Text

Draft SRO JPMs
	ML061650344
Person / Time
Site:	Robinson
Issue date:	04/26/2004
From:	; Division of Nuclear Materials Safety II
To:	;
References
	50-261/04-301
	Download: ML061650344 (109)
	v • d • e

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-I,-, i rii L-- cý ROBINSON AUG/SEPT 2004 EXAM 05000261/2004301 i

AUGUST 27, 2004 (written)

AUG. 30 - SEPT. 2004 (op)

Admin Initial Submittal

ES-301 Administrative Topics Outline Form ES-301-1 Final Facility: Robinson Date of Examination: 8/30/2004 Examination Level: SRO Operating Test Number: NRC Administrative Topic Describe activity to be performed.

(see Note) 2.1.2 (4.0) Knowledge of operator responsibility during all modes of plant Conduct of Operations operation JPM Al JPM: Monitor nuclear instrumentation during refueling operations (New JPM) 2.1.25 (3.1) Ability to obtain and interpret station reference materials such Conduct of Operations as graphs, monographs, and tables which contain performance JPM A2 data.

JPM: Perform a SDM calculation lAW FMP-012.

(Modified JPM) 2.2.6 (3.3) Knowledge of the process for making changes in procedures Equipment Control as described in the safety analysis report.

JPM A3 JPM: Review/approve a temporary change to a procedure lAW PRO-NGGC-0204.

(New JPM)

Radiation Control 2.3.11 (3.2) Ability to control radiation releases JPM A4 JPM: Adjust a radiation monitor setpoint (New JPM) 2.4.44 (4.0) Knowledge of emergency plan protective action Emergency Plan recommendations.

JPM A5 JPM: Given a set of conditions, determine the Emergency Action Level (EAL) and make a Protective Action Recommendation (PAR) IAW the Emergency Plan.

(New JPM)

NOTE: All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

NUREG-1021, Draft Revision 9

Appendix C Page 1 of 5 Form ES-C-1 PERFORMANCE INFORMATION Facility: HB ROBINSON Task No.: 1072101201 Task

Title:

Adjust the High Alarm Setpoint for JPM No.: 2004 NRC JPM SRO Radiation Monitor R-18 A4 K/A

Reference:

2.3.11 (2.7/3.2)

Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: X Actual Performance:

Classroom Simulator Plant X READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

The plant is at 100% power.

A Liquid Waste Release from Monitor Tank "A" has been prepared.

Task Standard: Satisfactory simulation of critical steps.

Required Materials: NONE General

References:

OP-920, RADIATION MONITORING SYSTEM, Revision 28 Handouts: Completed, approved EMP-023 -ATTACHMENT 10.3 Initiating Cue: You are the BOP Operator. The CRSS has assigned you to adjust the High Alarm setpoint for RMS Channel R-18 to the value on the approved release form in accordance with OP-920, Section 8.1. The drawer can be pulled out and returned to the correct position but simulate all other actions. Inform the licensed operators before opening the drawer and again after it is closed.

Time Critical Task: N/A Validation Time: 10 Minutes 2004 NRC JPM SRO A4 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 2 of 5 Form ES-C-1 PERFORMANCE INFORMATION (Denote Critical Steps with an asterisk)

Examiner's Cue: Provide a completed, approved copy of EMP-023, ATT. 10.3 Performance Step: I Review ATT. 10.3 Standard: Verifies and determines setpoint Comment:

Performance Step: 2 Obtain procedure for changing the setpoint Standard: References OP-920, Step 8.0 Comment:

OP-920, Step 8.1.1 Performance Step: 3 Adjusting the High Alarm Setpoint for Monitors R-1, R-2, R-3, R-4, R-5,R-6, R-7, R-8, R-9, R-11, R-12, R-15, R-16, R-17, R-18, R-20, R-21,R-30, R-31A, R-31B, R-31C and R-33 Standard: Identifies R-18 on list and 8.1.1 as the correct step to implement Comment:

OP-920, Step 8.1.1 Performance Step: 4 Slide out the drawer for the desired monitor.

Standard:

" Informs licensed operators

" Slides out drawer for R-18

Comment:

2004 NRC JPM SRO A4 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 3 of 5 Form ES-C-1 PERFORMANCE INFORMATION OP-920, Step 8.1.1 Performance Step: 5 Adjust the thumbwheels to desired setpoint.

Standard:

Points out thumbwheels

Discusses adjusting thumbwheels to 1.12E+04 Examiner's Cue: The thumbwheels are adjusted to the setpoint (

identified by the Candidate.

Comment:

OP-920, Step 8.1.1 Performance Step: 6 Push Alarm/Reset button and verify setpoint appears on ratemeter display.

Standard: Simulates pushing the Alarm/Reset button and points out ratemeter display.

Examiner's Cue: The setpoint identified by the Candidate ( ) is displayed.

Comment:

OP-920, Step 8.1.1 Performance Step: 7 Return the monitor drawer to its proper position.

Standard:

Slides drawer back in and secures *

" Informs licensed operators Comment:

Terminating Cue: The evaluator can inform the Candidate that evaluation on this JPM is complete at any time after the drawer has been returned to the proper position.

STOP TIME:

2004 NRC JPM SRO A4 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 4 of 5 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2004 NRC JPM SRO A4 Examinee's Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiner's Signature: Date:

2004 NRC JPM SRO A4 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 5 of 5 Form ES-C-1 JPM CUE SHEET INITIAL CONDITIONS:

The plant is at 100% power.

" A Liquid Waste Release from Monitor Tank "A" has been prepared.

INITIATING CUE: You are the BOP Operator. The CRSS has assigned you to adjust the High Alarm setpoint for RMS Channel R-1 8 to the value on the approved release form in accordance with OP-920, Section 8.1. The drawer can be pulled out and returned to the correct position but simulate all other actions. Inform the licensed operators before opening the drawer and again after it is closed.

2004 NRC JPM SRO A4 NUREG 1021, Revision 9, Supplement 1

M CP&L Multiple Use CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 3 PART 2 OPERATING PROCEDURE OP-920 RADIATION MONITORING SYSTEM REVISION 28 OP-920 Rev. 28 Page 1 of 59

SUMMARY

OF CHANGES PRR 68035 STEP/SECTION REVISION COMMENTS 1.1 Added the Yokogawa VR204 View Recorder to the purpose statement.

2.27 Added reference to RST-029, Calibration of R-24A, B, & C.

2.37 Added reference to EMP-034, Operation of R-24A, B, & C.

2.47 Added reference to Yokogawa Instruction Manual for VR200 Wide View Recorder.

4.9 Added precaution to address R-24A, B, & C indication in that the readings are not valid for quantification until the monitor has been adjusted per EMP-034 during an actual primary to secondary tube leak and that the readings are not power compensated with Reactor Power less than 40% and the readings will increase with power until 40% is reached.

5.2.4.3 & 5.2.7.4 Revised step for checking the RM-80 battery to only check voltage if the batteries were disconnected, and if they were disconnected to replace the battery pack if voltage was less than 3.2 volts, and a step to verify the battery leads are connected prior to proceeding.

Added note about recharging the replaced battery pack at a constant current of 10OMa for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months . Vendor manual discusses disconnecting battery leads if power secured for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to preserve the battery but does not require a battery voltage check if power secured for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months as the step originally stated. Additional scheduled maintenance is performed to check the battery voltage every 18 months and replace the battery every 3 years.

Deleted old steps 5.2.4.9 (R-19's) & 5.2.7.16 (R-37) for connecting battery leads as this is now performed in step 5.2.4.3.

5.2.6 Added new section for placing the R-24 monitors in service; addressing checking the RM-80 batteries connected, setting the RM-80 switch settings, powering up the RM-80 unit, and reloading the RM-80 database by E&C if necessary.

5.4 Added new section to place the R-24A/B/C Yokogawa recorder in service. Instructions provided to power up the recorder, check the alarm settings are correct, and set the date and time.

6.1.4 Added instructions to check the R-24A/B/C Yokogawa recorder alarm settings.

6.2.4 Added step to address the R-24A, R-24B, and R-24C source check is performed by E&C lAW EMP-034.

6.5 Added section to address normal operation of the R-24A/B/C Yokogawa recorder with instructions to review past (historical) measured data.

7.2.4.4 & 7.2.7.6 Revised step to disconnect RM-80 battery leads from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months per Vendor manual.

7.2.6 Added steps to remove R-24A1B/C from service.

7.4 Added steps to remove the R-24A/B/C Yokogawa recorder from service.

8.3 Added section for adjusting the R-24A/B/C Yokogawa recorder alarm setpoint, with a preceding note stating "Normally the R-24 setpoints for each channel are set to 5 gpd. If primary to secondary Steam Generator leakage is confirmed, the setpoint may be raised as directed by E&C to establish a new threshold value for the alarm:.

Attachment 9.1 Added information of applicable procedures for R-24A/B/C Removed reference to OMM-001-12, which does not perform a channel check, and added reference to OST-020, OST-021, OST-022, and OST-023 for the procedure to reference for the channel check as applicable for the monitor.

I Removed reference to EMP-024 for calibration and source checks of various monitos.

OP-920 Rev. 28 Page 2 of 59

TABLE OF CONTENTS SECTION PAGE 1.0 P UR P O S E ..................................................................................................... 4 2.0 R E FE R E NC E S ................................................................................................ 4 3.0 P R E R EQ UIS ITES ............................................................................................ 7 4.0 PRECAUTIONS AND LIMITATIONS ................................................................ 8 5 .0 S TA RT UP .................................................................................................... .. 11 5.1 Placing the Area Radiation Monitors in Service ........................................ 11 5.2 Placing the Process Radiation Monitors in Service .................................... 13 5.3 Placing the Westronics Series 3000 Recorder in Service ......................... 23 5.4 Placing the R-24A/B/C Yokogawa VR204 Recorder in Service ................ 25 6.0 NO R M A L O PERA TIO N .................................................................................. 28 6.1 Checking the High Alarm Setpoint ........................................................... 28 6.2 Source Check of Radiation Monitors ......................................................... 31 6.3 Changing R-11/R-12 Sampling Point ....................................................... 32 6.4 Normal Operation of the Westronics Series 3000 Recorder ..................... 33 6.5 Normal Operation of the R-24A/B/C Yokogawa VR204 Recorder ............ 34 7.0 S HUT DOW N .................................................................................................. 35 7.1 Removing the Area Radiation Monitors from Service ............................... 35 7.2 Removing the Process Radiation Monitors from Service ........................... 36 7.3 Removing the Westronics Series 3000 Recorder from Service ................ 40 7.4 Removing the R-24A/B/C Yokogawa VR204 Recorder from Service ........... 41 8.0 INFREQ UENT O PERATIO N........................................................................... 42 8.1 Adjusting a Radiation Monitor High Alarm Setpoint ................................... 42 8.2 Infrequent Operation of the Westronics Series 3000 Recorder ................ 43 8.2.1 Dedicating Recorder Digital Display to a Single Point ......................... 43 8.2.2 Bypassing and Restoration of an RR-1 Alarm Set point ...................... 44 8.2.3 Changing an RR-1 Hi Alarm Setpoint ................................................... 46 8.3 Adjusting the R-24A/B/C Yokogawa VR204 Recorder Alarm Setpoint ......... 49 8.4 Radiation Monitor R-11/R-12 and R-20/R-21 Warning Lights ................... 52 8.5 Radiation Monitor Recorder Warning Alarm .............................................. 54 9.0 A TTA C HME NTS ............................................................................................ 55 9.1 PROCEDURE REFERENCE TABLE ....................................................... 56 9.2 WESTRONICS SERIES 3000 RECORDER PROGRAM PRINTOUT ..... 57 OP-920 Rev. 28 Page 3 of 59

1.0 PURPOSE 1.1 Provide instructions for Placing In-Service, Normal Operation, Removing from Service, and Infrequent Operation of the Area Monitors, Process Monitors, the Westronics Series 3000 Recorder, and the Yokogawa VR204 View Recorder.

2.0 REFERENCES

2.1 Improved Technical Specification LCO 3.3.6 and LCO 3.4.15 2.2 ODCM 3.10 and 3.11 2.3 SD-019, Radiation Monitoring System 2.4 OMM-001-12, Minimum Equipment List and Shift Relief 2.5 OMM-014, Radiation Monitor Setpoints 2.6 OP-001, Reactor Control and Protection System 2.7 OP-406, Steam Generator Blowdown/Wet Layup System 2.8 OP-509-1, Condensate Polishing System 2.9 OP-603, Electrical Distribution 2.10 OP-903, Service Water System 2.11 OP-917, Secondary Sampling System 2.12 OST-021, Daily Surveillances 2.13 OST-924-1, Area Radiation Monitoring System 2.14 OST-924-2, Process Radiation Monitoring System 2.15 RST-001, Radiation Monitor Source Checks 2.16 RST-008, Calibration of Radiation Monitor System, Monitors R-1 through R-8 2.17 RST-009, Calibration of Radiation Monitor System, Monitors R-9, R-30, R-31A, B, C and R-33 OP-920 Rev. 28 Page 4 of 59

2.18 RST-010, Calibration of Radiation Monitoring System, Monitor R-1 1 2.19 RST-011, Calibration of Radiation Monitoring System, Monitors R-12, R-20 and R-21 2.20 RST-012, Calibration of Radiation Monitoring System, Monitor R-14 2.21 RST-013, Calibration of Radiation Monitoring System, Monitor R-15 2.22 RST-014, Calibration of Radiation Monitoring System, Monitor R-16 2.23 RST-015, Calibration of Radiation Monitoring System, Monitor R-17 2.24 RST-016, Calibration of Radiation Monitoring System, Monitor R-18 2.25 RST-017, Calibration of Radiation Monitoring System, Monitors R-37 and R-19A, B, and C 2.26 RST-020, Verification of Electronic Calibration of Radiation Monitoring System Monitors R-32A & B 2.27 RST-029, Calibration of R-24A, B & C 2.28 EMP-013, Operation of R-14 and F-14 2.29 EMP-020, Operation of R-22 and R-38 2.30 EMP-022, Gaseous Waste Release Permits 2.31 EMP-023, Liquid Waste Release and Sampling 2.32 EMP-024, RETS Surveillance 2.33 EMP-026, Calibration of R-22 and R-38 2.34 EMP-027, Operation of GA Monitors R-37 and R-19A, B, and C 2.35 EMP-028, Process Monitor Setpoint Determination 2.36 EMP-031, Operation of the NMC AM-221F (R-23) Radwaste Building Effluent Monitor OP-920 Rev. 28 Page 5 of 59

2.37 EMP-034, Operation of R-24A, B & C 2.38 MST-901, Radiation Monitoring System 2.39 CM-738, Configuration of Radiation Monitoring System Channel Ratemeters 2.40 LP-256, Containment High Range Radiation Monitor (Area) RMS 32A & 32B 2.41 PIC-024, Plant Stack Radiation Monitor Channels R-14D and R-14E Dual Interface Card 2.42 PLP-037, Conduct of Infrequently Performed Tests or Evolutions 2.43 Nuclear Measurements Corporation Stack Monitor Operating Instructions (CP&L #728-523-61 and CP&L #736-879-23) 2.44 Updated FSAR, Section 11.5 2.45 GA Technologies Instruction Manual (CP&L #731-040-02 and CP&L #739-126-51) 2.46 Westronics Series 3000 Digital Data Recorder Users Manual 2.47 Yokogawa Instruction Manual for VR200 Wide View Recorder 2.48 Nuclear Research Corporation Operation and Maintenance Manual (CP&L #736-761-32 and CP&L #736-761-57) 2.49 EE 93-184, PPS Solenoid Valve Containment Integrity Concern 2.50 CR 94-01841, Operation of R-1 1/R-12 Pressure Switch 2.51 CR 97-00542, R-1 1 OOS When Filter Paper Changed 2.52 NCR 24351, R-1 1 Count Increase 2.53 NCR 25812, Evaluate Creating an RR-1 Procedure 2.54 NCR 60636, R-1 1 & R-12 Radiation Monitor Operability Question 2.55 OMM-007, Equipment Inoperable Record OP-920 Rev. 28 Page 6 of 59

3.0 PREREQUISITES 3.1 The Electrical Distribution System is in service lAW OP-603.

3.2 The Service Water System is in service lAW OP-903.

3.3 Steam Generator Blowdown is in service lAW OP-406.

3.4 The Secondary Sampling System is in service lAW OP-917.

3.5 The Reactor Protection and Control System is in service lAW OP-001.

3.6 The Condensate Polishing System is in service lAW OP-509-1.

I OP-920 Rev. 28 Page 7of59

4.0 PRECAUTIONS AND LIMITATIONS 4.1 Operate the Radiation Monitor Recorder continuously when the Radiation Monitoring System is in operation.

4.2 The Radiation Monitor Recorder points are divided into the following eight groups for the different Radiation Monitors:

- Group 1: R-9, 11, 12, 14A, 14B, 14C, 15, and 18

- Group 2: R-1, 3,4,6, and 8

- Group 3: R-2, 7, 32A, and 32B

- Group 4: R-16 and 17

- Group 5: R-5 and 21

- Group 6: R-19A, 19B, 19C, 31A, 31B, and 31C

- Group 7: R-14D, 14E, and 33

- Group 8: R-20 and 30 Group 1 radiation monitor recorder points are recorded continuously. For groups 2 through 8, the monitor points in the affected group will start recording when one of the monitors in that group exceeds the Warning Alarm setpoint established by the recorder.

4.3 Normally, changes to RR-1 Warning Alarm Set points is required due to monitor background changes and is requested by E&RC. Following any recorder program changes, a program printout should be obtained and compared to Attachment 9.2 to verify point parameters.

4.4 R-1 1/12 filter failure indication only detects the absence of paper at the alarm switch. Problems associated with the paper drive would not be detected by this alarm if the paper were still intact on the rollers. Paper drive problems may be characterized by a slowly increasing radiation trend without any other indications of a problem in the plant.

OP-920 I Rev. 28 Page 8 of 59

4.5 The following precautions apply when changing R- 11/12 Filter Paper.

(IA 97-OP-39)

- R-1 1/12 Vacuum Pump shall be secured when the Filter Paper is removed.

- R-1 1 and R-12 are inoperable with the filter housing cover open.

- R-1 1 is inoperable with the filter paper removed.

- R-12 is inoperable with the Vacuum Pump stopped.

4.6 R-11 shall be declared Inoperable (ITS LCO 3.3.6 & 3.4.15 / ODCM Table 3.10-1

& Table 3.11-1) for a minimum of 25 minutes following operation of the filter drive in fast speed OR following filter paper change out. (NCR 24351) 4.7 For planned or routine activities requiring a radiation monitor to be removed from service where the monitor will be returned to service prior to the end of shift, an EIR is not required. However, any compensatory actions required by the EIR, such as notification to the E&C shift technician of monitor inoperability, must still be performed. Additionally, the component and any compensatory actions should still be logged in the SSO log for tracking purposes.

4.8 The following is the status of indicating lights for RCV-014 located at the Waste Disposal Boron Recycle Panel (WDBR) for various operating conditions:

During Release - Valve is normally throttled, red light is illuminated, and green and white lights are extinguished.

No Release in Progress - Valve is closed, green light is illuminated, and red and white lights are extinguished.

High Alarm on R-14C - Valve closed, white and green lights are illuminated, and red light is extinguished.

To reopen RCV-014 after an R-14C alarm, the Valve Control Wheel (on WDBR panel) must be positioned to "0" and the alarm cleared. The valve can then be positioned as desired.

I OP-920 Rev. 28 Page 9 of 59

4.9 The following applies to R-24A, B & C indication:

Readings are NOT valid for quantification of a S/G tube leak until the monitor has been adjusted per EMP-034 to match the actual measured primary to secondary leakage, but can be used for trending to determine if leakage is present AND if leakage is increasing.

Readings are NOT power compensated when power is less than 40%.

Readings will increase with power below 40%. Trend information may only be useful if the plant is at a constant Reactor power when below 40% power.

4.10 The principles of ALARA shall be used in planning and performing work and operations in the Radiation Control Area.

4.11 This procedure has been screened in accordance with PLP-037 criteria and determined Not Applicable to PLP-037.

IOP-920 Rev. 28 Page 10 of 59

REFERENCE USE Section 8.1 Page 1 of 1 8.0 INFREQUENT OPERATION 8.1 Adjusting a Radiation Monitor High Alarm Setpoint NOTE: The Setpoint Log and Change Record in OMM-014 is used to record adjustment of the High Alarm setpoint of Radiation Monitors R-1, R-2, R-3, R-4, R-5, R-6, R-7, R-8, R-9, R-30, R-31A, R-31B, R-31C, R-32A, R-32B and R-33.

8.1.1 Adjusting the High Alarm Setpoint for Monitors R-1, R-2, R-3, R-4, R-5, R-6, R-7, R-8, R-9, R-11, R-12, R-15, R-16, R-17, R-18, R-20, R-21, R-30, R-31A, R-31B, R-31C and R-33:

1. Slide out the drawer for the desired monitor.

2. Adjust the thumbwheels to desired setpoint.

3. Push Alarm/Reset button and verify setpoint appears on ratemeter display.

4. Return the monitor drawer to its proper position.

8.1.2 Adjustment of the High Alarm Setpoint for Monitors R-14A, R-14B, R-14C, R-14D, and R-14E will be performed by E&C lAW the applicable sections of EMP-022 and/or EMP-028.

8.1.3 Adjustment of the Alert and High Alarm Setpoints for Monitors R-19A, R-19B, and R-19C will be performed by E&C lAW the applicable sections of EMP-023.

8.1.4 Adjustment of the Alert and High Alarm Setpoints for Monitors R-32A and R-32B will be performed by I&C lAW the applicable sections of LP-256.

8.1.5 Adjustment of the High Alarm Setpoint for Monitor R-37 will be performed by E&C lAW the applicable portions of EMP-023.

OP-920 Rev. 28 Page 42 of 59

Appendix C Job Performance Measure Forri ES-C-1 Worksheet Facility: HB ROBINSON Task No.: 01000108705 Task

Title:

Monitor Nuclear Instrumentation During JPM No.: 2004 NRC JPM SRO Al Refueling Operations K/A

Reference:

G2.2.303.5 / 3.3 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

initial Conditions:

I ne plant is in a refueling outage.

" Core off-load is in progress. You are in communication with the Refueling SRO

Annunciator APP-005-A1 has just alarmed.

Task Standard: " Refueling terminated.

Audio Count Rate selected to N32.

Required Materials: None General

References:

APP-005-A1, Revision 26 OWP-01 1 TS 3.9.2 Handouts: None Initiating Cue: You are the only available licensed Operator in the Control Room. Respond as necessary.

Time Critical Task: N/A Validation Time: 10 minutes 2004 INRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

Appendix C Job Performance Measure Form ES-C-1 Worksheet SIMULATOR SETUP

1. IC 199.

2. Audio Count Rate Channel selected to N31.

3. START UP RATE Channel selected to N31.

4. Insert Malfunction NIS04A to 0 volts

5. FREEZE.

6. RUN when Candidate takes the watch.

2004 NRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

Appendix C Page 3 of 7 Form ES-C-1 PERFORMANCE INFORMATION (Denote Critical Steps with an asterisk)

Performance Step: 1 Responds to alarm APP-005-A1.

Standard: References procedure APP-005-AI.

Comment:

APP-005-A1 Action 1 Performance Step: 2 If the detector has failed then perform the following:

Standard:

Monitors control board indication and may check NIS Panel.

Determines N-31 has failed.

Comment:

NOTE: The JPM is written as if all actions of APP-005-Al are completed before N31 is removed from service (OWP-01 1).

If the candidate chooses to complete OWP-011 before completing APP-005-Al then go to JPM Performance Step 6.

Return to JPM Performance Step 4 when OWP-01 I has been completed. The JPM is complete when both APP-005-Al and OWP-011 have been completed.

APP-005-A1, Action 1.a Performance Step: 3 Remove the failed Source Range from service in accordance with OWP-01 1.

Standard: Refers to OWP-01 1 or completes the actions of APP-005-A1 before proceeding to OWP-01 1.

Examiner's Note: It is acceptable to perform the remaining actions of APP-005-Al before removing the channel from service in accordance with OWP-01 1.

Comment:

2004 NRC JPM SRO All NUREG 1021, Revision 9, Supplement 1

Appendix C Page 4 of 7 Form ES-C-1 PERFORMANCE INFORMATION APP-005-A1, Action 1.b Performance Step: 4 Refer to Tech Specs 3.8.1 and Table 3.5-2 (ITS Table 3.3.1-1 and ITS LCO 3.9.2).

Standard: Informs CRSS.

Examiner's Cue: Acknowledge as CRSS.

Comment:

APP-005-A1, Action 2 Performance Step: 5 IF the detector has failed AND refueling operations are in progress, THEN in addition to Step 1, perform the following:

a. Terminate fuel movement.

b. Log N51 OR N52 as a replacement channel for Control Room indication.

c. Verify the operable Source Range channel is selected to the AUDIO COUNT RATE Drawer.

d. WHEN desired, THEN recommence fuel movement.

Standard:

a. Contacts the Refueling SRO or CRSS to terminate fuel movement. *

b. Discusses a Control Room log entry to identify N51 OR N52 as a replacement channel for Control Room indication.

c. Selects the AUDIO COUNT RATE to N32.

NOTE: Selecting the AUDIO COUNT RATE to N32 is critical in either APP-005 or OWP-011.

Booth Operator Cue: a. If called as Refueling SRO: Acknowledge need to terminate fuel movement Examiner's Cue: a. As CRSS: Acknowledge need to terminate fuel movement

b. Acknowledge need to make log entry.

Comment:

2004 NRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

Appendix C Page 5 of 7 Form ES-C-1 PERFORMANCE INFORMATION OWP-01 1 Table Examiner Cue: Prior to START of OWP-01 1:

" Assume that you have verified the latest revision of OWP-011.

" The CRSS will assign a Work Request Number. Proceed to the VALVE/BREAKER/SWITCH lineup.

Performance Step: 6 a. REMOVE NI-31 from ERFIS SCAN: NIN0031A.

b. AUDIO COUNT RATE CHANNEL - CHANNEL SELECTOR SWITCH

c. START UP RATE CHANNEL SELECT Switch

d. NIS CHANNEL SELECTOR NR 45 PEN 1

e. NIS CHANNEL SELECTOR NR 45 PEN 2

f. LEVEL TRIP Switch

g. NIS TRIP BYPASS NI-31 Status Light

h. HIGH FLUX AT SHUTDOWN Switch Standard: a. Uses DR command to remove NIN0031A from processing

b. Selects N32 for the AUDIO COUNT RATE CHANNEL using the Channel Selector Switch

c. Selects N32 on the START UP RATE CHANNEL SELECT Switch d, e.Verifies N32 is selected on at least one pen

f. Turns the LEVEL TRIP Switch to the BYPASS position on the front of the N-31 drawer

,. Verifies status light on RTGB is illuminated

h. Turns the HIGH FLUX AT SHUTDOWN Switch to the BLOCK position on the front of the N-31 drawer Comment:

Terminating Cue: When fuel movement has been terminated and N-31 removed from service, inform Candidate that evaluation on this JPM is complete.

STOP TIME:

2004 NRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

Appendix C Page 6 of 7 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2004 NRC JPM SRO Al Examinee's Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiner's Signature: Date:

2004 NRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

Appendix C Page 7 of 7 Form ES-C-1 JPM CUE SHEET INITIAL CONDITIONS: e The plant is in a refueling outage.

Core off-load is in progress. You are in communication with the Refueling SRO

Annunciator APP-005-Al has just alarmed.

INITIATING CUE: You are the only available Licensed Operator in the Control Room. Respond as necessary.

2004 NRC JPM SRO Al NUREG 1021, Revision 9, Supplement 1

APP-005-Al ALARM SR DET LOSS OF DC *** WILL REFLASH AUTOMATIC ACTIONS

1. None Applicable CAUSE

1. Loss of DC voltage to Detector

2. Source Range Trip Blocked (expected alarm)

3. Loss of instrument power OBSERVATIONS

1. Source Range Instruments

2. Source Range Detector Volts

3. Source Range Instrument Power Fuses ACTIONS

1. IF the detector has failed, THEN perform the following:

a. Remove the failed Source Range from service in accordance with OWP-01 1.

b. Refer To Tech. Specs. 3.8.1 and Table 3.5-2 (ITS Table 3.3.1-1 and ITS LCO 3.9.2).

2. IF the detector has failed AND refueling operations are in progress, THEN in addition to Step 1, perform the following:

a. Terminate fuel movement.

b. Log N51 OR N52 as a replacement channel for Control Room indication.

c. Verify the operable Source Range channel is selected to the AUDIO COUNT RATE Drawer.

d. WHEN desired, THEN recommence fuel movement.

DEVICE/SETPOINTS

1. N31 / x 100 volts below normal voltage

2. N32 / x 100 volts below normal voltage POSSIBLE PLANT EFFECTS

1. Possible entry into Tech Spec LCO Action REFERENCES

1. ITS Table 3.3.1-1, Item 4 and ITS LCO 3.9.2

2. CWD B-190628, Sh 441 Cable AF, 443 Cable AF

3. OWP-01 1, Nuclear Instrumentation (NI)

I APP-005 Rev. 26 F Page 4 of39

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 3 PART 10 owP-o11 NUCLEAR INS TRUMENTA TION (NI)

REVISION 16 I OWP-011 Rev. 16 Page 1of27

SUMMARY

OF CHANGES PRR 106772 STEP # REVISION COMMENTS NI-1, NI-2, NI-3, NI-4, Added to the 2 nd note on LIU for maintenance: restoring NI-41 (42, 43, 44) to ERFIS scan for I&C calibration activities may adversely effect CAOC/DELTA Flux and R-24A, B and C, R-24A, B and C should be declared inoperable when the NI is on ERFIS scan and calibration activities are in progress.

OWP-01 1 Rev. 16 Page 2 of 27

CONTINUOUS USE OWP

Title:

NI-5 Page 1 of 2 NI-31, Source Range

1. This revision has been verified to be the latest revision available.

(Print)

Name Signature Date

2. System: NI Work Request No:__

3. Component: NI-31, Source Range

4. Scope of Work:

Perform maintenance on Nuclear Instrument NI-31.

5. Testing required on redundant equipment prior to rendering component inoperable:

N/A

6. Precaution:

1) Refer to ITS Table 3.3.1-1 for Source Range applicability and operability requirements when not in the Refueling condition (MODE 6)

2) Reference ITS LCO 3.9.2 during Refueling Operations (MODE 6).

3) Removal of control power fuses below P-6 will cause a reactor trip signal.

4) This OWP has been screened in accordance with PLP-037 criteria and determined to be a Case Three activity.

7. Valve/Breaker/Switch lineup has been completed. /

Signature Date

8. Clearance Issued (If applicable) Clearance No:

9. I&C Maintenance lineup complete. N/A / N/A Signature Date

10. Clearance removed and Valve/Breaker/

Switch lineup restored to normal.

/SignatureDate

11. Source Range NI-31 has been declared Signature Date operable.

Signature Date OWP-01 1 Rev. 16 1 Page 20 of 27

CONTINUOUS USE OWP

Title:

NI-5 Page 2 of 2 VALVE, BREAKER, SWITCH LINEUP COMPONENT POSITION FOR RESTORED DESCRIPTION MAINTENANCE POSITION SOURCE RANGE CHANNEL NI-31 INIT INIT REMOVE NI-31 from ERFIS SCAN: NIN0031A REMOVED RESTORED AUDIO COUNT RATE CHANNEL - CHANNEL SELECTOR Switch Selected to SR 32 START UP RATE CHANNEL SELECT Switch

NI NIS CHANNEL SELECTOR NR 45 PEN 1

NI NIS CHANNEL SELECTOR NR 45 PEN 2

NI LEVEL TRIP Switch BYPASS NORMAL NIS TRIP BYPASS NI-31 Status Light ILLUM EXTNG NORMAL IF SHUTDOWN HIGH FLUX AT SHUTDOWN Switch BLOCK OR BLOCK Switch should be selected to any NI which is NOT removed from service.

OWP-011 Rev. 16 Page 21 of 27

Nuclear Instrumentation 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO 3.9.2 Two source range neutron flux monitors shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required source A.1 Verify one Post 15 minutes range neutron flux Accident Monitor monitor inoperable (PAM) source range neutron flux monitor provides indication in the Control Room.

AND A.2 Log indicated PAM 30 minutes source range neutron monitor count rate. AND Once per 30 minutes thereafter (continued)

HBRSEP Unit No. 2 3.9-2 Amendment No. t-76,8O.190 I

Nuclear Instrumentation 3.9.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Actions and B.1 Suspend CORE Immediately Completion Times of ALTERATIONS.

Condition A not met.

AND B.2 Suspend operations Immediately that would cause introduction into the RCS, coolant with boron concentration less than required to meet boron concentration of LCO 3.9.1.

C. Two required source C.1 Initiate action to Immediately range neutron flux restore one source monitors inoperable, range neutron flux monitor to OPERABLE status.

AND C.2 Suspend CORE Immediately ALTERATIONS.

AND C.3 Suspend positive Immediately reactivity additions.

AND C.4 Perform SR 3.9.1.1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter HBRSEP Unit No. 2 3.9-3 Amendment No. V67,2i0.190

Nuclear Instrumentation 3.9.2 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.9.2.2 -------------------NOTE --------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perfrm.HANNL.CLIBRTIO....18....months......

Perform CHANNEL CALIBRATION. 18 months HBRSEP Unit No. 2 3.9 -3a Amendment No. ilk,18O,190 I

Appendix C Page 1 of 6 Form ES-C-1 PERFORMANCE INFORMATION Facility: HB ROBINSON Task No.: 01001101601 Task

Title:

Perform a Manual Shutdown Margin JPM No.: 2004 NRC JPM SRO Boron Concentration Calculation A2 Followina a Reactor Trip K/A

Reference:

G2.1.25 2.8/3.1 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom X Simulator Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

15 minutes ago an automatic reactor trip occurred when an RCP breaker tripped.

Control Bank D Rod D-08 failed to insert.

Following are the steady state conditions prior to the trip:

100% power

Bank D: 218 Steps

RCS Boron Concentration 190 PPM

Tavg: 574 'F

Cycle Exposure: 16400 MWD/MTU Task Standard: Calculations within the band specified.

Required Materials: FMP-012 Plant Curve Book Calculator 2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 2 of 6 Form ES-C-1 PERFORMANCE INFORMATION General

References:

FMP-012, Manual Determination of Shutdown Margin Boron Concentration, Revision 21 Plant Curve Book Handouts: NONE Initiating Cue: You are an extra operator. The CRSS has assigned you to complete a Shutdown Margin Boron Concentration Calculation in accordance with FMP-012. The plant will be maintained at 547 0 F. Powertrax is unavailable Time Critical Task: N/A Validation Time: 30 minutes 2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 3 of 6 Form ES-C-1 PERFORMANCE INFORMATION (Denote Critical Steps with an asterisk)

Performance Step: I Obtain a current copy of the procedure.

Standard: Candidate locates and reviews the procedure.

Examiner Cue: Provide a copy of FMP-012 andlor tell the Candidate that the procedure obtained/provided is the current copy.

Comment:

FMP-012, Step 6.2.1 Performance Step: 2 Most recent steady state critical conditions:

1. Using the most recent steady state Critical Data Stamp, COMPLETE Lines 1 through 5 on Attachment 7.2.

2. Record the current Cycle Exposure on Line 6 of Section I on ATTACHMENT 7.2.

Standard: Candidate completes ATT. 7.2, Lines 1-6, using the data in the Initial Conditions.

Comment:

The following steps are under 6.2.2: Boron Concentration Required to Maintain a Minimum of 1.77% k/k Shutdown Margin at 547°F:

FMP-012, Step 6.2.2.1 Performance Step: 3 Using Curve 1.11 and/or Table 1.11 of the Station Curve Book, locate the boron concentration corresponding to the current Cycle Exposure and 547°F, and record in Step 11.1, ATTACHMENT 7.2.

Standard: Using either Curve or Table 1.11, determines 1.77% SDM boron concentration to be 350-370 PPM and records it on Att. 7.2, Line 11.1.

Examiner's Note: Table 1.11 is the most accurate - 358 PPM.

Comment:

2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 4 of 6 Form ES-C-1 PERFORMANCE INFORMATION FMP-012, Step 6.2.2.2 Performance Step: 4 Using the Curve Book, Table 1.5 or Curve 1.5, determine the Differential Boron Worth based on boron concentration determined in Step 11.1, current Cycle Exposure and 547°F, and record in Step 11.2, ATTACHMENT 7.2.

Standard: Using Curve 1.5 (EOL 547 Curve line) or Table 1.5, determines Boron Worth to be between 8.9 pcm/PPM and 9.1 pcm/PPM and records it on ATT. 7.2, Line 11.2.

Comment:

FMP-012, Step 6.2.2.3 Performance Step: 5 IF there are Inoperable/Untrippable control rods, THEN determine Inoperable/Untrippable rod worth in ppm by multiplying the number of Inoperable/Untrippable control rods by the most reactive rod worth from Table 1.15 of the Station Curve Book and dividing by Differential Boron Worth determined in Step 11.2, and record in Step 11.3, ATTACHMENT 7.2.

Standard: Calculates Inoperable/Untrippable rod worth to be between 184 PPM and 188 PPM and records it on ATT. 7.2, Line 11.3.

Comment:

FMP-012, Step 6.2.2.4 Performance Step: 6 Calculate the compensated minimum RCS Boron concentration by adding the Inoperable/Untrippable rod worth from Step 11.3 to the minimum RCS Boron Concentration and record in Step 11.4, ATTACHMENT 7.2.

Standard: Determines the compensated minimum RCS Boron concentration to be between 534 PPM and 558 PPM and records it on ATT. 7.2, Line 11.4.

Comment:

Terminating Cue: When Attachment 7.2 sections I and II are complete and returned to the CRSS, inform the Candidate that evaluation on this JPM is complete.

STOP TIME:

2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 5 of 6 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2004 NRC JPM SRO A2 Examinee's Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiner's Signature: Date:

2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 6 of 6 Form ES-C-1 JPM CUE SHEET INITIAL CONDITIONS: 15 minutes ago an automatic reactor trip occurred when an RCP breaker tripped.

Control Bank D Rod D-08 failed to insert.

Following are the steady state conditions prior to the trip:

0 100% power 0 Bank D: 218 Steps 0 RCS Boron Concentration 190 PPM

Tavg: 574 'F 0 Cycle Exposure: 16400 MWD/MTU INITIATING CUE: You are an extra operator. The CRSS has assigned you to complete a Shutdown Margin Boron Concentration Calculation in accordance with FMP-012. The plant will be maintained at 547 0 F.

Powertrax is unavailable 2004 NRC JPM SRO A2 NUREG 1021, Revision 9, Supplement 1

IFMP-012 Rev. 19 Pa1g of30W

SUMMARY

OF CHANGES PRR 100951 REVISION # REVISION COMMENTS 19 Converted to XP Steps 3.2 and 4.8: Editorial correction of CRSS title.

Step 5.2.2: Added discussion of additional boration required to ensure the reactor is 1 % subcritical in Modes 3 through 5 if Shutdown Bank A is inserted instead of Shutdown Bank B.

Steps 6.2.2.4, 6.2.3.5, Att 7.2 Step 11.4, Att 7.2 Step 111.5: Revised the step to include the worth of H-1 0 if the H-1 0 ARPI is out of service or if Shutdown Bank A will be inserted instead of Shutdown Bank B.

FMP-012 I Rev. 19 1 Page 2 of 30

TABLE OF CONTENTS SECTION PAGE 1.0 P UR P O S E ..................................................................................................... 4 2.0 R E FER E NC E S ................................................................................................ 4 3.0 R ES PO NSIB ILITIES ....................................................................................... 5 4.0 DEFINITIONS/ABBREVIATIONS .................................................................... 5 5 .0 G E NE RA L ...................................................................................................... 6 6.0 P R O C E D U R E ................................................................................................ 13 7.0 A TTA C HMENTS ........................................................................................... 20 7.1 SHUTDOWN MARGIN MODES 1, 2 DATA FORM ................................... 21 7.2 SHUTDOWN MARGIN BORON CONCENTRATION MODES 3, 4, 5 D AT A F O R M............................................................................................ . . 24 7.3 SHUTDOWN MARGIN BORON CONCENTRATION MODE 6 DATA F O R M ...................................................................................................... . . 28 7.4 CONTROL ROD WORTHS ....................................................................... 29 FMP-012 I Rev. 19 Page 3 of 3

1.0 PURPOSE 1.1 To identify the Technical Specification Shutdown Margin requirements and describe how H. B. Robinson complies with those requirements.

1.2 To provide instructions for determining Shutdown Margin in the event of misaligned control rod(s) or inoperable/untrippable control rod(s).

1.3 To provide the necessary information to manually determine a Shutdown Margin boron concentration in order to comply with Technical Specifications for Modes 1, 2, 3, 4, 5, and 6 in the event Powertrax can not be used.

NOTE: PLP-037 is not applicable to this procedure.

2.0 REFERENCES

2.1 H. B. Robinson - Unit 2 Station Curve Book 2.1.1 Curve 1.1 and Table 1.1, Critical Boron Letdown Curve 2.1.2 Curve 1.3 and Table 1.3, Power Defect vs. Power 2.1.3 Curve 1.5 and Table 1.5, Differential Boron Worth vs. Boron Concentration 2.1.4 Curve 1.9 and Table 1.9, Rod Insertion Limits 2.1.5 Curve 1.11 and Table 1.11, Boron Concentration Required to Maintain A Minimum of 1.77% Ak/k Shutdown Margin 2.1.6 Curve 1.14 and Table 1.14, Boron Concentration Required to Maintain 4.0% Ak/k Shutdown Margin 2.2 Technical Specifications ITS LCO 3.1.1, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, LCO 3.4.5, LCO 3.4.6, LCO 3.9.1, ITS SR 3.1.1.1, SR 3.4.5.6, SR 3.9.1.1 FMP-012 Rev. 19 Page 4 of 30

2.3 ACR 92-316 2.4 PLP-100, Technical Requirements Manual (TRM) 2.5 FMP-001, Core Operating Limit Report (COLR) 2.6 EC 47804, Cycle 22 Core Design and Analysis 2.7 CR 81249, Incorrect MRR value in FMP-012 2.8 EMF-2781 (P), Robinson Nuclear Plant Cycle 22 Safety Analysis Report 2.9 RNP-F/NFSA-0083, RNP Cycle 22 H-10 Rod Worth 3.0 RESPONSIBILITIES 3.1 Manual calculation of shutdown margin boron concentration shall be performed by either Reactor Systems Engineering or Operations personnel.

3.2 The Superintendent - Shift Operations, OR Control Room Shift Supervisor, OR the Supervisor - Reactor Systems shall review and approve each manual shutdown margin calculation prior to its being considered valid.

4.0 DEFINITIONS/ABBREVIATIONS 4.1 Steady state, as used in this procedure, is the point at which power level has not changed for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.2 MWD/MTU - Megawatt-Days/Metric Ton Uranium, a unit of cycle exposure.

4.3 SDM - Shutdown Margin 4.4 RIL - Rod Insertion Limits 4.5 ITS - Improved Technical Specifications 4.6 COLR - Core Operating Limit Report 4.7 SSO - The Superintendent - Shift Operations 4.8 CRSS - Control Room Shift Supervisor FMP-012 Rev. 19 Page 5 of 30-1

5.0 GENERAL 5.1 SDM is the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition less the untrippable rod(s), if all control rods fully inserted except for the single control rod of highest reactivity worth which is assumed to remain fully withdrawn. SDM is initially maintained by control rods, the existing Reactor Coolant System (RCS) boron concentration and the presence of fission product poisons (xenon and samarium). Additional boron must be added to the RCS to offset the positive reactivity inserted by the decay of the fission product poisons in order to maintain the SDM.

5.2 Determination of the Boron Concentration Required to Provide Shutdown Margin 5.2.1 Modes 1 and 2 If the reactor is critical (Mode 1 and Mode 2 with Keff > 1.0), then adequate SDM can be verified by checking that the control rods are above the RILs specified in the COLR. If the control rods are above the RILs and all are trippable, then adequate SDM exists and no other verifications are necessary. However, if any control rod banks are below the RILs, or if any control rod(s)are misaligned and below the RILs, or if any control rod(s) are untrippable, then a SDM verification must be performed by determining the available SDM and comparing it to the required SDM. The available SDM is calculated by first determining the Total Rod Worth based on cycle exposure. The Total Rod Worth includes allowances for an unknown stuck rod (N-i), Power Shape effects, a 10% rod worth design uncertainty and other uncertainties.

Once the Total Rod Worth is determined, the Power Defect and inserted D Bank worth for the current plant conditions as well as the worth of any misaligned control rod bank(s) and the worth of any Inoperable/

Untrippable control rod(s) are subtracted from the Total Rod Worth to determine the available SDM. If the available SDM is less than the required SDM, then a potential exists for not having adequate SDM and the RCS must be borated to restore SDM.

FMP-012 Rev. 19 Page6of30

5.2.1 (Continued)

During startup (Mode 2 with Keff < 1.0), ITS SR 3.1.6.1 requires verification that the estimated critical control bank position is above the RILs. Adequate SDM is maintained as long as the RCS boron concentration is greater than or equal to the boron concentration which would result in criticality with the control rods at or above the RILs.

5.2.2 Modes 3 through 5 The boron concentration determined from Station Curve Book Curve 1.11 or Table 1.11 or from the POWERTRAX program for a specific RCS temperature will maintain the reactor 1 % Ak/k subcritical (keff =

0.99) at that temperature with no xenon, equilibrium samarium and all control rods fully inserted except for Shutdown Bank A and the control rod with the highest reactivity worth. Shutdown Bank A is verified during the core design process to have sufficient negative reactivity to make the reactor subcritical by at least an additional 0.77% Ak/k over the entire life of the core. Therefore, the boron concentration determined from Curve 1.11 or Table 1.11 of the Station Curve Book or from the POWERTRAX program will provide at least 1.77% Ak/k SDM.

For Cycle 22, if it is elected to insert Shutdown Bank A instead of Shutdown Bank B in Modes 3 through 5 then additional boration will be required to ensure that the reactor is 1 % Ak/k subcritical (keff = 0.99) since the worth of Shutdown Bank A is less than Shutdown Bank B. The worth of the H-10 control rod will be used to conservatively bound the difference in the worth between Shutdown Bank A and Shutdown Bank B.

The boron concentration determined from Station Curve Book, Curve 1.14 or Table 1.14 for a specific RCS temperature will provide a SDM as specified in the COLR at that temperature with no xenon, equilibrium samarium and the single control rod of highest reactivity worth fully withdrawn.

The boron concentration required to maintain a specified SDM decreases as the cycle exposure increases; therefore, the boron concentration required for a lower cycle exposure will typically conservatively bound the boron concentration required to maintain that same SDM at a higher cycle exposure. However, this may not always be true at the beginning of a cycle with large amounts of burnable poison in the core. Always verify the shape of the SDM curves versus exposure.

I FMP-012 Rev. 19 Page 7 of 30

5.2.2 (Continued)

The boron concentration required to maintain a specified SDM typically decreases as the temperature increases. However, during conditions when the boron concentration in the RCS is very high, it is possible that the opposite is true. This is typical in the refueling mode when high boron concentrations are required. So it should not be assumed that the boron concentration required for a lower temperature will conservatively bound the boron concentration required to maintain that same SDM at a higher RCS temperature. A verification of SDM curves within the range of the temperatures and boron concentrations expected during the modes of operation should be performed in order to determine if the boron concentration based on the lower RCS temperature will bound the SDM requirements.

5.2.3 Mode 6 The boron concentration provided in the COLR is verified during the core design process to maintain at least the SDM as specified in the COLR and ITS Bases at temperatures up to and including 140'F.

5.3 Shutdown Margin Requirements The amount of SDM required by ITS as specified in the COLR varies with plant conditions and core life. The SDM requirements and how H. B. Robinson meets those requirements are described below:

5.3.1 Power Operation, (Mode 1)

When the reactor is in Power Operation (Keff > 0.99 and Power > 5%),

the SDM requirements are specified in the COLR. The maximum SDM requirement, typically, occurs at the end of core life (low boron concentration) and is based on the value used in the analysis of the hypothetical steamline break accident. The additional SDM is required to suppress the positive reactivity inserted by an uncontrolled cooldown at the end of core life. ITS 3.1.1 is not applicable in Mode 1. However, there are LCOs which require a verification of SDM during certain control rod configurations in Mode 1. The following is a description of such requirements.

FMP-012 Rev. 19 Page 8 of 30

5.3.1 (Continued)

LCO 3.1.4 requires for all the control rods to be operable. If one or more rods becomes inoperable then perform required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If one rod is not within alignment limits then perform required actions B.2.1.1, Verify SDM is within limits specified in the COLR, OR B.2.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , AND perform required action B.2.3, Verify SDM is within limits specified in the COLR once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If more than one rod is not within alignment limit then perform required actions D.1.1, Verify SDM is within limits specified in the COLR, OR D.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

LCO 3.1.5 requires each shutdown bank shall be within insertion limits specified in the COLR. If one or both shutdown banks not within limits as specified in the COLR, then perform required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

LCO 3.1.6 requires control banks shall be within the insertion, sequence, and overlap limits specified in the COLR. If control bank insertion limits are not met, then perform the following required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If control bank sequence or overlap limits are not met, then perform required actions B.1.1, Verify SDM is within limits specified in the COLR, OR B.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

FMP-012 Rev. 19 Page 9 of 30

5.3.2 Startup, (Mode 2)

When the reactor is in Startup (Keff > 0.99 and Power < 5%), the SDM requirements are specified in the COLR. The maximum SDM requirement, typically, occurs at the end of core life (low boron concentration) and is based on the value used in the analysis of the hypothetical steamline break accident. The additional SDM is required to suppress the positive reactivity inserted by an uncontrolled cooldown at the end of core life. ITS 3.1.1 is not applicable in Mode 2 with Keff >

1.0. However, ITS 3.1.1 in mode 2 with Keff < 1.0 is met by maintaining the required SDM as specified in the COLR during the entire fuel cycle and by performing SR 3.1.1.1, Verify SDM every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In addition to the above requirements there are LCOs which require a verification of SDM during certain control rod configurations. The following is a description of such requirements.

LCO 3.1.4 in Mode 2 requires for all the control rods to be operable. If one or more rods becomes inoperable then perform required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If one rod is not within alignment limits then perform required actions B.2.1.1, Verify SDM is within limits specified in the COLR, OR B.2.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , AND perform required action B.2.3, Verify SDM is within limits specified in the COLR once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If more than one rod is not within alignment limit then perform required actions D.1.1, Verify SDM is within limits specified in the COLR, OR D.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

LCO 3.1.5 in Mode 2 with any control bank not fully inserted requires each shutdown bank shall be within insertion limits specified in the COLR. If one or both shutdown banks are not within limits as specified in the COLR, then perform required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

FMP-012 Rev. 19 Page 10 of 30

5.3.2 (Continued)

LCO 3.1.6 in Mode 2 with Keff > 1.0 requires control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.

If control bank insertion limits are not met, then perform the following required actions A.1.1, Verify SDM is within limits specified in the COLR, OR A.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If control bank sequence or overlap limits are not met, then perform required actions B.1.1, Verify SDM is within limits specified in the COLR, OR B.1.2, Initiate boration to restore SDM to within limit, and these actions need to be performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

5.3.3 Hot Standby, (Mode 3)

When the reactor is in Hot Standby (Keff < 0.99 and Tavg > 350 F), the SDM requirements are specified in the COLR. The SDM requirements are based on the Boron Concentration in the RCS and the number of reactor coolant loops in operation. The maximum SDM requirement for

> 2 reactor coolant loops in operation typically occurs at the end of core life (low boron concentration) and is based on the value used in the analysis of the hypothetical steamline break accident. The additional SDM is required to suppress the positive reactivity inserted by an uncontrolled cooldown at the end of core life. ITS 3.1.1 is met by providing required boron concentrations in Curve Book Curve 1.11 which maintain a SDM of > 1.77% Ak/k during the entire fuel cycle, and by verifying that the actual RCS boron concentration is greater than or equal to the required boron concentration ( ITS SR 3.1.1.1).

With less than two reactor coolant loops in operation AND the rod control system capable of rod withdrawal AND the reactor trip breakers closed AND the lift disconnect switches for all control rods not fully withdrawn closed then additional SDM is required. ITS LCO 3.4.5.d requires that a SDM as specified in the COLR be maintained. ITS LCO 3.4.5.d is met by providing required boron concentrations in Curve Book Curve 1.14 which provide the required SDM and by verifying that the actual RCS boron concentration is greater than or equal to the required boron concentration. (ITS SR 3.4.5.6)

FMP-012 Rev. 19 F Page 11 of 30

5.3.4 Hot Shutdown, (Mode 4)

When the reactor is in Hot Shutdown (Keff < 0.99 and 200°F < Tavg <

3500 F), the SDM requirements are specified in the COLR. The SDM requirements are based on the Boron Concentration in the RCS and the number of reactor coolant loops in operation. The maximum SDM requirement for _>2 reactor coolant loops in operation typically occurs at the end of core life (low boron concentration) and is based on the value used in the analysis of the hypothetical steamline break accident. The additional SDM is required to suppress the positive reactivity inserted by an uncontrolled cooldown at the end of core life. ITS 3.1.1 is met by providing required boron concentrations in Curve Book Curve 1.11 which maintain a SDM of _>1.77% Ak/k during the entire fuel cycle, and by verifying that the actual RCS boron concentration is greater than or equal to the required boron concentration ( ITS SR 3.1.1.1).

With less than two reactor coolant loops in operation AND the rod control system capable of rod withdrawal AND the reactor trip breakers closed AND the lift disconnect switches for all control rods not fully withdrawn closed then additional SDM is required. ITS LCO 3.4.5.d requires that a SDM as specified in the COLR be maintained. ITS LCO 3.4.5.d is met by providing required boron concentrations in Curve Book Curve 1.14 which provide the required SDM and by verifying that the actual RCS boron concentration is greater than or equal to the required boron concentration. (ITS SR 3.4.5.6)

With less than one loop or train consisting of RCS loops or residual heat removal (RHR) trains in operation AND the rod control system capable of rod withdrawal, ITS LCO 3.4.6, Action C.1 requires that all operations involving a reduction of RCS boron concentration to be suspended immediately.

FMP-012 Rev. 19 Page 12 of 30

5.3.5 Cold Shutdown, (Mode 5)

When the reactor is in Cold Shutdown (Keff < 0.99 and Tavg <-2000 F),

ITS 3.1.1 requires a SDM as specified in the COLR. ITS 3.1.1 is met by administratively requiring a SDM of > 1.77% Ak/k for all RCS temperatures and by performing SR 3.1.1.1, Verify SDM every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.3.6 Refueling, (Mode 6)

When the reactor is in the refueling mode (one or more reactor vessel head closure bolts less than fully tensioned and RCS temperatures

< 140°F {TRM 1.1}), ITS 3.9.1 requires a Boron Concentration as specified in the COLR. The core design process verifies that an RCS boron concentration as specified in the COLR will provide adequate SDM at RCS temperatures _ 140 0 F. Therefore, ITS 3.9.1 is met by verifying that the RCS boron concentration is at least as specified in the COLR and is conducted by performing SR 3.9.1.1, Verify SDM every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

6.0 PROCEDURE NOTE: If it is desired to determine the SDM boron concentration for a particular plant mode only, then perform either sections 6.1, 6.2, or 6.3 and complete either ATTACHMENT 7.1, 7.2, or 7.3 respectively and N/A all others.

Section 6.1 is performed only as a response to the LCO required action.

NOTE: During Cycle 22, while H-10 ARPI is inoperable, H-10 control rod worth will not be included in the determination of SDM in Modes 1 - 5. Also, Step 6.1.1 and Sections 1,11, and III of Attachment 7.1 shall not be performed while H-10 ARPI is OOS 6.1 Mode 1 and 2 6.1.1 If there are only misaligned control rod(s) which are located in Control Banks D or C, then determine current reactor conditions and complete Lines 1 through 6,Section I on Attachment 7.1, otherwise N/A Sections 1,11, and Ill in Attachment 7.1 and continue with Sep 6.1.4 and Section IV of Attachment 7.1.

FMP-012 Rev. 19 1 Page 13 of 30

6.1.2 Using the COLR or Curve Book, Table 1.9 or Curve 1.9 and the power level recorded in Attachment 7.1,Section I, Step 2, determine the RIL for Control Banks D and C and record in Section II, Attachment 7.1.

6.1.3 Determine if all of the control rods in Control Banks C and D are above the rod insertion limit and circle the appropriate response in Section III, Attachment 7.1.

1. If the response is yes, then N/A Sections IV through XI of Attachment 7.1, and continue with Step 6.1.13.

2. If the response is no then continue with Step 6.1.4 and complete the rest of Attachment 7.1.

6.1.4 Determine current reactor conditions and complete Lines 1 through 4,Section IV on Attachment 7.1 .

6.1.5 Determine the Total Power Defect based on the Latest Available RCS Boron Concentration, Power Level, and exposure recorded in Section IV, Attachment 7.1, using Curve Book, Table 1.3 or Curve 1.3, and record in Section V of Attachment 7.1.

6.1.6 Determine the total integral inserted/misaligned rod(s) worth by recording the inches of the lowest inserted rod in Control Banks D and C and lowest misaligned rod within the bank for each misaligned bank in Section VI, Attachment 7.1, then converting the steps of the insertion/misalignment of each bank into worths by using the Table 1 of Attachment 7.4, and record in Section VI, Attachment 7.1, then totaling up the worths and recording in Section VI, Attachment 7.1.

6.1.7 Determine the Total Rod Worths based on the current cycle exposure recorded in Attachment 7.1,Section IV, Step 1, using Table 2 of Attachment 7.4, and record in Section VII, Attachment 7.1.

6.1.8 Determine the number of inoperable/untrippable control rods.

1. If there are inoperable/untrippable control rods, then calculate the worth by multiplying the number of untrippable/inoperable rods by the worth of the Most Reactive Rod, and record in Section VIII, Attachment 7.1, otherwise N/A.

FMP-012 I Rev. 19 Page 14 of 30

6.1.9 Calculate the available SDM by subtracting the worths of any Inserted/Misaligned Bank(s), Power Defect, any Inoperable/ Untrippable Rod(s), and H-10 rod worth (during Cycle 22 when H-10 ARPI is OOS) from the Total Rod Worths, and record in Section IX, Attachment 7.1.

6.1.10 Determine the required SDM based on the Latest Available RCS boron concentration recorded in Attachment 7.1,Section IV, Step 4, using Figure 5.0 of the COLR (FMP-001), and record in Section X, Attachment 7.1.

6.1.11 Determine if adequate SDM exists by comparing the available SDM calculated in Section IX, Attachment 7.1 to the required SDM determined in Section X, Attachment 7.1.

1. If the available SDM is greater than required SDM, then Adequate SDM exists, circle YES, N/A the rest of Section XI, Attachment 7.1 and proceed to Step 6.1.13.

2. If not, then circle NO and perform Step 6.1.12.

6.1.12 Determine the amount of boron that is needed in order to re-establish the available SDM above the required SDM by subtracting the required SDM determined in Section X, Attachment 7.1 from the available SDM calculated in Section IX, Attachment 7.1 and dividing by the Differential Boron Worth based on the Latest Available RCS Boron Concentration, exposure, and Tavg, and using the Curve Book, Table 1.5 or Curve 1.5.

Record in Section XI, Attachment 7.1.

6.1.13 Have the SSO, or CRSS, or Supervisor- Reactor Systems review and approve ATTACHMENT 7.1.

6.1.14 Send the completed ATTACHMENT 7.1 to the Vault for permanent storage.

FMP-012 I Rev. 19 Page 15 of 30

6.2 Modes 3, 4, 5 6.2.1 Most recent steady state critical conditions:

1. Using the most recent steady state Critical Data Stamp, COMPLETE Lines 1 through 5 on Attachment 7.2.

2. Record the current Cycle Exposure on Line 6 of Section I on ATTACHMENT 7.2.

NOTE: If it is not desired to determine the SDM boron concentration for a particular plant condition, then that section of ATTACHMENT 7.2 may be marked N/A.

6.2.2 Boron Concentration Required to Maintain a Minimum of 1.77% Ak/k Shutdown Margin at 547 0 F:

1. Using Curve 1.11 and/or Table 1.11 of the Station Curve Book, locate the boron concentration corresponding to the current Cycle Exposure and 547 0 F, and record in Step 11.1, ATTACHMENT 7.2.

2. Using the Curve Book, Table 1.5 or Curve 1.5, determine the Differential Boron Worth based on boron concentration determined in Step 11.1, current Cycle Exposure and 547°F, and record in Step 11.2, ATTACHMENT 7.2.

3. IF there are Inoperable / Untrippable control rods, THEN determine Inoperable / Untrippable rod worth in ppm by multiplying the number of Inoperable / Untrippable control rods by the most reactive rod worth and dividing by Differential Boron Worth determined in Step 11.2, and record in Step 11.3, ATTACHMENT 7.2.

4. IF H-1 0 ARPI is declared Out of Service OR IF Shutdown Bank A will be inserted instead of Shutdown Bank B, THEN determine H-10 rod worth in ppm by dividing H-10 rod worth by Differential Boron Worth determined in Step 11.2, and record in Step 11.4, ATTACHMENT 7.2.

5. Calculate the compensated minimum RCS Boron concentration by adding to the minimum RCS Boron Concentration, Inoperable /

Untrippable rod worth from Step 11.3, H-10 rod worth from Step 11.4, and record in Step 11.5, ATTACHMENT 7.2.

FMP-012 Rev. 19 Page 16 of 30

6.2.3 Boron Concentration Required to Maintain a Minimum of 1.77% Ak/k Shutdown Margin at reduced Reactor Coolant System temperatures:

1. Determine the lowest anticipated RCS temperature during the shutdown, and record in Step 111.1, ATTACHMENT 7.2.

2. Using Curve 1.11 and/or Table 1.11 of the Station Curve Book, locate the boron concentration corresponding to the current cycle exposure and the lowest RCS temperature expected during the Shutdown, and record in Step 111.2, ATTACHMENT 7.2.

3. Using the Curve Book, Table 1.5 or Curve 1.5, determine the Differential Boron Worth based on boron concentration determined in Step 111.2, current Cycle Exposure and lowest RCS temperature expected during the Shutdown, and record in Step 111.3, ATTACHMENT 7.2.

4. IF there are Inoperable / Untrippable control rods, THEN determine Inoperable / Untrippable rod worth in ppm by multiplying the number of Inoperable / Untrippable control rods by the most reactive rod worth and dividing by Differential Boron Worth determined in Step 111.3, and record in Step 111.4, ATTACHMENT 7.2.

5. IF H-10 ARPI is declared Out of Service OR IF Shutdown Bank A will be inserted instead of Shutdown Bank B, THEN determine H-10 rod worth in ppm by dividing H-10 rod worth by Differential Boron Worth determined in Step 111.3, and record in Step 111.5, ATTACHMENT 7.2.

6. Calculate the compensated minimum RCS Boron concentration by adding to the minimum RCS Boron Concentration, Inoperable /

Untrippable rod worth from Step 111.4, H-10 rod worth from Step 111.5, and record in Step 111.6, ATTACHMENT 7.2.

FMP-012 Rev. 19 - Page 17 of 3

6.2.4 Boron Concentration Required to Maintain a Minimum of 4% Ak/k Shutdown Margin:

1. Determine the lowest RCS temperature expected to be achieved during the shutdown with < 2 RCPs in operation, and the rod control system capable of withdrawal (the reactor trip breakers closed and lift coil disconnects closed), and record in Step IV.1, ATTACHMENT 7.2.

2. Using Curve 1.14 and/or Table 1.14 of the Station Curve Book, locate the boron concentration corresponding to the current cycle exposure and the lowest RCS temperature anticipated while the reactor trip breakers are closed, and record in Step IV.2, ATTACHMENT 7.2.

3. Using the Curve Book, Table 1.5 or Curve 1.5, determine the Differential Boron Worth based on boron concentration determined in Step IV.2, current Cycle Exposure and, the lowest RCS temperature anticipated while the reactor trip breakers are closed and record in Step IV.3, ATTACHMENT 7.2.

4. IF there are Inoperable / Untrippable control rods, THEN determine Inoperable / Untrippable rod worth in ppm by multiplying the number of Inoperable / Untrippable control rods by the most reactive rod worth and dividing by Differential Boron Worth determined in Step IV.3, and record in Step IV.4, ATTACHMENT 7.2.

5. IF H-1 0 ARPI is declared Out of Service, THEN determine H-1 0 rod worth in ppm by dividing H-1 0 rod worth by Differential Boron Worth determined in Step IV.3, and record in Step IV.5, ATTACHMENT 7.2.

6. Calculate the compensated minimum RCS Boron concentration by adding to the minimum RCS Boron Concentration, Inoperable I Untrippable rod worth from Step IV.4, H-10 rod worth from Step IV.5, and record in Step IV.6, ATTACHMENT 7.2.

6.2.5 Have the SSO, or CRSS, or Supervisor - Reactor Systems review and approve ATTACHMENT 7.2.

6.2.6 Send the completed ATTACHMENT 7.2 to the Vault for permanent storage.

FMP-012 Rev. 19 Page 18of3

6.3 Mode 6 6.3.1 Refer to the current COLR for the appropriate refueling boron concentration, and record in Section I, Attachment 7.3.

6.3.2 Have the SSO, or CRSS, or Supervisor - Reactor Systems review and approve ATTACHMENT 7.3.

6.3.3 Send the completed ATTACHMENT 7.3 to the Vault for permanent storage.

FMP-012 Rev. 19 Page 19 of 3

7.0 ATTACHMENTS 7.1 Shutdown Margin Modes 1,2 Data Form 7.2 Shutdown Margin Boron Concentration Modes 3,4,5 Data Form 7.3 Shutdown Margin Boron Concentration Mode 6 Data Form 7.4 Control Rod Worths FMP-012 Rev. 19 Page20of30

ATTACHMENT 7.1 Page 1 of 3 SHUTDOWN MARGIN MODES 1, 2 DATA FORM I. Current reactor critical conditions:

1. Date/Time conditions recorded /

2. Reactor Power % Full Power

3. Demand D Bank Position steps

4. Demand C Bank Position steps

5. Record RPI indication for Control Bank D and convert to steps below Rod H-04 D-08 H-12 M-08 H-08 Inches Steps (1.6*Inches)

6. Record RPI indication for Control Bank C and convert to steps below Rod K-04 F-04 D-06 D-10 F-12 K-12 M-10 M-06 Inches Steps (1.6*Inches)

II. Based on the Power Level and using Curve Book, Table 1.9 or Curve 1.9, the RIL for Control Bank D is steps Control Bank C is --- steps III. Are the control rods in Control Banks C and D above the RIL, CIRCLE ONE YES Adequate SDM exists and no further verification is warranted, and N/A Sections IV through XI, Attachment 7.1.

NO Further verification of SDM is warranted, complete Sections IV through XI, Attachment 7.1.

FMP-012 Rev. 19 Page 21 of 30

ATTACHMENT 7.1 Page 2 of 3 SHUTDOWN MARGIN MODES 1, 2 DATA FORM IV. Record the following

1. Current Cycle exposure (from Control Room Status Board)

MWD/MTU

2. Reactor Power Level %

3. Tavg degrees F

4. Latest Available RCS Boron Concentration - - ppm Sample Time ___ Date / /

V. Based on the Latest Available RCS Boron Concentration, Power Level, and exposure and using Curve Book, Table 1.3 or Curve 1.3, the Total Power Defect is pcm NOTE: Data entered into the table below will be based on the lowest indicated RPI in the bank. An untrippable rod should not be counted as a misaligned rod.

VI. Determine the RPI position of the lowest rod in control banks D and C and enter into the table below. If a misaligned rod(s) is in CBB, CBA, SBB, SBA, or if the bank(s) are below RIL, then determine the RPI position of the lowest rod(s) within that bank and enter into the table below. Using Table 1, Attachment 7.4, determine the integral bank worth of the inserted/misaligned rod(s) by filling out the table below:

CBD CBC CBB CBA SBB SBA Total Worth Lowest Indicated RPI [Inches]

WORTH [pcm] I I II VII. Based on the current cycle exposure and using the Table 2, Attachment 7.4, the Total Rod Worth is pcm FMP-012 Rev. 19 Page 22of30

ATTACHMENT 7.1 Page 3 of 3 SHUTDOWN MARGIN MODES 1, 2 DATA FORM VIII. Number of inoperable/untrippable control rods If there are inoperable/untrippable rods then calculate the worth by performing the following, otherwise N/A.

1856 pcm = pcm

of rods Most Reactive Inop./Untrip.

Rod Rod Worth IX. The available SDM is calculated by:

- 804 or 0 (NOTE A) "- pcm Total Rod Inserted/ Power Inop/Untrip H-10 Available Worth Misaligned Defect Rod Worth Rod Worth Shutdown Worth Worth Margin (Step VII) (Step VI) (Step V) (Step VIII)

(NOTE: A) Use 804 pcm if H-10 ARPI is OOS, otherwise use 0. Circle the value used in the step above.

X. Based on the Latest Available boron concentration and Figure 5.0, Cycle 22 COLR (FMP-001), the required SDM is _ %

1000 pcm = pcm XI. Is the available SDM greater than the required SDM?

CIRCLE ONE YES Adequate Shutdown Margin Exists NO Adequate Shutdown Margin does not exist; perform the following:

1) Based on the current exposure, Tavg, and latest available Boron Concentration, and using Curve Book, Table 1.5 or Curve 1.5, the Boron Worth is (-) pcm/ppm

2) Borate to restore available SDM. Need to borate at least

( pcm - ___ pcm)/ (-) pcm/ppm = --------- ppm Available SDM Required SDM Boron Worth Amount to borate (Step IX) (Step X) (Step XI.1)

Performed By: Date:

Approved By: Date:

SSO or CRSS or Supervisor - Reactor Systems FMP-012 I Rev. 19 Page 23of30

ATTACHMENT 7.2 Page 1 of 4 SHUTDOWN MARGIN BORON CONCENTRATION MODES 3, 4, 5 DATA FORM Most recent steady state critical conditions:

1. Date/Time conditions recorded /

2. Reactor Power % Full Power

3. D Bank Position steps

4. RCS Boron Concentration ppm

5. Tavg -- OF

6. Cycle Exposure (from Control Room Status Board) ---------- MWD/MTU II. Boron Concentration Required to Maintain a Minimum of 1.77% Ak/k Shutdown Margin at 5470F:

1) Minimum Reactor Coolant System Boron Concentration = ppm

2) Differential Boron Worth (DBW) = (-)p pcm/ppm (Based on RCS Boron Concentration from Step 11.1, Tavg -5477F, and Cycle Exposure from Step 1.6)

IF there are Inoperable / Untrippable rods THEN perform Step 11.3, OTHERWISE N/A

3) Inoperable / Untrippable Rod(s) Worth

(-) 1856 pcm / (-) .... pcm/ppm = - ppm

rods Most Reactive DBW Rod (Step 11.2)

IF ARPI H-10 is OOS OR IF Shutdown Bank A will be inserted instead of Shutdown Bank B THEN perform Step 11.4, OTHERWISE N/A

4) H-10 Rod Worth

(-) 804 pcm / (-). pcm/ppm = - ppm H-10 DBW Rod Worth (Step 11.2)

FMP-012 Rev. 19 1 Page24of30

ATTACHMENT 7.2 Page 2 of 4

5) Compensated Min. RCS Boron Concentration ppm + ppm + ------ ppm = -p Min RCS Inop./Untrip H-1 0 Boron Rod Rod Worth Conc. Worth (Step 11.1) (Step 11.3) (Step 11.4)

IFMP-012 Rev. 19 Page 2of30

ATTACHMENT 7.2 Page 3 of 4 III. Boron Concentration Required to Maintain a Minimum of 1.77% Ak/k Shutdown Margin at reduced Reactor Coolant System temperatures:

1) Lowest Anticipated RCS Temperature = -F

2) Minimum Reactor Coolant System Boron Concentration = ppm

3) Differential Boron Worth (DBW) = (-) ___ pcm/ppm (Based on RCS Boron Concentration from Step 111.2, Tavg from Step 111.1, and Cycle Exposure from Step 1.6)

IF there are Inoperable / Untrippable rods THEN perform Step 111.4, OTHERWISE N/A

4) Inoperable / Untrippable Rod(s) Worth

(-) 1856 pcm / (-) -pcm/ppm = -------- ppm

rods Most Reactive DBW Rod (Step 111.3)

IF ARPI H-10 is OOS OR IF Shutdown Bank A will be inserted instead of Shutdown Bank B THEN perform Step 111.5, OTHERWISE N/A

5) H-10 Rod Worth

(-) 804 pcm / (-)------- pcm/ppm = ppm H-10 DBW Rod Worth (Step 111.3)

6) Compensated Min. RCS Boron Concentration

-ppm + ppm +- -------- ppm =-ppm Min RCS Inop./Untrip H-10 Boron Rod Rod Worth Conc. Worth (Step 111.2) (Step 111.4) (Step 111.5)

I FMP-012 Rev. 19 Page 26 of 30

ATTACHMENT 7.2 Page 4 of 4 IV. Boron Concentration Required to Maintain a Minimum of 4% Ak/k Shutdown Margin:

1) Lowest Anticipated RCS Temperature =F

2) Minimum Reactor Coolant System Boron Concentration = .--- ppm

3) Differential Boron Worth (DBW) = (-).p pcm/ppm (Based on RCS Boron Concentration from Step IV.2, Tavg from Step IV.1, and Cycle Exposure from Step 1.6)

IF there are Inoperable / Untrippable rods THEN perform Step IV.4, OTHERWISE N/A

4) Inoperable / Untrippable Rod(s) Worth

(-) 1856 pcm / (-) - pcm/ppm = ------- ppm

rods Most Reactive DBW Rod (Step IV.3)

IF ARPI H-10 is OOS THEN perform Step IV.5, OTHERWISE N/A

5) H-10 Rod Worth

(-) 804 pcm / (-)_ pcm/ppm = ------- ppm H-10 DBW Rod Worth (Step IV.3)

6) Compensated Min. RCS Boron Concentration

- ppm + -ppm + pppm =- - ppm Min RCS Inop./Untrip H-10 Boron Rod Rod Worth Conc. Worth (Step IV.2) (Step IV.4) (Step IV.5)

Performed By:--------- ------------------------- Date:

Approved By: - Date:

SSO or CRSS or Supervisor - Reactor Systems FMP-012 Rev. 19 Page 27 of 3

ATTACHMENT 7.3 Page 1 of 1 SHUTDOWN MARGIN BORON CONCENTRATION MODE 6 DATA FORM Current Refueling Outage_

Minimum Boron Concentration ppm Performed By: ----------- Date:

Approved By: Date:

SSO or CRSS or Supervisor - Reactor Systems FMP-012 Rev. 19 Page 28 of 30

ATTACHMENT 7.4 Page 1 of 2 CONTROL ROD WORTHS NOTE: This data is only valid for HB Robinson Unit 2, Cycle 22 due to the use of Cycle specific parameters.

TABLE 1 Inserted/Misalianed Bank Worths Steps - Ilnches Control Control ' Control Control ,Shutdown Shutdown Bank D Bank C Bank B BankA Bank B Bank A (pcm) (pcm) (pcm) (pcm) (pcm) (pcm) 225 141 0 0 0 0 0 0 213 133 43 47 30 51 54 47 201 126 165 246 57 228 263 224 189 118 322 479 115 437 501 422 177 111 462 680 167 616 700 586 165 103 569 849 208 767 865 722 153 96 657 987 236 888 997 828 141 88 720 1093 255 979 1102 905 129 81 764 1178 268 1050 1184 962 117 73 791 1243 275 1099 1247 1001 105 66 809 1291 280 1132 1292 1025 93 58 821 1322 282 1154 1321 1040 81 51 827 1341 284 1167 1339 1048 69 43 861 1352 285 1175 1349 1053 57 36 909 1360 285 1228 1356 1056 45 28 940 1364 285 1302 1360 1076 33 21 961 1369 285 1342 1360 1091 21 13 970 1369 285 1360 1366 1097 9 6 974 1369 286 1366 1366 1099 0 0 975 1369 286 1367 1366 1100 FMP-012 Rev. 19 Page 29 of 30

ATTACHMENT 7.4 Page 2 of 2 CONTROL ROD WORTHS NOTE: This data is only valid for HB Robinson Unit 2, Cycle 22 due to the use of Cycle specific parameters.

TABLE 2 Total Rod Worth Exposure Total Rod Worth (MWD/MTU) (pcm) 0 5932 1000 5919 2000 5906 3000 5893 4000 5880 5000 5868 6000 5855 7000 5842 8000 5829 9000 5816 10000 5803 11000 5790 12000 5777 13000 5764 14000 5752 15000 5739 16000 5726 17000 5713 18000 5700 FMP-012 Rev. 19 Page 30of3

Appendix C Page 1 of 6 Form ES-C-1 PERFORMANCE INFORMATION Facility: HB ROBINSON Task No.: 02341101603 Task

Title:

Review (For Approval) A Completed JPM No.: 2004 NRC JPM SRO A3 Temporary Procedure Change Form K/A

Reference:

G2.2.6 3.3 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance:

Classroom X Simulator Plant X READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions:

The unit is in Mode 3.

" EDG Surveillance testing is scheduled.

Task Standard: Identifies all errors and returns Attachment 6 unapproved.

Required Materials: NONE General

References:

PRO-NGGC-0204, Procedure Review and Approval, Revision 5 OST-409-1, EDG "A" Fast Speed Start, Revision 20 Handouts:

PRO-NGGC-0204, Attachment 6, completed to SSO signature

OST-409-1 with "a" added after REVISION and Rev. on the title page and the following added on Page 9 after P&L 5.4: "if the unit is in Mode 1 or Mode 2, as defined in ITS" and "a" added after Rev.

Initiating Cue: You are acting as the Superintendent-Shift Operations. Review the PRO-NGGC-0204, Attachment 6, Temporary Change Form that is being submitted for your approval.

2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 2 of 6 Form ES-C-1 PERFORMANCE INFORMATION Time Critical Task: N/A Validation Time: 10 Minutes 2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 3 of 6 Form ES-C-1 PERFORMANCE INFORMATION (Denote Critical Steps with an asterisk)

PRO-NGGC-0204 Attachment 6, OST-409-1 Handout Examiner's Cue: Provide the Candidate with the completed PRO-NGGC-0204 Attachment 6 and OST-409-1 Handout.

Performance Step: I Reviews handout.

Standard:

Obtains/refers to a copy of PRO-NGGC-0204, Section 9.3.

" Obtains/refers to a copy of OST-409-1.

" He/she may refer to ITS.

Comment:

PRO-NGGC-0204 Attachment 6 Performance Step: 2 Compares PRO-NGGC-0204 Section 9.3 requirements to the completed Attachment 6.

Standard:

Determines that "Temp Chg Expires on (Date)" is incorrect. A date no later than 21 days from the "Interim Approval Date" must be entered.

Examiner's Cue: If necessary: "I will change the date. Review the rest of Attachment 6".

Comment:

PRO-NGGC-0204 Attachment 6 Performance Step: 3 Compares PRO-NGGC-0202 Section 9.3 requirements to the completed Attachment 6.

Standard:

Rejects the change for either or both the following reasons:

" Concurrent testing would make both Diesels inoperable at certain points of the testing.

Concurrent testing would subject the Diesels to a possible "common mode failure".

Comment:

2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 4 of 6 Form ES-C-1 PERFORMANCE INFORMATION Terminating Cue: When the document has been returned, inform the candidate that evaluation on this JPM is complete.

STOP TIME:

2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 5 of 6 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2004 NRC JPM SRO A3 Examinee's Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiner's Signature: Date:

2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 6 of 6 Form ES-C-1 JPM CUE SHEET INITIAL CONDITIONS: 0 The unit is in Mode 3.

9 EDG Surveillance testing is scheduled.

INITIATING CUE: You are acting as the Superintendent-Shift Operations. Review the PRO-NGGC-0204, Attachment 6, Temporary Change Form that is being submitted for your approval.

2004 NRC JPM SRO A3 NUREG 1021, Revision 9, Supplement 1

ATTACHMENT 6 PAGE 1 of 1 TEMPORARY CHANGE FORM FORM PRO-NGGC-0204-6 (4103)

Descriotion Procedure No. Revision No Minor Ch No Temp ChgExpwesn(Dte)

Procedure Title NCR No.

Type of Action (Check Applicable Box) Affected Page Nos.

[E Temp Change Permanent to Follow

[2temp Change No Permanent to Follow Descrption of Procedure Action: " 4, 2 /,,/4:c-7-6,¶ *51 Basis for the Procedure Action:

Originator (Prnt .Dat Name)

Job Supervisor (Print Nkme)4V" Date -

1. 1-/'2-*:,,*m Interim Approval 1st Approver (Print) .provy (Sign) . Date

~ /74 ~A.

J,~

51 70 2nd Approver (SSO) (Print) 2d Approver (SSO) (Sign) Date Tech and REG-NGGC-0010 Reviews Technical Reviewer (Print) Technical Reviewer (Sign) Date Completed REG-NGGC-0010 (Check Applicable Box) Date Completed El Exempt [I Screening E] Evaluation Final Approval [BNP, HNP]

Final Approval Required by (Date) El Approved El Rejected Final Approval (Print) Final Approval (Sign) Date PNSC Chairman (Print) (if applicable) PNSC Chairman (Sign) Date Removal/Early Expiration [ALL]

Early Expiration Date Approval (Sign) Date QA RECORD PRO-NGGC-0204 Rev. 5 Page 42 of 56

P~mgs EneWg H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 3 PART 9 OST-409-1 EDG "A" FAST SPEED START REVISION 20,,

SOST-409-1 I Rev. 20,A Page1 of 65

5.0 PRECAUTIONS AND LIMITATIONS 5.1 This test can be performed during any phase of plant operation.

5.2 CCW Pump "B" will not auto start during Low CCW Header Pressure condition if EDG "A" Output Breaker is shut. Manual Operator action may be required to start a second CCW pump.

5.3 While EDG "A"output breaker is shut, MDAFW Pump "A" is out of service and will not auto start on Lo-Lo SIG level, AMSAC or both MFP breakers open. Manual action will be required to start the affected MDAFW pump.

5.4 Onl one DiTsel shall be tested at a time/ i4 "1e , / /r) /1)Mt .46rdfX 6, -.5 5.5 Maintain a minimum fuel oil storage of 19,000 gallons in the Diesel Fuel Oil Tank plus an additional 15,000 gallons in the I.C. Turbine Fuel Oil Tank.

(ESR 96-00375 & ITS LCO 3.8.3) 5.6 Coolant discharge pressure fluctuations on the jacket water system shall be observed. Fluctuations of greater than 3 psig indicate a possible water leak between the jacket water system and the cylinder liner. A water leak of this type could lead to erosion of both the cylinder liner and pistons.

5.7 The performance of this OST must be coordinated with other plant evolutions such that the minimum equipment operability requirements of TECH SPECS are met.

5.8 When taking the vibration readings, use the magnetic holder instead of the straight probe whenever possible.

5.9 When Raising or Lowering the load on the EDG using the Speed Control Lever on the Generator Panel allow approximately 15 seconds between Raise or Lower actions to enable the Governor time to respond to the new load demand especially when approaching the 2500 KW rating so as not to exceed it.

5.10 Diesel Generator load shall not exceed ratings of 2500 KW for continuous operation.

- Do NOT exceed 2750 KW.

- Do NOT operate at 2750 KW for more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months within a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

- Do NOT exceed 4,000 amps on the Generator.

IOST-409-1 I Rev. 20,a! Page 9 of 65

Appendix C Page 1 of 9 Form ES-C-1 PERFORMANCE INFORMATION Facility: HB ROBINSON Task No.: 02344100403 Task

Title:

Classify an Event/PAR JPM No.: 2004 NRC JPM SRO A5 K/A

Reference:

G2.4.44 4.0 Examinee: NRC Examiner:

Facility Evaluator: Date:

Method of testing:

Simulated Performance: Actual Performance: X Classroom Simulator X Plant READ TO THE EXAMINEE I will explain the initial conditions, which steps to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this Job Performance Measure will be satisfied.

Initial Conditions: 1. The plant tripped when an automatic safety injection occurred.

2. CV pressure relief was in progress. Pressure relief valves failed to close

3. Containment pressure peaked at 27 psig and is now reading 12 psig, slowly lowering.

4. RCS leakage is much greater than charging pump capacity.

5. The steam generators and steam side of the plant are intact.

6. Core Exit Thermocouples are reading approximately 225 0 F.

7. Containment water level is 287 inches.

8. There is a RED PATH on CSF-4 (RCS Integrity) and an ORANGE PATH on CSF-5 (Containment).

9. Bus E-2 de-energized on electrical fault and has not been restored.

10. Bus E-1 is powered from off-site.

11. RHR Pump A did not start.

12. It is August 31. The weather is sunny and clear.

13. Wind is from 140° @ 8 mph.

14. Off-site dose projections are not available at this time.

15. Following is the status of selected Radiation Monitors. R-1 1 and R-12 are aligned to the CV. All are in alarm:

" R 52 R/hr

" R 55 R/hr 2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 2 of 9 FormI ES-C-1 PERFORMANCE INFORMATION

R 3.0 E6 CPM

R 8.5 E4 CPM

R-32A - 48 R/hr

R-32B - 45 R/hr

R-14E - 3.5 E5 CPM Task Standard: Classifies as GE and makes PAR.

Required Materials: Copies of EAL-1/2 EPCLA-01 for Candidate General

References:

EAL-1/2 Revision 12 EPCLA-01 Revision 15 Handouts: EPCLA-01 Initiating Cue: You are the SSO. Classify this event and enter the appropriate procedure.

Time Critical Task: 15 Minutes to Classify Validation Time: 10 Minutes 2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 3 of 9 Form ES-C-1 PERFORMANCE INFORMATION (Denote Critical Steps with an asterisk)

CLASSIFICATION TIME START Performance Step: 1 Implement EAL-1 Flow Chart.

Standard: Enters EAL-1 at "OFF NORMAL CONDITION INDICATED OR OBSERVED".

Comment:

EAL-1 Performance Step: 2 Determines Fuel FPB Status.

Standard: Answers YES based on "R-11 RAD MONITOR >1M CPM" and "R-12 RAD MONITOR > 40K". Marks FUEL BREACHED on FPB Status Board.

Comment:

EAL-1 Performance Step: 3 Determine RCS FPB Status.

Standard: Answers YES based on "RCS LEAKAGE / CHARGING CAPABILITY" initial conditions. Indicates SAE on EAL Status Board and marks RCS BREACHED on FPB Status Board.

Comment:

EAL-1 Performance Step: 4 PRIMARY TO SECONDARY LEAKAGE > TECH SPECS?

Standard: Answers NO, from INITIAL CONDITIONS.

Comment:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 4 of 9 Form ES-C-1 PERFORMANCE INFORMATION EAL-1 Performance Step: 5 Determine CV FPB Status.

Standard: Answers YES based on R-14E reading and CV pressure relief valves open to "PATHWAY EXISTS FROM CV ATMOSPHERE TO ENVIRONMENT". Indicates CV BREACHED on FPB Status Board.

Comment:

EAL-1 Performance Step: 6 3FPB's BREACHED OR JEOPARDIZED?

Standard: Answers YES, declares GE and exits to EPCLA-01.

CLASSIFICATION TIME STOP Comment:

Performance Step: 7 Obtain EPCLA-01.

Examiner's Cue:

Provide a copy of EPCLA-01.

For the purpose of this JPM assume EPCLA-01 has been completed through Step 8.1.3.9. Beginning at Step 8.1.3.10, perform the steps of EPCLA-01.

Standard: Enters and reviews EPCLA-01.

Comment:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 5 of 9 Form ES-C-1 PERFORMANCE INFORMATION EPCLA-01 Step 8.1.3.10 This step and any other announcements can be simulated by pointing out the necessary actuating device(s) and discussing the announcement.

Performance Step: 8 Sound applicable alarms and perform a PA announcement with the "VLC" switch in "Emergency" position:

a. Announce "Attention all personnel, attention all personnel, at (state time of declaration) a(n) (give emergency declared) has been declared."

If Emergency Response Facilities are being activated, then announce:

"All EOF/TSCIOSC and JIC personnel report to your designated facility."

If external hazards require sheltering on site, then announce directions for taking shelter and isolating and/or placing the facility ventilation in the emergency mode.

b. Repeat announcement(s) and alarm (if sounded).

Standard: Sounds alarm and announces GE.

Examiner's Cue:

The EOF/TSC/OSC and JIC personnel have been notified to report.

Assume that on site sheltering actions have been completed.

Comment:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 6 of 9 Form ES-C-1 PERFORMANCE INFORMATION EPCLA-01 Step 8.1.3.11 Examiner's Cue: A site evacuation has already been initiated.

Performance Step: 9 If a Site Area Emergency or General Emergency has been declared, then a site evacuation is mandatory unless doing so will jeopardize the safety of plant personnel. To evacuate the site, sound the site evacuation alarm for approximately 15 seconds, and announce "All Non-Emergency Response personnel report to (give appropriate upwind location) immediately.

- Repeat announcement(s) and alarm (if sounded).

- If a site evacuation has been ordered at an earlier event declaration, it is not necessary to order another site evacuation.

To avoid confusion, a site evacuation should only be initiated once.

Standard: No action required.

Comment:

Performance Step: 10 If a General Emergency has been declared, formulate a Protective Action Recommendation (PAR).

a. Use guidance in Attachments 8.1.5.1, Initial Protective Action Recommendation Flowchart and Attachment 8.1.5.3, PAR Affected Zones Based on Wind Direction to formulate the initial recommendation and zones to be evacuated based on wind direction.

Standard: Completes Attachment 8.1.5.1.

" Evacuate Sectors A-0, A-i, D-1 and E-1.

" Shelter Sectors A-2, B-2, C-2, D-2, E-2, B-i, C-1.

Comment:

Performance Step: 11 Signs as SEC.

Standard: Provides Attachment 8.1.5.1 to Evaluator.

Comment:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 7 of 9 Form ES-C-1 PERFORMANCE INFORMATION Terminating Cue: When Attachment 8.1.5.1 has been completed, inform the Candidate that evaluation on this JPM is complete.

STOP TIME: TIME CRITICAL STOP TIME:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 8 of 9 Form ES-C-1 VERIFICATION OF COMPLETION Job Performance Measure No.: 2004 NRC JPM SRO A5 Examinee's Name:

Date Performed:

Facility Evaluator:

Number of Attempts:

Time to Complete:

Question Documentation:

Question:

Response

Result: SAT UNSAT Examiner's Signature: Date:

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

Appendix C Page 9 of 9 Form ES-C-1 JPM CUE SHEET INITIAL CONDITIONS: 1. The plant tripped when an automatic safety injection occurred.

2. CV pressure relief was in progress. Pressure relief valves failed to close

3. Containment pressure peaked at 27 psig and is now reading 12 psig, slowly lowering.

4. RCS leakage is much greater than charging pump capacity.

5. The steam generators and steam side of the plant are intact.

6. Core Exit Thermocouples are reading approximately 2251F.

7. Containment water level is 287 inches.

8. There is a RED PATH on CSF-4 (RCS Integrity) and an ORANGE PATH on CSF-5 (Containment).

9. Bus E-2 de-energized on electrical fault and has not been restored.

10. Bus E-1 is powered from off-site.

11. RHR Pump A did not start.

12. It is August 31. The weather is sunny and clear.

13. Wind is from 140° @ 8 mph.

14. Off-site dose projections are not available at this time.

15. Following is the status of selected Radiation Monitors. R-11 and R-12 are aligned to the CV. All are in alarm:

R 52 R/hr

R 55 R/hr

R 3.0 E6 CPM

R 8.5 E4 CPM

R-32A - 48 R/hr

R-32B - 45 R/hr

R-14E - 3.5 E5 CPM INITIATING CUE: You are the SSO. Classify this event and enter the appropriate procedure.

2004 NRC JPM SRO A5 NUREG 1021, Revision 9, Supplement 1

j Progress Energy H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 PLANT OPERATING MANUAL VOLUME 2 PART 5 EPCLA-01 EMERGENCY CONTROL REVISION 15 I EPCLA-01 Rev. 15 Page 1 of 21

SUMMARY

OF CHANGES PRR 96981 Step Description of Change 8.1.3.7.d.2 Changed the location of the alternate EOF assembly area from the National Guard Armory to the Darlington County Emergency Operations Center.

EPCLA-01 Rev. 15 Page 2 of 21

TABLE OF CONTENTS SECTION PAGE CR EMERGENCY CONTROL QUICK START GUIDE 4 8 .1.1 P U R P O S E ............................................................................................ .. 5 8.1.2 RESPONSIBILITIES ............................................................................ 5 8.1.3 INSTRUCTIONS ................................................................................... 5 8 .1.4 R E C O R D S .......................................................................................... . . 11 8.1.5 ATTACHMENTS ................................................................................. 12 8.1.5.1 Initial Protective Action Recommendation Flowchart ................ 13 8.1.5.2 EPA Protective Action Guide (PAGs) for the Early Phase ...... 15 8.1.5.3 PAR Affected Zones Based on Wind Direction ......................... 16 8.1.5.4 Turnover Checklist .................................................................... 17 8.1.5.5 Plant-based Protective Action Recommendations .................... 20 EPCLA-01 Rev. 15 Page 3 of 21

CR EMERGENCY CONTROL QUICK START GUIDE NOTE: This is a summary level guide and does not replace the procedure steps. EPCLA-01 is to be used with this guide.

1. Implement EALs as necessary. It is the expectation that the time between exceeding an EAL and declaration of event will not exceed 15 minutes unless extraordinary conditions prevail. Annotate time of the off normal condition on the top of the EAL board. Continue through the flowpath until a General Emergency has been identified or until the end of the flowpath.

2. Direct an Emergency Communicator to report to the Control Room at this time.

This will support communication activities and augmentation of the ERO.

3. The EAL board will direct you to EPCLA-01, "Emergency Control" or to AP-030 if there is no event classification. EPCLA-01,Section 8.1.3 provides guidance for classifying emergencies and control.

4. Declare the highest event classification identified by announcing the event to the Control Room and your assuming role as the SEC. This ends the 15 minute clock for the event declaration, and starts the 15 minute clock to notify the appropriate State and County agencies. Announce classification to the Site per EPCLA-01.

NOTE: The development of the Emergency Notification Form should include the status of the fission product barriers.

5. Develop, approve, and FAX/communicate the Emergency Notification Form.

Notify State and County agencies via Selective Signaling System or an alternate means. The notification clock stops after the first voice contact is established with an approved form. This is the time entered on Attachment 8.1.5.1 of EPNOT-01 page 2 of 7.

6. Fill out the Emergency Notification Form. Detailed instructions are in EPNOT-01, Attachment 8.1.5.1, page 3 through 7.

Click on "Emergency Preparedness Function Menu.

Click on "Log into Network data Base and log in as CRSS.

Click on Declare Event, then OK.

At top of screen type ER to bring up Environmental Data and print screen.

Click on Event Notification Form (ENF).

Click "ADD" on ENF.

7. Assess EALs for changing plant conditions. Attachment 8.1.5.4 in EPCLA-01 contains the checklist for turnover to the TSC.

EPCLA-01 Rev. 15 Page4of 21

8.1.1 PURPOSE

1. To provide consolidated guidance for classifying emergencies from the Control Room or Technical Support Center (TSC).

8.1.2 RESPONSIBILITIES

1. The Site Emergency Coordinator (SEC) has immediate and unilateral authority to implement this procedure.

2. The SEC may not delegate:

a. The decision to notify offsite authorities;

b. Making offsite Protective Action Recommendations (PAR);

and

c. Reclassifying or terminating the emergency.

3. The responsibility to notify offsite authorities and making offsite Protective Action Recommendations transfer to the Emergency Response Manager (ERM) upon activation of the Emergency Operations Facility (EOF).

4. The SEC may authorize exposure in excess of routine yearly limits for saving of life or protecting valuable equipment per EPOSC-04, Emergency Work Control.

8.1.3 INSTRUCTIONS

1. Enter the Emergency Action Level (EAL) flowpath, EAL-1, at the first step and determine the appropriate classification.

2. Declare or validate the highest classification of emergency determined.

a. Announce to Control Room or TSC personnel that you are assuming the position of SEC.

EPCLA-01 Rev. 15 1 Page5of 21

8.1.3 (Continued)

3. Direct the Emergency Communicator to prepare for communication activities in accordance with EPNOT-01, CR/EOF Emergency Communicator.

4. Determine if there are any personnel injuries;

a. Give priority to lifesaving activities over radiological exposure control; authorize exposures in excess of normal limits if required.

b. Refer to EPSPA-02, First Aid and Medical Care, for additional guidance on first aid and transportation of contaminated injured personnel.

5. Determine if onsite protective actions are necessary;

a. Evaluate radiological, chemical and other situations which may require evacuation or sheltering.

b. If evacuation or administration of potassium iodide is necessary, implement EPSPA-01, Evacuation and Accountability, or EPSPA-03, Administration of Potassium Iodide, respectively. If sheltering is required onsite (such as for external gas hazard); Make a plant announcement directing personnel to shelter in the nearest facility. Ensure ventilation is isolated/secured in the OSC and other facilities/buildings that are not equipped with emergency/re-circulation modes (Control Room, TSC/EOF). (AR #57330)

c. Evaluate possible severe weather protective actions.

(CR 22292)

6. Request any offsite assistance necessary;

a. The Unit 2 Control Room should contact Darlington County 911 Center for fire, police or ambulance service.

b. Logistics personnel may contact the 911 Center if Control Room staff are unable to request assistance.

EPCLA-01 Rev. 15 Page 6 of 21

8.1.3.6 (Continued)

c. Contact other agencies as necessary, selected offsite agency numbers are maintained in the Emergency Response Organization (ERO) phone book.

7. Activate appropriate Emergency Response Facilities (ERFs) as noted below:

a. IF all of the following occurs;

- The Start-up Transformer is lost.

- Backfeed through the Auxiliary Transformer is possible.

- Only 1 (one) Emergency Diesel is powering its respective bus.

THEN staff all of the onsite Emergency Response Facilities to assist with back feed logistics.

b. For Unusual Event - no activation is required; facilities may be activated at SEC discretion.

c. For Alert or above activate TSC, EOF, OSC and JIC.

d. Consider the following when choosing facilities to activate.

1. Alternate TSC is Control Room

2. Alternate EOF Assembly Area is the Darlington County Emergency Operations Center, 1625 Harry Byrd Highway (Highway 151), Darlington, SC.

3. Alternate/Back-up OSC is as defined in EPOSC-01.

4. Remote Facility may be activated for any event, normally for Security Events where reporting to the site may not be safe for the ERO.

8. Determine habitability of facilities for directing ERO personnel to the primary or alternate location via PA, pager code, etc.

9. For an Alert only, if the casualty has abated prior to or during notification of offsite agencies, ERO pagers and facilities need not be activated.

a. If no facility activation is desired, modify the upcoming Public Address (PA) announcement with DO NOT activate the Emergency Response Facilities.

IEPCLA-01 Rev. 15 Page7of 217

8.1.3 (Continued)

10. Sound applicable alarms and perform a PA announcement with the "VLC" switch in "Emergency" position;

a. Announce "Attention all personnel, attention all personnel, at (state time of declaration) a(n) (give emergency declared) has been declared."

NOTE: Discretion should be exercised when announcing the cause of the emergency due to a security event.

The cause of the emergency is If Emergency Response Facilities are being activated, then announce:

"All EOFITSC/OSC and JIC personnel report to your designated facility."

If external hazards require sheltering on site, then announce directions for taking shelter and isolating and/or placing the facility ventilation in the emergency mode.

b. Repeat announcement(s) and alarm (if sounded).

11. If a Site Area Emergency or General Emergency has been declared, then a site evacuation is mandatory unless doing so will jeopardize the safety of plant personnel. To evacuate the site, sound the site evacuation alarm for approximately 15 seconds, and announce "All Non-Emergency Response personnel report to (give appropriate upwind location) immediately.

- Repeat announcement(s) and alarm (if sounded).

- If a site evacuation has been ordered at an earlier event declaration, it is not necessary to order another site evacuation. To avoid confusion, a site evacuation should only be initiated once.

EPCLA-01 Rev. 15 Page 8 of 21

8.1.3.11 (Continued)

Designated locations are: (others may be used if necessary)

East - Building 110 next to Lake Robinson or parking lot.

West - Unit 2 Administrative Building Cafeteria or parking lot.

12. If a General Emergency has been declared, formulate a protective Action Recommendation (PAR).

a. Use guidance in Attachments 8.1.5.1, Initial Protective Action Recommendation Flowchart and Attachment 8.1.5.3, PAR Affected Zones Based on Wind Direction to formulate the initial recommendation and zones to be evacuated based on wind direction.

b. Use guidance in Attachment 8.1.5.5, Plant -Based Protective Action Recommendations, to recommend extended protective action recommendations based on plant conditions.

c. Subsequent PARs are made by comparing dose projections and environmental monitoring results to Attachment 8.1.5.2, Protective Action Guidelines (PAG) and upgrading the initial recommendations as necessary.

d. If conditions indicate the PAR needs upgrading, then the 15 minute notification standard applies as this will be a new initial message.

13. Develop and transmit an initial Emergency Notification Form to at least one State and County agency within 15 minutes of emergency declaration.

a. Follow up notifications are required at least every 30-60 minutes.

14. Within one hour of an Alert (or above) declaration, activate the Emergency Response Data System (ERDS) as noted below:

a. If the ERDS is not currently operational (ERDS = NORMAL is not displayed at the bottom of an ERFIS terminal), the SEC will ensure that ERDS is activated. Any problems should be reported to Information Technology personnel.

EPCLA-01 Rev. 15 Page 9 of 21

8.1.3 14 (Continued)

b. Display the ERDS activation screen by:

- Depressing the ERDS key on the ERFIS keyboard, or

- Typing the Turn-On-Code "ERDS" at the input field, or

- Selecting ERDS from the EP Menu.

c. When the ERDS Control and Status Display window appears, click on the green "Start ERDS" button.

- An "Are You Sure" message is displayed. Click yes to initiate ERDS, click no to cancel.

- Observe the "Start ERDS" button changes to a yellow "Starting..." button.

- When ERDS connects to the NRC Operations Center the yellow "Starting..." button will change to a red "Stop ERDS" button.

- Other buttons are provided to review system status and data transmissions.

- It may take several minutes for the system status in the Control and Status Display window or at the bottom of the screen to update.

d. Within five minutes after activation, the ERDS function should become operational. This is determined by ERDS -

NORMAL message displayed at the bottom of an ERFIS terminal.

e. If ERDS fails to become operational (ERDS = NORMAL is not displayed on an ERFIS Terminal) within five minutes, stop the ERDS function by clicking the red "Stop ERDS" button and notify onsite Information Technology.

EPCLA-01 Rev. 15 Page 10 of 21

8.1.3 (Continued)

15. If the Emergency Response Facility Information System/Electronic Display System (ERFIS/EDS) is out of service initiate manual transfer of safety parameter and other relevant data.

a. Forms for recording data are located in EPNOT-01, "Notification and Emergency Communications.

16. Continue to assess the plant status against the EALs to confirm, upgrade or downgrade the emergency classification.

a. If the State and County facilities have been activated, they should be consulted prior to any downgrade of emergency classification.

17. If the TSC is activating, perform a turnover with the TSC SEC.

a. A turnover checklist is provided as Attachment 8.1.5.4, Turnover Checklist.

18. Perform PA announcements periodically to update personnel in the field of any changing plant conditions.

19. When appropriate based on plant conditions, coordinate with any offsite agencies which have activated and terminate the emergency.

a. Direct the Emergency Communicator to make termination notifications to all agencies.

- Termination, as a change in classification, has a 15 minute time requirement.

b. If not previously terminated by the Nuclear Regulatory Commission (NRC), coordinate the termination of ERDS.

8.1.4 RECORDS N/A EPCLA-01 Rev. 15 Page 11 of 21

8.1.5 ATTACHMENTS 8.1.5.1 Initial Protective Action Recommendation Flowchart 8.1.5.2 EPA Protective Action Guide (PAGs) for the Early Phase 8.1.5.3 PAR Affected Zones Based on Wind Direction 8.1.5.4 Turnover Checklist 8.1.5.5 Plant-Based Protective Action Recommendations EPCLA-01 Rev. 15 1 Page 12 of 211

ATTACHMENT 8.1.5.1 Page 1 of 2 INITIAL PROTECTIVE ACTION RECOMMENDATION FLOWCHART EVACUATE 2 MILE RADIUS AND 5 MILES DOWNWIND.

SHELTER ALL REMAINING SECTORS IN 10 MILE RADIUS EVALUATE DOSE ASSESSMENTS AGAINST PAGS TO DETERMINE ADDITIONAL SECTORS TO EVACUATE EPCLA-01 Rev. 15 Page 13 of 21

ATTACHMENT 8.1.5.1 Page 2 of 2 INITIAL PROTECTIVE ACTION RECOMMENDATION FLOWCHART PAR REFERENCE GUIDE AND DOCUMENTATION FORM RULES FOR PROTECTIVE ACTION RECOMMENDATIONS

1. SHELTER ALL REMAINING SECTORS IN THE 10 MILE RADIUS NOT EVACUATED.

2. A PROTECTIVE ACTION RECOMMENDATION MAY NOT BE REDUCED FROM THE INITIAL RECOMMENDATION FOR ANY SECTOR UNTIL THE RELEASE IS TERMINATED AND THE DECISION IS COORDINATED WITH THE STATE AND COUNTIES.

3. A PROTECTIVE ACTION REQUIRED FOR ANY PORTION OF A SECTOR REQUIRES THAT ACTION BE IMPLEMENTED FOR THE ENTIRE SECTOR.

RECOMMENDATION PLACE A V IN THE APPROPRIATE BLANK FOR EACH SECTOR.

2 MILE RADIUS------

10 MILE RADIUS------ -------

EVACUATE SHELTER SECTOR EVACUATE SHELTER SECTOR A-0 -- -A-2 5 MILE RADIUS------

B-2

-- -A-1 C-2 B-1 D-2 C-1 E-2 D-1 E-1 RECOMMENDED BY /DATE@TIME:

RCD OR RCM APPROVED BY /DATE@TIME:

SEC OR ERM EPCLA-01 Rev. 15 Page 14 of 21

ATTACHMENT 8.1.5.2 Page 1 of 1 EPA PROTECTIVE ACTION GUIDE (PAGS)

FOR THE EARLY PHASE*

PROTECTIVE PAG COMMENTS ACTION Evacuate 1 Rem TEDE Change any sheltering subzones/sectors to evacuate if the Total Effective Dose Equivalent dose within any area exceeds PAG.

Evacuate 5 Rem CDE Change any sheltering subzones/sectors to evacuate if the Committed Dose Equivalent dose to the thyroid within any area exceeds PAG.

The Early Phase is the time between the beginning of an incident and when the incident source and releases have been brought under control.

Reference:

EPA 400-R-92-001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents," U.S. Environmental Protection Agency, Washington, D.C., May 1992 EPCLA-01 Rev. 15 Page15of 21

ATTACHMENT 8.1.5.3 Page 1 of 1 PAR AFFECTED ZONES BASED ON WIND DIRECTION (EVACUATION TIME IN MINUTES)'

WINTER WINTER SUMMER WINTER WEEKDAY, WEEKNIGHT, WEEKDAY, WEEKDAY, POTENTIALLY1 FAIR FAIR FAIR ADVERSE WIND FROM AFFECTED SECTORS WEATHER WEATHER WEATHER WEATHER North A-0, B-i, B-2, C-1, C-2, 225 180 210 295 (3380 - 022-) D-i, D-2 Northeast A-0, C-1, C-2, D-i, D-2, 225 180 210 295 (0230 - 067-) E-1, E-2 East A-0, D-i, D-2, E-1, E-2 225 180 210 295 (068°- -112°)

Southeast A-0, A-i, A-2, D-1, E-i, 225 180 210 295 (113°- -157-) E-2 South A-0, A-i, A-2, B-i, B-2, 225 180 210 295 (158°- -202°) E-1, E-2 Southwest A-0, A-i, A-2, B-i, B-2, 225 180 210 295 (2030 - 247-) E-i, E-2 West A-0,A-1, A-2, B-i, B-2, 225 180 210 295 (2480 - 292-) C-1, C-2 Northwest A-0, B-i, B-2, C-1, C-2, 225 180 210 295 (2930 - 337-) D-2 ALL ZONES 240 180 215 315 (10 MILE RADIUS)

1. Minimum recommendation for General Emergency is A-0 (2 mile radius) and affected (downwind) 5 mile radius sectors. Shelter all remaining sectors in the 10 mile radius.

2. Times listed are estimates based on evacuation times listed in the Emergency Plan.

NOTE: Conditions identified represent most limiting conditions.

EPCLA-01 Rev. 15 Page 16 of 21

ATTACHMENT 8.1.5.4 Page 1 of 3 TURNOVER CHECKLIST This checklist is guidance for turning over Emergency Response activities from one facility to another or between personnel holding Emergency Response positions.

NOTE: Blanks are provided for place keeping V's only, logs are the official record.

A. ONSITE SITUATION

1. Review Emergency Classification, basis for declaration, and mitigating actions. Suspend turnover if plant conditions exist that change the classification, notification, or PARs.

a. Review status of safety equipment and systems.

b. Review status of fission product barriers.

c. Review condition/stability of reactor.

d. Review any Emergency Action Levels exceeded.

e. Review cause, history, initiating events leading to declaration of emergency.

2. Review onsite protective actions taken.

a. Assembly

b. Shelter

c. Evacuations (Local, Protected Area, Site, Exclusion Area)

NOTE: If there is a Site Evacuation, Unit 1 may need to continue operating.

d. Potassium Iodide Administration

e. Complete PLP-015 Overtime Form for ERO as appropriate.

EPCLA-01 Rev. 15 Page 17of 21

ATTACHMENT 8.1.5.4 Page 2 of 3 TURNOVER CHECKLIST

3. Review status of offsite assistance requested for the site.

a. Fire Department

b. Rescue Squad

c. Local Law Enforcement Agency B. OFFSITE SITUATION

1. Review Status of Offsite Notifications.

- State and County initial and any follow-up messages

- NRC (including status of ERDS activation)

- Other: ANI, INPO, Westinghouse

- Any needed notifications that have not been made

2. Review Protective Action Recommendations made and notifications made to the State and Counties.

3. Review any status received from the State or Counties regarding activation, readiness, protective actions, or requests for information.

4. Review data on any projected or actual radiological releases.

5. Review the time and content of any press releases or media briefing.

EPCLA-01 Rev. 15 Page 18 of 21

ATTACHMENT 8.1.5.4 Page 3 of 3 TURNOVER CHECKLIST C. EMERGENCY RESPONSE

1. Review status of Emergency Response Organization Activation.

- Notifications made to off-duty and offsite personnel.

- Emergency Response Facilities that are activated.

- Emergency Response Facilities that will be activated.

- Other notifications needed.

2. Review outside organizations requested to mobilize.

3. Review assistance needed.

4. After the TSC-SEC assumes responsibilities for the event declaration, the CR-SEC maintains responsibility to keep the TSC updated of changing conditions and the urgency of declaring events based on the changing conditions.

D. TURNOVER COMPLETED EPCLA-01 Rev. 15 Page 19 of 21

ATTACHMENT 8.1.5.5 Page 1 of 2 Plant-Based Protective Action Recommendations General Emergency No Declared? No PAR S Required I

SYes U Condition 1 No Substantial Core Damage is imminent or has occurred.

{ Yes Condition 2 No A significant loss of reactor coolant is imminent or has occurred.

Yes M Condition 3 No Containment Failure (Primary or S/G) is imminent or has occurred.

Evacuate 5 Mile Radius 8 Evacuate 2 Mile Radius &

10 Miles Downwind. 5 miles downwind.

Shelter all Remaining Shelter all Remaining Sectors. Sectors.

I EPCLA-01 Rev. 15 Page 20 of 21

ATTACHMENT 8.1.5.5 Page 2 of 2 Plant-Based Protective Action Recommendations

1. Substantial core damage is imminent or has occurred. Indications that substantial core damage is imminent or has occurred include:

a. Core damage estimates greater than 1 % Melt.

b. Core Exit Thermocouples readings > 2300 degrees FO.

c. Core uncovered > 30 minutes.

2. A significant loss of reactor coolant is imminent or has occurred. Indications that a significant loss of reactor coolant is imminent or has occurred include:

a. Containment Radiation Monitors reading >10,000 R/hr with no containment spray or >4,000 R/hr with containment spray on.

b. Containment hydrogen gas concentration >1%.

c. Rapid vessel depressurization.

d. A large break loss of coolant accident.

3. Containment failure (primary or S/G) is imminent or has occurred. Indications that containment failure (primary or S/G) is imminent or has occurred include:

a. A release of radioactivity cannot be maintained below General Emergency EAL criteria.

b. Primary Containment pressure cannot be maintained below the design basis pressure of 42 psig.

c. Primary containment H2 gas concentration cannot be maintained below combustible limits of 4% by volume.

d. Faulted/ruptured steam generator with a relief valve open.

4. Accidents which result in a direct release pathway to the environment (for example, a faulted and ruptured S/G with water level below the tube bundles, S/G Narrow Range < 25% normal containment conditions or < 40% adverse containment conditions, and a relief valve open would provide such a pathway) will most likely be thyroid dose limiting. For circumstances involving this type of accident sequence:

a. Consider any Fuel Breach sufficient to warrant the determination that substantial core damage has occurred.

b. Consider any RCS Breach sufficient to warrant the determination that a significant loss of reactor coolant has occurred.

Containment monitors can provide indication of both core damage and RCS breach.

Monitor values used to determine a specific amount of core damage are dependent on plant conditions, power history and time after shutdown. Monitor readings used to quantify an amount of damage or coolant leakage should be complimented by other indications and engineering judgment.

EPCLA-01 Rev. 15 Page21of 21

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