENERATOR A(B) may be started and run using only one air receiver tank. (Ref. TSI-015)

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CONTINUOUS USE 2.41 EGA-C4A (C5A)(C4B)(C5B) will not operate in AUTO if its associated start air receiver pressure is less than 30 psig. Placing the control switch to RUN will allow the compressor to start. This should only be used for initial startups and repressurizing. DO NOT repressurize the air receiver should pressure fall to less than 25 psig while the diesel generator is loaded, this could cause a trip or uncontrolled loading of the diesel generator.

2.42 When operating in the NORMAL mode, if a TRIP annunciator(s) should come in, and the diesel does not trip, immediately check the amber UNIT TRIPPED light. If the UNIT TRIPPED light is ON, STOP the diesel. If the light is OFF, evaluate the annunciator(s) via other instrumentation. If the trip condition does NOT exist or can not be verified, attempt to RESET the annunciators(s). If the alarm can not be reset and the diesel has run for 2 minutes, manually stop the diesel. (Ref. 7.12) 2.43 The diesel generator will continue to run without control air pressure in emergency conditions. Upon complete loss of Control air pressure, it should not be restored until the diesel is shutdown.

2.44 If the diesel is paralleled with the grid and a Ground Fault Trip/Lockout occurs, the diesel will not Auto start on an Emergency Auto signal. Refer to Section 4.7. (Ref. 7.13) 2.45 When operating the diesel generator at reduced loads, care should be exercised to avoid reverse power trips.

2.46 While in MODES 1, 2 & 3 placing an OPERABLE Emergency Diesel Generator into

'MAINTENANCE' mode causes the diesel to be INOPERABLE. STP-000-0102, Power Distribution Alignment Check shall be performed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> unless the diesel is restored to OPERABLE status in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. (Ref. 7.29) 2.47 During movement of recently irradiated fuel assemblies in the Primary Containment or Fuel Building AND while in Modes 4 and 5, placing an operable Diesel Generator into MAINTENANCE mode causes the Diesel Generator to be inoperable. Tech Spec 3.8.2, AC Sources - Shutdown, shall be immediately referred to and the action requirements complied with. (Ref. 7.29) 2.48 Non-essential 125 VDC controls are fed from the BYS batteries. The BYS battery chargers are lost on a LOCA or LOP. Use AOP-0014, Loss of 125 VDC for what is lost if the BYS batteries are not available. Safety-related control functions are not affected.

2.49 Non-essential 120 VAC controls are fed from SCA-PNL15A1(B1). In the event of a LOP, EGS-TI64A(B), multi-point (Doric) temperature indicator will be lost. Therefore upon loss of this indicator the operator should use local thermometer readings. Refer to Attachment 5, Engine Parameters for alternate indications. Safety-related control functions are not affected.

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CONTINUOUS USE 2.50 Loss of FORWARD DC power prevents the following indicator lights from coming on:

• UNIT AVAIL EMERGENCY STATUS

• DC CONTROL POWER ON

• UNIT TRIPPED

• READY TO LOAD 2.51 The electric signal that actuates the Maintenance mode is momentary. MAINTENANCE mode is retained by a self-sealing pilot on pneumatic control valve EGS-PNL3A(B)-P2.

There is no electrical seal-in. The RETURN TO OPERATIONAL signal energizes a solenoid valve, which opens a vent path to break this pneumatic seal-in, and P-2 defaults to the OPERATIONAL position. If while in MAINTENANCE mode, the control panel is depressurized, the P-2 self seal-in is lost, and the control system will come up in the OPERATIONAL mode when pressure is restored.

2.52 While running in the test mode, any manual EMERGENCY START signal will activate the governor and voltage regulator pre-position circuits which return the frequency to near 59.7 Hz (Div I DG) or 60 Hz (Div II DG) and voltage to near 4160 v. The output breaker is not signaled to trip. If the DG happens to be synchronous and loaded, the net affect will be a loss of kw and kvar, over a 6 to 10 second period.

2.53 The Diesel Engine oil sump level dip stick indications are as follows:

STANDBY (Diesel Generator not running with keep warm oil pumps running):

• Maintain the oil sump level greater than or equal to the T7 Mark per Tech Spec 3.8.3 and less than or equal to the FULL Mark.

• As long as the oil level is maintained greater than the LOW STBY mark, the Diesel is capable of safely starting and running if required for an emergency.

RUNNING:

• Maintain oil sump level greater than or equal to the LOW RUN Mark.

• The T6 and T7 Marks are not used for oil sump level when the diesel is running.

C 2.54 Special requirements for restoring from Diesel Engine Maintenance/Tagouts:

• If the engine fuel oil system has been tagged out and drained for maintenance, the fuel oil lines shall be refilled by manually operating the DC Fuel Oil Pump for 1 to 2 minutes. This should be done promptly after releasing the tagout to ensure the lines are full before any diesel starts.

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CONTINUOUS USE 2.55 When ENS-SWG1A or B is deenergized such as during a bus outage, the undervoltage relays generate a Loss of Power (LOP) signal to start the associated Standby Diesel Generator. If the diesel is tagged out, the LOP signal will seal in and can go undetected until the tagout is cleared. Therefore, to prevent an auto start of the diesel upon diesel restoration, prior to restoring the diesel, relays R3A and R3B in the side panels at EGS-PNL3A(B) should be checked to ensure that the LOP start is not present. The relay is tripped if the red button in the middle of the relay is not flush (recessed) with the case.

Refer to Section 6.5 of this procedure to reset the relays. (Ref. 7.37) 2.56 When reviewing this procedure for pending operations or system configuration realignments, ensure vulnerabilities to common cause and common mode failures are evaluated for current plant conditions to protect safety sources and safety trains. (SOER 03-1 Recommendation 2 Emergency Power Reliability)(Ref. 7.41) 2.57 When the Diesel is operating synchronized to the grid, the diesel generator shall be declared inoperable. This is because if a Loss Offsite Power were to occur during operations when synchronized to the grid the resultant operations with the diesel powering the Div I(II) bus will cause the diesel frequency to be outside TS 3.8.1.2 and 3.8.1.7 frequency limits.

(Ref 7.39)

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CONTINUOUS USE 2.58 Starting 4.16kV and certain 480VAC loads while the DG is parallel to off-site power can result in the diesel output breaker tripping on overload condition (Ref. 7.42).

2.58.1. To prevent exceeding the maximum load rating of 3130kW when EGS-EG1A is paralleled to the off-site power supply, manual start of equipment on the following switchgears should not be permitted:

• ENS-SWG1A

• EJS-SWG1A

• EJS-SWG2A Additionally if NNS-SWG1A is being powered from RTX-XSR1C, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1A

• NNS-SWG4A (and NNS-SWG4B if cross-tied)

Additionally if NNS-SWG1C is being powered from NNS-SWG1A AND NNS-SWG1A is being powered from RTX-XSR1C, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1C

• E22-S004 Auto start of 4.16kV loads during parallel load testing may result in overload of EGS-EG1A or trip of ENS-ACB07, STBY D/G A OUTPUT BRKR.

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CONTINUOUS USE 2.58.2. To prevent exceeding the maximum load rating of 3130kW when EGS-EG1B is paralleled to the off-site power supply, manual start of equipment on the following switchgears should not be permitted:

• ENS-SWG1B

• EJS-SWG1B

• EJS-SWG2B Additionally if NNS-SWG1B is being powered from RTX-XSR1D, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1B

• NNS-SWG4B (and NNS-SWG4A if cross-tied)

Additionally if NNS-SWG1C is being powered from NNS-SWG1B AND NNS-SWG1B is being powered from RTX-XSR1D, manual start of equipment on the following switchgears should not be permitted:

• NNS-SWG1C

• E22-S004 Auto start of 4.16kV loads during parallel load testing may result in overload of EGS-EG1B or trip of ENS-ACB27, STBY D/G B OUTPUT BRKR.

2.59 The GERB viscous damper is not required for operability of the Div I or Div II Diesel Generator. Issues with the GERB should be identified and reported via the Condition Report process. (Ref. 7.43) 2.60 Annunciator, P877-31A(32A)-C03, ENS*SWG1A(B) SPLY OR DIST BRKR INOPERATIVE may alarm momentarily whenever ENS-ACB07(27), STBY D/G A(B)

OUTPUT BREAKER is manipulated.

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CONTINUOUS USE 3 PREREQUISITES 3.1 The Fire Protection Water System to the Standby Diesel Generator EGS-EG1A(B) Room is in service per SOP-0037, Fire Protection Water System Operating Procedure.

3.2 The Makeup Water System is available for makeup to the Jacket Water Standpipe per SOP-0099, Makeup Water System.

3.3 The Normal Service Water System is operating per SOP-0018, Normal Service Water.

3.4 The Standby Service Water System is operable per SOP-0042, Standby Service Water System.

3.5 The following electrical systems are operable:

3.5.1. 4160VAC per SOP-0046, 4.16 KV System (except on loss of power start) 3.5.2. 480VAC per SOP-0047, 480 VAC System 3.5.3. 120VAC per SOP-0048, 120 VAC System 3.5.4. 125VDC per SOP-0049, 125 VDC System 3.6 Diesel Generator Building HVAC in operation per SOP-0061, Diesel Generator Building Ventilation.

3.7 Obtain copy of PEP-0026, Diesel Generator Operating Logs for use in all start attempts.

3.8 The Instrument Air System is operable per SOP-0022, Instrument Air System.

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CONTINUOUS USE 5.4.16. IF fuel oil spillage has occurred in the berme, THEN drain per Section 5.7 of this procedure.

5.4.17. IF fuel oil spillage did not occur AND Section 5.7 was not performed, THEN realign the Berme Drain System to the Storm Drain System as follows:

1. Open SRW-V3001, D/G BLDG FUEL OIL OFFLOAD TRUCK BERME ISOLATION VALVE.

2. Open SRW-V3003, D/G BLDG BERME DRAIN TO STORM DRAIN SYSTEM.

5.4.18. Log completion of the fuel oil offload and manipulated device line up along with independent verification in the MCR log.

5.4.19. After offloading diesel fuel oil, notify the Control Room/Work Management Center to order another load of diesel fuel oil if required.

5.5 Cross-Connecting Air Receivers Within a Single Division 5.5.1. Division 1 Air System

1. Close EGA-V3130(V3122), FORWARD (REAR) AIR START SUPPLY ISOLATION VALVE for the inoperable compressor.

2. Open EGA-V3124, REAR AIR START SUPPLY CROSS TIE ISOLATION VALVE.

3. Open EGA-V3169, FORWARD AIR START SYSTEM CROSS TIE ISOLATION VALVE.

4. Start the operable air compressor by placing EGA-C4A(C5A),

REAR(FORWARD) START AIR COMPRESSOR to RUN.

5. WHEN the desired pressure is reached in the cross tied air receivers, THEN stop the air compressor by placing EGA-C4A(C5A), REAR(FORWARD)

START AIR COMPRESSOR to OFF.

6. Close EGA-V3169, FORWARD AIR START SYSTEM CROSS TIE ISOLATION VALVE.

7. Close EGA-V3124, REAR AIR START SUPPLY CROSS TIE ISOLATION VALVE.

8. Open EGA-V3130(V3122), FORWARD (REAR) AIR START SUPPLY ISOLATION VALVE for the inoperable compressor.

9. Place the operable air compressor in AUTO.

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CONTINUOUS USE

10. Verify the restorations are independently verified and logged.

5.5.2. Division 2 Air System

1. Close EGA-V3148(V3140), FORWARD (REAR) AIR START SUPPLY ISOLATION VALVE for the inoperable compressor.

2. Open EGA-V3142, REAR AIR START SUPPLY CROSS TIE ISOLATION VALVE.

3. Open EGA-V3170, FORWARD AIR START SYSTEM CROSS-TIE ISOLATION VALVE.

4. Start the operable air compressor by placing EGA-C4B(C5B),

REAR(FORWARD) START AIR COMPRESSOR to RUN.

5. WHEN the desired pressure is reached in the cross tied air receivers, THEN stop the air compressor by placing EGA-C4B(C5B), REAR(FORWARD)

START AIR COMPRESSOR to OFF.

6. Close EGA-V3170, FORWARD AIR START SYSTEM CROSS-TIE ISOLATION VALVE.

7. Close EGA-V3142, REAR AIR START SUPPLY CROSS TIE ISOLATION VALVE.

8. Open EGA-V3148(V3140) FORWARD (REAR) AIR START SUPPLY ISOLATION VALVE for the inoperable compressor.

9. Place the operable air compressor in AUTO.

10. Verify the restorations are independently verified and logged.

SOP-0053 REV - 327 PAGE 49 OF 115

RJPM-NRC-M14-P3 Rev 0 NUCLEAR PLANT OPERATOR JOB PERFORMANCE MEASURE SRO RO ALTERNATE PATH TITLE: Align Instrument Air System (IAS) to Safety Relief Valve Air System (SVV)

OPERATOR: DATE:

EVALUATOR: EVALUATOR SIGNATURE:

CRITICAL TIME FRAME: Required Time (min): NA Actual Time (min): NA PERFORMANCE TIME: Average Time (min): 20 Actual Time (min):

JPM RESULTS*: (Circle one)

• SAT UNSAT Refer to Grading Instructions at end of JPM EVALUATION METHOD: EVALUATION LOCATION:

Perform X Plant X Simulate Simulator Control Room Prepared: Dave Bergstrom Date: September 11, 2013 Reviewed: Jeff Reynolds Date: January 22, 2014 (Operations Representative)

Approved: Joey Clark Date: January 27, 2014 (Facility Reviewer)

River Bend Station Initial License Exam Page 1 of 14

RJPM-NRC-M14-P3 Rev. 0 EXAMINER INFO SHEET Task Standard: Backup air is supplying at least 101 psig air pressure to the SVV system using AOP-0050, Attachment 6.

Synopsis: During a station blackout event, power is lost to the SVV Compressors. This JPM provides a method for providing a backup source of air to the SRVs.

This task will align the Instrument Air System (IAS) to the Safety Relief Valve Air System (SVV) using AOP-0050, Station Blackout. An Alternate Path is taken when the SVV Header fails to pressurize.

NOTE: If in the Plant or the Control Room, Caution the operator NOT to MANIPULATE the controls, but to make clear what they would do if this were not a simulated situation.

1) Read to the operator:

I will provide the initial conditions and initiating cues to you. I may also provide cues during the performance and ask follow-up questions at the conclusion of this JPM.

When you complete the task successfully, the objective for this JPM will be satisfied.

Inform me when you have completed the task.

2) Initiating Cues:

The CRS has directed you to provide backup air from the Instrument Air System (IAS) to the SRVs per Attachment 6 of AOP-0050.

3) Initial Conditions:

A station blackout is in progress.

SRVs are required to be cycled to stabilize reactor pressure The IAS Diesel Air Compressor is operating, lined up to supply air to the Instrument Air System.

4) Solicit and answer any questions the operator may have.

RJPM-NRC-M14-P3 Rev 0 Page 2 of 14

RJPM-NRC-M14-P3 Rev. 0 DATA SHEET TASK

Title:

Task Number K&A SYSTEM: K&A RATING:

Align the Instrument Air System (IAS) 278001004004 218000 A2.03 3.4 / 3.6 to the Safety Relief Valve Air System 300000 K4.02 3.0 / 3.0 (SVV) in accordance with AOP-0050, 295003 AK1.06 3.8 / 4.0 Attachment 6.

REFERENCES:

APPLICABLE OBJECTIVES AOP-0050, Rev 48 RLP-HLO-541, Obj 7 REQUIRED MATERIALS: SAFETY FUNCTION:

AOP-0050, Rev 48, Attachment 6 __8__

SIMULATOR CONDITIONS & SETUP:

1. NA - This is an In Plant JPM.

CRITICAL ELEMENTS: Items marked with an

• are Critical Steps and are required to be performed. Failure to successfully complete a Critical Step requires the JPM to be evaluated as Unsatisfactory.

TASK STANDARD: Backup air is supplying at least 101 psig air pressure to the SVV system using AOP-0050, Attachment 6 RJPM-NRC-M14-P3 Rev 0 Page 3 of 14

RJPM-NRC-M14-P3 Rev. 0 PERFORMANCE:

START TIME:

AOP-0050, Station Blackout Attachment 6, IAS Diesel Air Compressor Backup to SVV Header Procedure Step: 1. Contact the Aux Control Room, verify Diesel Air Compressor is lined up and operating to supply air to the IAS System.

Standard NA Cue Notes Applicant has no actions to perform in this step per initial conditions.

1. Procedure Step: 2. Proceed to Turbine Building 95 ft el, northeast corner, next to the Auxiliary Building door, and perform the following:

Standard Applicant arrived at the designated location.

Cue Notes Results SAT UNSAT

2. Procedure Step: 2.1 Open AOP-0050, Station Blackout, IAS Diesel Air Compressor Backup to SVV Header Supply Kit, and verify the kit contains the following equipment:

Standard Applicant located and identified the kit and its contents.

Cue Do not allow the applicant to open the kit.

Notify the applicant that all required equipment is present and that movement of equipment will be simulated.

Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 4 of 14

RJPM-NRC-M14-P3 Rev. 0

3. Procedure Step: 2.2 Move the needed contents of this kit to the Aux Building 141 ft el.

Standard Applicant arrived at the designated location.

Cue Notes Results SAT UNSAT PROCEDURE NOTE It is preferable to supply the SVV header via the IAS system per Section 3 of this attachment rather than supplying each SRV division separately from IAS per Section 4.

4. Procedure Step: 3.1 On IAS-V345, IAS SUPPLY ROOT, behind HVR-FLT2, next to door to SBGT Train A, install a 90° elbow with tee fitting.

Standard Applicant located/identified IAS-V345 and simulated installing the elbow and T-fitting.

Cue Notes Valve is located about 7 foot off the floor.

Results SAT UNSAT

5. Procedure Step: 3.2 At SVV-V3000, AIR HEADER X-CONNECT DRAIN, located five feet northeast of PVLCS Skid A Accumulator Tank, install a quick connect fitting Standard Applicant located/identified SVV-V3000 and simulated installing the quick connect fitting.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 5 of 14

RJPM-NRC-M14-P3 Rev. 0

6. Procedure Step: 3.3 Make a hose connection between IAS-V345, IAS SUPPLY ROOT and SVV-V3000, AIR HEADER X-CONNECT DRAIN.

Standard Applicant stated/simulated connecting a hose between the IAS-V345 and SVV-V3000.

Cue Notes Results SAT UNSAT

7. Procedure Step: 3.4 Close the bleed valve on the tee fitting at IAS-V345, IAS SUPPLY ROOT.

Standard Applicant located/identified stated that the bleed valve on the tee fitting has been closed by turning in the clockwise direction until motion stopped.

Cue Notes Results SAT UNSAT

8. Procedure Step: 3.5 OPEN IAS-V345, IAS SUPPLY ROOT.

Standard Applicant has stated that valve IAS-V345 has been opened by turning the handwheel in the counter-clockwise direction until valve motion stopped.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 6 of 14

RJPM-NRC-M14-P3 Rev. 0

9. Procedure Step: 3.6 OPEN SVV-V3000, AIR HEADER X-CONNECT DRAIN Standard Applicant has stated that valve SVV-V3000 has been opened by turning the handwheel in the counter-clockwise direction until valve motion stopped.

Cue Notes Results SAT UNSAT

10. Procedure Step: 3.7 At the SVV Dryer Panel, observe SVV header pressure on SVV-PI38A or 38B. IF SVV header pressure can NOT be maintained greater than 101 psig, THEN disconnect the hookup per steps 3.8.1 through 3.8.6 and proceed to section 4 of this attachment.

Standard Applicant located/identified SVV-PI38A or 38B.

{Cue}

Applicant transitions to step 3.8.1 Cue Indicate a pressure of 90 psig Notes Results SAT UNSAT ALTERNATE PATH:

11. Procedure Step: 3.8.1 Close IAS-V345, IAS SUPPLY ROOT.

Standard Applicant has stated that valve IAS-V345 has been closed by turning the handwheel in the clockwise direction until valve motion stopped.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 7 of 14

RJPM-NRC-M14-P3 Rev. 0

12. Procedure Step: 3.8.2 Close SVV-V3000, AIR HEADER X-CONNECT DRAIN Standard Applicant has stated that valve SVV-V3000 has been closed by turning the handwheel in the clockwise direction until valve motion stopped.

Cue Notes Results SAT UNSAT

13. Procedure Step: 3.8.3 At IAS-V345, IAS SUPPLY ROOT open the bleed valve on the tee fitting.

Standard Applicant located/identified stated that the bleed valve on the tee fitting has been opened by turning in the counter-clockwise direction until motion stopped.

Cue Notes Results SAT UNSAT

14. Procedure Step: 3.8.4 Disconnect the hose from IAS-V345, IAS SUPPLY ROOT and SVV-V3000, AIR HEADER X-CONNECT DRAIN.

Standard Applicant stated/simulated disconnecting a hose between the IAS-V345 and SVV-V3000.

Cue Notes Results SAT UNSAT

15. Procedure Step: 3.8.5 Remove the tee fitting from IAS-V345, IAS SUPPLY ROOT.

Standard Applicant simulated removing the elbow and T-fitting.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 8 of 14

RJPM-NRC-M14-P3 Rev. 0

16. Procedure Step: 3.8.6 At SVV-V3000, AIR HEADER X-CONNECT DRAIN, remove the quick connect fitting.

Standard Applicant simulated removing the quick connect fitting.

Cue Notes Results SAT UNSAT Section 4 Supplying each SRV Division Separately from IAS.

17. Procedure Step: 4.1 On IAS-V345, IAS SUPPLY ROOT, behind HVR-FLT2, next to door to SBGT Train A, install a 90° elbow with cross fitting.

Standard Applicant located/identified IAS-V345 and simulated installing the elbow and cross fitting.

Cue Notes IAW step 10 of this JPM, applicant performed procedure steps 3.8.1 through 3.8.6 then proceeds to step 4.

Results SAT UNSAT

18. Procedure Step: 4.2 At SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION, located 4 feet southeast of LPCS Injection Valve, E21-MOVF005, install a quick connect fitting.

Standard Applicant located/identified SVV-V48 and simulated installing the quick connect fitting.

Cue Notes Best access for the valve is using the LPCS pit ladder; can be seen best from the top of the other pit (to the east).

Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 9 of 14

RJPM-NRC-M14-P3 Rev. 0

19. Procedure Step: 4.3 At SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION, located 2 feet south of HPCS Injection Valve E22-MOVF004, install a quick connect fitting.

Standard Applicant located/identified SVV-V51 and simulated installing the quick connect fitting.

Cue Notes Results SAT UNSAT

20. Procedure Step: 4.4 Make a hose connection between IAS-V345, IAS SUPPLY ROOT and SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

Standard Applicant stated/simulated connecting a hose between the IAS-V345 and SVV-V48.

Cue Notes Results SAT UNSAT

21. Procedure Step: 4.5 Make a hose connection between IAS-V345, IAS SUPPLY ROOT and SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

Standard Applicant stated/simulated connecting a hose between the IAS-V345 and SVV-V51.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 10 of 14

RJPM-NRC-M14-P3 Rev. 0

22. Procedure Step: 4.6 On the cross fitting at IAS-V345, IAS SUPPLY ROOT, close the bleed valve.

Standard Applicant stated that the bleed valve has been closed by turning in the clockwise direction until motion stopped.

Cue Notes Results SAT UNSAT

23. Procedure Step: 4.7 Open IAS-V345, IAS SUPPLY ROOT.

Standard Applicant stated that valve IAS-V345 has been opened by turning the handwheel in the counter-clockwise direction until valve motion stopped.

Cue Notes Results SAT UNSAT

24. Procedure Step: 4.8 Verify the shutoff valves on the cross fittings are closed.

Standard Applicant stated that the shutoff valves on the cross fittings at SVV-V48 and SVV-V51 are verified shut by turning the knob to the clockwise position until motion stopped.

Cue Notes Results SAT UNSAT RJPM-NRC-M14-P3 Rev 0 Page 11 of 14

RJPM-NRC-M14-P3 Rev. 0

25. Procedure Step: 4.9 Open SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

Standard Applicant simulated opening SVV-V48 by turning the handwheel in the counter-clockwise direction until motion stopped.

Cue Notes Results SAT UNSAT

26. Procedure Step: 4.10 Open SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

Standard Applicant simulated opening SVV-V51 by turning the handwheel in the counter-clockwise direction until motion stopped.

Cue Notes Results SAT UNSAT

27. Procedure Step: 5.5.2.10 Slowly open one or both shutoff valves on the cross fitting to supply either or both headers as needed. IF sufficient pressure can not be obtained with both shutoff valves open, THEN open only one shutoff at a time to develop sufficient pressure and flow to maintain the SRV pressure greater than 101 psig.

Standard Applicant has stated that the Shutoff valves on cross fitting are open Cue Due to space constraints in the area of the pressure gauge, it is acceptable to verbally provide the pressure reading versus pointing to the value.

When applicant has located SVV-PI38A or SVV-PI38B notify him that the gauge reads 115 psig.

Notes Results SAT UNSAT Terminating Cue: Backup air is supplying at least 101 psig air pressure to the SVV system using AOP-0050, Attachment 6.

This completes this JPM.

STOP TIME:

RJPM-NRC-M14-P3 Rev 0 Page 12 of 14

RJPM-NRC-M14-P3 Rev. 0 JPM COMMENT SHEET CRITERIA FOR SATISFACTORY EVALUATION

1. 100% of critical elements/steps identified in the JPM successfully completed.

2. Critical Time Frame is met if applicable.

3. No actual safety violation (radiological or industrial) requiring evaluator intervention.

CRITERIA FOR UNSAT EVALUATION

1. Any critical element/step is graded as "UNSAT"

2. Critical Time Frame is not met if applicable. *

3. Actual safety violation (radiological or industrial) requiring evaluator intervention.

4. Operator's actions would have damaged plant equipment, created a personnel safety hazard, or otherwise reduced the level of safety of the plant RJPM-NRC-M14-P3 Rev 0 Page 13 of 14

RJPM-NRC-M14-P3 Rev. 0 OPERATOR CUE SHEET INITIAL CONDITIONS:

A station blackout is in progress.

SRVs are required to be cycled to stabilize reactor pressure The IAS Diesel Air Compressor is operating, lined up to supply air to the Instrument Air System.

INITIATING CUE:

The CRS has directed you to provide backup air from the Instrument Air System (IAS) to the SRVs per Attachment 6 of AOP-0050.

RJPM-NRC-M14-P3 Rev 0 Page 14 of 14

CONTINUOUS USE

• G12.1.7 RIVER BEND STATION STATION OPERATING MANUAL

• ABNORMAL OPERATING PROCEDURE

• STATION BLACKOUT PROCEDURE NUMBER: *AOP-0050 REVISION NUMBER: *048 Effective Date: *09/24/2013 NOTE : SIGNATURES ARE ON FILE.

• INDEXING INFORMATION

CONTINUOUS USE TABLE OF CHANGES LETTER DESIGNATION TRACKING NUMBER DETAILED DESCRIPTION OF CHANGES AOP-0050 REV - 048 PAGE 1 OF 84

CONTINUOUS USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE/DISCUSSION ..........................................................................................................3 2 SYMPTOMS ..............................................................................................................................3 3 AUTOMATIC ACTIONS ..........................................................................................................3 4 IMMEDIATE OPERATOR ACTIONS .....................................................................................4 5 SUBSEQUENT OPERATOR ACTIONS ..................................................................................5 6 REFERENCES ...........................................................................................................................18 ATTACHMENT 1 - EMERGENCY MANUAL START OF DIESEL GENERATORS ................19 ATTACHMENT 2 - INJECTION INTO RPV WITH FIRE WATER SYSTEM DURING STATION BLACKOUT...................................................................................................28 ATTACHMENT 3 - DC LOAD SHEDDING...................................................................................31 ATTACHMENT 4 - CONTROL ROOM PANEL AIR CIRCULATION ........................................34 ATTACHMENT 5 - CONTAINMENT/DRYWELL/STEAM TUNNEL ACTIONS......................35 ATTACHMENT 6 - IAS DIESEL AIR COMPRESSOR BACKUP TO SVV HEADER................43 ATTACHMENT 7 - SUPPRESSION POOL TEMPERATURE DETERMINATION ....................47 ATTACHMENT 8 - EMERGENCY MAKEUP OF WATER TO THE SUPPRESSION POOL ................................................................................................................................51 ATTACHMENT 9 - USING THE DIV 3 DIESEL GENERATOR TO SUPPLY POWER TO ENS-SWG1A..............................................................................................................55 ATTACHMENT 10 - USING DIV 3 DIESEL GENERATOR TO SUPPLY POWER TO ENS-SWG1B ....................................................................................................................63 ATTACHMENT 11 - RELAY LOCATION AND TERMINAL IDENTIFICATION IN E22-S004 AUXILIARY COMPARTMENT....................................................................71 ATTACHMENT 12 - RHR C OR RCIC ROOM ENTRY TO CLOSE DOORS AB-078-01 AND AB-095-03..........................................................................................................72 ATTACHMENT 13 - TIME CRITICAL ACTIONS ........................................................................74 ATTACHMENT 14 - ALTERNATE POWER TO HYDROGEN IGNITORS................................79 AOP-0050 REV - 048 PAGE 2 OF 84

CONTINUOUS USE 1 PURPOSE/DISCUSSION 1.1 This procedure provides instructions in event of a loss of all offsite power combined with a failure of both Division 1 and 2 Diesel Generators to start and/or supply their emergency buses.

1.2 For a listing of instruments which remain in service, refer to AOP-0042, Loss of Instrument Bus.

1.3 Upon restoration of Division 1 or 2 Diesel Generator, refer to AOP-0004, Loss of Offsite Power.

2 SYMPTOMS 2.1 Loss of all equipment except that supplied by the DC buses. Division 3 equipment may or may not be available.

3 AUTOMATIC ACTIONS 3.1 Reactor scram occurs as a result of loss of power to the RPS buses.

3.2 Full MSIV isolation occurs as a result of loss of power to the RPS buses.

3.3 RCIC initiates when RPV level lowers to Level 2.

3.4 SRVs cycle to augment RCIC for RPV pressure control.

3.5 A Diesel Fire Pump starts at 110 psig fire header pressure.

AOP-0050 REV - 048 PAGE 3 OF 84

CONTINUOUS USE ATTACHMENT 6 PAGE 1 OF 4 IAS DIESEL AIR COMPRESSOR BACKUP TO SVV HEADER DISCUSSION During a station blackout event, power is lost to the SVV Compressors. This attachment provides for a means of providing a backup source of air to the SRVs, allowing continued use of the SRVs for pressure control. This meets the intent of TRM 3.6.1.8 for both PVLCS subsystems inoperable.

INSTRUCTIONS:

1 Contact the Auxiliary Control Room, verify Diesel Air Compressor is lined up and operating to supply air to the IAS system.

2 Proceed to Turbine Building 95 ft el, northeast corner, next to the Auxiliary Building door, and perform the following:

2.1 Open AOP-0050, Station Blackout, IAS Diesel Air Compressor Backup To SVV Header Supply Kit, and verify the kit contains the following equipment:

• 250 feet of air hose, with quick connect fittings & whip restraints

• Two 24 inch pipe wrenches

• Two 3/4 inch quick connect fittings

• One 90° elbow to cross fitting with 2 quick connect fittings and bleed/shutoff valves

• One 90° elbow to tee fitting with 1 quick connect fitting and a bleed valve

• One extra miscellaneous 3/4 inch and 1 inch fittings 2.2 Move the needed contents of this kit to Auxiliary Building 141 ft el.

NOTE It is preferable to supply the SVV header via the IAS system per Section 3 of this attachment rather than supplying each SRV division separately from IAS per Section 4.

3 Supplying the SVV Supply Header from IAS.

3.1 On IAS-V345, IAS SUPPLY ROOT, behind HVR-FLT2, next to door to SBGT Train A, install a 90° elbow with tee fitting.

3.2 At SVV-V3000, AIR HEADER X-CONNECT DRAIN, located five feet northeast of PVLCS Skid A Accumulator Tank, install a quick connect fitting.

AOP-0050 REV - 048 PAGE 43 OF 84

CONTINUOUS USE ATTACHMENT 6 PAGE 2 OF 4 IAS DIESEL AIR COMPRESSOR BACKUP TO SVV HEADER 3.3 Make a hose connection between IAS-V345, IAS SUPPLY ROOT and SVV-V3000, AIR HEADER X-CONNECT DRAIN.

3.4 Close the bleed valve on the tee fitting at IAS-V345, IAS SUPPLY ROOT.

3.5 Open IAS-V345, IAS SUPPLY ROOT.

3.6 Open SVV-V3000, AIR HEADER X-CONNECT DRAIN.

3.7 At the SVV Dryer Panel, observe SVV header pressure on SVV-PI38A or 38B. IF SVV header pressure can NOT be maintained greater than 101 psig, THEN disconnect the hookup per Steps 3.8.1 through 3.8.6 and proceed to Section 4 of this attachment.

3.8 WHEN normal operation of the SVV compressors is restored and/or both PVLCS subsystems are operating, THEN shutdown the backup IAS supply as follows:

3.8.1. Close IAS-V345, IAS SUPPLY ROOT.

3.8.2. Close SVV-V3000, AIR HEADER X-CONNECT DRAIN.

3.8.3. At IAS-V345, IAS SUPPLY ROOT, open the bleed valve on the tee fitting.

3.8.4. Disconnect the hose from IAS-V345, IAS SUPPLY ROOT and SVV-V3000, AIR HEADER X-CONNECT DRAIN.

3.8.5. Remove the tee fitting from IAS-V345, IAS SUPPLY ROOT.

3.8.6. At SVV-V3000, AIR HEADER X-CONNECT DRAIN, remove the quick connect fitting.

3.8.7. Return all items to AOP-0050, Station Blackout, IAS Diesel Air Compressor Backup To SVV Header Supply Kit.

4 Supplying each SRV Division Separately from IAS.

4.1 On IAS-V345, IAS SUPPLY ROOT, behind HVR-FLT2 next to door to SBGT Train A, install a 90° elbow with cross fitting.

4.2 At SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION, located 4 feet southeast of LPCS Injection Valve, E21-MOVF005, install a quick connect fitting.

4.3 At SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION, located 2 feet south of HPCS Injection Valve, E22-MOVF004, install a quick connect fitting.

AOP-0050 REV - 048 PAGE 44 OF 84

CONTINUOUS USE ATTACHMENT 6 PAGE 3 OF 4 IAS DIESEL AIR COMPRESSOR BACKUP TO SVV HEADER 4.4 Make a hose connection between IAS-V345, IAS SUPPLY ROOT EL and SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

4.5 Make a hose connection between IAS-V345, IAS SUPPLY ROOT and SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

4.6 On the cross fitting at IAS-V345, IAS SUPPLY ROOT, close the bleed valve.

4.7 Open IAS-V345, IAS SUPPLY ROOT.

4.8 Verify the shutoff valves on the cross fitting are closed.

4.9 Open SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

4.10 Open SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

4.11 Slowly open one or both shutoff valves on the cross fitting to supply either or both headers as needed. IF sufficient pressure can not be obtained with both shutoff valves open, THEN open only one shutoff at a time to develop sufficient pressure and flow to maintain the SRV pressure greater then 101 psig.

4.12 WHEN normal operation of the SVV compressors is restored and/or both PVLCS subsystems are operating, THEN shutdown the backup IAS supply as follows:

4.12.1. Close IAS-V345, IAS SUPPLY ROOT.

4.12.2. Close SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

4.12.3. Close SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

4.12.4. At IAS-V345, IAS SUPPLY ROOT, open both shutoff valves on the cross fitting.

4.12.5. At IAS-V345, IAS SUPPLY ROOT, open the bleed valve on the cross fitting.

4.12.6. Disconnect the hose from between IAS-V345, IAS SUPPLY ROOT and SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION.

4.12.7. Disconnect the hose from between IAS-V345, IAS SUPPLY ROOT and SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION.

4.12.8. Remove the cross fitting from IAS-V345, IAS SUPPLY ROOT.

AOP-0050 REV - 048 PAGE 45 OF 84

CONTINUOUS USE ATTACHMENT 6 PAGE 4 OF 4 IAS DIESEL AIR COMPRESSOR BACKUP TO SVV HEADER 4.12.9. At SVV-V48, HDR A OUTBOARD LEAKAGE MONITORING CONNECTION, remove the quick connect fitting.

4.12.10. At SVV-V51, HDR B OUTBOARD LEAKAGE MONITORING CONNECTION, remove the quick connect fitting.

4.12.11. Replace any pipe caps removed while implementing for this procedure.

4.12.12. Return all items to AOP-0050, Station Blackout, IAS Diesel Air Compressor Backup To SVV Header Supply Kit.

AOP-0050 REV - 048 PAGE 46 OF 84

Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: 1 IC No.: 210 Examiners: Theresa Buchanan Operators: _____________________________

Steve Garchow _____________________________

Mike Bloodgood _____________________________

Initial Conditions: 100% reactor power.

STP-257-0201 is in progress Turnover Shift priorities: 1) Perform STP-509-0101.

Event Malf. Event Event No. No. Type* Description 1 NA N (SRO, ATC) Perform STP-509-0101, Bypass Valve Operability Test 2 NMS011F I (SRO, ATC) APRM F fails upscale (Tech Spec) 3 GMC002A I (SRO, ATC) Stator Cooling Water Pump A trips, B fails to auto start GMC001B B21005 C (SRO, BOP) 4 B21-PTN078A, (Reactor Pressure) fails high (Tech Spec) 5 MSS005N C (SRO, BOP) Safety Relief Valve 51G fails open (Tech Spec) 6 NA M (ALL) Reactor scram with uncontrolled pressure drop 7 FWS004B C (SRO, ATC) FW master controller output fails high (triggered 5 minutes after Mode Switch) 8 MSS018 C (SRO, BOP) B21-AOVF022A & B21-F028A fail to close MSS019 with MSS004 Steam Leak outside primary containment

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (6) APRM F, Stator Cooling trip, Rx Pressure Instrument Failure SRV open, FW Master controller failure, MSIV failure Malfunctions after EOP entry (1-2) (2) FW Master controller, B21-AOVF022A & B21-AOVF028A Abnormal Events (2-4) (4) AOP-0035, AOP-0001, AOP-0002, AOP-0003 Major Transients (1-2) (1) Reactor scram with uncontrolled pressure drop EOPs entered (1-2) (2) EOP-0001, EOP-0002 EOP contingencies (0-2) (1) Enter EOP-0002 Critical Tasks (2-3) (2) Mode switch to Shutdown by 105°F; Close Main Steam Stop Valve, B21-F098A

Number: *RSMS-NRC14-1 RIVER Revision: 01 BEND STATION Page 1 of 15 Approximate Time: 1 Hour(s)

SIMULATOR SCENARIO Record Type: *Z01.24 TRAINING PROGRAM:

SIMULATOR TRAINING LESSON PLAN:

• Loss of Stator Cooling / Stuck Open SRV /

Steam Leak Outside Containment REASON FOR REVISION:

NRC March 2014 exam PREPARE / REVIEW:

Angie Orgeron 1538 10-07-2013 Preparer KCN Date Dave Bergstrom 0257 10-28-2013 Technical Review (SME) KCN Date Jeff Reynolds 0358 1-22-2014 Operations Representative KCN Date Joey Clark 0260 1-27-2014 Facility Reviewer KCN Date

• Indexing Information

I. DESCRIPTION OF SCENARIO This scenario begins with the plant at 100% power.

Events for this scenario:

Perform STP-509-0101, Bypass Valve Op Test.

APRM F Upscale failure (Tech Specs)

Stator Cooling Water Pump A trips with no auto start for standby pump B.

Reactor Pressure Instrument Fails High (Tech Specs)

SRV 51G fails open (Tech Specs); requiring power reduction Suppression Pool Temperature rises to the point of requiring a SCRAM; Continued pressure drop.

Operators close MSIVs but 22A & 28A stuck open.

FWS Master Controller output fails high 3 minutes after Mode Switch to SD.

Steam Leak Outside Primary Containment requires closing the 98A.

II. TERMINAL OBJECTIVES

1. Establish safe and stable plant conditions following a stuck open SRV and steam leak outside containment per plant procedures.

RSMS-NRC14-1 Page 2 OF 15

IV. INITIAL CONDITIONS/SHIFT TURNOVER INITIAL TRAINING EQUIPMENT STATUS REQUIRED CONDITION FOCUS DOCUMENTS IC #210 Power: 100% STP-509-0101 Core: Xenon equilibrium Marked up STP-257-0201 Equipment OOS: None STPs Due: 509-0101, Turbine Bypass Valve Cycle Test LCOs: None Evolutions in progress: 257-0201, SBGT Op Test.

Note: this is a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run and will not be completed during this scenario Problem/Lit annunciators: None RSMS-NRC14-1 Page 3 OF 15

V. GENERAL INSTRUCTIONS Event Number MFS-OR-REM-SCH Expected Operator Actions Simulator Setup Check Boards for Malfunctions Equip Tags GMC001B, Stator Cooling Pump B fails to Auto Start Check procedures and hard cards for marks MSS018, B21-AOVF022A fails to close.

MSS019, B21-AOVF028A fails to close Check Gauges/Meters for marks.

Overrides Remote Functions Make marked-up copies T6 LO_B21-F051G-R (and -G) OFF, T12 Encl 12, RPS of STPs available. (simulates removing fuses)

T16 Encl 16, Instrument Air Check that the EVENT TRIGGERS T18 Encl 18, Level 8 Jumpers Shutdown Plan is appropriate for this T2 NMS011F APRM F Fails Upscale scenario. T3 GMC002A, Stator Cooling Water Pump A Trip Check power <3090 MWth T4, B21005, B21-PTN078A Fails Upscale T5 MSS005N SRV 51G sticks open Bring up Insight -

MSTun Temperature T7 FWS004A Master Controller output fails high (3 min after Mode Switch to SD).

T8 MSS004 Steam Leak Outside Primary Containment (0.45% triggered to MSIV 22A switch)

RSMS-NRC14-1 Page 4 OF 15

Event Number MFS-OR-REM-SCH Expected Operator Actions Simulator Setup Event 0 RUN CREW: Board walk down / Turnover.

Event 1 SRO Direct the ATC to perform STP-509-0101.

Perform STP-509-0101, Turbine Bypass Valve Cycle Test.

Event initiated by ATC Accept the direction to perform STP-509-0101.

crew from turnover Perform STP-509-0101.

[Simul DR for C85AM6 (#2 BPV current)]

sheet.

RSMS-NRC14-1 Page 5 OF 15

Event 2 T2 NMS011F APRM F Fails Upscale ATC Recognize and report APRM failure APRM F fails upscale. Verify no individual rod scrams Event trigger T2 ROLE PLAY: As work management, SRO Accept report from ATC of APRM failure.

initiated at Lead maintenance, and/or Reactor Engineering, Notify maintenance of APRM failure; Evaluator discretion. accept report of failed APRM.

complete OSP-0046 notifications; notify Reactor Engineering.

Recognize Potential LCO:

o TS 3.3.1.1 Condition A o TS 3.3.2.1 Direct ATC to place APRM F to bypass and reset half-scram.

UO Verify indications of APRM F in Backpanel ATC When directed, place APRM F to bypass and reset half scram.

RSMS-NRC14-1 Page 6 OF 15

GMC001B, Stator Cooling Pump B fails UO Event 3 Recognize and report Stator Cooling Water Pump to Auto Start Auto Trip annunciator.

Stator Cooling Pump T3 GMC002A, Stator Cooling Water trip with no auto start Pump A Trip Recognize Standby Pump (B) failed to auto start of standby pump. and take actions to start the B pump.

Event trigger T3 Refer to ARP-870-54A-D01.

initiated at Lead ROLE PLAY Dispatch turbine building operator to investigate.

Evaluator discretion.

As the turbine building operator, accept SRO Accept the report of the tripped pump and direct, if Time the direction to investigate the Stator Water Cooling Pump. necessary, the manual start of the standby pump.

Call Back ________ Call back in 5 minutes to report that the Notify work management or maintenance of pump bearing on the pump end of the motor is trip and request OSP-0046 notifications.

hot to the touch.

ATC Recognize and report Turbine Runback (until the B Stator Water Cooling Pump is started)

If B pump not started promptly, reduce power as necessary to preclude a reactor SCRAM (using Recirc Flow).

RSMS-NRC14-1 Page 7 OF 15

Event 4 T4, B21005, B21-PTN078A Fails Upscale ATC Recognize and report Reactor pressure indication Reactor Pressure failure Transmitter Instrument Refer to ARP Failure. ROLE PLAY: Report 1/2 Scram and no single rod scram Event trigger T4 As back panel operator: When requested, SRO Accept report initiated at Lead report that the trip unit indicates full Enter Tech Spec 3.3.1.1 A Evaluator discretion. upscale and trip light on. (Place in trip condition in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

Enter Tech Spec 3.3.6.1 A As I&C technician, report the failure (Place in trip condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) appears to be the transmitter.

Make notifications per OSP-46 Request Work Control address instrument failure.

UO Verify conditions in backpanels Report that B21-PTN078A is in gross fail RSMS-NRC14-1 Page 8 OF 15

Event 5 T5 MSS005N SRV 51G sticks open ATC When directed, reduce power to 90%

SRV opens/sticks open.

UO Recognize and report the open SRV Event T5 initiated at Lead Evaluator Perform action of AOP-0035:

ROLE PLAY discretion. Backpanel - actions to attempt closing Place SRV 51G switch to OPEN.

SRV 51G:

Accept request to cycle switch at P631 to Make a Plant Announcement OPEN and then to OFF. When power 90%, take control switch for SRV 51G to OFF; cycle switch to OPEN and back to Accept request to cycle switch at P631 to OFF.

OPEN and then to OFF.

Go to backpanel (phone) and follow instructions in Accept request to remove fuses from P628, step 5.5.3 of AOP-0035 to cycle switch twice.

Bay B, B21C-F113A and F114A. Monitor Suppression Pool Temperature.

T6 LO_B21-F051G-R OFF, Remove fuses IAW step 5.7 of AOP-35 Attach. 1.

T6 LO_B21-F051G-G OFF, Determine that SRV does not close by using steam (simulates removing fuses) flows, etc.

Accept request to remove fuses from P631, SRO Accept report about the SRV.

Bay B, B21C-F113B and F114B. Assign AOP-0035, Stuck Open SRV.

Direct the ATC to reduce reactor power to 90%.

When available, direct UO to place RHR A or B in Supp Pool Cooling using the Hard Card.

Refer to EIP-2-001 for possible applicability.

Enter EOP-0002, on suppression pool level.

RSMS-NRC14-1 Page 9 OF 15

Event 6 CRITICAL TASK: Insert a Manual SRO Give Crew Brief in anticipation of manual Reactor Reactor Scram BEFORE Suppression Pool SCRAM due to High Supp Pool Temperature.

Reactor SCRAM Temperature reaches 105°F.

Direct the ATC to insert a manual reactor scram.

Uncontrolled Pressure Drop after Reactor Accept the SCRAM Report SCRAM Enter EOP-1 and direct EOP-1 actions:

o Verify mode switch in S/D (SCRAM Report) o ATC - restore and maintain RPV water level from -20 to 51 inches with Feed & Condensate.

o ATC - stabilize reactor pressure below 1090 psig, then give band of 500-1090 psig.

ATC - assigned AOP-0001, Rx Scram and AOP-0002, Turbine Trip.

Monitor HCTL.

Re-enter EOP-0002 on suppression pool temperature; direct EOP-2 actions:

o UO - initiate Suppression Pool Cooling.

RSMS-NRC14-1 Page 10 OF 15

Event 6 (contd) ATC When directed, insert a manual SCRAM.

Provide a SCRAM Report.

Restore and maintain RPV water level:

T12 Encl 12, RPS -20 to 51 inches with Feed & Condensate.

T16 Encl 16, Instrument Air Recognize and Report an uncontrolled pressure T18 Encl 18, Level 8 Jumpers drop in the RPV due to the stuck open SRV.

When below 500 psig, notify the CRS that we are out of pressure band low.

Complete actions of AOP-1, Reactor Scram and AOP-2, Turbine Trip.

NOTE: When T8 is activated, then SRO Direct UO to CLOSE MSIVs (outboard preferred)

REMOVE the T5 MSS005N malfunction. by 600 psig.

This will close SRV 51G.

UO When directed to CLOSE MSIVs, operator placed outboard MSIV switches to CLOSE.

Recognize and report the failure of the A Inboard MSS018, B21-AOVF022A fails to close. MSIV to close.

Place switches for Inboard MSIVs to CLOSE.

MSS019, B21-AOVF028A fails to close Recognize and report the failure of the A Outboard MSIV to close.

RSMS-NRC14-1 Page 11 OF 15

Event 7 T7 FWS004A Master Controller output ATC Recognize and report the failure of the master FWS Master fails high. (3 min after controller.

Controller output fails mode switch to S/D)

Place Feedwater level control into manual.

high.

Manually control reactor level within the given band Event initiated upon of -20 to 51 inches.

Mode Switch operation with a Recognize and report the level 8 trip and restart one 3 minute delay. feed pump when level <51 and FWLC in Manual SRO Accept report of FWLC failure Direct ATC to take manual control of the feedwater level control valves to restore and maintain RPV level -20 to 51 inches.

Event 8 T8 MSS004 Steam Leak Outside Primary ALL Recognize and report the steam leak.

Steam Leak outside Containment. (0.45%)

SRO Direct UO to close the 98A, Main Steam Stop primary containment. CRITICAL TASK: Close the 98A MSSV Valve.

Event initiated upon NOTE: When T8 is activated, then UO Accept direction and close the 98A MSSV.

switch operation of REMOVE the T5 MSS005N malfunction.

MSIV. This will close SRV 51G.

Role Play:

Backpanel - MS Tunnel Temperature Termination is at the FREEZE Critical Task Review:

discretion of the Chief 1. Mode Switch to S/D before Suppression Pool reaches 105°F.

Examiner. 2. Close the 98A Main Steam Stop Valve.

RSMS-NRC14-1 Page 12 OF 15

VI. TERMINATION CRITERIA:

The exercise should be terminated when the performance objectives have been achieved or the operators are unable to diagnose and respond effectively to the scenario.

The following conditions provide an indication of performance objective achievement for this scenario; Critical Tasks are indicated with an *:

APRM F Bypassed; Tech Specs referenced 1/2 Scram reset Standby Stator Water Cooling Pump B manually started.

Actions taken for RPV Pressure Transmitter failure; Tech Specs referenced Actions taken for stuck open SRV IAW AOP-0035 Level control is established

• Mode Switch to Shutdown before Suppression Pool Temperature reaches 105°F.

Master FWLC in Manual Control.

• Main Steam Stop Valve (98A) closed to stop a steam leak outside primary containment.

RSMS-NRC14-1 Page 13 OF 15

VII. REFERENCES A. Plant Procedures

1. STP-257-0201, SBGT Filter Train A Monthly Test

2. STP-509-0101, Turbine Bypass Valve Cycle Test

3. AOP-0035, Stuck Open SRV

4. AOP-0001, Reactor Scram

5. AOP-0002, Turbine Trip

6. EOP-1, RPV Control

7. EOP-2, Primary Containment Control

8. OSP-0053, Emergency and Transient Response Support Procedure RSMS-NRC14-1 Page 14 OF 15

Offgoing OSM: Oncoming OSM: Off-Going Shift N D (Print) KCN (Print) KCN Date STP-257-0201, Standby Gas Treatment Filter Train A Monthly is in progress.

STP-509-0101, Turbine Bypass Valve Cycle Test is due.

SIGNIFICANT LCO STATUS EOOS STATUS 10.0 Green EQUIPMENT STATUS PROTECTED EQUIPMENT Div 2 Night Orders Standing Orders Board Walkdown Temp Alts (Signature: Oncoming OSM Review Completed) KCN RSMS-NRC14-1 Page 15 OF 15

Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: 2 IC No.: 207 Examiners: Theresa Buchanan Operators: _____________________________

Steve Garchow _____________________________

Mike Bloodgood _____________________________

Initial Conditions: Reactor power 65%, FWS-P1A tagged out, 2 Feed Reg Valves in service, Fuel shuffle in Spent Fuel Pool ongoing Turnover Shift priorities: 1) Place 3rd Feedwater Reg Valve in service.

2) Raise power with control rods.

Event Malf. Event Type* Event No. No. Description rd 1 NA N (SRO, ATC) Place 3 Feedwater Regulating Valve in service.

2 NA R (ATC) Raise power with control rods to 65%.

3 p870_56a:e-2 C (SRO, BOP) SFC-LI28B, Spent Fuel Pool Level Out of Spec Low (Tech Spec)

AO_SFC-LI28A-M AO_SFC-LI28B-M 4 p601_16a:h-5 I (SRO, BOP) E22-N654C fails downscale (CST Low Level) - (Tech Spec)

LO_E22-D2-A Manual action to align HPCS Suction to Supp Pool 5 RPS003A C(ALL) Loss of RPS A 6 ED002B Loss of NPS-SWG1B/Loss of Feedwater RPS001B M (ALL) Failure to Scram-Auto RPS001C Failure to Scram-Manual ARI effective 7 RCIC001 C (SRO, BOP) RCIC turbine trip 8 HPCS005 C (SRO, BOP) HPCS suction strainer blockage (delay 1:20)

HPCS001 HPCS pump trip (delay 2:50)

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (7) Spent Fuel Pool Level Issue, CST Level Instrument Failure, Loss of RPS, Loss of NPS SWG1B, Failure to Scram, RCIC, HPCS Malfunctions after EOP entry (1-2) (2) RCIC, HPCS Abnormal Events (2-4) (4) AOP-0010, AOP-0001, AOP-0002, AOP-0003 Major Transients (1-2) (1) Loss of NPS-SWGR1B/Loss of Feed EOPs entered (1-2) (1) EOP-0001 EOP contingencies (0-2) (1) Alternate Level Control Critical Tasks (2-3) (2) Maintain level >-186, Initiate ARI

Number: *RSMS-NRC14-2 RIVER Revision: 01 BEND STATION Page 1 of 13 Approximate Time: 1 Hour(s)

SIMULATOR SCENARIO Record Type: *Z01.24 TRAINING PROGRAM:

SIMULATOR TRAINING LESSON PLAN:

• Loss of RPS-A ; Loss of NPS-B/Feed ; ATWS - No Hi Pressure Feed REASON FOR REVISION:

NRC March 2014 exam PREPARE / REVIEW:

Angie Orgeron 1538 10-07-2013 Preparer KCN Date Dave Bergstrom 0257 10-30-2013 Technical Review (SME) KCN Date Jeff Reynolds 0358 1-22-2014 Operations Representative KCN Date Joey Clark 0260 1-27-2014 Facility Reviewer KCN Date

• Indexing Information

I. DESCRIPTION OF SCENARIO This scenario begins with the plant at 63% power.

Events for this scenario:

Place 3rd Feed Reg Valve in Service.

Raise Power with Control Rods.

Spent Fuel Pool Level Low (Tech Spec) (stop fuel shuffle in spent fuel pool)

E22-N654C failure upscale (Tech Spec) (swap HPCS suction from CST to Sup Pool)

Loss of RPS A.

Loss of NPS-SWG 1B / Loss of Feedwater.

Failure to Scram - ARI is effective.

RCIC turbine trip.

HPCS suction strainer blockage - pump trip.

II. TERMINAL OBJECTIVES

1. Establish safe and stable plant conditions following a loss of all high pressure feed per plant procedures.

RSMS-NRC14-2 Page 2 OF 13

IV. INITIAL CONDITIONS/SHIFT TURNOVER INITIAL TRAINING EQUIPMENT STATUS REQUIRED CONDITION FOCUS DOCUMENTS IC #207 Power: 63% STP-000-0102 (potential)

Core: Xenon equilibrium Equipment OOS: FWS-P1A tagged out FWRV A not in service STPs Due: None LCOs: None Evolutions in progress:

Power ascension in progress.

Fuel Shuffle in the Spent Fuel Pool.

Problem/Lit annunciators: None RSMS-NRC14-2 Page 3 OF 13

V. GENERAL INSTRUCTIONS Event Number MFS-OR-REM-SCH Expected Operator Actions Simulator Setup Check Boards for Malfunctions Equip Tags Tagout for FWS-P1A:

RPS001B, Failure to Scram auto signals LO_FWS-P1A-A Off Tagout for FWS-P1A RPS001C, Failure to Scram manual LO_FWS-P1A-G Off ARI is effective LO_FWS-P1A-R Off Check procedures and LO_FWS-P1A-W Off hard cards for marks Remote Functions LO_FWSMOV26A-G Off T10 NIS001, Reset NIs LO_FWSMOV26A-R Off Check Gauges/Meters for marks. T11 RPS004, Reset EPA Breakers OVERRIDES Check that the EVENT TRIGGERS Reactivity Control Plan T3 AO_SFC-LI28B-M, low alarm setpoint T3 p870_56a:e-2, spent fuel pool low level is appropriate for this NOTE: SFC-LI28A-M is close, but not in alarm T4 p601_16a:h-5 E22-N655C fails scenario. T4 LO_E22-D2-A, postage stamp - gross fail upscale T5 RPS003A, Loss of RPS A Ensure FB Ventilation lined up for Fuel T6 ED002B Loss of NPS-SWG1B movement. Loss of FW T8 RCIC001, RCIC Turbine Trip T9 HPCS005 HPCS suction strainer blockage T9 HPCS001 HPCS Pump Trip (delay 1 min 10 sec after breaker closed).

RSMS-NRC14-2 Page 4 OF 13

Simulator Setup Event 0 RUN CREW: Board walk down / Turnover.

Event 1 SRO Direct the ATC to place the third Feed Reg Valve in Place third Feed Reg service in accordance with SOP-0009.

Valve in Service.

ROLE PLAY ATC Accept direction to place 3rd FRV in service and Event initiated by As the turbine building operator take crew from turnover place FRV-A in service using SOP-0009, Reactor direction to monitor C33-LVF001A Feedwater System, Section 4.11:

sheet. IAW step 4.11.5 of SOP-0009.

When asked, report smooth operation of Direct the Turbine Building Operator to monitor FRV-A. FRV-A for smooth operation as it is stroked from the MCR.

Event 2 SRO Direct ATC to perform RCP-18-015 to raise Raise Reactor Power Reactor Power.

using Control Rods per Act as Reactivity SRO for Control Rod movement.

RCP.

ATC Accept the direction to perform power ascension.

Event initiated by crew from turnover Perform step 90 of RCP-18-015.

sheet.

UO Act as the peer checker for performing control rod movement.

RSMS-NRC14-2 Page 5 OF 13

T3 p870_56a:e-2, on UO Event 3 Recognize and report Spent Fuel Pool Hi/Low T3 AO_SFC-LI28B-M, failed low Level Alarm and B level indicator low-off-scale.

Spent Fuel Pool level transmitter for SFC- Refer to ARP-870-56A-E02.

ROLE PLAY:

LI28B fails low.

As Reactor Building operator, accept Direct Fuel Handling Team to cease fuel movement Event trigger T3 direction to check spent fuel pool Direct RB operator to investigate SFP level and initiated at Lead level/add makeup water to SFP. take actions to fill SFP to appropriate level.

Evaluator discretion.

As Fuel Handling Team, accept direction to Report the downscale failure of meter SFC-LI28B.

stop all fuel movement and place fuel in a safe location. SRO Accept report of possible instrument failure Reference Tech Specs 3.7.6 As I&C, accept direction to investigate instrument failure. Direct Fuel Handling Team to cease fuel movement Request assistance from Work Control / I&C As Reactor Building Operator, report back that Spent Fuel Pool level is low and fill has been commenced using condensate transfer.

RSMS-NRC14-2 Page 6 OF 13

T4 p601_16a:h-5, Annunciator - HPCS UO Event 4 Recognize and report HPCS System INOP System INOP annunciator.

Instrument for Cond T4 LO_E22-D2-A, (postage stamp: hpcs Storage Tank Level trip unit in cal gr fail) Refer to ARP-601-16A-H05.

fails downscale.

Investigate backpanels for cause.

Event trigger T4 ROLE PLAY Report the upscale failure of E22-N654C.

initiated at Lead Backpanel: indicate that E22-N654C for Evaluator discretion. SRO Accept the report of the trip unit failure.

CST level is reading pegged high and has the gross fail light illuminated. Notify work management/maintenance of (STP-000-0001 step 37) instrument failure and request OSP-0046 notifications.

Enter Tech Spec 3.3.5.1 Condition A1, Immediately enter Condition D.

Enter Tech Spec 3.3.5.1 Condition D.1, Declare HPCS INOP within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Enter Tech Spec Condition D.2.2, Align HPCS pump suction to the suppression pool within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Direct the UO to align the HPCS suction to the Suppression Pool IAW SOP-0030, HPCS System UO Accept the direction to align HPCS to the Suppression Pool per SOP-0030, section 5.3.1 RSMS-NRC14-2 Page 7 OF 13

T5 RPS003A, Loss of RPS A ATC Event 5 Recognize and report Loss of RPS-A.

Loss of RPS-A.

Event T5 initiated at Lead Evaluator ROLE PLAY UO Transfer RPS-A power to available power source Backpanel - actions to perform AOP-10 on Panel 610 (Backpanel).

discretion.

Report that RPS-A has been transferred T10 NIS001, Reset NIs (immediate actions of AOP-10)

Accept direction to own AOP-0010, Loss of RPS.

Complete the subsequent steps of AOP-0010.

SRO Accept report about Loss of RPS-A.

Assign AOP-0010, Loss of RPS.

Accept the report about completion of immediate actions of AOP-10.

If necessary, prompt the UO to not delay opening IAS-MOV106.

RSMS-NRC14-2 Page 8 OF 13

Event 6 T6 ED002B Loss of NPS-SWG1B ATC Recognize and report loss of feed.

Loss of FW Loss of NPS-B, Initiate a manual Scram causing a Loss of Upon Loss of Feed - RPV level will drop to (ATC may note that Auto Scram did not occur)

Feedwater. the Scram setpoint quickly, but will not o Mode Switch to S/D scram the plant.

Event T6 initiated at o Arm/Depress Manual Scram P/B Lead Evaluator Operator Actions will not scram the plant discretion. until ARI is initiated. o Arm/Depress ARI Pushbuttons Provide a SCRAM Report RPS001B, Failure to Scram auto signals RPS001C, Failure to Scram manual Verify Immediate Actions of AOP-1.

ARI is effective Complete Actions for AOP-1, 2, and 6.

CRITICAL TASK: Initiate ARI Accept the pressure band of 500-1090 psig SRO Accept the report about loss of feed.

Accept SCRAM report Enter EOP-1 and direct EOP-1 actions:

Verify Mode Switch in S/D (Scram Report)

Direct UO to restore and maintain RPV water level from 10 to 51 inches with RCIC.

Direct UO to install Enclosure 33, Defeating RCIC High Area Temperature Isolation Interlocks.

Direct the ATC to stabilize reactor pressure below 1090 psig, then give band of 500-1090 psig.

Assign the ATC: AOP-0001, Rx Scram, AOP-0002, Turbine Trip, and AOP-0006, Condensate Feedwater Failures.

RSMS-NRC14-2 Page 9 OF 13

Event 6 (contd) NOTE: Level 2 would normally auto initiate UO Initiate RCIC.

Loss of Feed RCIC and HPCS, but an inserted When directed, install Encl 33, RCIC Isolation.

malfunction will defeat RCIC auto initiation.

T7 RCIC001, RCIC Turbine Trip Recognize and report trip of RCIC Event 7 Recognize and report clogged HPCS strainer.

RCIC Trips.

Recognize and report HPCS pump trip.

Event 8 T8 HPCS005 HPCS suction strainer Perform EOP-1 actions as directed by SRO HPCS Suction blockage Strainer Blockage Maximize CRD, Inhibit ADS, Inject with SLC T8 HPCS001 HPCS Pump Trip followed by HPCS When directed, open 7 ADS SRVs (delay 1 min 10 sec after Pump trip.

breaker closed).

SRO Accept the report about RCIC Trip.

Accept the report about HPCS Pump trip.

CRITICAL TASK: Emerg. Depressurize When RPV level drops below -143 (Level 1), then Direct UO/ATC to inhibit ADS.

When level cannot be restored and maintained above -162, then enter the Alternate Level Control Leg of EOP-0001 When level cannot be maintained above -186, direct UO to Emergency Depressurize Termination is at the FREEZE Critical Task Review:

discretion of the Chief 1. Manually Arm & Depress ARI Pushbuttons.

Examiner. 2. Emergency Depressurize.

RSMS-NRC14-2 Page 10 OF 13

VI. TERMINATION CRITERIA:

The exercise should be terminated when the performance objectives have been achieved or the operators are unable to diagnose and respond effectively to the scenario.

The following conditions provide an indication of performance objective achievement for this scenario; Critical Tasks are indicated with an *:

FRV-A is in service.

Reactivity Control Plan performed to raise power with rods.

Tech Specs referenced for Spent Fuel Pool Level- All Fuel Handling in the SFP secured Tech Specs referenced for CST level instrument - CST suction transferred to Suppression Pool All appropriate actions of AOP-0010 are performed.

• Manual initiation of ARI.

Alternate Level Control Leg of EOP-1 entered.

• Emergency Depressurization RSMS-NRC14-2 Page 11 OF 13

VII. REFERENCES A. Plant Procedures

1. SOP-0009, Reactor Feedwater System

2. SOP-0071, Rod Control and Information System

3. ARP-870-56A, Spent Fuel Pool Water Level High/Low

4. ARP-601-16A, HPCS System Inoperative

5. SOP-0030, HPCS System

6. AOP-0010, Loss of One RPS Bus

7. AOP-0001, Reactor Scram, AOP-0002, Turbine Trip

8. AOP-0006, Condensate/Feedwater Failures

9. EOP-1, RPV Control

10. OSP-0053, Emergency and Transient Response Support Procedure RSMS-NRC14-2 Page 12 OF 13

Offgoing OSM: Oncoming OSM: Off-Going Shift N D (Print) KCN (Print) KCN Date PART I - TO BE REVIEWED PRIOR TO ASSUMING THE SHIFT UNIT STATUS MODE 1 RX POWER 63%

EVOLUTIONS (COMPLETED / IN PROGRESS / PLANNED); GENERAL INFORMATION Fuel shuffle in progress in the Spent Fuel Pool.

Place Feedwater Regulating Valve A in service per SOP.

Raise power with control rods in accordance with RCP 18-015 for sequence exchange.

SIGNIFICANT LCO STATUS EOOS STATUS EQUIPMENT STATUS PROTECTED EQUIPMENT FWS-P1A is tagged Out for Corrective Maintenance Night Orders Standing Orders Board Walkdown Temp Alts (Signature: Oncoming OSM Review Completed) KCN RSMS-NRC14-2 Page 13 OF 13

Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: 3 IC No.: 208 Examiners: Theresa Buchanan Operators: _____________________________

Steve Garchow _____________________________

Mike Bloodgood _____________________________

Initial Conditions: Reactor power 70%.

CRD-A, HDL-P1B & C, CCP-P1A & B running NNS-C being supplied from NNS-B Turnover Shift priorities 1) Rotate RPCCW pumps,

2) Lower Reactor Power Event Malf. Event Type* Event No. No. Description 1 NA N (SRO, BOP) Rotate RPCCW pumps.

2 p601-19-A02 I (SRO, BOP) MSL High Flow Instrument Failure - Low (Tech Spec) 3 NA R (ATC) Lower reactor power with control rods to 65%.

3 ED003A C (ALL) Loss of NNS-SWGR1A 4 CRD016 C (SRO, ATC) CRD suction filter clogging and pump trip (Tech Spec) trigger initiated on CRD-P1B pump start (ramped to 97% over 2 min, 30sec)

{Remove Malfunction after ATWS - role play}

5 CRD014 M (ALL) Anticipated Transient without Scram (Hydraulic Lock) [83%]

6 TMS003 C (SRO, ATC) Main Turbine High Vibration [ 20 mils over 4 minutes] - Manual Turbine Trip 7 WCS004 C (SRO, BOP) G33-MOVF004 and G33-MOVF001 fail to isolate.

WCS005

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (6) SBGT Instrument Failure, Loss of NNS-SWGR1A, CRD suction filter, ATWS, Turbine Vibration, G33-MOV failures Malfunctions after EOP entry (1-2) (2) Turbine Vibration, G33-MOV failures Abnormal Events (2-4) (3) AOP-0001, AOP-0002, AOP-0003 Major Transients (1-2) (1) ATWS EOPs entered (1-2) (2) EOP-0001, EOP-0002 EOP contingencies (0-2) (1) EOP-0001A Critical Tasks (2-3) (2) Lower level to <-56, Begin control rod insertion

Number: *RSMS-NRC14-3 RIVER Revision: 01 BEND STATION Page 1 of 14 Approximate Time: 1 Hour(s)

SIMULATOR SCENARIO Record Type: *Z01.24 TRAINING PROGRAM:

SIMULATOR TRAINING LESSON PLAN:

• Loss of NNS-A ; CRD Pump Trip ; ATWS - High Turbine Vibration REASON FOR REVISION:

NRC March 2014 exam PREPARE / REVIEW:

Angie Orgeron 1538 10-07-2013 Preparer KCN Date Dave Bergstrom 0257 10-31-2013 Technical Review (SME) KCN Date Jeff Reynolds 0358 1-22-2014 Operations Representative KCN Date Joey Clark 0260 1-27-2014 Facility Reviewer KCN Date

• Indexing Information

I. DESCRIPTION OF SCENARIO This scenario begins with the plant at 70% power.

Events for this scenario:

Rotate RPCCW Pumps MSL High Flow Instrument Failure Lower Power with Control Rods Loss of NNS A CRD suction filter clogged CRD pump trip ATWS High Turbine Vibration requiring manual trip RWCU Valves - failure to isolate II. TERMINAL OBJECTIVES

1. Establish safe and stable plant conditions following a loss of CRD and ATWS per plant procedures.

RSMS-NRC14-3 Page 2 OF 14

IV. INITIAL CONDITIONS/SHIFT TURNOVER INITIAL TRAINING EQUIPMENT STATUS REQUIRED CONDITION FOCUS DOCUMENTS IC #208 Power: 70%

Core: Xenon equilibrium Equipment OOS:

STPs Due:

LCOs: None Evolutions in progress: Down Power in progress.

Problem/Lit annunciators: None RSMS-NRC14-3 Page 3 OF 14

V. GENERAL INSTRUCTIONS Event Number MFS-OR-REM-SCH Expected Operator Actions Simulator Setup Check Boards for Malfunctions Equip Tags CRD014, ATWS-Hydraulic Lock (83%) EQUIPMENT STATUS WCS004, G33-MOVF004 fails to isolate Ensure the following are running:

Check procedures and CCP pumps A & B hard cards for marks WCS005, G33-MOVF001 fails to isolate CRD pump A HDL Pumps A & D Check Gauges/Meters Overrides NNS-C supplied from NNS-B for marks.

Check that the Remote Functions Reactivity Control Plan is appropriate for this EVENT TRIGGERS scenario.

T2 p601_19a:A02, MSL High Flow Provide marked up Instrument Failure copy of GOP-0002.

T4 ED003A Loss if NNS-SWGR1A

?Check that NNS C is T5 CRD016, CRD Suction Filter Clogging tied to NNS B (not A) tied to pump start (97% over 2 min, 30sec)

T7 TMS003 Turbine High Vibration (15 mils over 5 minutes)

RSMS-NRC14-3 Page 4 OF 14

Simulator Setup Event 0 RUN CREW: Board walk down / Turnover.

Event 1 SRO Direct the UO to alternate CCP Pumps to A & C running Rotate CCP Pumps. in accordance with SOP-0016.

ROLE PLAY Event initiated by As the reactor building operator take crew from turnover UO Accept direction to alternate CCP Pumps using direction to perform steps 5.1.1 and 5.1.2 of sheet. SOP-0016 for the C Pump. SOP-0016, Reactor Component Cooling Water Report back in 2 minutes that steps are System, Section 5.1:

complete. Direct the Reactor Building Operator to perform steps 5.1.1 & 5.1.2 in SOP-0016 for CCP Pump C.

Event 2 T2 p601_19a:A02, MSL High Flow UO Recognize and report the alarm.

Instrument Failure - Instrument Failure MSL High Flow Reference ARP-601-19-A02 T2 GTS-AOD22B, open Event trigger T2 Investigate backpanel ROLE PLAY initiated at Lead Evaluator discretion. BackPanel-Report that the flow transmitter SRO Accept the report and notify work management of E31-ESN688D for MSL C is reading low- failure and request OSP-0046 notifications.

off scale, with the gross fail light on. Enter Tech Spec 3.3.6.1 Condition A.1 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to place channel in trip).

As I&C Tech, after 15 minutes report that the pressure transmitter E31-ESN688D for MSL C High Flow has failed and you will be going back to the shop to write up a work package.

RSMS-NRC14-3 Page 5 OF 14

Event 3 SRO Direct ATC to perform RCP-18-008 to lower Lower Reactor Power Reactor Power.

using Control Rods per Act as Reactivity SRO for Control Rod movement.

RCP.

Event 3 initiated by ATC Accept the direction to perform power decrease.

crew from turnover Perform step 07 of RCP-18-008.

sheet.

UO Act as the peer checker for performing control rod movement.

RSMS-NRC14-3 Page 6 OF 14

Event 4 T4 ED003A Loss if NNS-SWGR1A UO Recognize and report Loss of NNS-A.

Loss of NNS-A. o Recognize loss of CCS-A Event trigger T4 ROLE PLAY o Recognize loss of CRD-A.

initiated at Lead Accept phone call to make notifications and Evaluator discretion. for maintenance support. ATC Recognize and report Loss of NNS-A.

o Recognize loss of HDL Pump C As turbine building operator, accept report o Recommend starting HDL Pump D to restart TB Chill Water System (B) per SOP-116. SRO Accept the report of the electrical failure.

Notify work management/maintenance of electrical failure and request OSP-0046 notifications.

Direct the UO to Start CRD Pump B IAW the ARP (or the SOP).

Direct the ATC to Start HDL Pump D IAW the SOP.

When second accumulator fault alarms, Enter Tech Spec 3.1.5 Condition B.1, Restore charging water header pressure 1540 psig within 20 minutes.

When CRD pump B trips, Enter Tech Spec 3.1.5 Condition D.1, Place mode switch in Shutdown immediately.

Direct ATC to refer to AOP-0006, Condensate/Feedwater Failures and AOP-0007, Loss of Feedwater Heating.

RSMS-NRC14-3 Page 7 OF 14

Event 5 T5 CRD016, CRD Suction Filter Clogging UO When directed, Start CRD Pump B.

tied to pump start CRD Suction filter clog Recognize and report CRD low suction/clogged (97% over 2 min, 30sec) strainer.

CRD pump trip. trigger: zlo5(1256)=1 Recognize and report CRD Pump B trip.

Event T5 initiated at the Pump B start. ATC Monitor for CRD accumulator faults ROLE PLAY Perform actions IAW ARP680-07A-C03.

As Reactor Building Operator, accept When 2 faults received, note the time for the 20 direction to inspect CRD.

minute clock for the LCO.

After 5 minutes,RB Operator, indicate that SRO Accept the report of the CRD pump trip the strainer discharge valve handwheel broke when attempting to open. Perform a control room brief to insert a manual scram while not exceeding the 20 minute LCO.

After 10 minutes, notify the MCR that Notify work management/maintenance of electrical mechanical maintenance is on station failure and request OSP-0046 notifications.

preparing to replace suction filter media.

May recommend placing CRD Strainer in service (bypassing Filters). (RDS-STRD013)

RSMS-NRC14-3 Page 8 OF 14

Event 6 CRD014, ATWS-Hydraulic Lock (83%) ATC When directed, Initiate a manual Scram ATWS o Mode Switch to S/D After trip of CRD B, the 2nd accumulator o Arm/Depress Manual Scram P/B Event 6 initiated when fault begins the 20 minute LCO o Arm/Depress ARI Pushbuttons the ATC inserts a will require inserting a scram.

manual scram. Provide an ATWS Report.

When directed, trip both Recirc Pumps.

• CRITICAL TASK:

When directed, terminate injection with feedwater Lower RPV Level to < -56 inches and lower RPV level to -60 to -140 inches.

SRO Direct the ATC to take the mode switch to S/D.

Accept the ATWS report.

Enter EOP-1 and transition to EOP-1A Direct EOP-1A actions:

o ATC verify ARI initiation o ATC trip both reactor recirc pumps o UO terminate and prevent injection with HPCS o UO inhibit ADS o UO install EOP-5 enclosures 16 and 24 o ATC terminate injection with feedwater and lower reactor water level to -60 to -140 o UO install EOP-5 enclosures 12 and 14.

RSMS-NRC14-3 Page 9 OF 14

Event 6 CRD014, ATWS - 83% hydraulic lock UO When directed, inhibit ADS (continued) When directed, terminate and prevent injection with Remotes: HPCS.

T12 EOP012B Encl 12 ARI jumpered T13 EOP012A Encl 12 RPS jumpered When directed, install EOP-5 enclosures 16 and 24.

T14 EOP014, Encl 14 RC&IS Interlocks When directed, initiate SLC.

T16 EOP016, Encl 16 Containment IAS Isolation Interlocks jumpered When directed, install EOP-5 enclosures 12 and 14.

T20 EOP020, Encl 20 DW Cooling When directed re-start CRD pump B per the ARP.

T24 EOP024, Encl 24 RPV level 1 MSIV and Drains Isol Intlk jumpered SRO Accept the report about CRD availability.

Direct the UO to start CRD B per the ARP.

Role Play:

5 minutes after the scram is initiated, delete When CRD B restored, direct the ATC to drive CRD016, CRD suction filter clogged. control rods per Enclosure 14.

As mechanical Maintenance, report that the ATC When directed, insert control rods per Encl 14.

CRD filters have been replaced and CRD B can be restored.

RSMS-NRC14-3 Page 10 OF 14

Event 7 T7 TMS003 Turbine HighVibration ATC Recognize and report rising turbine vibration.

(15 mils over 5 minutes) When directed OR when turbine vibration exceeds Turbine Vibration Note: Ramp vibration back down after 12 mils, Trip the main turbine.

Event trigger T7 initiated at Lead turbine is tripped over a 20 minute period. Perform the actions of AOP-0002, Turbine Trip.

Evaluator discretion.

Perform the actions of AOP-0001, Reactor Scram.

Control Reactor pressure within given band Event 8 (950-1090 psig).

RWCU Isolation Valves Fail to Auto WCS004, G33-MOVF004 fails to isolate SRO Accept the report about turbine vibration.

Isolate. WCS005, G33-MOVF001 fails to isolate Direct the ATC with a contingency for tripping the main turbine.

Direct the ATC to control reactor pressure between 950-1090 psig using bypass valves and drains.

UO Accept ownership of AOP-0003, Isolations Recognize and report the failure to close for RWCU valves and take action to close them.

Termination is at the FREEZE Critical Task Review:

discretion of the Chief 1. Lower reactor water level to < -56.

Examiner. 2. Begin Control Rod Insertion.

RSMS-NRC14-3 Page 11 OF 14

VI. TERMINATION CRITERIA:

The exercise should be terminated when the performance objectives have been achieved or the operators are unable to diagnose and respond effectively to the scenario.

The following conditions provide an indication of performance objective achievement for this scenario; Critical Tasks are indicated with an *:

CCP Pumps swapped.

Tech Spec entered for Main Steam Line High Flow Instrument.

Reactivity Control Plan performed to lower power with rods.

Actions taken to restore CRD.

Tech Spec entered for CRD accumulator faults Reactor Scram inserted per LCO.

EOP-1A actions taken.

• Reactor water level lowered to < -56 inches.

• Control Rod Insertion has begun.

Action taken to investigate RWCU valves that failed to isolate.

RSMS-NRC14-3 Page 12 OF 14

VII. REFERENCES A. Plant Procedures

1. SOP-0016, Reactor Plant Component Cooling Water System

2. GOP-0002, Power Decrease/Plant Shutdown

3. AOP-0001, Reactor Scram

4. AOP-0002, Turbine Trip

5. EOP-1, RPV Control

6. EOP-1A, RPV Control, ATWS

7. EOP-5, Enclosures

8. OSP-0053, Emergency and Transient Response Support Procedure

9. Tech Specs RSMS-NRC14-3 Page 13 OF 14

Offgoing OSM: Oncoming OSM: Off-Going Shift N D (Print) KCN (Print) KCN Date PART I - TO BE REVIEWED PRIOR TO ASSUMING THE SHIFT UNIT STATUS MODE 1 RX POWER 70%

EVOLUTIONS (COMPLETED / IN PROGRESS / PLANNED); GENERAL INFORMATION Shutting Down to Hot Standby.

Rotate RCCP Pumps from A,B to A,C.

Lower power with control rods in accordance with RCP 18-008 for forced outage 14-01.

SIGNIFICANT LCO STATUS EOOS STATUS EQUIPMENT STATUS PROTECTED EQUIPMENT NNS-C supplied from NNS-B Night Orders Standing Orders Board Walkdown Temp Alts (Signature: Oncoming OSM Review Completed) KCN RSMS-NRC14-3 Page 14 OF 14

Appendix D Scenario Outline Form ES-D-1 Facility: River Bend Station Scenario No.: 4 IC No.: 209 Examiners: Theresa Buchanan Operators: _____________________________

Steve Garchow _____________________________

Mike Bloodgood _____________________________

Initial Conditions: Reactor Power 90%, RCIC tagged out.

Turnover Shift Priorities: 1) Raise reactor power with Recirc flow.

Event Malf. Event Type* Event No. No. Description 1 NA R (ATC) Raise power with Recirculation flow.

2 RMS0165 I (SRO, BOP) RMS-RE16B-PAMs Instrument-Failure Rx Bldg Radiation (Tech Spec) 3 RHR008B C (SRO) Div 2 ECCS line fill pump trips. (Tech Spec) 4 NA C (ALL) Seal leak on HDL-P1A.

5 CNM006 C (ALL) Condensate Filter High dp / Loss of Feedwater / Reactor Scram 6 MSS001 M (ALL) Drywell Leak (on Mode Switch - 550 gpm ramped over 3 minutes) 2 MSC007 Bypass Leakage (when DW d/p = 2.3 psid; 12.56 in )

RHR002A RHR A Trip (when DW d/p = 1.7 psid) 7 E22MOVF004 C (SRO, BOP) HPCS Injection Valve breaker trips (on HPCS pump breaker closure) 8 LPCS002 C (SRO, BOP) Low Pressure Core Spray Injection Valve fails to automatically open.

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Total Malfunctions (5-8) (7) RMS-RE16B Instrument Failure, Div 2 line fill pump, HDL-P1A leak, Condensate Filter dp, Drywell Leak, HPCS Inj Vlv, LPCS Inj Vlv Malfunctions after EOP entry (1-2) (2) HPCS Inj Vlv, LPCS Inj Vlv Abnormal Events (2-4) (3) AOP-0001, AOP-0002, AOP-0003 Major Transients (1-2) (1) Drywell Leak EOPs entered (1-2) (2) EOP-0001, EOP-0002 EOP contingencies (0-2) (2) Alternate Level Control, Emergency Depressurization Critical Tasks (2-3) (2) Emergency Depressurize, Restore and maintain level >-186.

Number: *RSMS-NRC14-4 RIVER Revision: 01 BEND STATION Page 1 of 14 Approximate Time: 1 Hour(s)

SIMULATOR SCENARIO Record Type: *Z01.24 TRAINING PROGRAM:

SIMULATOR TRAINING LESSON PLAN:

• ECCS Line Fill Trip ; Loss of Feed ; Steam Leak in DW REASON FOR REVISION:

NRC March 2014 exam PREPARE / REVIEW:

Angie Orgeron 1538 10-07-2013 Preparer KCN Date Dave Bergstrom 0257 11-01-2013 Technical Review (SME) KCN Date Jeff Reynolds 0358 1-22-2014 Operations Representative KCN Date Joey Clark 0260 1-27-2014 Facility Reviewer KCN Date

• Indexing Information

I. DESCRIPTION OF SCENARIO This scenario begins with the plant at 90% power and RCIC tagged out.

Events for this scenario:

Raise Power with Recirc Flow.

RMS-RE16B-Gas Instrument Failure.

Div 2 ECCS Line Fill Pump Trips - requiring control fuse removal.

Seal Leak on HDL Pump A past an ODMI trigger point.

Condensate filter clogged - Loss of Feedwater - SCRAM.

Drywell Steam Leak Occurs as a result of the Scram transient.

HPCS and RHR-A pumps trip upon initiation.

No High Pressure Feed will require ED.

LPCS Injection Valve failure to auto open requires operator action to inject.

II. TERMINAL OBJECTIVES

1. Establish safe and stable plant conditions following a loss of high pressure feed and a steam leak in the DW per plant procedures.

RSMS-NRC14-4 Page 2 OF 14

IV. INITIAL CONDITIONS/SHIFT TURNOVER INITIAL TRAINING EQUIPMENT STATUS REQUIRED CONDITION FOCUS DOCUMENTS IC #209 Power: 90%

Core: Xenon equilibrium Equipment OOS: RCIC STPs Due:

LCOs: T.S. 3.5.3 Action A1 complete; A2 due in 10 days Evolutions in progress: Power ascension in progress.

Problem/Lit annunciators: None RSMS-NRC14-4 Page 3 OF 14

V. GENERAL INSTRUCTIONS Event Number MFS-OR-REM-SCH Expected Operator Actions Simulator Setup Check Boards for Malfunctions Equip Tags LPCS002, LPCS Injection Valve fails to EQUIPMENT STATUS auto Open RCIC Tagged Out Check procedures and HDL Pump A running hard cards for marks Remote Functions Check Gauges/Meters for marks. T4 ECCS005, RHR B racked out (delay 2 min 30sec)

Check that the T4 ECCS006, RHR C racked out Reactivity Control Plan (delay 4 min) is appropriate for this scenario. T12 EOP12A, Encl 12 RPS signals Tagout for RCIC. T16 EOP16, Encl 16 Instrument Air T20 EOP20, Encl 20 DW Cooling EVENT TRIGGERS T2 RMS0165, RMS-RE16B-PAMs Instr.

Failure, pegged high T3 RHR008B, Div 2 Line Fill Trip T5 CNM006 Condensate Filter High d/p 99% ramped over 1:30 Loss of Feed RSMS-NRC14-4 Page 4 OF 14

Event Number MFS-OR-REM-SCH Expected Operator Actions EVENT TRIGGERS (continued)

NOTE:

T6 MSS001 Steam Leak in DW triggers T6, T7, and T8 are all part of event 6 and are triggered (triggered by Mode Switch) automatically (550 gpm over 3 minutes)

T7 MSC007 Bypass Leakage (triggered when DW d/p >2 psid) 12.56 in2, ramped in 30 sec T8 RHR002A RHR-A trips (triggered (when DW d/p >1.7 psid)

T9 E22MOVF004 HPCS Inj Valve trips (triggered on breaker closure)

RSMS-NRC14-4 Page 5 OF 14

Simulator Setup Event 0 RUN CREW: Board walk down / Turnover.

Event 1 SRO Direct ATC to perform RCP-18-021 to raise Reactor Power.

Raise Reactor Power using Recirc per RCP. Act as Reactivity SRO for reactivity manipulation.

Event initiated by ATC Accept the direction to perform power increase.

crew from turnover Perform step 91A of RCP-18-021.

sheet.

UO Act as the peer checker for performing reactivity manipulation.

Event 2 UO Recognize and report alarm.

RMS-RE16B-PAMs T2 RMS0165, RMS-RE16B-PAMs, instrument pegged high. Instrument Pegged High SRO Accept the report of the rad alarm.

Event trigger T2 Notify work management/RP/maintenance of initiated at Lead possible instrument failure and request OSP-0046 Evaluator discretion. notifications.

Enter Tech Spec 3.3.3.1, A1 (30 days)

RSMS-NRC14-4 Page 6 OF 14

Event 3 T3 RHR008B, Div 2 Line Fill Trip UO Recognize and report Loss Div 2 Line Fill Pump.

Div 2 ECCS Line Fill o Perform actions of ARP-601-17A-C01 & C02 Pump trip. ROLE PLAY: o Check Backpanel 618 for ECCS pressures.

Event trigger T3 Backpanel - indicate RHR B and C disch o Direct the Reactor Building Operator to check initiated at Lead pressures are low (alarm at 28 psig).

the Div 2 Line Fill Pump.

Evaluator discretion. Accept phone call to make notifications and for maintenance support. o Direct the Control Building Operator to remove control power fuses for RHR-B and C Time Called: _______ As CB/RB Operator, accept direction to pump breakers.

investigate Div 2 ECCS Line Fill Pump Call Back: _________

As CB/RB Operator, report back that the SRO Accept the report of the Div 2 Line Fill Pump.

line fill pump motor bearings are extremely Notify work management/maintenance of electrical hot (too hot to touch). failure and request OSP-0046 notifications.

As CB/RB Operator, accept direction to Direct the UO to check Div 2 ECCS pressures.

remove control power fuses RHR-B and - When pressure has decayed, or when it is known Time Called: _______ C pump breakers. that the line fill pump will not be started quickly, direct that the control power fuses for the Div 2 Call Back: _________ REMOTE: ECCS pump breakers be removed.

T4 ECCS005, RHR B fuses pulled Enter Tech Spec 3.5.1 C.1 (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO)

(delay 2 min 30sec)

T4 ECCS006, RHR C fuses pulled (delay 3 min 15 sec)

Report back that control power fuses are pulled for RHR B and C.

RSMS-NRC14-4 Page 7 OF 14

Event 4 SRO Accept the report of the HDL pump seal ODMI trigger point.

Shift HDL Pumps due ROLE PLAY to report of seal Direct the ATC to swap HDL Pumps from A to B.

As the Aux Control Room Operator, leakage. inform the MCR that Condensate Filter C will be removed from service to perform ATC Accept the direction to swap HDL Pumps.

backflush operation.

Event 4 initiated at Perform actions IAW SOP-0010, MSR &

Lead Evaluator Feedwater Heating, Section 5.1 - Alternating discretion. As Duty Manager, call the MCR with a Heater Drain Pumps.

report that the seal leakage rate on HDL Pump A has reached the ODMI trigger Direct the Turbine Building Operator to perform Time Called: _______ point that requires swapping HDL Pumps pre-start checks of HDL-P1B (step 5.1.3).

from A to B.

Call Back: _________

As the turbine building operator, if Direct the Turbine Building Operator to perform requested report that all pre-start checks post-start checks of HDL-P1A and B (step 5.1.10).

are complete and SAT.

(Pump B not rotating)

Post start checks include pump A not rotating.

RSMS-NRC14-4 Page 8 OF 14

Event 5 T5 CNM006 Condensate Filter High d/p ATC Recognize and report feedwater failure.

Condensate Filter 99% ramped over 1 min Initiate a manual Scram.

High D/P Loss of Feed Loss of Feed Provide a Scram Report.

Scram ATC recognize and report no high pressure feed.

ROLE PLAY Event trigger T5 Perform actions of AOP-0001, and 2.

initiated at Lead As Aux Control Room, call MCR to notify Evaluator discretion. that Condensate Filter C is being removed from service to perform backwash SRO Accept the loss of feed report.

operation. (SOP-124 Sect 5.2 Accept the scram report.

Note: this notification may be performed during the previous event to set up for this Enter EOP-1 and direct EOP-1 actions:

triggered event. o ATC verify mode switch in S/D o UO/ATC stabilize pressure <1090 psig then give a pressure band of 500-1090 psig w/ bypass valves and drains o UO restore and maintain level between -20 T9 E22MOVF004 HPCS Inj Valve trips and 51 inches using HPCS.

(triggered on breaker closure)

Direct entry into AOP-0001, -2, and -6 UO Accept pressure band of 500-1090 psig with bypass valves and drains.

Accept level band of -20 to 51 using HPCS Initiate HPCS (if not already at level 2)

Recognize and report the trip of HPCS Inj Valve RSMS-NRC14-4 Page 9 OF 14

Event 6 T6 MSS001 Steam Leak in DW SRO Accept the report of HPCS inj valve trip.

(triggered by Mode Switch) Enter Alternate Level Control of EOP-1.

Steam Leak in the DW (550 gpm over 3 minutes) o Direct inhibiting ADS Bypass Leakage T7 MSC007 Bypass Leakage (triggered Recognize and report rising DW d/P.

RHR A trips when DW d/p >2 psid) 12.56 in2, ramped in 30 sec Direct installing Encl 16 Direct injection with SLC T8 RHR002A RHR-A trips (triggered (when DW d/p >1.7 psid) Enter EOP-0002, Primary Containment Control When Containment Pressure (due to bypass leakage) will cause entry into the unsafe region of CRITICAL STEP: Open 7 ADS SRVs the PSP curve, then Emergency Depressurization IS Required (CP-4)

T12 EOP12A, Encl 12 RPS signals Enter ED Leg of EOP-1 T16 EOP16, Encl 16 Instrument Air Direct opening 7 ADS SRVs T20 EOP20, Encl 20 DW Cooling ATC/ When directed, inhibit ADS UO When directed, initiate SLC.

When directed, install Encl 16

• CRITICAL STEP:

Open LPCS Inj Valve When directed open 7 ADS SRVs Recognize and report the trip of RHR-A Recognize and report the failure of LPCS inj valve and OPEN the valve.

Restore RPV level using LPCS Termination is at the FREEZE Critical Task Review:

discretion of the Chief 1. Open 7 ADS SRVs.

2. Restore reactor water level to -20 to 51 using LPCS.

RSMS-NRC14-4 Page 10 OF 14

Examiner.

RSMS-NRC14-4 Page 11 OF 14

VI. TERMINATION CRITERIA:

The exercise should be terminated when the performance objectives have been achieved or the operators are unable to diagnose and respond effectively to the scenario.

The following conditions provide an indication of performance objective achievement for this scenario; Critical Tasks are indicated with an *:

Reactivity Control Plan performed to raise power with Recirc Flow.

Tech Spec 3.3.3.1 Condition A1 entered for Containment Radiation PAM Tech Spec 3.7.1.E.1 entered.

RHR B and C control fuses pulled in response to line fill pump trip.

EOP-1 actions taken.

EOP-2 actions taken

• 7 ADS SRVs opened to emergency depressurize.

• Reactor water level restored with LPCS.

RSMS-NRC14-4 Page 12 OF 14

VII. REFERENCES A. Plant Procedures

1. SOP-0003, Reactor Recirculation System

2. SOP-0010, MSRs and Feedwater Heating System

3. AOP-0006, Condensate/Feedwater Failures

4. EOP-1, RPV Control

5. EOP-0002, Primary Containment Control

6. AOP-0001, Reactor Scram

7. AOP-0002, Turbine Trip

8. AOP-0003, Isolations

9. EOP-5, Enclosures

10. OSP-0053, Emergency and Transient Response Support Procedure RSMS-NRC14-4 Page 13 OF 14

Offgoing OSM: Oncoming OSM: Off-Going Shift N D (Print) KCN (Print) KCN Date PART I - TO BE REVIEWED PRIOR TO ASSUMING THE SHIFT UNIT STATUS MODE 1 RX POWER 90%

EVOLUTIONS (COMPLETED / IN PROGRESS / PLANNED); GENERAL INFORMATION Raise power with Recirc flow in accordance with RCP 18-021.

RCIC Governor maintenance ongoing.

SIGNIFICANT LCO STATUS EOOS STATUS T.S. 3.5.3 Action A2 due in 10 days EQUIPMENT STATUS PROTECTED EQUIPMENT RCIC tagged out for maintenance HPCS ODMI in place for HDL Pump A seal leakage.

Night Orders Standing Orders Board Walkdown Temp Alts (Signature: Oncoming OSM Review Completed) KCN RSMS-NRC14-4 Page 14 OF 14