ML101590195

Initial Exam 2010-301 Final Administrative JPMs

ML101590195
Person / Time
Site:	Brunswick
Issue date:	05/03/2010
From:	NRC/RGN-II
To:	Progress Energy Carolinas
References
50-324/10-301, 50-325/10-301
Download: ML101590195 (102)

Text

LESSON TITLE: LESSON NUMBER: REVISION NO: PROGRESS ENERGY -CAROLINAS BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE ADMIN NRC ADMIN C002 (SRO) Evaluate Refueling Technical Specifications SOT -OJT -J P-305-CO 1 01 Evaluate Refueling Technical Specifications SAFETY CONSIDERATIONS:

None EVALUATOR NOTES: (Do not read to trainee) 1. The applicable procedure section WILL be provided to the trainee. 2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021, Appendix E, or similar to the trainee. 3. This JPM may be administered in the simulator, control room, or classroom setting. 4. Ensure examinee has access to Technical Specifications, Bases, and POM documents.

Read the following to trainee. TASK CONDITIONS:

The following plant conditions exist during a fuel shuffle on Unit Two: RHR Shutdown Cooling (SDC) Loop A in service RHR Shutdown Cooling (SDC) Loop B under clearance Reactor Recirculation Loops A and B under clearance Fuel Pool Cooling is in service -Supplemental Spent Fuel Pool Cooling is in service The Refuel Floor SRO requests that SDC be secured for the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to support retrieving a Reactor Pressure Vessel Head nut from within the vessel using the Auxiliary Platform. (This is the only available opportunity to get the Foreign Material out of the RPV) No work or maintenance will be performed on SDC Loop A while it is removed from service. INITIATING CUE: The Shift Manager has directed you to; 1. Evaluate Technical Specifications to remove Shutdown Cooling from operation for the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and 2. Determine if in-vessel fuel movements are allowed to continue without SDC in service after the Foreign Material has been retrieved.

SOT-OJT-JP-305-C01 Page 2 of 9 REV. 01 Evaluate Refueling Technical Specifications ANSWER KEY Technical CONDITION REQUIRED COMPLETION Fuel Movements Specification ACTION TIME Allowed LCO 3.9.7 C. No RHR C.1 Verify reactor 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from YE shutdown coolant circulation discovery of no cooling by an alternate reactor coolant subsystem in method. circulation operation.

AND AND Once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> NO thereafter C.2 Monitor reactor coolant temperature.

Once per hour SOT -OJT -J P-305-CO 1 Page 3 of 9 REV. 01 Evaluate Refueling Technical Specifications PERFORMANCE CHECKLIST NOTE: Sequence is not essential to completing the task. TIME START: _____ _ Step 1 -Obtain current revision of Unit Two Tech Specs Current copy of Unit Two Tech Specs is obtained SAT/UNSAT*

Step 2 -Determine LCO 3.9.7 applies to current plant conditions (may use TOC to determine)

LCO 3.9.7 determined to be applicable to current plant conditions.

• CRITICAL STEP*

• SAT/UNSAT*

Step 3 -Determine LCO 3.9.7 requires one RHR shutdown cooling subsystem to be Operable and in operation Determine that LCO 3.9.7 requires one RHR subsystem to be Operable and in operation.

SAT/UNSAT*

Step 4 -Determine LCO 3.9.7 is modified by a note allowing the required RHR shutdown cooling subsystem to be removed from operation for up to 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. Determine that LCO 3.9.7 is modified by a note aI/owing the required RHR shutdown cooling subsystem to be removed from operation for up to 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> in a 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period SAT/UNSAT*

NOTE: Candidate may propose allowing SDC removal from service for up to 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> without entering the Tech Spec ACTION statement.

PROMPT: If necessary, Direct candidate to address Technical Specifications for removing SDC from service for 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />. SOT -OJT -J P-305-CO 1 Page 4 of 9 REV. 01 Evaluate Refueling Technical Specifications Step 5 -Determine LCO 3.9.7 will not be met if RHR shutdown cooling is removed from service for 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> (more than the allowable two hour time) Determine that LCO 3.9.7 will not be met if RHR shutdown cooling is removed from service for 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> (more than the allowable two hour time) *CRITICAL STEP*

• SAT/UNSAT*

Step 6 -Determine LCO 3.9.7 Condition A does not apply because the RHR shutdown cooling loop is only being removed from service and will not be made inoperable Determines LCO 3.9.7 Condition A does not apply *CRITICAL STEP* SAT/UNSAT*

Step 7 -Determine LCO 3.9.7 Condition B does not apply since Condition A does not apply. Determine LCO 3.9.7 Condition B does not apply *CRITICAL STEP* SAT/UNSAT*

Step 8 -Determine LCO 3.9.7 Condition C does apply since no RHR shutdown cooling subsystem will be in operation Determine that LCO 3.9.7 Condition C does apply *CRITICAL STEP*

• SAT/UNSAT*

Step 9 -Determine required action C.1 applies and requires verification of reactor coolant circulation by an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of no reactor coolant circulation and once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> thereafter Determine that required action C. 1 applies and requires verification of reactor coolant circulation by an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of no reactor coolant circulation and once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> thereafter

• CRITICAL STEP*

• SAT/UNSAT*

Step 10 -Determine required action C.2 also applies and requires monitoring of reactor coolant temperature once per hour Determine that required action C.2 applies and requires monitoring of reactor coolant temperature once per hour *CRITICAL STEP*

• SAT/UNSAT*

SOT -OJT -J P-305-CO 1 Page 5 of 9 REV. 01 Evaluate Refueling Technical Specifications NOTE: Candidate must evaluate RHR system operability and determine not in operation while still Operable will still allow for in-vessel fuel movement during high water level conditions.

Step 11 -Determine if in-vessel fuel movement is still allowed. Determines in-vessel fuel movement is still allowed. *CRITICAL STEP*

• SAT/UNSAT*

Step 13 -Inform Shift Manager of results. SM notified.

SAT/UNSAT*

TERMINATING CUE: When the results are reported to the SM this JPM is complete.

TIME COMPLETED:

_____ _

• Comments required for any step evaluated as UNSAT. SOT -OJT -J P-305-CO 1 Page 6 of 9 REV. 01 Evaluate Refueling Technical Specifications LIST OF REFERENCES RELATED TASKS: 342000 B3 02 Approve requests to remove plant equipment from operation.

KIA REFERENCE AND IMPORTANCE RATING: Generic Conduct of Operations 2.1.40 Knowledge of refueling administrative requirements.

RO I SRO 2.8 I

3.9 REFERENCES

Unit 2 Technical Specifications Amendment 233 OAP-22 TOOLS AND EQUIPMENT:

None SAFETY FUNCTION (from NUREG 1123, Rev 2 Supp. 1): Generic -Conduct of Operations REASON FOR REVISION:

Revised to the proper KIA reference, which had changed in supplement

1. SOT-OJT-JP-305-C01 Page 7 of 9 REV. 01 Evaluate Refueling Technical Specifications Time Required for Completion:

.2!L Minutes (approximate).

APPLICABLE METHOD OF TESTING Performance:

Simulate Actual l Unit: _2_ Setting: In-Plant Simulator Admin X Time Critical:

Yes No -2L Time Limit N/A Alternate Path: Yes No --2L EVALUATION Trainee: -------------------------------------------------

JPM: Pass __ Fail __ Remedial Training Required:

Yes No __ Did Trainee Verify Procedure Revision?

Yes No ____ _ (Each Student should verify one JPM per evaluation)

Comments:

Evaluator Signature:

____________________________

_ Date: __________

_ SOT-OJT-JP-305-C01 Page 8 of 9 REV. 01 TASK CONDITIONS:

The following plant conditions exist during a fuel shuffle on Unit Two: RHR Shutdown Cooling (SDC) Loop A in service RHR Shutdown Cooling (SDC) Loop B under clearance Reactor Recirculation Loops A and B under clearance Fuel Pool Cooling is in service Supplemental Spent Fuel Pool Cooling is in service The Refuel Floor SRO requests that SDC be secured for the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to support retrieving a Reactor Pressure Vessel Head nut from within the vessel using the Auxiliary Platform. (This is the only available opportunity to get the Foreign Material out of the RPV) No work or maintenance will be performed on SDC Loop A while it is removed from service. INITIATING CUE: The Shift Manager has directed you to; 1. Evaluate Technical Specifications to remove Shutdown Cooling from operation for the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and 2. Determine if in-vessel fuel movements are allowed to continue without SDC in service after the Foreign Material has been retrieved.

LESSON TITLE: LESSON NUMBER: REVISION NO: PROGRESS ENERGY CAROLINAS BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE ADMINISTRATIVE NRC ADMIN EP (SRO) Classify An Emergency Per PEP-02.1.

SOT-ADM-JP-301-A 15 02 Classify An Emergency Per PEP-02.1.

SAFETY CONSIDERATIONS:

None. EVALUATOR NOTES: (Do not read to performer)

1. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021, Appendix E, or similar to the performer.

2. The examinee will be provided a copy of OPEP-02.1 flowchart.

3. A copy of OPEP-02.2.1, Emergency Action Level Technical Bases, should be available if requested.

4. A clock must be available in the classroom and visible to examiner and examinees.

5. Copies of all remaining PEPs should be available upon request. 6. Task standards (Le. pass/fail criteria) for each JPM step are ITALICIZED below the step. 7. Emphasize to candidates that this is a Time Critical JPM and that following cue sheet review the evaluator will designate the START TIME on the board and stop the JPM at the applicable critical time. SOT-AOM-JP-301-A 15 Page 2 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

Read the following to the J PM performer.

TASK CONDITIONS:

• • This is a Time Critical JPM** Time begins when directed by the evaluator

1. Unit One is operating at 100% power. 2. Unit Two is operating at 100% power with DG4 under clearance when the following event occurs (consider all items that exceed EAL thresholds occur at the same time). Unit Two Event Description A seismic event with a magnitude of 0.12g is felt on plant site and is verified I validated by the National Earthquake Center results in a loss of Corporate Telephone Communications System (Voicenet), Commercial Telephones, and NRC Emergency Telecommunications System. The Main Turbine Trips and a majority of the control rods remain withdrawn.

A Manual Scram has been inserted, the Mode Switch has been placed in Shutdown, ARI has been initiated, and reactor power indicates 20%. LEP-02 (Alternate Control Rod Insertions) actions and SLC injection are in progress.

Current indicated reactor power is 12%. Reactor Water level is being maintained in a band of +60 to +90 inches. Reactor Pressure is being controlled by EHC at 945 psig. NO radiological releases are in progress at this time. There are no indications of an Onsite Security Event. INITIATING CUE: You are to evaluate the above event as the Control Room Site Emergency Coordinator (SEC) and determine the highest required classification including the EAL Identifier for Unit Two ONLY: 1. Write the required Classification along with the EAL identifier in the table below. 2. Raise your hand when complete to have the evaluator stop the evaluation Time and collect your cue sheet. CLASSIFICATION EAL IDENTIFIER SOT-ADM-JP-301-A 15 Page 3 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

ANSWER KEY CLASSIFICATION EAL IDENTIFIER Site Area Emergency SS2.1 SOT -ADM-JP-301-A 15 Page 4 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless denoted in the Comments.

Step 1 -Obtain a current revision of PEP-02.1 if required.

Current Revision of PEP-02. 1 obtained.

SATIUNSAT*

PROMPT: Ensure a clock is visible for candidates.

Announce and Write the Start Time on the board. Add 15 minutes to the Start Time and write that in the JPM Completion Time. If all candidates have not Declared a classification, this is the time to STOP all work, put pencils/pens down, and collect all remaining cue sheets. I NOTE: Declaration of event must be made in 15 minutes from the Start Time. TIME START ____ _ NOTE: Loss of Corporate Telephone Communications System (Voicenet), Commercial Telephones, and NRC Emergency Telecommunications System does not reach EAL classification .threshold.

Step 2 -Determine required Classification threshold and associated EAL Number(s) o Alert -HA 1. 1 Seismic event identified by any two of the following:

-Earthquake felt in plant -National Earthquake Center AND Control Room indication of degraded performance of systems required for the safe shutdown of the plant (ATWS provides this) Seismic event> Operating Basis Earthquake (0.08 g) per analysis SAT/UNSAT*

SOT-ADM-JP-301-A 15 Page 5 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

NOTE: Automatic Scram fails to shutdown the reactor and manual actions taken from the control console are not successful in shutting down.the reactor. Step 3 -Determine required Classification threshold and associated EAL Number(s) o Site Area Emergency

-SS2. 1 Automatic scram fails to reduce reactor power < 2% (APRM downscale)

AND Manual scram actions taken at the reactor control console (Manual PBs, Mode Switch, ARI) do not shutdown the reactor as indicated by reactor power> 2% **CRITICAL STEP** SA TlUNSAT* DECLARATION TIME ___ _ Step 4 -Classification made within required the required time (Declaration Time minus Start Time 15 minutes).

o Classification declared 15 minutes of Start Time. **CRITICAL STEP** SAT/UNSAT*

TERMINATING CUE: When the event is classified with applicable EAL identifier(s) in the table, this JPM is complete.

TIME COMPLETED

____ _ *NOTE: Comments required for any step evaluated as UNSAT. SOT-ADM-JP-301-A 15 Page 6 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

RELATED TASKS: 3442568502 Direct Initial Emergency Actions Including Emergency Classification per OPEP-02.1 KIA REFERENCE AND IMPORTANCE RATING: GEN 2.4.29 3.1/4.4 Knowledge of the Emergency Plan

REFERENCES:

OPEP-02.1, Rev 51 TOOLS AND EQUIPMENT:

None. ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2., Supp.1): Admin -Emergency Procedures 1 Plan REASON FOR REVISION:

2010-1 NRC License Exam -Modified to reflect NEI 99-01 implementation.

SOT-ADM-JP-301-A 15 Page 7 of 8 REV. 02 Classify An Emergency Per PEP-02.1.

Validation Time: 15 Minutes (approximate).

Time Taken: ----APPLICABLE METHOD OF TESTING Performance:

Simulate Actual X Unit: 2 Setting: In-Plant Simulator Admin X Time Critical:

Yes X No Time Limit 15 min Alternate Path: Yes No X EVALUATION Performer:

_______________________

_ JPM: Pass --Fail __ Remedial Training Required:

Yes No __ Comments:

Comments reviewed with Performer Evaluator Signature:

____________

_ Date: ----SOT-ADM-JP-301-A 15 Page 8 of 8 REV. 02 TASK CONDITIONS:

• • This is a Time Critical JPM** Time begins when directed by the evaluator

1. Unit One is operating at 100% power. 2. Unit Two is operating at 100% power with DG4 under clearance when the following event occurs (consider all items that exceed EAL thresholds occur at the same time). Unit Two Event Description A seismic event with a magnitude of 0.12g is felt on plant site and is verified / validated by the National Earthquake Center results in a loss of Corporate Telephone Communications System (Voicenet), Commercial Telephones, and NRC Emergency Telecommunications System. The Main Turbine Trips and a majority of the control rods remain withdrawn.

A Manual Scram has been inserted, the Mode Switch has been placed in Shutdown, ARI has been initiated, and reactor power indicates 20%. LEP-02 (Alternate Control Rod Insertions) actions and SLC injection are in progress.

Current indicated reactor power is 12%. Reactor Water level is being maintained in a band of +60 to +90 inches. Reactor Pressure is being controlled by EHC at 945 psig. NO radiological releases are in progress at this time. There are no indications of an Onsite Security Event. INITIATING CUE: You are to evaluate the above event as the Control Room Site Emergency Coordinator (SEC) and determine the highest required classification including the EAL Identifier for Unit Two ONLY: 1. Write the required Classification along with the EAL identifier in the table below. 2. Raise your hand when complete to have the evaluator stop the evaluation Time and collect your cue sheet. CLASSIFICATION EAL IDENTIFIER LESSON TITLE: REVISION NO: CAROLINA POWER & LIGHT COMPANY BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE IN-PLANT NRC ADMIN EC (RO/SRO) Develop a Clearance Boundary -RBCCW Pump 2C. 01 Develop a Clearance Boundary -RBCCW Pump 2C. SAFETY CONSIDERATIONS:

None EVALUATOR NOTES: (Do not read to trainee) 1. The applicable procedure section WILL be provided to the trainee. 2. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021, Appendix E, or similar to the trainee. 3. If desired, the evaluator may provide the examinee with a copy of OPS-NGGC-1301, 20P-21, and the referenced prints. 4. The evaluator should have available copies of prints 0-02538 Sheets 1 & 2 and LL-09241 sheets 8 & 24 to support performance of JPM, or perform JPM in a location where a print machine is available.

5. Provide examinee a marked up copy of Attachment 4 of OPS-NGGC-1301 with "N/A" in the Clearance No. and Action column. Read the following to trainee. TASK CONDITIONS:

1. You are an operator in the Work Control Center. PASSPORT (Equipment Tag Out) is not available for use. No historical clearances are available for review. 2. A Clearance has been requested by maintenance to place RBCCW Pump 2C under clearance for pump packing replacement.

3. RBCCW Pumps 2A and 2B will remain running. INITIATING CUE: The WCC SRO directs you to propose a Clearance Boundary for RBCCW Pump 2C by completing Attachment 4 of OPS-NGGC-1301, Equipment Clearance procedure.

LOT -OJT -J P-20 1-E04 Page 2 REV. 01 Develop a Clearance Boundary -RBCCW Pump 2C. ANSWER KEY ATTACHMENT 4 Sheet 1 of 1 OPERATIONS CLEARANCE TAG SHEET Clearance No. N/A PAGE _1_of_1 INT NAME (PRINT) INT NAME (PRINT)

• Independent Verification Required?

Yes/No If No, N/A the Blocks Seq Action Type Tag Id Position EquipmenUComponent Completed By 1 N/A CIT 2-RCC-CS-449 OFF or RBCCW Pump 2C HUNG Control Switch 2 N/A RED 2-2XE-EA7-52 OPEN or RBCCW Pump 2C ** OFF Circuit Breaker 3 N/A RED 2-RCC-V34 CLOSED RBCCWPumpC

• • Discharge Valve 3 N/A RED 2-RCC-V30 CLOSED RBCCW Pump C ** Suction Valve 4 N/A RED 2-RCC-V128 OPEN RBCCW Pump C Suction Drain Valve ** 4 N/A RED 2-RCC-V129 OPEN RBCCW Pump C Discharge Drain Valve 4 N/A RED 2-RCC-V301 OPEN RBCCW Pump C Casing ** Vent Valve Sequence must have either switch off or breaker open prior to operating discharge/suction valves, and discharge/suction valves closed before operating drain/vent valves. ** Minimum requirements for satisfactory boundary are breaker, discharge valve, suction valve, at least one drain valve, and vent valve. IV By LOT -OJT -JP-201-E04 Page 3 REV. 01 Develop a Clearance Boundary -RBCCW Pump 2C. PERFORMANCE CHECKLIST I NOTE: If desired, the examiner may providea copy OPS-NGGC-1301 to the examinee.

NOTE: Sequence is assumed unless denoted in the Comments.

Step 1 -Obtain a current revision of OPS-NGGC-1301.

Current Revision of OPS-NGGC-1301 obtained.

SAT/UNSAT*

TIME START ____ _ NOTE: NOTE: If desired, the examiner may provide a copy of the required prints to the examinee (0-02538, Sheets 1 & 2, LL-9241, Sheets 8 & 24) and 20P-21 lineups. CITs can have position OFF or In-Place I HUNG Step 2 -Obtain copies of required prints (0-02538, Sheets 1 & 2, LL-09241, Sheets 8 & 24). Drawings 0-02538, Sheets 1 & 2, LL-09241, Sheets 8 & 24 obtained.

SAT/UNSAT*

Step 3 -Identify control switch 2-RCC-CS-449 should be placed to OFF. Determine 2-RCC-CS-449 should be placed to OFF. SAT/UNSAT*

Step 4 -Identify breaker 2XE Compt EA7 should be placed to OFF. Determine 2XE Compt EA 7 should be placed to OFF. ** CRITICAL STEP ** SAT/UNSAT*

LOT -OJT -JP-201-E04 Page 4 REV. 01 NOTE: Develop a Clearance Boundary -RBCCW Pump 2C. Sequence of suction or discharge valve closing .is not critical so long as the pump motor is de:..energized.

Step 5 -Identify discharge valve 2-RCC-V34 should be closed. Determine 2-RCC-V34 should be closed. ** CRITICAL STEP ** SA T1UNSAT* Step 6 -Identify suction valve 2-RCC-V30 should be closed. NOTE: NOTE: Determine 2-RCC-V30 should be closed. ** CRITICAL STEP ** SAT/UNSAT*

Either2-RCC-V128or 2-RCC-V129is critical.

Both are notrequired to meet critical task standards.

Sequence of drain or vent valve opening is not critical.

Step 7 -Identify drain valve 2-RCC-V128 should be open. Determine 2-RCC-V128 should be open. ** CRITICAL STEP ** SAT/UNSAT*

Step 8 -Identify drain valve 2-RCC-V129 should be open. Determine 2-RCC-V129 should be open. ** CRITICAL STEP ** SAT/UNSAT*

Step 9 -Identify vent valve 2-RCC-V301 should be open. Determine 2-RCC-V301 should be open. ** CRITICAL STEP ** SAT/UNSAT*

LOT -OJT -J P-20 1-E04 Page 5 REV. 01 Develop a Clearance Boundary -RBCCW Pump 2C. Step 10 -Submit proposed boundary to WCC SRO. Proposed boundary submitted to wee SRO. SAT/UNSAT*

TERMINATING CUE: When the proposed boundary.

has been submitted, thisJPM is complete.

.. .. TIME COMPLETED

____ _

• Comments required for any step evaluated as UNSAT. LOT -OJT -JP-201-E04 Page 6 REV. 01 Develop a Clearance Boundary -RBCCW Pump 2C. RELATED TASKS: 299020B301, Develop A Clearance per OPS-NGGC-1301 KIA REFERENCE AND IMPORTANCE RATING: 2.2.13 (4.1/4.3)

Knowledge of tagging and clearance procedures

REFERENCES:

OPS-NGGC-1301 20P-21 0-02538, Sheets 1 & 2 LL-09241, Sheets 8 & 24 TOOLS AND EQUIPMENT:

Referenced prints and procedures.

SAFETY FUNCTION (from NUREG 1123, Rev 2.): A.2 -Equipment Control REASON FOR REVISION:

Corrected typos and added NGGC-1301, Attachment

4. LOT -OJT -J P-20 1-E04 Page 7 REV. 01 Develop a Clearance Boundary -RBCCW Pump 2C. Time Required for Completion: Minutes (approximate).

Time Taken: ------APPLICABLE METHOD OF TESTING Performance:

Simulate X Actual Unit: 2 Setting: In-Plant Simulator Admin _X:...:-__ Time Critical:

Yes No X Time Limit N/A Alternate Path: Yes No EVALUATION Trainee: ---------------------------

JPM: Pass __ Fail __ Remedial Training Required:

Yes No __ Did Trainee Verify Procedure?

Yes __ No __ _ (Each Student should verify one JPM per evaluation set) Comments:

Comments reviewed with Student Evaluator Signature:

Date: -----LOT-OJT-JP-201-E04 Page 8 REV. 01 ATTACHMENT 4 Sheet 10f 1 Operations Clearance Checklist Clearance No. --------Page __ of __ Checklist Type: Hang; Uft; Boundary Change (Circle one) INT NAME (PRINT) INT NAME (PRINT) " Independent Veriiication R,equired?

'YESINO If NO, r'lI.A. Ule Blocks Seq Action Type Tag Id Position Equipment/Component Completed.

Verified By By .. Continued Y l N QARECORD I OPS-NGGC-'130 1 Rev. 20 Page 95 of '121 TASK CONDITIONS:

1. You are an operator in the Work Control Center. PASSPORT (Equipment Tag Out) is not available for use. No historical clearances are available for review. 2. A Clearance has been requested by maintenance to place RBCCW Pump 2C under clearance for pump packing replacement.

3. RBCCW Pumps 2A and 2B will remain running. INITIATING CUE: The WCC SRO directs you to propose a Clearance Boundary for RBCCW Pump 2C by completing Attachment 4 of OPS-NGGC-1301, Equipment Clearance procedure.

ATTACHMENT 1 Page 1 of 2 Reactor Building Closed Cooling Water System Electrical Lineup PERSON(S)

PERFORMING OR VERIFYING LINEUP AND INITIALS (PRINT) REMARKS: Started: Completed:

Approved by: DOCUMENT any components NOT in required position and reason below. Page may be duplicated as necessary.

Date ________ _ Time ________ _ Date ________ _ Time ________ _ / SRO Date Time C Continuous Use 120P-21 Rev. 65 Page 49 of 79 )

ATTACHMENT 1 Page 2 of 2 Reactor Building Closed Cooling Water System Electrical Lineup Number Description Position/

Checked Indication Control Building -Emergency 120V AC Distribution Panel2A -EI23' -Cable Spread Area 16 CRD Pump 2A Cooling Water Inlet Valve ON RCC-V74 Solenoid 24 RBCCW Discharge Header Pressure Switch ON RCC-PS-672 26 RBCCW Inlet To Penetration Cooling System ON Valve RCC-V18S Solenoid Control Building -Emergency 120V AC Distribution Panel2B -EI23' -Cable Spread Area 16 CRD Pump 2B Cooling Water Inlet Valve ON RCC-V73 Solenoid 24 RBCCW Inlet To Water Chiller Isolation Valve ON RCC-VS1 Solenoid Reactor Building -Emergency 120V AC Distribution PaneI2A-RX

-EI20' -North -CRD-HCU Area 29 DWEDT Cooler Inlet Isolation Valve RCC-VS4 ON Solenoid Reactor Building -480V MCC 2XC -EI 20' -East DT6 RBCCW Drywell Supply Header Valve, ON RCC-VS2 DT7 RBCCW Drywell Discharge Header Valve, ON RCC-V28 Reactor Building -480V MCC 2XF -EI SO' -East ED7 RB Closed Cooling Water Pump B ON Reactor Building -480V MCC 2XE -EI SO' -West EA1 RB Closed Cooling Water Pump A ON 0.A7 J RB Closed Cooling Water Pump C ON Radwaste Building -MCC RWD-EI 23' -South BR9 RBCCW Pump 2D ON Radwaste Building -Between FDF Area and Waste Precoat Area-EI 23' 2-RCC-2D-PMP-DIS RBCCW Pump 2D Disconnect Switch OFF 120P-21 Rev. 55 Page 50 of 791 ATTACHMENT 2 Page 1 of 2 Reactor Building Closed Cooling Water System Panel Lineup PERSON(S)

PERFORMING OR VERIFYING LINEUP AND INITIALS (PRINT) REMARKS: Started: Completed:

Approved by: DOCUMENT any components NOT in required position and reason below. Page may be duplicated as necessary.

Date ________ _ Time ________ _ Date ________ _ Time ________ _ I SRO Date Time C Continuous Use \20P-21 Rev. 65 Page 51 of 791 ATTACHMENT 2 Page 2 of 2 Reactor Building Closed Cooling Water System Panel Lineup Number Description Position/

Checked Indication Control Building -RTGB Panel XU2 -EI49' -Control Room RCC-CS-447 RBCCW Pump 2A OFF RCC-CS-448 RBCCW Pump 2B OFF RCC-CS-449 RBCCW Pump 2C OFF RCC-V51 RBCCW to Sample Cooler OPEN RCC-V185 RBCCW to Pen Clg OPEN RCC-SV-1222B RBCCW Sample Valve from Recirc Pump A OPEN (Pen. X77-B) RCC-SV-1222C RBCCW Sample Valve from Recirc Pump B OPEN (Pen. X77-C) RCC-V28 RBCCW to OW 1501 Vlvs OPEN RCC-V52 RBCCW to OW 1501 Vlvs OPEN Radwaste Control Room -Panel HB9 RCC-V75 Cooling Wtr Sply OPEN 1 2 0 P-21 Rev. 65 Page 52 of 791 ATTACHMENT 3 Page 1 of 22 Reactor Building Closed Cooling Water System Valve Lineup PERSON(S)

PERFORMING OR VERIFYING LINEUP AND INITIALS (PRINT) REMARKS: Started: Completed:

Approved by: DOCUMENT any components NOT in required position and reason below. Page may be duplicated as necessary.

Date ________ _ Time ________ _ Date ________ _ Time ________ _ / SRO Date Time C Continuous Use 120P-21 Rev. 65 Page 53 of 791 ATTACHMENT 3 Page 2 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -North RHR Room -EI-17' -North RCC-V76 RBEDT Cooler Inlet Air Operated Valve CLOSED RCC-V76 RBEDT Cooler Inlet Air Operated Valve Manual NEUTRAL Actuator RIA-IV-215 Instrument Air Isolation Valve To RCC-V76 OPEN RIA-IV-115 Instrument Air Supply Valve To RCC-V76 OPEN RCC-V297 RBEDT Cooler Inlet Isolation Valve OPEN RCC-V98 RBEDT Cooler Shell Side Drain Valve CLOSED RCC-V5 RBEDT Cooler Outlet Throttle Valve LOCKED OPEN RCC-V97 RBEDT -Cooler Outlet Test Valve CLOSED RCC-FE-2258-1 High Pressure Root Valve To RCC-FE-2258 CLOSED RCC-FE-2258-2 Low Pressure Root Valve To RCC-FE-2258 CLOSED Reactor Building -Northeast Pipe Chase -EI 20' -Northeast RCC-FE-1822-1 High Pressure Root Valve To RCC-FE-1822 CLOSED RCC-FE-1822-2 Low Pressure Root Valve To RCC-FE-1822 CLOSED RCC-FE-2257

-1 High Pressure Root Valve To RCC-FE-2257 CLOSED RCC-FE-2257

-2 Low Pressure Root Valve To RCC-FE-2257 CLOSED Reactor Building -North RHR Pump Room (Overhead) -EI-17 RIA-IV-114 Instrument Air Supply Valve To RCC-V75 OPEN RIA-IV-214 Instrument Air Isolation Valve To RCC-V75 OPEN RIA-V401 Instrument Air Root Valve to RCC-V75 OPEN RCC-V6 RBCCW Return From Radwaste Isolation Valve OPEN Reactor Building -CRD Pump Area -EI-17' -South RCC-V144 CRD Pump B Cooling Water Inlet Test Valve CLOSED RCC-V73 CRD Pump B Cooling Water Inlet Air Operated CLOSED Valve RCC-V73 CRD Pump B Cooling Water Inlet Air Operated NEUTRAL Valve Manual Actuator 120P-21 Rev. 65 Page 54 of 791 ATTACHMENT 3 Page 3 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -CRD Pump Area -EI-17' -South RCC-U-V4 CRD Pump B Outboard Bearing Cooling OPEN Isolation Valve RCC-U-V5 CRD Pump B Inboard Bearing Cooling Isolation OPEN Valve RCC-U-V6 CRD Pump B Lube Oil Cooling Isolation Valve OPEN RCC-V142 CRD Pump B Cooling Water Outlet Test Valve CLOSED RCC-V7 CRD Pump B Cooling Water Outlet Valve OPEN RCC-V295 CRD Pump B Cooling Water Inlet Valve OPEN RCC-V294 CRD Pump A Cooling Water Inlet Valve OPEN RCC-V145 CRD Pump A Cooling Water Inlet Test Valve CLOSED RCC-V74 CRD Pump A Cooling Water Inlet Air Operated CLOSED Valve RCC-V74 CRD Pump A Cooling Water Inlet Air Operated NEUTRAL Valve Manual Actuator RNA-IV-552 Instrument Air Isolation Valve To RCC-V74 OPEN RNA-IV-488 Instrument Air Supply Valve To RCC-V74 OPEN RNA-IV-551 Instrument Air Isolation Valve To RCC-V73 OPEN RNA-IV-487 Instrument Air Supply Valve To RCC-V73 OPEN RCC-U-V1 CRD Pump A Outboard Bearing Cooling OPEN Isolation Valve RCC-U-V2 CRD Pump A Inboard Bearing Cooling Isolation OPEN Valve RCC-U-V3 CRD Pump A Lube Oil Cooling Isolation Valve OPEN RCC-V8 CRD Pump A Cooling Water Outlet Valve OPEN RCC-V143 CRD Pump A Cooling Water Outlet Test Valve CLOSED 120P-21 Rev. 65 Page 55 of 791 ATTACHMENT 3 Page 4 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -Post Accident Sampling Skid -EI 20' -South RCC-V315 RBCCW To Post Accident Sampling System LOCKED OPEN RCC-V316 RBCCW From Post Accident Sampling System LOCKED OPEN Reactor Building -Southeast Drywell Wall -EI 20' -Southeast RNA-IV-239 Instrument Air Isolation Valve To RCC-V51 OPEN RNA-IV-141 Instrument Air Supply Valve To RCC-V51 OPEN RNA-V144 Instrument Air Root Valve To RCC-V51 OPEN RCC-V29 Water Chiller Cooling Water Outlet Valve LOCKED OPEN RCC-FE-1824-1 High Pressure Root Valve To RCC-FE-1824 CLOSED RCC-FE-1824-2 Low Pressure Root Valve To RCC-FE-1824 CLOSED Reactor Building -Penetration X77 -EI. 20' -Az 295 0 RCC-IV-2374 Test Valve For Sample Valve RCC-SV-1222B LOCKED CLOSED RCC-IV-2375 Test Valve For Sample Valve RCC-SV-1222C LOCKED CLOSED Reactor Building -RBCCW HX Area -EI 50' -North RCC-V44 RBCCW HX C Inlet Valve OPEN RCC-V158 RBCCW HX C Inlet Test Valve CLOSED RCC-V115 RBCCW HX C Inlet Vent Valve CLOSED RCC-V116 RBCCW HX C Shell Side Drain Valve CLOSED RCC-V124 RBCCW HX C Outlet Vent Valve CLOSED RCC-V161 RBCCW HX C Outlet Test Valve CLOSED RCC-V47 RBCCW HX C Outlet Valve CLOSED RCC-V43 RBCCW HX B Inlet Valve OPEN RCC-V157 RBCCW HX B Inlet Test Valve CLOSED 1 2 0 P-21 Rev. 65 Page 56 of 791 ATTACHMENT 3 Page 5 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -RBCCW HX Area -EI 50' -North RCC-V112 RBCCW HX B Inlet Vent Valve CLOSED RCC-V113 RBCCW HX B Shell Side Drain Valve CLOSED RCC-V123 RBCCW HX B Outlet Vent Valve CLOSED RCC-V160 RBCCW HX B Outlet Test Valve CLOSED RCC-V46 RBCCW HX B Outlet Valve CLOSED RCC-V42 RBCCW HX A Inlet Valve OPEN RCC-V156 RBCCW HX A Inlet Test Valve CLOSED RCC-V109 RBCCW HX A Inlet Vent Valve CLOSED RCC-V110 RBCCW HX A Shell Side Drain Valve CLOSED RCC-V122 RBCCW HX A Outlet Vent Valve CLOSED RCC-V159 RBCCW HX A Outlet Test Valve CLOSED RCC-V45 RBCCW HX A Outlet Valve CLOSED RCC-V190 RBCCW HX Inlet Sample Root Valve OPEN RCC-V40 RBCCW Header Discharge Pressure Root Valve OPEN To RCC-PT-671, RCC-PS-672, And RCC-PSL-673 RCC-PT -671-3 RCC-PT-671 Instrument Isolation Valve OPEN RCC-PT -671-6 RCC-PT-671 Instrument Drain Valve CLOSED RCC-PS-672-3 RCC-PS-672 Instrument Isolation Valve OPEN RCC-PS-672-6 RCC-PS-672 Instrument Drain Valve CLOSED RCC-PS-672-7 RCC-PT -671, RCC-PS-672, And RCC-PSL-673 CLOSED Instrument Line Drain Valve RCC-PSL-673-3 RCC-PSL-673 Instrument Isolation Valve OPEN RCC-PSL-673-6 RCC-PSL-673 Instrument Drain Valve CLOSED RCC-FE-2256-1 High Pressure Root Valve To RCC-FE-2256 CLOSED RCC-FE-2256-2 Low Pressure Root Valve To RCC-FE-2256 CLOSED 120P-21 Rev. 65 Page 57 of 791 ATTACHMENT 3 Page 6 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -RBCCW Pump Area -EI 50' -North RCC-V102 Chemical Addition Tank Inlet Valve CLOSED RCC-V103 Chemical Addition Tank Outlet Valve CLOSED RCC-V104 Chemical Addition Tank Vent Valve CLOSED RCC-V105 Chemical Addition Tank Drain Valve CLOSED RCC-V326 RBCCW Return From Post Accident Sampling LOCKED System Valve OPEN RCC-V317 RBCCW Supply To Post Accident Sampling LOCKED System Valve OPEN '(RCC-V30 RBCCW Pump C Suction Valve OPEN RCC-V58 RBCCW Pump C Suction Test Valve CLOSED /(

RBCCW Pump C Suction Drain Valve CLOSED )

RBCCW Pump C Casing Vent Valve CLOSED (' vi RBCCW Pump C Discharge Drain Valve CLOSED _"..;1" RCC-V117 RBCCW Pump C Discharge Pressure Root OPEN Valve To RCC-PI-668 RCC-PI-668-3 RCC-PI-668 Instrument Isolation Valve OPEN RCC-PI-668-6 RCC-PI-668 Instrument Drain Valve CLOSED (:" "RCC-V34 -) RBCCW Pump C Discharge Valve OPEN RCC-V31 RBCCW Pump B Suction Valve OPEN RCC-V63 RBCCW Pump B Suction Test Valve CLOSED RCC-V133 RBCCW Pump B Suction Drain Valve CLOSED RCC-V302 RBCCW Pump B Casing Vent Valve CLOSED RCC-V134 RBCCW Pump B Discharge Drain Valve CLOSED RCC-V118 RBCCW Pump B Discharge Pressure Root OPEN Valve To RCC-PI-669 1 2 0 P-21 Rev. 65 Page 58 of 791 ATTACHMENT 3 Page 7 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -RBCCW Pump Area -EI 50' -North RCC-PI-669-3 RCC-PI-669 Instrument Isolation Valve OPEN RCC-PI-669-6 RCC-PI-669 Instrument Drain Valve CLOSED RCC-V36 RBCCW Pump B Discharge Valve OPEN RCC-V32 RBCCW Pump A Suction Valve OPEN RCC-V67 RBCCW Pump A Suction Test Valve CLOSED RCC-V138 RBCCW Pump A Suction Drain Valve CLOSED RCC-V303 RBCCW Pump A Casing Vent Valve CLOSED RCC-V139 RBCCW Pump A Discharge Drain Valve CLOSED RCC-V119 RBCCW Pump A Discharge Pressure Root OPEN Valve To RCC-PI-670 RCC-PI-670-3 RCC-PI-670 Instrument Isolation Valve OPEN RCC-PI-670-6 RCC-PI-670 Instrument Drain Valve CLOSED RCC-V38 RBCCW Pump A Discharge Valve OPEN RCC-V61 RBCCW Suction Pressure Root Valve To OPEN RCC-PI-666 RCC-PI-666-3 RCC-PI-666 Instrument Isolation Valve OPEN RCC-PI-666-6 RCC-PI-666 Instrument Drain Valve CLOSED RCC-V300 Pass Coolers 604 & 605 Cooling Return Isolation LOCKED Valve/Chemical Addition Injection Isolation Valve OPEN RCC-V39 RCC-TV-695 Inlet Valve OPEN RCC-V41 RCC-TV-695 Outlet Valve OPEN RNA-IV-243 Instrument Air Isolation Valve To RCC-TIC-695 OPEN RNA-IV-145 Instrument Air Supply Valve To RCC-TIC-695 OPEN And RCC-TV-695 RCC-FE-1821-1 High Pressure Root Valve To RCC-FE-1821 CLOSED RCC-FE-1821-2 Low Pressure Root Valve To RCC-FE-1821 CLOSED 120P-21 Rev. 65 Page 59 of 791 ATTACHMENT 3 Page 8 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position!

Checked Indication Reactor Building -RBCCW Pump Area -EI 50' -North RNA-IV-555 Instrument Air Isolation Valve To RCC-TV-695 OPEN RNA-V179 Instrument Air Root Valve To RCC-TI-695 And OPEN RCC-TV-695 Reactor Building -RWCU Pump and Heat Exchanger Room -EI 50' RCC-V149 RWCU Pump B Cooler Inlet Test Valve CLOSED RCC-V304 RWCU Pump B Cooler Inlet Valve OPEN RCC-V5087 RWCU Pump B Gear Box Cooling Isolation OPEN Valve RCC-V5088 RWCU Pump B Cooler Outlet Throttle Valve OPEN RCC-V306 RWCU Pump B Cooler Outlet Isolation Valve OPEN RCC-V147 RWCU Pump B Cooler Outlet Test Valve CLOSED RCC-V172 RWCU Pump B Cooler Outlet Sample Root OPEN Valve RCC-FE-2888-1 High Pressure Root Valve To RCC-FE-2888 CLOSED RCC-FE-2888-2 Low Pressure Root Valve To RCC-FE-2888 CLOSED RCC-V84 RWCU Non-Regenerative HX Shell Side Outlet CLOSED Drain Valve RCC-V150 RWCU Non-Regenerative HX Shell Side Outlet CLOSED Test Valve RCC-V3 RWCU Non-Regenerative HX Shell Side Outlet OPEN Valve RCC-V4 RWCU Non-Regenerative HX Shell Side Outlet LOCKED Throttle Valve THROTTLED 13.25 turns from full open RCC-V174 RWCU Non-Regenerative HX Outlet Sample OPEN Root Valve 120P-21 Rev. 65 Page 60 of 791 ATTACHMENT 3 Page 9 of 22 Reactor Building Closed Cooling Water System Valve Lineup Number Description Position/

Checked Indication Reactor Building -RWCU Pump and Heat Exchanger Room -EI 50' RCC-V83 RWCU Non-Regenerative HX A Shell Side Vent CLOSED Valve RCC-V86 RWCU Non-Regenerative HX Shell Side Drain CLOSED Valve RCC-V151 RWCU Non-Regenerative HX Inlet Test Valve CLOSED RCC-V50 RWCU Non-Regenerative HX Inlet Valve OPEN RCC-V183 RWCU Non-Regenerative HX Shell Side Sample OPEN Root Valve RCC-V85 RWCU Non-Regenerative HX B Shell Side Vent CLOSED Valve RCC-V148 RWCU Pump A Cooler Inlet Test Valve CLOSED RCC-V307 RWCU Pump A Cooler Inlet Valve OPEN RCC-V5089 RWCU Pump A Gear Box Cooling Isolation OPEN Valve RCC-V5090 RWCU Pump A Cooler Outlet Throttle Valve OPEN RCC-V5091 RWCU Pump A Casing Cooling Isolation Valve OPEN RCC-V309 RWCU Pump A Cooler Outlet Isolation Valve OPEN RCC-V146 RWCU Pump A Cooler Outlet Test Valve CLOSED RCC-V170 RWCU Pump A Cooler Outlet Sample Root OPEN Valve RCC-FE-2887

-1 High Pressure Root Valve To RCC-FE-2887 CLOSED RCC-FE-2887

-2 Low Pressure Root Valve To RCC-FE-2887 CLOSED Reactor Building -RWCU Precoat Pump Area -EI 80' -East RCC-V181 RBCCW Outlet Test Valve From RWCU Precoat CLOSED Pump RCC-V121 RBCCW Outlet Isolation Valve To RWCU OPEN Precoat Pump 1 2 0 P-21 Rev. 65 Page 61 of 791 CAROLINA POWER & LIGHT COMPANY BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE IN-PLANT NRC ADMIN RC (RO/SRO) LESSON TITLE: Determine Stay Time Limitations in High Radiation Area LESSON NUMBER: LOT-ADM-JP-102-A03 REVISION NO: 01 Determine Stay Time Limitations in High Radiation Area SAFETY CONSIDERATIONS:

None. EVALUATOR NOTES: (Do not read to performer)

1. The attached Radiological Survey WILL be provided to the performer.

2. If this is the first JPM of the JPM set, read the JPM briefing contained in NUREG 1021, Appendix E, or similar to the performer.

3. Task standards (i.e. pass/fail criteria) for each JPM step are ITALICIZED below each step. LOT-ADM-JP-102-A03 Page 2 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area Read the following to the JPM performer.

TASK CONDITIONS:

Two workers will be required to perform an emergent leak repair on the Unit Two Steam Packing Exhauster 2A. Worker #1 has accumulated 500 mrem this year (Progress Energy dose only). Worker #2 has accumulated 1615 mrem this year (Progress Energy dose only). The following times for each worker have been estimated for performance of the job. 1. Staging time in access area directly outside the A SJAE Room 45 minutes 2. Staging time in area directly inside A SJAE room access door 15 minutes 3. Work time at the 2A SPE 1.5 hours0.208 days <br />0.0298 weeks <br />0.00685 months <br /> Following completion of the job, an additional 15 mrem per worker will be received during staging activities.

INITIATING CUE: Using the information above and the radiological survey provided, evaluate the leak repair job for possible performance under current plant conditions to include: 1. Determine the projected dose each worker would receive to perform the leak repair. (Assume the same task times for both workers) 2. Based upon the projected total accumulated dose, determine if any Progress Energy administrative dose limitations will be exceeded.

LOT-ADM-JP-102-A03 Page 3 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area ANSWER KEY WORKER #1 WORKER #2 INITIAL DOSE 500 mrem 1615 mrem PROJECTED DOSE 510.25 mrem 510.25 mrem PROJECTED TOTAL DOSE 1010.25 mrem 2125.25 mrem BNP ADMIN DOSE LIMIT EXCEEDED YES NO (Circle Yes or No) LOT -ADM-JP-1 02-A03 Page 4 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless denoted in the Comments.

NOTE: Provide the examinee the JPM attached survey map. TIME START ___ _ Step 1 -Determines dose for each worker as follows: a. Staging time in access area directly outside the SJAE room (45 min) 0.75 Hr X 5 mrlhr = 3.75 mrem Estimate 3.75 mrem dose accumulation.

• • CRITICAL STEP** SAT/UNSAT*

b. Staging time in area directly inside SJAE room access door (15 min) 0.25 Hr X 16 mrlhr = 4 mrem Estimate 4 mrem dose accumulation.

• • CRITICAL STEP** SAT/UNSAT*

c. Work time at the SPE 2A 1.5 Hr X 325 mrlhr = 487.5 mrem Estimate 487.5 mrem dose accumulation.

• • CRITICAL STEP** SAT/UNSAT*

d. An additional 15 mr will be accumulated once the job is done for de-staging activities.

• • CRITICAL STEP** SAT/UNSAT*

LOT -ADM-JP-1 02-A03 Page 5 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area e. Total = 3.75 + 4 + 487.5 + 15 = 510.25 mrem Determines Total dose for job completion

-510.25 mrem. ** CRITICAL STEP** SAT/UNSAT*

Step 2 -Determines that Worker #2 would exceed the PGN administrative limit of 2 REM per calendar year if the estimated dose were accumulated.

Worker #1 Worker #2 500 mr + 510.25 mr = 1010.25 mr << 2R limit) 1615 mr + 510.25 mr = 2125.25 mr (> 2R limit) ** CRITICAL STEP** SAT/UNSAT*

TERMINATING CUE: When the total dose for each worker has been determined and the administrative limits addressed, the JPM is complete.

TIME COMPLETED

____ _ *NOTE: Comments required for any step evaluated as UNSAT. LOT-ADM-JP-102-A03 Page 6 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area RELATED TASKS: (Specify)

KIA REFERENCE AND IMPORTANCE RATING: 2.3.7 Ability to comply with radiation work permit requirements during normal and abnormal conditions 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

REFERENCES:

HPS-NGGC-0003 DOS-NGGC-0004 TOOLS AND EQUIPMENT:

Calculator.

JPM Survey Map SAFETY FUNCTION (from NUREG 1123): A.3 Radiation Control REASON FOR REVISION:

2010-1 NRC Exam -Modified to different location and rates. LOT -ADM-JP-1 02-A03 Page 7 of 9 Rev.01 Determine Stay Time Limitations in High Radiation Area Validation Time: __

Minutes (approximate).

Time Taken: ___ _ APPLICABLE METHOD OF TESTING Performance:

Simulate Setting: In-Plant Time Critical:

Yes Alternate Path: Yes Actual Simulator No L No EVALUATION Unit: 2 Admin X Time Limit N/A Performer:

___________________

_ JPM: Pass Fail --Remedial Training Required:

Yes No __ (Reference T AP-411 for evaluation guidance)

Comments:

Comments reviewed with Performer Evaluator Signature:

____________

_ Date: -----LOT -ADM-J P-1 02-A03 Page 8 of 9 Rev.01 2T06 BRUNSWICK NUCLEAR PLANT RADIOLOGICAL SURVEY SURVEY NO 1717 -003 -DATE:

0600 DOSE RATE CONTAMINATION TIME: INST. TYPE Telelector IJ"D aD IJ"D aD SURVEYOR Joe H. Physics SERIAL NO. T191A L-l77 M52 pCC' -;t:

INSTISN SIGNATURE 100% EFF 0.10 0.'18 -;'1' RX POVVER RE'y'IEWED H2SCFM 3*3 BKGD. 40 CPM 22 CPM 2 TB 20' "An STEAM JET AIR EJECTOR ROOM UNIT __ BLDG --EL --AREA ______________

_ RWPNO. Cor .... 1PONENT OR TAG NO. SPE 2A WR/JO 002seS62 WORK PERFOR!.,.1ED

_:-_e"'_h_R_e;..p"_ir

________________________

_

PlJo.TFOR\f I 103 1182 1--1 AF"l"ER CD COt..:JE"iG*=-:

CD 325 X I I ! Work Area 1 'Note: This is High nOise area ' .... hile operating.

5 RA -R!o.DiA 710N AREA -HIGH RADIA 71C1N AREA LMRA -LOCK::D H0H AREA RCA C:)NT::;:OL CA HeA -HIG'i COIFAMINAnCN AREA AA -AiRBORNE AREA R8PE L::"

E ..

SMEAR RESULTS IN DPMI100 eM" B=9ETA IH mRAD/HRI!OOCM

M=CF'f,l.'MASLIN NO. RESUL:S *:MD.L, A TASK CONDITIONS:

Two workers will be required to perform an emergent leak repair on the Unit Two Steam Packing Exhauster 2A. Worker #1 has accumulated 500 mrem this year (Progress Energy dose only). Worker #2 has accumulated 1615 mrem this year (Progress Energy dose only). The following times for each worker have been estimated for performance of the job. 1. Staging time in access area directly outside the A SJAE Room 45 minutes 2. Staging time in area directly inside A SJAE room access door 15 minutes 3. Work time at the 2A SPE 1.5 hours0.208 days <br />0.0298 weeks <br />0.00685 months <br /> Following completion of the job, an additional 15 mrem per worker will be received during staging activities.

INITIATING CUE: Using the information above and the radiological survey provided, evaluate the leak repair job for possible performance under current plant conditions to include: 1. Determine the projected dose each worker would receive to perform the leak repair. (Assume the same task times for both workers) 2. Based upon the projected total accumulated dose, determine if any Progress Energy administrative dose limitations will be exceeded.

LESSON TITLE: LESSON NUMBER: REVISION NO: PROGRESS ENERGY CAROLINAS BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE ADMIN NRC ADMIN C001 (RO/SRO) Core Performance Parameter Check LOT-ADM-201-D02 01 Core Performance Parameter Check SAFETY CONSIDERATIONS:

1. NONE. EVALUATOR NOTES: (Do not read to examinee)

1. The applicable procedure WILL be provided to the examinee.

2. If this is the first JPM of the JPM set, read the JPM briefing contained in NUREG 1021, Appendix E, or similar to the examinee.

3. This JPM may be administered in the simulator, control room, or classroom setting. 4. Performance Frequency A (every 24 hrs when >25% RTP) is required.

Read the following to examinee.

TASK CONDITIONS:

1. The CODSR requires performance of 2PT-01.11, Core Performance Parameter Check, on Unit Two. 2. All applicable prerequisites of 2PT-01.11 are met. 3. No APRM gain adjustments will be performed during this test. 4. If independent verification, notifications, or other actions are required during performance, assume the action is complete as applicable by providing the following in the comments area of this cue sheet.

• The Step number followed by a short description of the Action. 5. Cover page information will be filled out by the Unit SCO. INITIATING CUE: The Unit SCO has directed you to complete 2PT-01.11, Core Performance Parameter Check, utilizing the Core Performance Log provided and provide the results below upon completion.

2PT-01.11, Core Performance Parameter Check, results are (circle choice): SAT UNSAT Comments (if necessary):

_______________________

_ LOT -ADM-JP-201-D02 Page 2 of 12 Rev. 01 Core Performance Parameter Check ANSWER KEY 2PT-01.11, Core Performance Parameter Check, results are (circle choice): UNSAT Comments (if necessary):

______________________

_ Step 7.1.7 -APRM 1 gain adjustment factor is greater than 1.00 Step 7.1.11 -CMFLPD is greater than 1.00, Unit SCO notified.

Step 7.1.11.1 -CMFLPD is greater than 1.00, On-Shift Reactor Engineer notified to take action to restore CMFLPD to an acceptable value. Step 7.1.7 -Notify Unit SCO test is complete.

LOT -ADM-JP-201-D02 Page 3 of 12 Rev. 01 Core Performance Parameter Check PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless denoted in the Comments.

Step 1 -Obtain current revision of 2PT-01.11 and verify if applicable.

Current revision of 2PT-01. 11 obtained and verified as required.

SAT/UNSAT*

TIME START ___ _ PROMPT: Provide the Core Performance Log attached to this JPM to be evaluated by candidate.

NOTE: APRM gain adjustment and PPC usage are not being evaluated by this JPM. Step 2 -OBTAIN an edit of the process computer core performance program. Core Performance Log obtained from Evaluator.

SAT/UNSAT*

NOTE: No failed sensors, this step may be initialed or marked N/A. Step 3 -IF there are failed inputs on the failed sensor list, THEN ENSURE correct values have been substituted where appropriate.

Determines no failed sensors exist and marks this step as NIA. SATIUNSAT*

Step 4 -Evaluates value of NSSFFLG is equal to 1. Determines NSSFFLG is equal to 1 and proceeds to Step 7.1.4. SATIUNSAT*

NOTE: steps 7.1.3.1.aand b (NSSFFLG is NOTequal to 1) should be marked N/A. LOT -ADM-J P-20 1-D02 Page 4 of 12 Rev. 01 Core Performance Parameter Check Step 5 -Evaluates value of WFWFFLG, NSSCUFLG, and NSSCRFLG all equal to 1. NOTE: Determines WFWFFLG, NSSCUFLG, and NSSCRFLG are all equal to 1 and proceeds to Step 7.1.4. SAT/UNSAT*

Steps 7.1.4.1.a and b (WFWFFLG OR NSSCUFLG OR NSSCRFLG are NOT equal to 1) should be marked N/A. Step 6 -Determine if criteria listed in Section 6.1 of 2PT-1.11 are met. NOTE: NOTE: Checks core performance log and verifies value for CMFLCPR :::1.00. ** CRITICAL STEP ** SAT/UNSAT*

Checks core performance log and verifies value for CMAPRA T :::1.00. ** CRITICAL STEP ** SA TlUNSAT* Checks core performance log and verifies value for CMFDLRX :::1.00. ** CRITICAL STEP ** SAT/UNSAT*

Steps 7.1.6 and 7.1.6.1 (limits specified inSeclion 6.1 are NOT satisfied) should be marked N/A. Report of APRM1 GAF >1.00 is not required to satisfy Critical Step. Step 7 -Determine if the acceptance criteria listed in Section 6.2 are satisfied (at least three APRM gain adjustment factors are less than or equal to 1.00). Determines APRMS 2, 3, and 4 are within limits and APRM 1 is above the limit. ** CRITICAL STEP ** SAT/UNSAT*

PROMPT: If informed that APRM 1 GAF is > 1.00, acknowledge the report and inform the trainee the APRM will be adjusted later. Direct the trainee to continue the PT. LOT-ADM-JP-201-D02 Page 5 of 12 Rev. 01 Core Performance Parameter Check NOTE: Steps 7.1.8 and 7.1:9 (APRMgain adjustment) should be marked N/A. Step 8 -Determine if CMFLPD is 1.00. Checks core performance log and recognizes CMFLPD is >1.00. ** CRITICAL STEP ** SAT/UNSAT*

NOTE: Note prior to Step 7.1.10 -Reference 2.5 states.that NRC expects LHGR monitoring (CMFLPD) to remain the same as CPR and APLHGR forGNF fuel (i.e., to restore LHGR within limits within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />, or be below 23% power within the next 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />). This information may be included in the comments, but is not required to satisfy the JPM. Step 9 -NOTIFY the Unit SCO that CMFLPD is greater than 1.00. Notifies Unit SCO that CMFLPD is greater than 1.00 ** CRITICAL STEP ** SAT/UNSAT*

Step 10-NOTIFY the On-Shift Reactor Engineer to take action to restore CMFLPD to an acceptable value. Notifies RE to restore CMFLPD to 1.0. ** CRITICAL STEP ** SAT/UNSAT*

Step 11 -ENSURE Acceptance Criteria listed in Section 6.0 have been satisfied.

Determines Acceptance Criteria has been satisfied.

• • CRITICAL STEP ** SAT/UNSAT*

NOTE: Attaching edits and coversheet information is not required for classroom performance.

Step 12 -ATTACH the collected edits. Attaches the collected edits to the procedure.

SAT/UNSAT*

LOT -ADM-JP-20 1-002 Page 6 of 12 Rev. 01 Core Performance Parameter Check NOTE: Section 7.2 should be marked N/A. NOTE: Attachments 2 and 3 may be discarded. , Step 13 -ENSURE the required information has been recorded on the cover page. Cover page information to be filled out by Unit SCQ per cue. SAT/UNSAT*

Step 14 -Notify the Unit SCQ that 2PT-01.11 is completed SAT with CMFLPO greater than 1.00 which required RE notification to restore 1.0 and APRM 1 requires adjustment.

Unit SCQ notified 2PT-01.11 is completed SAT with APRM 1 requiring gain adjustment to restore GAF to 1.0 and CMFLPD is > 1. O. RE has been notified to take action to restore CMFLPD to 1. O. ** CRITICAL STEP ** SAT/UNSAT*

TERMINATING CUE: 2PT..;01.11 acceptance criteria verified.

and the Unit SCQ informed the results are satisfactory.

TIME COMPLETED

____ _

• Comments required for any step evaluated as UNSAT. LQT -AOM-JP-20 1-002 Page 7 of 12 Rev. 01 Core Performance Parameter Check LIST OF REFERENCES RELATED TASKS: 2992018201:

Perform Daily Surveillance Report Per 01-3.1 or 01-3.2. KIA REFERENCE AND IMPORTANCE RATING: Generic 2.1.7 (3.7/4.4)

Ability to evaluate plant performance and make operational judgments based on operating characteristics 1 reactor behavior 1 and instrument interpretation.

REFERENCES:

2PT-01.11 TOOLS AND EQUIPMENT:

None. SAFETY FUNCTION (from NUREG 1123, Rev 2): 7 -Instrumentation (APRM System) REASON FOR REVISION:

Revised to allow classroom performance.

Unit 2 process computer core performance program edit will be provided to examinee.

LOT -ADM-J P-20 1-002 Page 8 of 12 Rev. 01 Core Performance Parameter Check Time Required for Completion: Minutes (approximate).

Time Taken: ______ _ APPLICABLE METHOD OF TESTING Performance:

Simulate Actual X Unit: 2 Setting: In-Plant Simulator Admin X Time Critical:

Yes No L Time Limit N/A Alternate Path: Yes No EVALUATION Trainee: ________________________

_ JPM: Pass Fail ---Remedial Training Required:

Yes No __ _ Did Trainee Verify Procedure as Authorized Copy?: Yes No __ _ (Each Student should verify one JPM per evaluation set) Comments:

Comments reviewed with Performer Evaluator Signature:

Date: ----------------

---

LOT-ADM-JP-201-D02 Page 9 of 12 Rev. 01 TASK CONDITIONS:

1. The CQDSR requires performance of 2PT-01.11, Core Performance Parameter Check, on Unit Two. 2. All applicable prerequisites of 2PT-01.11 are met. 3. No APRM gain adjustments will be performed during this test. 4. If independent verification, notifications, or other actions are required during performance, assume the action is complete as applicable by providing the following in the comments area of this cue sheet.

• The Step number followed by a short description of the Action. 5. Cover page information will be filled out by the Unit SCQ. INITIATING CUE: The Unit SCQ has directed you to complete 2PT-01.11, Core Performance Parameter Check, utilizing the Core Performance Log provided and provide the results below upon completion.

2PT-01.11, Core Performance Parameter Check, results are (circle choice): SAT UNSAT BRUNSWICK-2 WK-1016 10TODAY-07.00.27 8902 MWD/MTU TRIGR=2HR CORE PERFORMANCE LOG ---LONG EDIT TL 0 TLO TBVINS NSS NFWT BOC TO NEOC CTP CALCULATION

HEAT BALANCE SYMMETRY : FULL STATE CONDITIONS FLOW (MLB/HR) GMWE 973.2 WT 78.84(102.4%)

CMWT 2905.9 (99.4%) WDP 78.50 PR 1044.2 PSIA NSSFFLG 1. DHS 20.85 BTU/LB WFW 12.72 CRD 0.0931 WFWFFLG 1. CYEXP 8902.3 MWD/MT NSSCUFLG 1. CAVEX 22911.1 MWD/MT NSSCRFLG 1. RODLINE 97.9% WD 36.40 NON_EO XE -1.7% LOCATION 2 3 AXIAL REL POWER 0.40 0.94 1.14 REGION REL POWER 0.94 1.00 0.92 RING REL POWER 0.96 1.14 1.06 APRM GAFS 1.02 1.00 0.98 4 1.18 1.03 1.14 0.99 CORE PARAMTRS CMEO 0.2791 CAEO 0.1634 CAOA 0.1538 CAVF 0.4548 CAPD 58.934 RWL 187.5460 CDLP 25.0146 DPCC 30.3681 KEFF 0.9971 5 6 7 1 .18 1.20 1 .21 1.08 1.05 0.93 1.14 1.09 0.65 NUCLEAR P-PCS P-PCFC CMFLCPR CMAPRAT CMFDLRX CMFLPD CMINCPR 8 9 1. 11 1.07 1.03 0.96 ***************

NUCLEAR LIMITS BY REGION ***************

7 8 9 0.867 13-40 0.860 31-40 0.885 39-40 0.750 17-42-15 0.759 33-44-15 0.767 41-36-15 0.688 09-36-13 0.704 33-42-13 0.702 43-36-13 0.622 11 13 0.652 33-36-18 1.155 41-36-13 -----------------

-+------------------

-+----------------

--4 5 6 0.838 13-32 0.821 31-32 0.865 39-32 0.758 09-20-13 0.714 27-28-18 0.769 43-34-15 0.700 11 13 0.663 33-34-18 0.895 41-34-13 0.611 09-20-13 0.566 27-28-17 0.895 43-34-15 -----------------

-+------------------

-+----------------

--1 2 3 0.864 13-14 0.822 21 -14 0.852 39-14 0.749 11-18-15 0.748 19-10-13 0.736 41-18-13 0.691 09 13 0.693 19-12-13 0.684 43-18-13 0.619 11-18-13 0.602 19-10-13 0.613 41-18-13 REV=usep06 CYCLE: 19 LIMITS LOCATION -0.06 33-36-19 -0.06 33-36-19 0.885 39-40 0.769 43-34-15 0.895 41-34-13 1 .155 41 13 1.528 39-40 10 11 12 0.99 0.90 0.68

• MFLCPR *

• MAPRAT *

• MFDLRX *

• MFLPD * **********

• AXIAL REL POWER* NODE NOTCH REL-POW EXP(GWD/MT) 25 0.187 24 00 0.608 23 02 0.784 D 22 04 0.939 21 06 1.035 20 08 1.088 19 10 1.102 18 12 1.097 17 14 1.078 C 16 16 1.035 -15 18 1.023 14 20 1. 111 13 22 1.116 12 24 1.128 11 26 1.150 B 10 28 1.181 9 30 1.203 8 32 1 .211 7 34 1.216 6 36 1.214 5 38 1.201 A_4 40 1.166 3 42 1.077 2 44 0.846 1 46 0.242 # NODES W/ LIMITS> 1 FLCPR APRAT FDLRX FLPD o o o 1 3.842 11 .154 18.881 23.791 27.184 29.224 30.344 30.901 31.147 30.565 30.935 29.084 29.780 30.578 31.298 31.895 32.380 33.056 33.871 34.675 35.119 34.452 31.119 23.396 6.163 ***************

THE 10 MOST LIMITING BUNDLES *****************

FLCPR LOC CPR LIMIT APRAT LOC APLHGR LIMIT FDLRX LOC LPD LIMIT FLPD LOC LPD LIMIT

---

0.909 39-40 1.518 1.381 0.867 43-34-15 8.01 9.25 0.895 41-34-13 9.53 0.00 1 . 1 55 41 13 8.45 13.40 0.905 39-36 1.525 1.381 0.866 41-36-15 8.09 9.35 0.704 33-42-13 9.43 0.00 1 .058 35 13 8.36 13.40 0.894 39-32 1.544 1.381 0.854 35-42-15 7.55 8.84 0.702 43-36-13 9.41 0.00 0.895 43-34-15 8.33 13.40 0.893 35-40 1.546 1.381 0.840 33-44-15 7.78 9.26 0.694 35-44-13 9.30 0.00 0.652 33-36-18 8.19 13.40 0.892 31-40 1.548 1.381 0.833 33-36-18 7.72 9.26 0.691 41-42-05 9.27 0.00 0.610 33-44-13 8.18 13.40 0.888 37-38 1.554 1.381 0.833 41-28-17 9.49 11.38 0.687 35-36-18 9.20 0.00 0.606 35-34-18 8.12 13.40 0.883 39-28 1.563 1.381 0.825 25-42-17 9.35 11.34 0.685 39-40-05 9.18 0.00 0.600 27-42-17 8.04 13.40 0.880 37-34 1.568 1.381 0.818 35-34-18 7.38 9.02 0.676 39-36-13 9.05 0.00 0.598 41-28-16 8.01 13.40 0.876 37-30 1.576 1.381 0.816 39-42-06 7.13 8.73 0.672 41-44-05 9.00 0.00 0.593 43-38-13 7.95 13.40 0.872 43-32 1.583 1.381 0.816 41-32-15 7.21 8.84 0.671 35-40-13 8.99 0.00 0.591 41-32-15 7.92 13.40 BRUNSWICK-2 WK-1016 10TODAY-07.00.27 8902 MWD/MTU TRIGR=2HR REV=usep06

CONTROL ROD DATA *************

02 06 10 14 18 22 26 30 34 38 42 46 50 51 51 47 47 43 26 20 26 43 39 39 35 26 16 08 16 26 35 31 31 27 20 08 18 08 20 27 23 23 19 26 16 06 12 26 19 15 15 11 26 20 26 11 07 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 RODS OUT OF MIRROR SYMMETRY:

26-35 26-19 34-19 SUBST. RODS: **********

CALIBRATED LPRM READINGS ************

• LPRM FAILED SENSORS* *OTHER FAILED SENSORS* 45 32.7 43.5 44.4 38.9 26.9 LOCATION STATUS SENSOR STATUS 46.2 60.6 60.2 56.7 37.5 54.3 59.7 59.3 56.9 44.3 52.1 52.4 54.7 52.9 88.1 37 26.7 46.0 53.3 54.2 52.7 39.6 D 38.8 63.3 55.6 56.5 60.3 57.2 C 43.0 61.2 58.9 60.6 61.6 57.9 B 36.2 54.6 45.7 47.7 52.5 86.6 A 29 34.0 48.0 55.6 58.3 54.2 44.4 49.7 59.7 52.0 55.3 55.6 60.0 60.4 59.4 58.7 58.3 60.2 60.2 53.4 48.0 45.5 43.1 48.7 56.5 21 32.1 48.6 54.7 50.5 52.2 42.1 46.0 62.7 54.6 48.3 54.5 60.3 55.4 61.0 59.6 57.4 61.6 61.2 51.0 47.3 47.8 45.9 49.9 55.9 13 39.5 47.5 48.7 44.8 31.0 57.4 62.4 60.6 58.7 44.8 66.9 60.1 59.5 60.6 53.7 66.8 51.5 46.5 53.4 51.5 05 31.9 32.1 26.7 45.9 46.5 37.2 53.1 55.3 43.5 49.3 55.8 37.0 04 12 20 28 36 44 BRUNSWICK NUCLEAR PLANT DATE COMPLETED

_________

_ FREQUENCY:

c Continuous Use UNIT _2_ % PWR GMWE __ SUPERVISOR

___________

_ REASON FOR TEST (check one or more) A. Once/24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> when operating 23% rated thermal power _ Routine Surveillance

_W/O# ______________________

_ B. Within 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after thermal power is 23% of rated thermal power _ Other (explain)

PLANT OPERATING MANUAL VOLUME X PERIODIC TEST UNIT 2 2PT-01.11 CORE PERFORMANCE PARAMETER CHECK REVISION 7 1 2PT-0 1.11 Rev. 7 Page 1 of 14/

1.0 PURPOSE This PT provides a procedure for obtaining the basic core performance parameters required by Technical Specifications and calibrates APRM channels to read greater than or equal to actual core thermal power. The procedure satisfies Technical Specifications SR 3.2.1.1, SR 3.2.2.1, SR 3.2.3.1, and SR 3.3.1.1.3.

2.0 REFERENCES

2.1 Technical Specifications 2.2 NEDE-24011-P-A (GESTAR II), Amendment 19 2.3 NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications 2.4 Core Operating Limits Report 2.5 Letter, L. M. Quintana to B. A. Morgan, "Linear Heat Generation Rate Monitoring," October 9,1989, LMQ: 89-241 2.6 001-72, Plant Process Computer System Operating Instruction 2.7 00P-55, Process and ERFIS Computer System Operating Procedure 2.8 OPT-01.8C, Hand Calculation of AGAFs 2.9 OPT-01.8D, Core Thermal Power Calculation 2.10 OENP-24.19, Operation of the BWR Process Computer Backup Program 2.11 20P-09, Neutron Monitoring System Operating Procedure 2.12 GE SIL 516 Supplement 1, Recirculation Drive Flow/Core Flow Correlation 2.13 NEDO-32465-A, licensing Topical Report: Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applicability GE Nuclear Energy, August 1996. 2.14 EC 49527, Modify Thermal Limits Surveillance Method to Account for Powerplex CPR Convergence Failures.

/2PT-01.11 Rev. 7 Page 2 of 14/

3.0 PRECAUTIONS AND LIMITATIONS 3.1 Special care should be taken in verifying data. Errors on the nonconservative side could cause operation that might lead to a violation of Technical Specifications.

3.2 IF the value of a core performance parameter exceeds its limits, THEN the Unit SCQ should be immediately notified.

3.3 During APRM gain adjustments, the plant should be held at a steady state operating condition.

During APRM gain adjustments, the affected APRMs may be bypassed.

3.4 IF APRM gain adjustments are made, THEN Independent Verification is required.

4.0 PREREQUISITES Thermal power is greater than or equal to 23% 5.0 SPECIAL TOOLS AND EQUIPMENT None 6.0 ACCEPTANCE CRITERIA OTE: Attachment 3 contains definitions and abbreviations of terms. 6.1 This PT is acceptable when it is shown by the certifying signature that the parameters have been obtained correctly according to this instruction and these conditions exist: 6.1.1 CMFLCPR is less than or equal to 1.0 (See Attachment 2 for parameter location on the core performance edit). 6.1.2 CMAPRAT is less than or equal to 1.0 (See Attachment 2 for parameter location on the core performance edit). 6.1.3 CMFDLRX is less than or equal to 1.0. (See Attachment 2 for parameter location on the core performance edit; applicable to AREVA fuel bundles only) 6.2 At least three operable APRMs are adjusted such that the APRM gain adjustment factors (GAFs) are less than or equal to 1.00. The APRM gain adjustment factor is determined by either the periodic NSS Core Performance Log (Attachment 2), Display 820 (Heat Balance/Core Mon), or hand calculation of AGAFs (OPT-01.8C).

IF APRM gain adjustments are performed, THEN the postadjustment AGAFs are verified by a second Display 820 or OPT-01.8C.

/2PT-01.11 Rev. 7 Page 3 of 14/

7.0 PROCEDURALSTEPS Initials NOTE: Attachments 2 and 3 may be discarded after completion of test. NOTE: IF unable to obtain the required process computer edits, THEN the On-Shift Reactor Engineer should be contacted prior to performance of Section 7.2. 7.1 Using the Process Computer NOTE: Plant process computer operating instructions and report codes are contained within procedures 001-72 and 00P-55. 7.1.1 OBTAIN an edit of the process computer core performance program. 7.1 .2 IF there are failed inputs on the failed sensor list, THEN ENSURE correct values have been substituted where appropriate.

NOTE: Attachment 2 shows the location of NSSFFLG on the Core Performance Log. 7.1.3 IF NSSFFLG is equal to 1, THEN GO TO Step 7.1.4. 1. IF NSSFFLG is NOT equal to 1, THEN PERFORM the following:

a. NOTIFY the On-Shift Reactor Engineer to evaluate if core flow is accurate for thermal limit calculations.

12PT-01.11

b. IF core flow was changed by the On-Shift Reactor Engineer in Step 7.1.3.1.a, THEN OBTAIN a new Core Performance Log edit. Rev. 7 Page 4 of 14/

7.0 PROCEDURAL STEPS Initials NOTE: Attachment 2 shows the respective locations of WFWFFLG, NSSCUFLG, and NSSCRFLG on the Core Performance Log. 7.1.4 IF WFWFFLG AND NSSCUFLG AND NSSCRFLG are equal to 1, THEN GO TO Step 7.1.5. 1. IF WFWFFLG OR NSSCUFLG OR NSSCRFLG are NOT equal to 1, THEN PERFORM the following:

a. NOTIFY the On-Shift Reactor Engineer to evaluate accuracy of inputs to the heat balance calculation.

b. IF inputs to the heat balance were changed by the On-Shift Reactor Engineer in Step 7.1.4.1.a, THEN OBTAIN a new Core Performance Log edit. 7.1.5 DETERMINE from the core performance edit, if criteria listed in Section 6.1 are met. NOTE: Locations on the core performance edit where thermal limit parameters are found are indicated on Attachment

2. 7.1.6 IF limits specified in Section 6.1 are NOT satisfied, THEN NOTIFY the Unit SCO of the condition.

1. NOTIFY the On-Shift Reactor Engineer to take action to restore thermal limits to acceptable values. NOTE: Attachment 2 shows the location of the APRM GAFs. The APRMs are ordered: APRMs 1, 2, 3, 4. 7.1.7 DETERMINE, from the core performance edit, if Acceptance Criteria listed in Section 6.2 are satisfied (at least three APRM gain adjustment factors are less than or equal to 1.00). 7.1.8 IF an APRM gain adjustment is necessary, THEN PERFORM applicable section of 20P-09. 7.1.9 IF an APRM gain change was performed, THEN OBTAIN from the process computer a copy of Display 820, Heat Balance/Core Mon, OR PERFORM OPT-01.8C to ensure APRM GAFs satisfy the requirements of Acceptance Criteria, Section 6.2. / IndVer. 12PT-01.11 Rev. 7 Page 5 of 141 7.0 PROCEDURAL STEPS Initials NOTE: Reference 2.5 states that NRC expects LHGR monitoring (CMFLPD) to remain the same as CPR and APLHGR for GNF fuel (Le., to restore LHGR within limits within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />, or be below 23% power within the next 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />). 7.1.10 DETERMINE, from the core performance edit, if CMFLPD is less than or equal to 1.00 (Attachment 2 for CMFLPD location on the core performance edit). 7.1.11 IF CMFLPD is greater than 1.00, THEN NOTIFY the Unit SCO of the condition.

1. NOTIFY the On-Shift Reactor Engineer to take action to restore CMFLPD to an acceptable value. 7.1.12 ENSURE Acceptance Criteria listed in Section 6.0 have been satisfied.

7.1.13 ATTACH the collected edits. 7.2 Using a Process Computer Backup Program NOTE: This section is performed only when Section 7.1 can NOT be completed.

The On-Shift Reactor Engineer should be contacted to perform this section. 7.2.1 PERFORM OPT-01.8D to calculate core thermal power. NOTE: OENP-24.19 should be referenced for instructions on utilizing the backup program. 7.2.2 OBTAIN LPRM readings, control rod pattern, core flow, reactor pressure, and additional data as required by the computer backup program AND RUN the program. 7.2.3 OBTAIN the output of the backup program. 7.2.4 DETERMINE from the backup edit if criteria listed in Section 6.1 are met. /2PT-01.11 Rev. 7 Page 6 of 14/

7.0 PROCEDURAL STEPS 7.2.5 IF limits specified in Section 6.1 are NOT satisfied, THEN NOTIFY the Unit SCQ of the condition.

1. NOTIFY the On-Shift Reactor Engineer to take action to restore thermal limits to acceptable values. 7.2.6 PERFORM OPT-01.8C to determine if acceptance criteria listed in Section 6.2 are satisfied.

7.2.7 IF APRM gain adjustment is necessary, THEN PERFORM applicable section of 20P-09. 7.2.8 IF APRM gain change was performed, THEN PERFORM another OPT-01.8C to ensure APRM gain adjustment factor satisfies the requirements of Acceptance Criteria, Section 6.2. Initials / IndVer. NOTE: Reference 2.5 states that NRC expects LHGR monitoring (CMFLPD) to remain the same as CPR and APLHGR for GNF fuel (i.e., to restore LHGR within limits within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />, or be below 23% power within the next 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br />). 7.2.9 DETERMINE, from the backup edit, if CMFLPD is less than or equal to 1.00. 7.2.10 IF CMFLPD is greater than 1.00, THEN NOTIFY the Unit SCO of the condition.

1. INITIATE actions to restore CMFLPD to an acceptable value. 7.2.11 ENSURE Acceptance Criteria listed in Section 6.0 have been met satisfactorily.

7.2.12 ATTACH the backup program edits. 7.3 ENSURE the required information has been recorded on the cover page. 7.4 NOTIFY the Unit SCO when this test is complete or found to be unsatisfactory.

12PT-01.11 Rev. 7 Page 7 of 141 ATTACHMENT 1 Page 1 of 1 Certification and Review Form General Comments and Recommendations Initials Name (Print) Test procedure performed by __ _ Exceptions to satisfactory performance

________________

_ Corrective action required _____________________

_ Test procedure has been satisfactorily completed:

Unit SCO: Signature Date Test procedure has NOT been satisfactorily completed:

Unit seo: Signature Date Test has been reviewed by: Shift Manager:

__________

_ Signature Date 12PT-01.11 Rev. 7 Page 8 of 141 ATTACHMENT 2 Page 1 of 2 Limit Positions on the Core Performance Log -Long Edit BRUNSWICK-1 WK-0739 07SEP29-09.00.27 14505 MWO/MTU TRIGR=2HR REV=usep06 CORE PERFORMANCE LOG ---LONG EDIT TL 12 TLO TBPINS ALL MOC TO EOC OPTB POW DEP MCPR CTP CALCULATION

HEAT BALANCE SYMMETRY : FULL CYCLE: 16 *AXIAL REL POWER* FLOW (MLB/HR) CORE PARAMTRS NUCLEAR LIMITS LOCATION NODE NOTCH REL-POW EXP(GWD/MT)

WT 78.02(101.3%)

CMEO 0.3118 P-PCS 0.02 07-28-09 25 0.177 4.766 WOP 77.74 CAEO 0.1667 P-PCFC 0.02 07-28-09 24 GG 13.333 NSSFFLG 1. CADA 0.1538 8.909 39-18 23 02 0.784 18.881 WFW 12.72 CAVF 0.4773 P=O.OOO F=O.OOO 0 22 04 0.939 23.791 CMFLCPR WFWFFLG 1. CAPO 58.934 CMAPRAT", 8.867 41 82 17 -21 86 1.885 27.184 CMAPRAT <.,;yt:;u' 14bUb,;j

1. RWL 187.5460 P=O.OOO F=O.OOO 20 08 1.088 29.224 NSSCUFLG eAVEX 27956.1 MWOlfliLNSSCRFLG

1. CDLP 18.7146 CMFDLRX. G.7S141-30-17 19 10 1.102 30.344 00 34.22 DPCC 24.0078 41-30-17 18 12 1.097 38.981 NSSCRFLG KEFF 0.9933 CMINCPR"",-

1 17 14 LOCATION 1 2 3 4 5 6 7 8 AXIAL REL POWER 0.53 1.08 1.17 1.17 1.15 1.10 1.08 0.99 REGION REL POWER 0.94 1.06 0.96 0.85 1.13 1.08 0.90 1.06 RING REL POWER 1. 14 1. 12 1.14 1. 14 1.17 1. 03 0.65 1.078 1.035 APRM GAFS 0.99 1.00 1.000.99 APRM GAFS h**h***hh**

NUCLEAR LIMITS BY REGION ***************

B_ 0.842 0.756 0.667 0.667 789 13-36 0.868 31-36 0.909 39-36 17-38-19 0.830 27-38-20 0.817 35-38-19 13-44-04 0.723 33-38-19 0.688 35-36-19 13-44-04 0.723 33-38-19 0.688 35-36-19 -----------------

-+------------------

-+----------------

--456 0.845 11-20 0.893 29-20 0.905 39-28 0.724 11-22-08 0.854 27-24-19 0.867 41-32-17 0.657 11-20-08 0.780 27-22-18 0.781 41-32-17 0.657 11-20-08 0.780 27-22-18 0.781 41-30-17

• MFLCPR *

• MAP RAT *

• MFDLRX *

• MFLPD ------------------+-------------------+------------------

1 2 3 0.872 13-18 0.892 31-18 0.909 39-18 0.783 17-16-19 0.840 27-16-20 0.833 35-16-19 0.666 17-14-19 0.739 33-16-19 0.702 35-14-19 0.666 17-14-19 0.739 33-16-19 0.702 35-14-19 # NODES WI LIMITS > 1 FLCPR APRAT FDLRX FLPD o o o o ***************

THE 10 MOST LIMITING BUNDLES *****************

FLCPR LOC CPR LIMIT APRAT LOC APLHGR LIMIT FDLRX LOC LPD LIMIT FLPD LOC LPD LIMIT --------------------------0.909 39-18 1.518 1.381 0.867 41-32-17 8.01 9.25 0.781 41-30-17 10.47 13.40 0.781 41-30-17 10.47 13.40 0.905 39-28 1.5251.381 0.866 43-30-17 8.09 9.35 0.78027-22-18 10.46 13.40 0.780 27-22-18 10.46 13.40 0.894 39-22 1.544 1.381 0.854 27-24-19 7.55 8.84 0.739 33-16-19 9.90 13.40 0.739 33-16-19 9.90 13.40 0.893 29-20 1.546 1.381 0.840 27-16-20 7.78 9.26 0.73743-28-17 9.88 13.40 0.737 43-28-17 9.88 13.40 0.892 31-18 1.5481.381 0.833 35-16-19 7.72 9.26 0.737 43-22-17 9.87 13.40 0.737 43-22-17 9.87 13.40 0.888 37-20 1.554 1.381 0.833 41-30-17 9.49 11.38 0.734 45-30-17 9.83 13.40 0.73445-30-17 9.83 13.40 0.883 41-20 1.563 1.381 0.825 27-22-19 9.35 11.34 0.73427-14-20 9.83 13.40 0.734 27-14-20 9.83 13.40 0.880 41-16 1.5681.381 0.818 41-28-17 7.38 9.02 0.732 39-32-17 9.81 13.40 0.732 39-32-17 9.81 13.40 0.876 35-18 1.576 1.381 0.816 27-28-19 7.13 8.73 0.72429-24-19 9.70 13.40 0.724 29-24-19 9.70 13.40 0.872 39-14 1.583 1.381 0.816 33-14-19 7.21 8.84 0.721 29-20-19 9.66 13.40 0.721 29-20-19 9.66 13.40 12PT-01.11 Rev. 7 r Page 9 Of141 CMFDLRX CMINCPR ATTACHMENT 2 Page 2 of 2 Limit Positions on the Core Performance Log -Long Edit *************

CONTROL ROD DATA ************* 02 06 10 14 18 22 26 30 34 38 42 46 50 51 51 47 24 47 43 43 39 14 12 14 39 35 35 31 00 10 16 10 20 31 27 00 00 27 23 10 16 10 20 23 19 19 15 14 12 14 15 11 11 07 24 07 03 03 02 06 10 14 18 22 26 30 34 38 42 46 50 RODS OUT OF MIRROR SYMMETRY:

06-31 10-31 42-31 06-27 10-27 42-27 46-27 SUBST. RODS: **********

CALIBRATED LPRM READINGS ***********

• LPRM FAILED SENSORS* *OTHER FAILED SENSORS* 45 30.7 45.5 51.4 41.9 25.9 LOCATION STATUS SENSOR STATUS 41.2 54.6 61.2 50.7 32.5 52.3 59.7 57.3 58.9 39.3 12-21-B MAN 65.1 68.4 59.7 69.9 40.1 20-21-0 MAN 37 21.7 41.0 55.3 60.2 54.7 42.6 0 29.8 52.3 52.6 56.5 55.3 54.2 C 39.0 63.2 56.9 58.6 58.6 60.9 B 36.2 70.6 54.7 51.7 58.5 69.6 A 29 17.0 39.0 56.6 61. 3 59.2 52.4 23.7 48.7 54.0 59.3 57.6 65.0 31.4 57.4 58.7 57.3 57.2 56.2 21.4 54.0 51.5 44.1 50.7 59.5 21 25.1 43.6 57.7 62.5 58.2 49.1 36.0 53.7 53.6 58.3 57.5 59.3 49.4 65.0 58.6 59.4 58.6 55.2 52.0 65.3 51.8 48.9 54.9 59.9 13 40.5 54.5 59.7 52.8 34.0 53.4 54.4 57.6 55.7 43.8 64.9 59.1 60.5 60.6 52.7 69.8 59.5 55.5 62.4 63.5 05 34.9 41.1 28.7 44.9 50.5 35.2 49.1 44.3 40.5 57.3 43.8 43.0 04 12 20 28 36 44 12PT-01.11 Rev. 7 Page 10 of 14/

ATTACHMENT 3 Page 1 of 3 Definitions and Abbreviations Found on Core Performance Edit 1. 2. 3. APLHGR APLHGR LIMIT APRAT Nodal Average Planar Linear Heat Generation Rate (kw/ft) Nodal limiting value of APLHGR (kw/ft) Maximum fraction of limiting APHLGR rate = Maximum of APLHGR APLHGR LIMIT 4. APRM GAFS Average Power Range Monitor Gain Adjustment Factors 5. CMAPRAT Core Maximum APRAT (TIS 3.2.1). Maintaining this value ::; 1.00 ensures that the peak cladding temperature will be kept < 2200°F and, therefore, core geometry will be maintained during a LOCA. 6. CMINCPR The MCPR used to determine the requirement for RBM operability iaw Tech Spec 3.3.2.1. This is the lowest core CPR and corresponding bundle location.

7. CMFDLRX Core Maximum Fuel Design Limit Ratio X, similar to CMFLPD for GNF fuel, but only applies to AREVA fuel. 8. CMFLCPR Core Maximum Fraction of limiting CPR (TIS 3.2.2). Maintaining this value::; 1.00 ensures that departure from nucleate boiling will not occur. 9. CMFLPD Core Maximum Fraction of Limiting Power Density. Formerly a technical specification limit, maintaining this value::; 1.00 ensures that the fuel cladding does not exceed 1 % plastic strain. (GNF fuel bundles only) 10. CPR Critical Power Ratio 11. CPR LIMIT The limiting CPR 12. FDLRX Fuel Design Limit Ratio X (AREVA fuel bundles only) = LHGR LHGR Limit /2PT-01.11 Rev. 7 Page 11 of 14/

13. 14. ATTACHMENT 3 Page 2 of 3 Definitions and Abbreviations Found on Core Performance Edit FLCPR FLPD Maximum Fraction of Limiting CPR = Maximum of CPR LIMIT CPR Fraction of Limiting Power Density (kw/ft) = LHGR = LPD LHGR Limit LPD Limit 15. LHGR Power generation in 1 foot of a fuel rod (kw/ft) 16. LPD Limiting fuel rod Power Density (LHGR) 17. LPD LIMIT Fuel Rod Power Density Limit (LHGR limit) 18. NSSCRFLG The NSSCRFLG on the Core Performance Log indicates when CRD flow used in the heat balance calculation is valid. NSSCRFLG = 1: indicates that 8020 is valid. NSSCRFLG = 2 indicates that 8020 is invalid. The last good value is used to help prevent use of non-conservative values in the heat balance. Check page 2 of MON edit for Failed Sensors. 19. NSSCUFLG The NSSCUFLG on the Core Performance Log indicates when RWCU cleanup flow values used in the heat balance calculation are valid. /2PT-01.11 NSSCUFLG = 1: indicates that WCU1, WCU3, 8047, and 8048 are valid. NSSCUFLG = 2: indicates that WCU1, WCU3, 8047, or 8048 are invalid. The last good value is used to help prevent use of non-conservative values in the heat balance. Check page 2 of MON edit for Failed Sensors. Rev. 7 Page 12 of 14/

ATTACHMENT 3 Page 3 of 3 Definitions and Abbreviations Found on Core Performance Edit 20. NSSFFLG The NSSFFLG on the Core Performance Log indicates which core flow value is used for thermal limit calculations NSSFFLG = 1: indicates wrCF Calculated Total Core Flow representing the summation of the 20 single tap jet pumps. The comparison of wrCF with WOP is :s; 5%. NSSFFLG = 2: indicates WOP Core Plate Differential Core Flow, which is flow resulting from the Core DP and Reactor Power to wrCF correlation.

NSSFFLG = 3: indicates a manually inserted value for wrCF NSSFFLG = 4: indicates wrCF but the comparison with WDP is> 5% 21. WFWFFLG The WFWFFLG on the Core Performance Log indicates when instantaneous or smoothed feedwater flow values are used in the heat balance calculation.

\2PT-01.11 WFWFFLG = 1: indicates smoothed feedwater flow from U2NSSWFWA and U2NSSWFWB is used in U2CP AFWA and U2CP _AFWB respectively.

This condition is expected during steady state operation.

WFWFFLG = 2: indicates instantaneous feedwater flow from U2C32B022 and U2C32B023 is used in U2CP AFWA and U2CP _AFWB respectively.

This condition can be expected following a large change in reactor power. Rev. 7 Page 13 of 141 REVISION

SUMMARY

Revision 7 incorporates EC 70517, Unit 2 Core Reload. This resulted in deletion of old Precaution and Limitation 3.5, all references to Attachment 4 and the attachment itself, revision of Steps 7.1.6 and 7.2.5 and the addition of notes where appropriate.

Acceptance Criteria 6.1.3 was added for CMFDLRX. "Shift Superintendent" has been changed to "Shift Manager".

Revision 6 incorporates Child EC 69315 (Master EC 67951) for installation of Powerplex III. Revision 5 incorporates EC 62831 to provide new Attachment 4 for removal of overly conservative PPC thermal limits under certain operating conditions.

Revision 4 incorporates EC 62384, revising definition of WTSUB on Attachment 3, deleting Attachment 4, Precaution and Limitation 3.5, and reference to Precaution and Limitation 3.5 previously contained in Step 7.1.3.2, and steps referencing Attachment

4. Revision 3 adds rods 47-36 and 45-34 to Attachment 4 for MCPR monitoring when PPX CPR convergence fails. Revision 2 adds a method to account for PPX CPR convergence failures and allows the Reactor Engineer or PPX software to be used to obtain >LI>BU edits. Revision 1 incorporated EC, 47907, Unit 2 Extended Power Uprate Implementation, and EC 46730, Replace Unit 2 Power Range Neutron Monitoring System, required changes and updated to Word 2000 software.

Revision 0 was issued in accordance with ESR 00-00442, Unit 1 Power Range Neutron Monitoring Replacement, which required the OPT-01.11 be separated into unit specific procedures.

12PT-01.11 Rev. 7 Page 14 of 141 LESSON TITLE: LESSON NUMBER: REVISION NO: PROGRESS ENERGY CAROLINAS BRUNSWICK TRAINING SECTION JOB PERFORMANCE MEASURE ADMIN NRC ADMIN C002 (RO) Determination of Drywell Volumetric Average Temperature LOT-OJT-JP-002-A01 01 Determination of Drywell Volumetric Average Temperature SAFETY CONSIDERATIONS:

None EVALUATOR NOTES: (Do not read to performer)

1. This JPM is performed in the classroom or simulator for Unit Two. 2. If performed in the Simulator:

a. Ensure all simulator computer screens are OFF prior to commencing JPM. b. Reset the simulator to IC-11. c. Input override values to the following points to obtain the associated readings on TR-4426-1A11 B/2A12B

• Div 2A (Channel 3) -5823 40.0

• Div 2A (Channel 4) -5824 185.5

• Div 1A (Channel 4) -5822 185.5

• Div 2B (Channel 1) -5802 174.1

• Div 1 B (Channel 2) -5803 168.3

• Div 2B (Channel 2) -5804 162.7

• Div 1 B (Channel 3) -5805 139.3

• Div 2B (Channel 3) -5806 136.8

• Div 1 B (Channel 4) -5807 133.5

• Div 2B (Channel 4) -5808 40.0

• Div 1 B (Channel 5) -5809 124.5

• Div 1A (Channel 3) -5812 121.6

• Div 2B (Channel 6) -5813 119.2

• Div 1A (Channel 5) -5817 88.7

• Div 1A (Channel 6) -5818 40.0

• Div 2A (Channel 5) -5819 88.7

• Div 2A (Channel 6) -5820 88.7 3. If performed in the classroom, provide Recorder Attachment with readings to performer.

4. If this is the first JPM of the JPM set, read the JPM briefing contained NUREG 1021, Appendix E, or similar to the performer.

5. The trainee will need a calculator to perform this JPM. 6. The trainee will be given a copy of OPT-16.2.

7. Task standards (i.e. pass/fail criteria) for each JPM step are ITALICIZED below the step. LOT-OJT-JP-002-A01 Page 2 REV. 01 Determination of Drywell Volumetric Average Temperature Read the following to the JPM performer.

TASK CONDITIONS:

1. Today is the 15 th of the month. 2. Drywell air RTDs CAC-TE-1258-8, CAC-TE-1258-18 and CAC-TE-1258-23 are inoperable.

3. An LCO has been established that requires the performance of OPT-16.2, Drywell Volumetric Average Temperature, once per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. 4. The CAC-TY-4426-1 and 2 Microprocessors are not available for use. 5. The Process Computer and ERFIS are not available.

6. You have been given permission by the Unit CRS to perform this test. 7. The Unit CRS will record required cover page information INITIATING CUE: You have been directed by the Unit Two CRS to perform OPT-16.2, Drywell Volumetric Average Temperature, and record the Volumetric Average Temperature (Step 7.7.1.6) and the results of the test (circle choice) in the table below. DRYWELL VOLUMETRIC AVERAGE TEMPERATURE SAT OF UNSAT LOT-OJT-JP-002-A01 Page 3 REV. 01 Determination of Drywell Volumetric Average Temperature PERFORMANCE CHECKLIST NOTE: Sequence is assumed unless denoted in the Comments.

Step 1 -Obtain current revision of OPT-16.2 and verify if applicable.

Current revision of OPT-16.2 obtained and verified as required.

TIME START ____ _ NOTE: Figure 1 may be used as a worksheet but is NOT required to be retained.

NOTE: All temperatures should be rounded off to the nearest tenth of a degree. NOTE: Inoperable channels indicate downscale (40.00). PROMPT: If being performed in a classroom, provide Recorder Attachment to candidate.

Step 2 -Obtain drywell temperature readings for the locations listed in step 7.2.1 Records the following readings:

4426-2A Channel 3 -INOP 4426-2A Channel 4 -185.5 (:to. 1) 4426-1 A Channel 4 -185.5 (:to. 1) Step 3 -Add step 7.2.1 recorded temperatures and record in step 7.2.2 SATIUNSAT*

185.5 (:to. 1) and 185.5 (:to. 1) added totaling 371.0 (:to.2) and documented in step 7.2.2 **CRITICAL STEP** SAT/UNSAT*

Step 4 -Divide total obtained in step 7.2.2 by the number of temperatures used 371.0 (:to.2) divided by 2 equaling 185.5 (:to. 1) and documented in step 7.2.3 **CRITICAL STEP** SAT/UNSAT*

Step 5 -Obtain drywell temperature readings for the locations listed in step 7.3.1 Records the following readings:

4426-28 Channel 1 -174.1 (:to. 1) 4426-18 Channel 2 -168.3 (:to. 1) LOT-OJT-JP-002-A01 Page 4 SAT/UNSAT*

REV. 01 Determination of Drywell Volumetric Average Temperature Step 6 -Add step 7.3.1 recorded temperatures and record in step 7.3.2 174.1 (:!:.0.1) and 168.3 (:!:.0.1) added totaling 342.4 (:!:.0.2) and documented in step 7.3.2 **CRITICAL STEP** SAT/UNSAT*

Step 7 -Divide total obtained in step 7.3.2 by the number of temperatures used 342.4 (:!:.0.2) divided by 2 equaling 171.2 (:!:.0.1) and documented in step 7.3.3 **CRITICAL STEP** SAT/UNSAT*

Step 8 -Obtain drywell temperature readings for the locations listed in step 7.4.1 Records the following readings:

4426-18 Channel 3 -139.3 (:!:.O. 1) 4426-28 Channel 3 -136.8 (:!:.O. 1) 4426-18 Channel 4 -133.S (:!:.0.1) 4426-28 Channel 4 -INOP Step 9 -Add step 7.4.1 recorded temperatures and record in step 7.4.2 SAT/UNSAT*

139.3 (:!:.0.1), 136.8 (:!:.0.1) and 133.S (:!:.0.1) added totaling 409.6 (:!:.0.3) and documented in step 7.4.2 **CRITICAL STEP** SA TlUNSAT* Step 10-Divide total obtained in step 7.4.2 by the number of temperatures used 409.6 (:!:.0.3) divided by 3 equaling 136.S (:!:.0.1) and documented in step 7.4.3 **CRITICAL STEP** SATIUNSAT*

Step 11 -Obtain drywell temperature readings for the locations listed in step 7.5.1 Records the following readings:

4426-18 ChannelS -124.S (:!:.0.1) 4426-1A Channel 3 -121.6 (:!:.0.1) 4426-28 Channel 6 -119.2 (:!:.0.1)

LOT -OJT -JP-002-A01 Page 5 SAT/UNSAT*

REV. 01 Determination of Drywell Volumetric Average Temperature Step 12 -Add step 7.5.1 recorded temperatures and record in step 7.5.2 124.5 121.6 and 119.2 added totaling 365.3 and documented in step 7.5.2 **CRITICAL STEP** SAT/UNSAT*

Step 13 -Divide total obtained in step 7.5.2 by the number of temperatures used 365.3 divided by 3 equaling 121.8 and documented in step 7.5.3 **CRITICAL STEP** SAT/UNSAT*

Step 14 -Obtain drywell temperature readings for the locations listed in step 7.6.1 Records the following readings:

4426-1 A Channel 5 -88. 7

1) 4426-1A Channel 6 -INOP 4426-2A Channel 5 -88. 7

1) 4426-2A Channel 6 -88.7 (:!:.O. 1) Step 15 -Add step 7.6.1 recorded temperatures and record in step 7.6.2 SAT/UNSAT*

88. 7

88. 7 and 88. 7 added totaling 266.1 and documented in step 7.6.2 **CRITICAL STEP** SAT/UNSAT*

Step 16 -Divide total obtained in step 7.6.2 by the number of temperatures used 266.1 (:!:.O.3) divided by 3 equaling 88. 7 and documented in step 7.6.3 **CRITICAL STEP** SAT/UNSAT*

Step 17 -Transfer average readings documented in steps 7.2.3,7.3.3,7.4.3,7.5.3 and 7.6.3 into step 7.7.1 185.5 (:!:.O.1), 171.2 136.5 121.8 and 88. 7 transferred into step 7. 7. 1 SAT/UNSAT*

LOT-OJT-JP-002-A01 Page 6 REV. 01 Determination of Drywell Volumetric Average Temperature Step 18 -Multiply each reading in 7.7.1 by the designated factor and document in step 7.7.1 9.3 (:!:.0.1), 15.4 (:!:.0.1), 54.6 (:!:.0.1), 46.3 (:!:.0.1), and 7.1 (:!:.0.1) documented in step 7.7.1 **CRITICAL STEP** SAT/UNSAT*

Step 19 -Add weighted averages in step 7.7.1 and document as TOTAL in step 7.7.1.6 9.3 (:!:.0.1), 15.2 (:!:.0.1), 54.6 (:!:.0.1), 46.3 (:!:.0.1), and 7.1 (:!:.0.1) added totaling 132.7 (:!:.0.5) and documented in step 7. 7. 1 **CRITICAL STEP** SAT/UNSAT*

PROMPT: If asked the status of computer point C074, report it is unavailable.

Step 20 -Record computer point C074 value in step 7.7.2 Step 7.7.2 marked NIA due to unavailability Step 21 -Perform steps 7.7.3 and 7.7.4 (only on the first day of the month) Step 7.7.3 and 7.7.4 marked NIA (not required)

NOTE: Cover page information will be filled out by the Unit CRS. Step 22 -Perform step 7.7.5 Cover page information verified complete SAT/UNSAT*

SAT/UNSAT*

SATIUNSAT*

Step 23 -Determine whether the test is SAT or UNSAT based on the cue sheet instructions Applicant determines the test is SAT because calculated Drywell Avg Air Temp 132.7 (:!:.0.5) is less than 150°F (acceptance criteria)

• • CRITICAL STEP** SAT/UNSAT*

LOT-OJT-JP-002-A01 Page 7 REV. 01 Determination of Drywell Volumetric Average Temperature Step 24 -Perform step 7.7.6 Unit CRS notified test is complete SATIUNSAT TERMINATING CUE: When CRS is informed of test results, this JPM can be terminated.

TIME COMPLETED

____ _ NOTE: Comments required for any step evaluated as UNSA T. LOT-OJT-JP-002-A01 Page 8 REV. 01 Determination of Drywell Volumetric Average Temperature RELATED TASKS: 223*206*81*01 Obtain Air and Water RTD Temperatures from the SPTMS lAW OP-24 KIA REFERENCE AND IMPORTANCE RATING: GEN 2.1.20 4.6/4.6 Ability to interpret and execute procedure steps

REFERENCES:

OPT-16.2, Drywell Volumetric Average Temperature TOOLS AND EQUIPMENT:

Calculator Simulator or This JPMs Recorder Attachment ADMINISTRATIVE CATEGORY (from NUREG 1123, Rev 2. Supplement 1): Admin -Conduct of Operations REASON FOR REVISION:

Changed to allow classroom performance for HLC2010-1 NRC License Exam. LOT-OJT-JP-002-A01 Page 9 REV. 01 Determination of Drywell Volumetric Average Temperature Validation Time: Minutes (approximate)

Time Taken: ____ _ APPLICABLE METHOD OF TESTING Performance:

Simulate Setting: In-Plant Time Critical:

Yes Alternate Path: Yes Actual Simulator No X No X EVALUATION Unit: 2 Admin X Time Limit N/A Performer:

_______________________

_ JPM: Pass __ Fail --Remedial Training Required:

Yes No __ (Reference TAP-411 for evaluation guidance)

Comments:

Comments reviewed with Performer Evaluator Signature:

____________

_ Date: -----LOT-OJT-JP-002-A01 Page 10 REV. 01

tJ CD (") o 0-CD ""'I (") ::T 3 CD :::l -

0 CI) (") o a CI) ., (")

::r 3 CI) ::l -

TASK CONDITIONS:

1. Today is the 15 th of the month. 2. Drywell air RTDs CAC-TE-1258-8, CAC-TE-1258-18 and CAC-TE-1258-23 are inoperable.

3. An LCO has been established that requires the performance of OPT-16.2, Drywell Volumetric Average Temperature, once per 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />. 4. The CAC-TY-4426-1 and 2 Microprocessors are not available for use. 5. The Process Computer and ERFIS are not available.

6. You have been given permission by the Unit CRS to perform this test. 7. The Unit CRS will record required cover page information INITIATING CUE: You have been directed by the Unit Two CRS to perform OPT-16.2, Drywell Volumetric Average Temperature, and record the Volumetric Average Temperature (Step 7.7.1.6) and the results of the test (circle choice) in the table below. DRYWELL VOLUMETRIC AVERAGE TEMPERATURE SAT OF UNSAT Progress Energy BRUNSWICK NUCLEAR PLANT DATE COMPLETED

_________

_ FREQUENCY:

c Continuous Use UNIT __ % PWR, ____ GMWE ___ _ SUPERVISOR, ___________

_ REASON FOR TEST (check one or more): _ Routine sUNeillance Once every 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> when at least one SPTMS air RTD is inoperable in each division of SPTMS and the unit is in Mode 1, 2, or 3. _WO# ____________

_ _ Other (explain)

_________

_ PLANT OPERATING MANUAL VOLUME X PERIODIC TEST UNIT o OPT-16.2 DRYWELL VOLUMETRIC AVERAGE TEMPERATURE REVISION 35 IOPT-16.2 Rev. 35 Page 1 of 12 )

1.0 PURPOSE 2.0 1.1 This test is performed to determine the operability of the drywell in conformance with the requirements specified in Technical Specification SR 3.6.1.4.1.

1.2 2.1 2.2 2.3 2.4 2.5 2.6 This test involves averaging air temperatures within the drywell. REFERENCES Technical Specifications FSAR, Section 6.2 System Description SD-24, Containment Atmosphere Control System EER 93-578 (SPTMS Air RTD Divisions/Channels Defined) 2SP-95-208, Temperature Measurement for SPTMS RTD ESR 98-00387, Rev. 0, TS Change to ContainmentlDrywell Air Temperature Requirements 3.0 PREREQUISITES None Applicable 4.0 PRECAUTIONS AND LIMITATIONS None Applicable 5.0 SPECIAL TOOLS AND EQUIPMENT Calculator 6.0 ACCEPTANCE CRITERIA This test may be considered satisfactory when the drywell volumetric average temperature is less than or equal to 150°F. IOPT-16.2 Rev. 35 Page 2 of 12 )

7.0 PROCEDURAL STEPS Initials 7.1 NOTE: NOTE: NOTE: NOTE: IOPT-16.2 OBTAIN permission from Unit CRS to perform this test. Figure 1 may be used as a worksheet but is NOT required to be retained with th is PT. All temperatures used in this procedure should be rounded off to the nearest tenth of a degree. Primary containment temperatures may be obtained from the CAC-TY -4426-1 (2) microprocessor per OP-24, Operator Interface With The SPTMS Microprocessors, from the recorders in the Control Room, from ERFIS display screens 740 and 745 of the Validation Menu; or as a backup method, the designated process computer points may be used. Microprocessor point values may also be obtained by I&C manually taking temperature measurements of SPTMS RTDs. Technical Specifications requires at least one operable temperature detector in each channel for each location.

Rev. 35 Page 3 of 12) 7.0 PROCEDURAL STEPS 7.2 90' Elevation And Above 7.2.1 OBTAIN drywell temperature for the locations listed below AND RECORD in the spaces provided.

Recorder/ERFIS Display 4426-2A Channel 3 __ _ 4426-2A Channel 4 __ _ 4426-1A Channel 4 __ _ SPTMS (5823) __ (5824) __ (5822) __ CHAN II II I N/A N/A N/A 7.2.2 ADD the recorded temperatures AND RECORD the total in the following space: __________

-Recorder/ERFIS display total __________

-SPTMS total 7.2.3 DIVIDE the total by the number of temperatures used to obtain the average temperature for the 90' elevation and above in the following space: = total no. of temperatures average Initials I OPT-16.2 Rev. 35 Page 4 of 12 )

7.0 PROCEDURAL STEPS 7.3 Between 70' And 80' Elevation 7.3.1 OBTAIN drywell temperatures for the locations listed below AND RECORD in the spaces provided.

Recorder/ERFIS Display SPTMS CHAN PPC Unit 1 Only 4426-2B Channel 1 (5802) II W109 4426-1 B Channel 2 (5803) I F148 4426-2B Channel 2 (5804) II W110 Unit 2 Only 4426-2B Channel 1 (5802) II W109 4426-1 B Channel 2 (5803) I F148 NOTE: Notify BESS duty manager if any temperatures between 70' and 80' elevation exceed 240°F. 7.3.2 ADD the recorded temperatures AND RECORD the total in the following spaces: _________

-Recorder/ERFIS display total ________ -SPTMS total _________

-PPC total 7.3.3 DIVIDE the total by the number of temperatures used to obtain the average temperature for the 70' -80' elevation in the following space: = total no. of temperatures average Initials / IOPT-16.2 Rev. 35 Page 5 of 121 7.0 PROCEDURAL STEPS 7.4 Between 28' And 45' Elevation 7.4.1 OBTAIN drywell temperatures for the locations listed below AND RECORD in the spaces provided.

Recorder/ERFIS Display SPTMS CHAN PPC 4426-1 B Channel 3 (5805) I F149 4426-2B Channel 3 (5806) II W085 4426-1 B Channel 4 (5807) I F150 4426-2B Channel 4 (5808) II W086 7.4.2 ADD the recorded temperatures AND RECORD the total in the following spaces: ________ -Recorder/ERFIS display total ________ -SPTMS total ________ -PPC total 7.4.3 DIVIDE the total by the number of temperatures used to obtain the average temperature for the 28' -45' elevation in the following space: = total no. of temperatures average 7.5 Between 10' And 23' Elevation 7.5.1 OBTAIN drywell temperatures for the locations listed below AND RECORD in the spaces provided.

Recorder/ERFIS Display SPTMS CHAN PPC Initials 4426-1 B Channel 5 __ _ (5809) __ (5812) __ (5813) __ F151 ____ _ 4426-1A Channel 3 __ _ W105 ____ _ 4426-2B Channel 6 __ _ II W088 ____ _ IOPT-16.2 Rev. 35 Page 6 of 12/

7.0 PROCEDURAL STEPS 7.5.2 ADD the recorded temperatures AND RECORD the total in the following spaces: _________

-Recorder/ERFIS display total ________ -SPTMS total _________

-PPC total 7.5.3 DIVIDE the total by the number of temperatures used to obtain the average temperature for the 10' -23' elevation in the following space: = total no. of temperatures average 7.6 Below 5' Elevation 7.6.1 OBTAIN primary containment temperatures for the locations listed below AND RECORD in the spaces provided.

Recorder/ERFIS Display PPC Initials 4426-1A Channel 5 __ _ SPTMS (5817) __ (5818) __ (5819) __ (5820) __ CHAN I W106 ____ _ 4426-1A Channel 6 __ _ W107 ____ _ 4426-2A Channel 5 __ _ II II W115 ____ _ 4426-2A Channel 6 __ _ W116 ____ _ CAUTION If microprocessor is used to obtain data, failure to restore mode selector switch to the ERFIS (NORMAL) position will cause loss of SPTMS indications on the ERFIS Control Room displays.

7.6.2 ADD the recorded temperatures AND RECORD the total in the following spaces: IOPT-16.2

_________

-Recorder/ERFIS display total _________

-SPTMS total _________

-PPC total Rev. 35 Page 7 of 12 )

7.0 PROCEDURAL STEPS Initials NOTE: 7.7 7.6.3 DIVIDE the total by the number of temperatures used to obtain the average temperature for the below 5' elevation in the following space: = total no. of temperatures average The total in Step 7.7.1.6 is the volumetric average temperature for the drywell. If this temperature is greater than 150°F, actions should be taken as required by Technical Specification SR 3.6.1.4.1.

Volumetric Average 7.7.1 1. DETERMINE the volumetric average temperature by performing the following calculations:

Average above 90' elevation x 0.05 = ___ _ Step 7.2.3 2. Average 70' -80' elevation x 0.09 = ___ _ Step 7.3.3 3. Average 28' -45' elevation x 0.40 = ___ _ Step 7.4.3 4. Average 10' -23' elevation x 0.38 = ___ _ Step 7.5.3 5. Average below 5' elevation x 0.08 = ___ _ Step 7.6.3 6. Total ____________

_ Steps 7.7.1.1 through 7.7.1.5 IOPT-16.2 Rev. 35 Page 8 of 121 7.0 PROCEDURAL STEPS Initials NOTE: NOTE: Computer point C074 may only be used to satisfy EOP concerns and may NOT be used to satisfy Technical Specifications.

7.7.2 RECORD computer point C074 value (for information).

C074 ____ _ Performance of Steps 7.7.3 and 7.7.4 is only required once a month (perform on the first day of the month). 7.7.3 SUBTRACT the value obtained in Step 7.7.2 from the value obtained in Step 7.7.1.6. = Step 7.7.1.6 Step 7.7.2 Difference 7.7.4 IF the absolute value of the difference obtained in Step 7.7.3 is greater than 10°F, THEN ENSURE that Unit CRS is informed of the discrepancies in the computer point reading (EOP concern).

7.7.5 ENSURE required information has been recorded on the cover page. 7.7.6 NOTIFY Unit CRS when this test is complete or found to be unsatisfactory.

IOPT-16.2 Rev. 35 Page 9 of 121 ATTACHMENT 1 Page 1 of 1 Certification and Review Form General Comments and Recommendations

---

Initials Name (Print) Performed by: Exceptions to satisfactory performance

_______________________________

_ Corrective action required _____________________________

_ Test procedure has been satisfactorily completed:

Unit CRS/SRO: ____________

_ Signature Date Test procedure has NOT been satisfactorily completed:

Unit CRS/SRO: ____________

_ Signature Date Test has been reviewed by: Shift Manager: Signature Date IOPT-16.2 Rev. 35 Page 10 of 121 TIME 4426-2A (Channel 3)-5823 4426-2A (Channel 4)-5824 4426-1A (Channel 4)-5822 Unit 1 only 4426-2B (Channel 1 )-5802 4426-1 B (Channel 2)-5803 4426-2B (Channel 2)-5804 Unit 2 only 4426-2B (Channel 1 )-5802 4426-1 B (Channel 2)-5803 4426-1 B (Channel 3)-5805 4426-2B (Channel 3)-5806 4426-1 B (Channel 4)-5807 4426-2B (Channel 4)-5808 4426-1 B (Channel 5)-5809 4426-1A (Channel 3)-5812 4426-2B (Channel 6)-5813 4426-1A (Channel 5)-5817 4426-1A (Channel 6)-5818 4426-2A (Channel 5)-5819 4426-2A (Channel 6)-5820 AVERAGE TEMPERATURE FIGURE 1 Page 1 of 1 Containment Temperature Worksheet

___ 0.05 ___ 0.09 ___ 0.09 ___ 0040 ___ 0.38 ___ 0.08 I OPT-16.2 Rev. 35 REVISION

SUMMARY

Revision 35 incorporates temporary change EC 75096 which temporarily disables temperature indication 2CAC-TE-1258-4 (4426-2B [channel 2]-5804).

This revision also updated operations titles Shift Superintendent to Shift Manager and SCQ to CRS or CRS/SRQ as applicable.

Revision 34 restores 2CAC-TE-1258-20 (4426-2A [channel 6]-5820) that was temporarily removed lAW EC 63061. Revision 33 incorporates temporary change EC 63061 which temporarily disables temperature indication 2CAC-TE-1258-20 (4426-2A [channel 6]-5820).

Revision 32-Removes the Unit Specific items and associated notes that were added per Temporary Plant Modification ESR 99-00485.

Temperature Sensor 2-CAC-TE-1258-22 has been repaired during the refuel outage. This revision also revises terminology for PassPort Implementation.

Revision 31 -This revision removes the notes that specified equivalencies during implementation of ESRs 97-00125 and 97-00051, and changed points A, B, C, D, E, & F to channel 1, 2, 3, 4, 5, & 6. IOPT-16.2 Rev. 35 Page 12 of 121