ML15119A497

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Initial Exam 2015-301 Draft Administrative JPMs
ML15119A497
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 04/29/2015
From:
Division of Reactor Safety II
To:
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Download: ML15119A497 (235)


Text

u J Page 1 of 10 JPM TITLE: Blend to the RWST JPM NUMBER: 01046006100 REV. 1-0 TASK NUMBER(S) I 01046006100/

TASK TITLE(S): Blend to the RWST KIA NUMBERS: 2.1.29 KIA VALUE: RO 4.1ISRO4.0 Justification (FOR KIA VALUES <3.0): N/A TASK APPLICABILITY:

~RO ~SRO ~ STA 0 Non-Lie ~SRO CERT 0 OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: D Perform: [U EVALUATION LOCATION: In-Plant D Control Room: D Simulator: D Classroom: x I Lab: Other: I Time for Completion: 15 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No Date Date Training Program Owner Date TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/RO

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 2 of 10 JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 3 of 10 UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE Updated to fleet template; Updated for L-15-1 See cover page N/A 1-0 01982463 text/grammar changes NRC Exam See cover page N/A 1-1 1-2 1-3 1-4 1-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 4 of 10 SIMULATOR SET-UP: N/A Required Materials: 0-OP-046 (CVCS - Boron Concentration Control), upfront material and Section 7.5 (CVCS Manual Makeup to the RWST) - marked-up copy Unit 3 Plant Curve Book, Section 3 Calculator General

References:

0-OP-046, CVCS - Boron Concentration Control Unit 3 Plant Curve Book, Section 3 Task Standards: Demonstrate knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

For the given conditions, determine the required boric acid flow rate (38.7 gpm), boric acid volume (1,745 gallons), primary water volume (2,255 gallons), Boric Acid Flow Controller setting (7.74), and Primary Water Flow Controller setting (3.33)

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 5 of 10 I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

Annunciator G-8/2 (RWST TECH SPEC MIN LEVEL) is in alarm.

A manual makeup to the RWST is to be performed, to clear the low-level alarm and maintain the existing RWST concentration.

An adequate supply of boric acid and primary water are available, and plant conditions will allow system alignment, to support this evolution.

Boron concentrations in the BAST and RWST are 5,675 ppm and 2,475 ppm, respectively.

The desired primary-water flow rate is 50 gpm and the total volume to be added is 4,000 gallons.

All relevant prerequisites, precautions/limitations, and associated attachments in 0-OP-046 (CVCS - Boron Concentration Control) have been addressed.

Section 7.5 (CVCS Manual Makeup to the RWST) of 0-OP-046 has been completed through Step 7.5.2.1.

Initiating Cue:

You are directed to use Method 2 (Calculation) from Section 3 of the Plant Curve Book, to calculate the following parameters:

o Boric-acid flow rate: __________________ (to the nearest tenth of a gpm) o Boric-acid volume: __________________ (to the nearest gallon) o Primary-water volume: __________________ (to the nearest gallon)

Based on the available information, determine the potentiometer settings for the following controllers:

o Boric Acid Flow Controller (FC-3-113A): __________________

o Primary Water Flow Controller (FC-3-114A): __________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 6 of 10 JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

Provide examinee with a marked-up copy of 0-OP-046 (CVCS - Boron Evaluator Cue: Concentration Control), upfront material and Section 7.5 (CVCS Manual Makeup to the RWST).

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 7 of 10 Determine the appropriate boric acid and primary water flows and Performance Step: 2 volumes needed to obtain the desired blend concentration from the Critical: Yes boron change tables in Section 3 of the Plant Curve Book.

Obtain required materials.

Determine boric-acid flow rate, boric-acid volume, and primary-water volume, and record values on Turnover Sheet.

Standard:

o Boric-acid flow rate: 38.7 gpm (38 to 39 gpm) o Boric-acid volume: 1,745 gallons (1,727 to 1,753 gallons) o Primary-water volume: 2,255 gallons (2,247 to 2,273 gallons)

Primary-water flow rate (50 gpm) and total volume (4,000 gallons) were provided in the Initiating Cue.

From Section 3 of Plant Curve Book:

Boronppm = (Acidgpm)(BASTppm)/(Acidgpm + Watergpm) 2,475 = (Acidgpm)(5,675)/(Acidgpm + 50)

Acidgpm = 38.7 gpm For exact combined flow of 88.7 gpm (i.e., 50 gpm + 38.7 gpm):

Evaluator Note: Boric-acid volume = (4,000 gallons)(38.7/88.7) = 1,745 gallons Primary-water volume = (4,000 gallons)(50/88.7) = 2,255 gallons For rounded-down combined flow of 88 gpm (i.e., 50 gpm + 38 gpm):

Boric-acid volume = (4,000 gallons)(38/88) = 1,727 gallons Primary-water volume = (4,000 gallons)(50/88) = 2,273 gallons For rounded-up combined flow of 89 gpm (i.e., 50 gpm + 39 gpm):

Boric-acid volume = (4,000 gallons)(39/89) = 1,753 gallons Primary-water volume = (4,000 gallons)(50/89) = 2,247 gallons Evaluator Cue: Provide examinee with a copy of the Unit 3 Plant Curve Book, Section 3.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 8 of 10 Based on the available information, determine the potentiometer settings for the following controllers:

Performance Step: 3 Critical: Yes Boric Acid Flow Controller (FC-3-113A)

Primary Water Flow Controller (FC-3-114A)

Determine potentiometer settings:

Standard: Boric Acid Flow Controller (FC-3-113A): 7.74 (7.60 to 7.80)

Primary Water Flow Controller (FC-3-114A): 3.33 (3.32 to 3.34)

From Step 4.23 of 0-OP-046 (ratio of 5 gpm to 1), a boric-acid flow rate of 38.7 gpm/38 gpm/39 gpm is equivalent to a controller setting of 7.74/7.60/7.80 on the ten-turn potentiometer.

Evaluator Note:

From Step 4.24 of 0-OP-046 (ratio of 15 gpm to 1), a primary-water flow rate of 50 gpm is equivalent to a controller setting of 3.33 on the ten-turn potentiometer.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 3, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

01046006100, Blend to the RWST, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 9 of 10 Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

Annunciator G-8/2 (RWST TECH SPEC MIN LEVEL) is in alarm.

A manual makeup to the RWST is to be performed, to clear the low-level alarm and maintain the existing RWST concentration.

An adequate supply of boric acid and primary water are available, and plant conditions will allow system alignment, to support this evolution.

Boron concentrations in the BAST and RWST are 5,675 ppm and 2,475 ppm, respectively.

The desired primary-water flow rate is 50 gpm and the total volume to be added is 4,000 gallons.

All relevant prerequisites, precautions/limitations, and associated attachments in 0-OP-046 (CVCS - Boron Concentration Control) have been addressed.

Section 7.5 (CVCS Manual Makeup to the RWST) of 0-OP-046 has been completed through Step 7.5.2.1.

Initiating Cue:

You are directed to use Method 2 (Calculation) from Section 3 of the Plant Curve Book, to calculate the following parameters:

o Boric-acid flow rate: __________________ (to the nearest tenth of a gpm) o Boric-acid volume: __________________ (to the nearest gallon) o Primary-water volume: __________________ (to the nearest gallon)

Based on the available information, determine the potentiometer settings for the following controllers:

o Boric Acid Flow Controller (FC-3-113A): __________________

o Primary Water Flow Controller (FC-3-114A): __________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

Power Light Comp Turkey Point Nuclear Plant O-OP-046 CAUTION Performance of this procedure may affect core reactivity.

Title:

CVCS - Boron Concentration Control Continuous Use Safety Related Procedure Responsible Department: Operations Revision Number: 15 Issue Date: 10/16/14 Revision Approval Date: 10/16/14 PCRs 08-1698, 08-3495, 08-5850, 08-5630, 09-0019, 09-1830, 09-2712, 09-2070, 08-4893, 1614987, 1624961, 1634654, 1617791, 1672064, 1691756, 1691803, 1729195, 1729601, 1742261, 1778197, 1609494, 1667204, 1648086, 1793596, 1854397, 1863967, 1717296, 1921045, 191787~ 1958264, 1976490, 1995673 PC/Ms87-257, 87-258,89-494, 90-440,91-068, 90-423,90-424, 91-092,94-141, 95-040,95-102, 95-140,95-172, 95-081,00-016, 10-010,09-137, 09-139 ECs 242547, 247006, 247008, 249292 W201 O:TN M/fm/cls/cls

Procedure No.: Page:

Procedure

Title:

2 Approval Date:

O-OP-046 10/16/14 CVCS - Boron Concentration Control LIST OF EFFECTIVE PAGES (Rev. 15)

Revision Revision Revision Revision Page Date Page Date Page Date Page Date 1 10/16/14 31 02126114 61 10/15/13 91 10/15/13 2 10/16/14 32 03/08/13 62 10/15/13 92 10/15/13 3 10/16/14 33 03/08/13 63 10/15/13 93 10/15/13 4 10/16/14 34 03/08/13 64 02/26/14 94 10/15/13 5 10/16/14 35 03/08/13 65 10/15/13 95 10/15/13 6 10/16/14 36 03/08/13 66 10/15/13 96 10/15/13 7 03/08/13 37 03/08/13 67 10/15/13 97 10/15/13 8 03/08/13 38 03/08/13 68 10/15/13 98 10/15/13 9 03/08/13 39 03/08/13 69 10/15/13 99 10/15/13 10 05/06/13 40 10/15/13 70 10/15/13 100 10/15/13 11 03/08/13 41 03/08/13 71 10/15/13 101 10/15/13 12 03/08/13 42 03/08/13 72 10/15/13 102 10/15/13 13 03/08/13 43 10/15/13 73 10/15/13 103 10/15/13 14 03/08/13 44 10/15/13 74 10/15/13 104 10/15/13 15 03/08/13 45 10/15/13 75 10/15/13 105 10/15/13 16 03/08/13 46 10/15/13 76 10/15/13 106 10/15/13 17 02/26/14 47 10/15/13 77 10/15/13 107 10/15/13 18 02/26/14 48 10/15/13 78 10/15/13 108 10/15/13 19 03/08/13 49 05/05/14 79 10/15/13 109 10/15/13 20 03/08/13 50 05/05/14 80 10/15/13 110 10/15/13 21 02/26/14 51 05/05/14 81 10/15/13 111 10/15/13 22 02/26/14 52 05/05/14 82 10/15/13 112 10/15/13 23 02/26/14 53 10/15/13 83 10/15/13 113 10/15/13 24 02/26/14 54 10/15/13 84 10/15/13 114 10/15/13 25 02/26/14 55 10/15/13 85 10/15/13 115 10/15/13 26 03/08/13 56 10/15/13 86 10/15/13 116 10/15/13 27 03/08/13 57 10/15/13 87 10/15/13 117 10/15/13 28 02/26/14 58 10/15/13 88 02/26/14 118 10/15/13 29 02/26/14 59 10/15/13 89 10/15/13 119 10/15/13 30 02/26/14 60 10/15/13 90 10/15/13 120 10/15/13

Procedure No.: Page:

Procedure

Title:

3 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/16/14 LIST OF EFFECTIVE PAGES (Cont'd)

(Rev. 15)

Revision Revision Revision Page Date Page Date Page Date 121 10/15/13 151 10/15/13 181 10/16/14 122 10/15/13 152 10/15/13 182 10/16/14 123 10/15/13 153 10/15/13 183 10/16/14 124 10/15/13 154 10/15/13 184 10/16/14 125 10/15/13 155 10/15/13 185 10/16/14 126 10/15/13 156 07/08/14 186 10/16/14 127 10/15/13 157 07/08/14 187 10/16/14 128 10/15/13 158 07/08/14 188 10/16/14 129 10/15/13 159 10/16/14 189 10/16/14 130 10/15/13 160 10/16/14 190 10/16/14 131 10/15/13 161 10/16/14 191 10/16/14 132 10/15/13 162 10/16/14 192 10/16/14 133 10/15/13 163 10/16/14 193 10/16/14 134 10/15/13 164 10/16/14 194 10/16/14 135 10/15/13 165 10/16/14 195 10/16/14 136 10/15/13 166 10/16/14 196 10/16/14 137 10/15/13 167 10/16/14 197 10/16/14 138 10/15/13 168 10/16/14 139 10/15/13 169 10/16/14 140 10/15/13 170 10/16/14 141 10/15/13 171 10/16/14 142 10/15/13 172 10/16/14 143 10/15/13 173 10/16/14 144 10/15/13 174 10/16/14 145 10/15/13 175 10/16/14 146 10/15/13 176 10/16/14 147 10/15/13 177 10/16/14 148 10/15/13 178 10/16/14 149 10/15/13 179 10/16/14 150 10/15/13 180 10/16/14

Procedure No.: Procedure

Title:

Page:

4 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/16/14 TABLE OF CONTENTS Section 1.0 PURPOSE..................................................................................................................... 7

2.0 REFERENCES

/RECORDS REQUIRED/

COMMITMENT DOCUMENTS............................................................................... 7 3.0 PREREQUISITES....................................................................................................... 11 4.0 PRECAUTIONS/LIMITATIONS.............................................................................. 12 5.0 STARTUP/NORMAL OPERATION 5.1 Automatic Malceup............................................................................................. 16 5.2 Boration.............................................................................................................. 21 5.3 Dilution ... ... ................... .... ........ ... .. ...... ...... ................. .... .......... ... ... ...... .. ... ........ 27 5.4 Manual Makeup................................................................................................. 33 6.0 SHUTDOWN................................................................................................................ 37 7.0 INFREQUENT OPERATIONS 7.1 Boric Acid Batching .. .. .. .... .. ... .... .... ..... .. ... ... .. .... .. .. .. .. ... .. ..... .. ...... ... ... ..... .. .. ... .... 38 7 .2 Transferring Boric Acid from the Batching Tank to B BAST Using 3B Boric Acid Transfer Pump.............................................. 43 7 .3 Transferring Boric Acid from the Batching Tank to C BAST Using or 4B Boric Acid Transfer Pump ......................................... 49 7.4 A, B, and C BAST Recirculation Using a Unit 3 and Unit 4 Boric Acid Transfer Pump .. .. .. .... .. .. .. .. .. ... ... .... ... ... ... .. .... .... ... .... ....... 54 7.5 CVCS Manual Makeup to the RWST................................................................ 57 7.6 BAST Level Bubbler Lines - Blowdown........................................................... 62 7.7 Chemical Addition to the RCS - Unit 3 ............................................................. 66 7.8 Chemical Addition to the RCS - Unit 4 ............................................................. 72 7.9 Initial Cleaning or Draining of the Temporary Boric Acid Tank (TT-2) .......... 78 7.10 Boric Acid Batching in Temporary Boric Acid Tank........................................ 83 7 .11 Transferring Boric Acid from the Temporary Boric Acid Tank to the BA Batch Tank.......................................... 88

Procedure No.: Page:

Procedure

Title:

5 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/16/14 TABLE OF CONTENTS (Cont'd)

Section 7.12 Filling Temporary Boric Acid Tank with Primary Water................................. 95 7.13 Transferring Water Between CVCS Holdup Tanks Without Using the Gas Stripper Feed Pumps.................................................... 98 7.14 Recirculation of CVCS Holdup Tanks.............................................................. 101 7.15 Transferring Water Between CVCS Holdup Tanks Without Using the Holdup Tank Recirc Pump.................................................. 105 7 .16 Boration of RCS from the RWST...... .... .. ...... ... .... .... .... ..... .... ... ..... .... ..... .... ....... 108 7.17 Swapping CVCS Holdup Tanks........................................................................ 112 7 .18 Post Maintenance Venting Manual Boration Header *-3 56 ...... ........ ....... .. ....... 114 7.19 Dilution of RCS with CVCS-Boron Concentration Control System Out of Service for Maintenance............................................... 118 7 .20 Hydrazine Addition to the RCS - Unit 3 .... .......... ... .. .. .. ...... ... ... ..... .... ..... .. ..... .... 123 7.21 Hydrazine Addition to the PRZ - Unit 3...... .. .... .... .. .. ... .. ... .... .... .... ....... ..... .. .. .... 128 7 .22 Hydrazine Addition to the RCS - Unit 4.... ............... .. ... ............. .... ................ ... 134 7 .23 Hydrazine Addition to the PRZ - Unit 4..................... ... ... ... ..... .. .. .. .... ... ... .. .. .. .. . 140 7.24 Transferring Boric Acid from the Batching Tank to A BAST Using 3A Boric Acid Transfer Pump ............................................. 146 7.25 Transferring Boric Acid from the Batching Tank to B BAST Using 4A Boric Acid Transfer Pump.............................................. 151 7.26 Auto Makeup Feed and Bleed of RCS............................................................... 157 7 .27 Placing a BAST Level Hose in Service When a BAST Level Transmitter is Out of Service .. ..... ....... .... ....... .. ...... .. .. .. .. . 160 ENCLOSURES/ATTACHMENTS Enclosure 1 Nitrogen Supply to BAST Bubblers .................................................................. 166 Enclosure 2 Temporary Boric Acid Tank Configuration ...................................................... 168

Procedure No.: Procedure

Title:

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6 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/16/14 TABLE OF CONTENTS (Cont'd)

Section Enclosure 3 Tanl(er Vendor Drawing.................................................................................... 169 Enclosure 4 Tank Height -vs- Volume ........................................ .......................................... 170 Enclosure 5 Solubility of Boron (H 3B0 3) Versus Temperature............................................ 171 Enclosure 6 Guidance for Maintaining TAVE and T REF Matched at Steady State Full Power Using Primary Water Dilution................................ 173 Enclosure 7 Guidance for Temporary Boric Acid Tank Immersion Heater Controller Operation............................................................ 175 Attachment 1 CVCS - Boric Acid System Valve Alignment................................................... 176 Attachment 2 CVCS - Boric Acid Batch Tank Valve Alignment............................................ 183 Attachment 3 CVCS - Boric Acid System Breaker Alignment................................................ 185 Attachment 4 CVCS - Units 3 and 4 Boric Acid Evaporators and Gas Strippers System Boundary Valves ... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. 189 Attachment 5 Reactivity Worksheet......................................................................................... 192 Attachment 6 CVCS - Temporary Boric Acid Tank Valve Alignment................................... 195

Procedure No.: Procedure

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7 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 1.0 PURPOSE 1.1 This procedure provides instructions for the startup, normal operation, and infrequent operation of the CVCS Boron Concentration Control System.

2.0 REFERENCES

/RECORDS REQUIRED/COMMITMENT DOCUMENTS 2.1 References 2.1.1 Technical Specifications

1. Section 3/4.1.1, Boration Control
2. Section 3/4.1.2, Boration Systems
3. Section 3/4.4.7, Chemistry
4. Section 3/4.5.4, Refueling Water Storage Tank 2.1.2 FSAR
1. Section 9 .2, Chemical and Volume Control
2. FSAR Table 4.2-2, Reactor Coolant Water Chemistry Specifications 2.1.3 Plant Drawings
1. 5610-E-25, Sh 6 - Primary Water Makeup Pump
2. 5610-E-25, Sh 10 - Boric Acid Transfer Pump
3. 5610-E-25, Sh 46 - Boric Acid Tank Heaters A and C
4. 5610-E-25, Sh 46A - Boric Acid Tank Heaters B
5. 5610-E-25, Sh 82 - Boric Acid Flow to Blender, FCV-113A
6. 5610-E-25, Sh 83 - Demin Water Flow to Blender, FCV-114A
7. 5610-E-25, Sh 84 - Diluted Flow to VCT, FCV-114B
8. 561 O-E-25, Sh 85 - Blender Flow to CHP Suction Header, FCV-113B
9. 5610-E-25, Sh D - Control Switch Development
10. 5610-E-855, Breaker List
11. 5610-M-3046, Sh 1, CVCS - Boric Acid Systems
12. 5610-M-3061, Sh 1, Waste Disposal System - Liquid Waste Holdup and Transfer
13. 5610-T-D Boron Concentration Contra 1 System

Procedure No.: Procedure

Title:

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8 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 2.1.3 (Cont'd)

14. 5610-T-Ll, Sh 24B - Boric Acid Transfer Pumps 3A, 3B
15. 5610-T-Ll, Sh 24 - Primary System Pumps, Containment Sump and Safety Injection
16. 5613-M-3014, Sh 1 and 2, Condenser System
17. 5613-M-3020, Sh 1and2, Primary Water Makeup System
18. 5613-M-3036, Sh 1, Sample System - NSSS
19. 5613-M-3047, Sh 1, CVCS - Charging and Letdown System
20. 5613-M-3084, Sh 1, Auxiliary Steam System
21. 5614-M-3014, Sh 1 and 2, Condenser System
22. 5614-M-3020, Sh 1and2, Primary Water Makeup System
23. 5614-M-3036, Sh 1, Sample System - NSSS
24. 5614-M-3047, Sh 1, CVCS - Charging and Letdown System
25. 5614-M-3084, Sh 1, Auxiliary Steam System 2.1.4 Plant Procedures
1. 0-ADM-l 02, On-the-Spot Changes to Procedures
2. O-ADM-103, Temporary Procedure
3. O-ADM-215, Plant Surveillance Tracking Program
4. O-ADM-217, Conduct ofinfrequently Performed Tests or Evolutions
5. O-NCOP-041, Chemical Adjustments to the Reactor Coolant System and Pressurizer
6. 3/4-0NOP-046.4, CVCS - Malfunction of Boron Concentration Control System
7. 3/4-0P-020, Primary Water System
8. 3/4-0P-041.2, Pressurizer Operation
9. 3/4-0P-047, CVCS-Charging and Letdown
10. O-OP-061.13, Waste Disposal System - Transferring Water to the Portable Demineralizer Skid for Processing
11. O-OP-065.3, Nitrogen Gas Supply System

Procedure No.: Page:

Procedure

Title:

9 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 2.1.4 (Cont'd)

12. 3/4-0SP-041.1, Reactor Coolant System Leak Rate Calculation
13. 3/4-0SP-046.3, CVCS Boration Systems Flowpath Verification
14. O-OSP-200.1, Schedule of Plant Checks and Surveillances
15. MA-AA-101-1000, Foreign Material Exclusion Procedure
16. OP-AA-101-1000, Clearance and Tagging 2.1.5 Vendor/Technical Manuals
1. Westinghouse Operating Procedure, Coolant Chemistry Addition and Control S-3.3, Parts II and III 2.1.6 Miscellaneous Documents (i.e., PC/Ms, ECs, Correspondence)
1. NRC IE Bulletin 88-04, Potential for Safety Related Pump Loss
2. JPE-PTN-SELJ-88-027, Response to NRC IE Bulletin 88-04 on Potential for Safety Related Pump Loss, dated July 8, 1988
3. PC/M 87-257, Load Center 4H, MCC 4D and Transfer of Loads
4. PC/M 87-258, Load Center 3H, and Repowering ofMCC 3D
5. PC/M 89-494, Boric Acid Storage Tank Alarm Setpoint Change
6. PC/M 90-423, Boric Acid Transfer Pump Seal Replacement
7. PC/M 90-424, Boric Acid Transfer Pump Seal Replacement
8. PC/M 90-440, Boric Acid Concentration Reduction.
9. PC/M 91-092, Removal of Boric Acid Tank Header
10. PC/M 91-068, Removal of BIT Recirculation Header for the BAS Tank
11. PC/M 94-141, Boric Acid Evaporators and Gas Strippers Abandonment
12. PC/M 95-140, Abandonment of Boric Acid Recirculation Pumps and Related Piping
13. PC/M 95-172, Unit 4 BIT Bypass
14. PC/M 95-081, Abandonment of Auxiliary Steam System Desuperheater Stations, Condensate Recovery Transfer Pumps and Auxiliary Steam Components Inside of the Auxiliary and Radwaste Buildings
15. NRC Information Notice 96-69: Operator Actions Affecting Reactivity

Procedure No.: Page:

Procedure

Title:

10 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 5/6/13 2.1.6 (Cont'd)

16. PC/M 95-040, Abandonment of Various Boron Recycle System and Liquid Waste Disposal System Components
17. Plant Curve Book, Section 3, CVCS
18. PTN-ENG-SEMS-00-0008, Use of PRC-01 Resin for Removal of Cobalt-58 Contaminants in the Letdown Stream
19. PC/M 00-016, CRN M-10278, Reconfigure Primary Water Hose Connections in Unit 3 Cask Wash Area
20. CR 2006-34989, Hydrazine Potential to Create a Crud Burst in the RCS
21. EC 24 7006, PCM 09-13 7, Unit 3 EPU Instrument Setpoint/Indication Changes
22. EC 247008, PCM 09-139, EPU LAR Umbrella Doc only PC/M 2 .2 Records Required 2.2.1 The date, time, and section completed shall be entered in the Unit Narrative Log.

Also, problems encountered while performing the procedure should be entered; i.e., malfunctioning equipment, delays due to changes in plant conditions, etc.

2.2.2 Completed copies of the QA Record Pages for the below listed items are Quality Assurance records and shall be transmitted to QA Records for retention in accordance with Quality Assurance Records Program requirements:

1. Subsections 5.1 through 5 .4
2. Subsections 7.1 through 7.23
3. Attachments 1, 2, 3, 4, and 6 2.2.3 Completed copies of the below listed items shall be retained in the Shift Manager file until the next performance of that attachment:
1. Attachments 1, 2, 3, and 4 2.2.4 Completed attachments listed below, that have the Tag column ~hecked ('1), shall be copied and transmitted to the labeling Coordinator:
1. Attachments 1, 2, 3, and 4

2.3 Commitment Documents 2.3.l Memorandum L-88-295, Response to NRC IE Bulletin 88-04, Potential for Safety Related Pump Loss, dated July 11, 1988 2.3.2 INPO SOER 94-2, Boron Dilution Events in PWRs 2.3.3 CR-99-0603, Incorrect Boron Concentration Used for Dilution Calculation 2.3.4 AR 1695469, RCS Activity Increase Following Hydrazine Addition - CAPR 3.0 PREREQUISITES 3.1 The following systems are operable or m operation to support the CVCS - Boron Concentration Control System operation:

3 .1.1 Instrument Air 3 .1.2 Primary Water System 3.1.3 Chemical Volume Control System 3 .1.4 Primary Sampling System 3.1.5 Nitrogen Gas Supply System 3.2 All plant electrical systems are operable to supply power and control functions to support the CVCS - Boron Concentration Control System operation.

3.3 All instruments and control devices are in service for the CVCS - Boron Concentration Control System operation with no surveillances required and no outstanding PWOs, clearances, or temporary system alterations that affect system operability as per the following:

3.3.l O-ADM-215, Plant Surveillance Tracking Program, and O-OSP-200.1, Schedule of Plant Checks and Surveillances. (No surveillances have exceeded the date required on the missed surveillance sheet.)

3.3.2 Temporary System Alteration (TSA) Log 3 .3 .3 Clearance Log 3.3.4 Out-of-Service Log

Procedure No.: Page:

Procedure

Title:

12 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 318113 3.4 The CVCS - Boron Concentration Control System valve and breaker alignments have been verified by the completion of the following attachments as applicable:

3 .4 .1 Attachment 1 3.4.2 Attachment 2 3.4.3 Attachment 3 3.4.4 Attachment 4 3.5 For the use of the Temporary Boric Acid Tank:

3.5.l Temporary Power Supply for Temporary Boric Acid Pumps is operable.

3.5.2 The CVCS - Temporary Boric Acid Tank valve alignment has been verified by completing Attachment 6.

4.0 PRECAUTIONS/LIMITATIONS 4.1 Before changing system status, Technical Specifications should be consulted for system requirements for that plant mode.

4.2 During normal makeup operation, the BAST levels shall not be reduced below Technical Specifications Figure 3.1-2 for Operational Modes 1through4.

[Commitment - Step 2.3.l]

4.3 Technical Specifications require a flow path for boration to be maintained at all times when fuel is in the Reactor Vessel.

4.4 Except during RCS dilution or boration, the Reactor Makeup Selector switch should always be set on Auto with the RCS Makeup Control switch placed to the Start position to obtain a red start light.

4.5 Due to the inherent inaccuracies of the primary water and boric acid totalizers, the Pressurizer and the Reactor Coolant Loops should be sampled to determine if the desired boron concentrations have been achieved after boric acid or primary water additions.

4.6 To ensure effective blending, major boration changes shall be made with an RCP (preferably B or C) operating or RHR flow greater than 3,000 gpm.

4.7 Boron dilution shall not be used to bring the reactor critical except for low power physics or initial criticality after refueling.

4.8 The proper personnel protective equipment (PPE) should be worn while handling boric acid.

4.9 All work performed in the Radiation Controlled Area shall be performed in accordance with a Radiation Work Permit and the ALARA program.

4.10 Boric Acid pumps should not be run simultaneously in the recirculation mode to the same BAST due to having a common recirc line and the possibility of dead heading one of the pumps. [Commitment - Step 2.3 .1]

Procedure No.: Procedure

Title:

Page:

13 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 4.11 If a Boric Acid spill occurs, then the drains should be flushed as soon as possible with plain water.

4.12 With the BASTs in their normal configuration (all tied together), pumping any solution into a BAST will eventually equalize with the remaining BASTs.

4.13 If a BAST boron concentration or volume does not meet the requirements of Tech Spec Figure 3 .1-2, then increase boron concentration or volume until requirements of Figure 3 .1-2 are satisfied.

4.14 The Batching Tank or Temporary Boric Acid Tank boron concentration shall not exceed the concentration shown on Enclosure 5, 10°F margin line, based on the batching tank or Temporary Boric Acid Tank water temperature, as applicable.

4.15 Lithium concentration in the RCS must be maintained in accordance with O-NCOP-041, Chemical Adjustments to the Reactor Coolant System and Pressurizer.

4.16 Anticipate changes in reactivity whenever the reactor coolant boron concentration is altered. Observe TAVG or subcritical count rate to ensure that the desired change in RCS boron concentration has been achieved. If TAVG increases or decreases l .5°F from TREF or subcritical count doubles, then the change in boron concentration shall be stopped and cause found. [Commitment Step 2.3 .2]

4.17 Verify that all chemicals to be added to the RCS have been approved by Chemistry Department.

4.18 Always open the chemical addition mix tank drain valve prior to opening the chemical fill valve to avoid blowing chemicals on the operator.

4.19 Boric acid concentration should be between 5682 ppm and 6993 ppm prior to transferring the Temporary Boric Acid Tank contents to the BA Batch Tank.

4.20 Fluoride and Chloride concentration shall be less than 150 ppb prior to transferring the Temporary Boric Acid Tank contents to the BA Batch Tank or Decontamination Tan1c 4.21 The Temporary Boric Acid Tank has a maximum volume of 19,749 gallons.

4.22 Valves TBA-5 and TBA-6, TBA Pumps suction valves, fail closed on loss of hydraulic pressure.

4.23 The ratio of boric acid flow to setpoint on Boric Acid Flow Controller FC-*-113A is 5 gpm to 1. (i.e., 10 gpm is equal to a setpoint of2) 4.24 The ratio of primary water flow to setpoint on the Primary Water Flow Controller, FC-*-114A, is 15 gpm to 1, i.e., 30 gpm is equal to a setpoint of 2.

4.25 The ratio of primary water flow to setpoint on the Primary Water Auto Setpoint, HIC-*-114, is 15 gpm to 1, i.e., 30 gpm is equal to a setpoint of2.

4.26 During large dilutions or borations, monitor VCT level, and ensure that level is maintained between 20 and 70 percent by reducing the dilution/boration flow rate or increasing charging pump flow rate as required.

Procedure No.: Procedure

Title:

Page:

14 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 4.27 Accuracy of boron analysis (plus or minus 1 percent) and accuracy of the primary water and boric acid totalizers will typically result in some variation between actual and expected results.

4.28 Following large borations or dilutions of the RCS, intended boron concentration shall be compared to the sample results obtained by chemistry. If a significant deviation exists, the Shift Manager shall be notified.

4.29 The access manway cover of the Temporary Boric Acid Tank needs to be maintained closed or funnel covered to minimize the potential for introducing foreign material into the system, except for adding chemicals or for performing inspections.

4.30 Make-up to the RCS or Reactor Cavity with fuel in the reactor vessel is a positive reactivity addition if the make-up boron concentration is less than the RCS or the Reactor Cavity, i.e., the make-up boron concentration is 2100 ppm and RCS boron concentration is 2200 ppm. Applicable Technical Specifications apply.

4.31 An Independent Verification of boron reduction change calculations should be performed when below the point of adding heat. This allows an extra verification to ensure SDM requirements are not challenged.

4.32 The Chemical Mixing Tank capacity is 6 gallons. The tank should be flushed with 30 gallons of primary water.

4.33 The primary water supply orifice (R0-*-6692) to the Chemical Mixing Tank will deliver approximately 3.1 gpm of flow.

4.34 IfFIT-*-114 fails high, an erroneous high primary water flow on FR-*-113 will cause the Primary Water Totalizer to count up continuously.

4.35 Unit 3 primary water flow indication is less accurate at lower flow rates, i.e., at 20 gpm flow rate, there could be a 4.5% error which could result in 0.9 gpm variation in actual flow.

4.36 No demineralizer bypass is required for hydrazine addition if using PRC-01 resin.

4.37 Due to valve and positioner characteristics, there is no direct correlation between demand meter setting and actual flowrate for FC-*-113A/114A, i.e., setting demand to 20% with the controller in manual on FC-*-114A will not result in a flowrate of exactly 20% of scale or 30 gpm. When performing manual makeups or blends, the need to adjust the controller manually to achieve the desired flowrate after the system is started should be anticipated.

4.38 For normal steady state full power operations, Enclosure 6 may be used to maintain TAVE and reactor power. Enclosure 6 should be referenced during the dilution evolutions.

A laminated copy of Enclosure 6 may be used for placekeeping and maintained in the Unit 3 and Unit 4 Information Book.

4.39 Addition of hydrogen peroxide or hydrazine could cause crud bursts in the RCS resulting in a short-term increase in radiation levels in the RCS.

4.40 Borations and dilutions to the VCT outlet via FCV-*-113B have a more immediate reactivity effect than borations and dilutions to the VCT gas space via FCV-*-114B.

W201 O:TN:rn11111L,1;,1""'

Procedure No.: Page:

Procedure

Title:

15 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 3/8/13 4.41 Boration headers to the charging pump suction headers (i.e., MOV-*-350, FCV-*-113B, and *-356 headers) have high point vents to facilitate venting which is required using the guidance of O-ADM-222, Drain and Vent Rig Controls, following maintenance activities that drain the charging pump suction header. Boration headers are higher than the charging pump suction header and will drain when the charging pump suction header is drained.

4.42 Boric Acid Make Up Flow Deviation Alarm, Annunciator A 2/5, and Primary Water Make Up Flow Deviation Alarm, Annunciator A 216 are blocked for 22 seconds after start of make up to the VCT to avoid spurious initiation.

4.43 Minimum required PPE when handling boric acid is gloves, long sleeves, steel or composite shoes, and a dust mask. The use of the dust mask can be waved by Safety Department if BA concentration in the air is less than 5mg/m3.

4.44 Chemical addition pre-job briefing shall be face to face briefings with required personnel.

No designees or telecom use permitted. [Commitment - Step 2.3.4]

4.45 BAST boric acid concentration range will be administratively controlled between 3.25 wt% (5682 ppm) and 4.0 wt% (6993 ppm). A boric acid concentration below 3 .25 wt% invalidates the BAST inventory and alarm setpoint calculations and may result in a locked-in alarm condition.

4.46 With one boric acid tank out-of-service for maintenance, the minimum boron concentration in the other two boric acid tanks is 3.6 wt%.

4.4 7 The chemical addition funnel cover needs to be maintained closed to mm1m1ze the potential for introducing foreign material into the system, except during chemical additions or for performing inspections.

4.48 Proper foreign material controls are required for boric acid drums and other chemical addition equipment. Refer to MA-AA-101-1000, Foreign Material Exclusion Procedure.

Procedure No.: Procedure

Title:

Page:

57 Approval Date:

~

O-OP-046 CVCS- Boron Concentration Control 10/15/13 7.5 eves Manual Makeup to the RWST Date/Time Started: _ _ _ -fo_~--11-----=--/_f'J_()_W_ __

@ Initial Conditions e?f' An adequate supply of boron and primary water is available for the desired

/'\._/ level change.

~ Plant conditions will allow aligning the eves - Boron Concentration Control System to the RWST for the time required for the makeup.

@ Procedure Steps C. Mrf-0

\_/

I0 N Instrument uncerlainties for the Boric Acid and Primary Water flow transmitters can, result in the actual amount of Boric Acid or Primary Water added to be either more or Jess than the amount calculated. Thus, care is needed to ensure that excessive reduction in RCS boron concentration does NOT occur due to the uncerlainties.

r,,..-- -- -- -- -- ~- - -- -- -- -- - 1 18fl The RWST makeup shares the same line as Safety Injection Pump recirculation.

\./ RWST Operation of safety injection pumps with recircirculation flow aligned to the affected will affect RWST makeup flow rates.

I

~ VCT level is 14. 15 gallons per % level indication.

~ When the RWST boron concentration is greater than 2500 ppm, makeup may be accomplished by using primary water only. When this method is used, caution shall be I exercised to ensure that the RWST is NOT diluted below its boric acid concentration of I 2400 ppm.

Iff If planning to make up only primary water to the RWST, then steps designated with an 1

asterisk(*) can be marked NIA.

lff If Annunciator G 812, RWST TECH SPEC MIN LEVEL, is received prior to makeup, approximately 4000 gallons of makeup will be required to clear the alarm.

I

~The RWST Hi Level alarm setpoint is 332, 000 gallons with a margin of error of 1675 gallons. Makeup should be limited to a level of 330,000 gallons to avoid a locked I in RWST Hi Level alarm. If filling above 330, 000 gallons, RWST Hi Level is an expected alarm.

ffiecord the unit number on the QA Record Page.

~-- - --- -

2. Determine the approximate boric acid and primary water flows and volumes needed to obtam the desired blend concentration from the boron change tables in Section III of the Plant Curve Book.

W20i O:TNM/fm/cls/cls

Procedure No.: Procedure

Title:

Page:

58 Approval 'Date:

O-OP-046 CVCS - Boron Concentration Control 10/15/13 7.5.2 (Cont'd)

CAUTION TAvG and reactor power are required to be monitored for changes during makeup to the RWST. [Commitment - Step 2.3.2]

3. Verify Closed the following valves:
a. Blender to Charging Pump Suction, FCV-*-113B
b. Blender to VCT, FCV-*-114B
c. Manual Emerg Boration Isol, *-356
4.
  • Place the applicable control switches to Close for the following valves:
a. Blender to Charging Pump Suction, FCV-*- l 13B
b. Blender to VCT, FCV-*-l 14B
5. Verify Emergency Boration Valve, MOV-*-350, is Closed.
6. Place RCS Makeup Control switch to Stop.
7. Unlock AND Open Blender Disch to RWST Stop Vlv, *-365A.
8. Open Blender Disch to RWST, *-365B.
  • 9. Place Reactor Makeup Selector switch to Borate.
  • 10. Adjust Boric Acid Flow Controller, FCV-*-113A, Auto Setpoint to the value determined in Substep 7 .5 .2.2.
11. Place Primary Water Flow Controller, FCV-*-114A, to Manual AND adjust the output on the demand meter to zero.
  • 12. Place the control switch for Boric Acid to Blender, FCV-*-113A, to Auto.

W201 O:TN M/fm/cls/cls

  • Procedure No.: Page:

Procedure

Title:

59 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/15/13 INITIALS CK'D VERIF 7.5.2 (Cont'd)

  • 13. Set the Boric Acid Totalizer and the Primary Water Totalizer to the volume determined in Substep 7.5.2.2 by performing the following on each totalizer:
a. Press LilvllT 1.
b. Press CLR.
c. Enter desired amount using numeric keypad.
d. Press ENT.
e. Press COUNT A.
f. Press LIMIT 1 AND verify desired amount was properly entered.
g. Press COUNT A.

NOTES ---------,

I . The Boric Acid addition will stop automatically when the Boric Acid Totalizer reaches the pre-set value.

.- The Primary Water flow will continue and the Primary Water Totalizer will count as I long as FCV-*-114A is open.

L-----------------------1

15. Perform the following:
a. Tum the RCS Makeup Control switch to Start.
b. Verify Red Start light is Energized.
16. Place the control switch for Primary Water to Blender, FCV-*-114A, to Open.
17. Adjust Primary Water Flow Controller, FCV-*-114A, to the value determined in Substep 7.5.2.2.
18. Verify proper flow to the RWST by observing an increase in RWST level AND a decrease in BAST and PWST levels as applicable.

Procedure No.: Procedure

Title:

Page:

60 Approval Date:

O-OP-046 CVCS - Boron Concentration Control 10/15/13 INITIALS CK'D VER.IF 7.5.2 (Cont'd)

19. IF primary water addition is complete before boric acid addition is complete, THEN Close Primary Water to Blender Valve, FCV-*-114A.
  • 20. WHEN boric acid addition is complete, THEN flush the RWST fill line with primary water as follows:
  • a. Reset the Primary Water Totalizer.
  • b. Verify Open Primary Water to Blender Valve, FCV-*-114A.
  • c. WHEN approximately 100 gallons of additional primary water has been added, THEN Close Primary Water to Blender Valve FCV-*-114A. - -
21. Close AND lock Blender Disch to RWST Stop Valve, *-365A.
22. Close Blender Disch to RWST Valve *-365B.
23. Restore CVCS System alignment for auto makeup using Subsection 5 .1.

The RWST should be recirculated for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure a valid sample.

L----------------------

24. IF the R WST Purification Loop is available AND NOT in service, THEN place it in service using 3-NOP-033 or 4-0P-033, Spent Fuel Pit Cooling System, as applicable, to assure mixing of the RWST for sampling.
25. Have the RWST sampled to ensure Tech Spec requirements are met.
26. Ensure log entries specified in Subsection 2.2 are recorded.
27. Complete the QA Record Page for this subsection.

Procedure No.: Page:

Procedure

Title:

61 Approval Date:

O-OP-046 CVCS-Boron Concentration Control 10/15/13 QA RECORD PAGE II II (Page 1 of 1)

Procedure Revision Date ____/ Cc.tVYu1t;_ _ __

7.5 CVCS Manual Makeup to the RWST 7.5.2.1 Unit ----

b RWST Date/Time Started

- - - - -I- - - - Date/Time Completed _ _ _ ___.:_/_ _ __

PERFORMED BY (Print) INITIALS VERIFIED BY (Print) INITIALS I have reviewed this subsection and it has been satisfactorily performed. Deviations or TCs used to perform this procedure are listed under Remarks.

REVIEWED BY: _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ Date:- - - - - - -

Reactor Operator or Designee W201 O:TNM/fm/cls/cls

Section 3, Figure 4 24 Feb 2006 RJT BLENDED FLOW Page 2 of2 METHOD 2: CALCULATION The following equation represents the relation between the blended boron concentration and the flow rates of boric acid and primary water. This single equation, however, contains two unknowns, the boric acid flow rate AND the primary water flow rate. An appropriate method to solve this equation is to "assign" a value to one of the unknowns, and then solve for the other.

An example of this method appears below.

ACIDgpm x BASTppm BORONppm=

(ACIDgpm + WATERgpm)

WHERE: BORON ppm is the desired blended boron concentration in ppm BASTppm is the most recent average concentration of the Boric Acid Storage Tanks in ppm (A nominal value of 5664 ppm may be used.)

ACIDgpm is the Boric Acid Flow in gpm WATERgpm is the Primary Water Flow in gpm EXAMPLE:

Assume that the desired blended boron concentration is 1200 ppm. From the table on page 1, it can be seen that if, say 12 gpm of Boric Acid flow is "assigned" then the Primary Water flow will be calculated at just above 40 gpm. This of course assumes an average BAST concentration of 5245 ppm. Since the average BAST concentration is normally greater than that, it follows that the required Primary Water fiow will also be greater. For this example, let us say that the average BAST concentration is 5675 ppm.

The calculation is as follows:

ACIDgpm x BASTppm BORONµpm=

rearranging the equation to solve for WA TERgpm gives us:

ACIDgpm x BASTppm

- ACIDgpm = wATERgpm BORON ppm 12 gpm x 5675 ppm

- 12 gpm = 44.75 gpm 1200 ppm The calculated 44.75 gpm of Primary Water may be appropriately rounded.

M u J Page 1 of 11 JPM TITLE: Test Containment High Range Radiation Monitors JPM NUMBER: 01066003201 REV. 1-0 TASK NUMBER(S) I 01066003200/

TASK TITLE(S): Test Containment High Range Radiation Monitors KIA NUMBERS: 2.3.5 KIA VALUE: RO 2.9 Justification (FOR KIA VALUES <3.0): N/A (>2.5 for NRC LOIT Annual Exam)

TASK APPLICABILITY:

~ RO D SRO D STA D Non-Lie D SRO CERT D OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: D Perform: 0 EVALUATION LOCATION: In-Plant: D Control Room:

Simulator: D Classroom: x I Lab: Other: I Time for Completion: 10 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No Date Date Approved by: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

Training Program Owner TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/RO

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 2 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 3 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE Updated to fleet template; Updated for L-15-1 See cover page N/A 1-0 01982463 text/grammar changes NRC Exam See cover page N/A 1-1 1-2 1-3 1-4 1-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 4 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION SIMULATOR SET-UP: N/A Required Materials:

  • 3-OSP-204, Accident Monitoring Instrumentation Channel Checks -

marked-up copy General

References:

  • 3-OSP-204, Accident Monitoring Instrumentation Channel Checks
  • Demonstrate ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
  • Given channel-check data for CHRRMs, recognize results are UNSAT, Unit Supervisor must be notified immediately, and Section 5.2 (Documentation) must be completed TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 5 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • Unit 3 is at 100% steady-state power.

Initiating Cue:

  • You are directed to complete Section 4.8 (Containment High Range Area Radiation Monitoring

[CHRRMS] Channel Check) of 3-OSP-204 (Accident Monitoring Instrumentation Channel Checks), beginning at Step 12.

  • Based on the test results, determine the appropriate operator response and complete any required procedure sections.

Operator Action(s): ____________________________________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/RO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 6 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

Provide examinee with a marked-up copy of 3-OSP-204, Accident Evaluator Cue:

Monitoring Instrumentation Channel Checks.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 7 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Performance Step: 2 Calculate the differences between channel A and channel B Critical: Yes readings and record in Table 1.

Calculate the differences between channel A and channel B readings, as follows, and record in Table 1:

Standard:

  • RI-3-6311A (QR81) vs. RI-3-6311B (QR81): 2E7
  • RI-3-6311A (VPC) vs. RI-3-6311B (VPC): 5.5E7
  • RAR-3-6311A vs. RAR-3-6311B: 1.1E7 Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Performance Step: 3 Determine if acceptance criteria are SAT or UNSAT, identify the Critical: Yes need for supervisor notification, and document results.

  • Recognize that VPC data do not meet the acceptance criteria and mark results as UNSAT.
  • Per Step 4.1.1, the Unit Supervisor must be notified immediately.

Standard:

  • Per Step 4.1.1, document condition in Section 5.2 by checking and initialing acceptance criteria as UNSAT and summarizing the failure in the Remarks section.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 8 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Reset DCS DEGRADED CONTAINMENT (NOT EQ) CONDITION by Performance Step: 4 touching the Operator Reset of Degraded Containment - RESET Critical: No button on the Degraded Containment screen on the DCS.

Standard: No action required.

Evaluator Cue: Inform examinee that this step has been performed.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Performance Step: 5 If CHRRMS channel check is UNSAT, then go to 3-ONOP-094, Critical: No Alternate Methods for Containment Post Accident Monitoring.

Standard: Request/obtain required materials.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 5, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

01066003201, Test Containment High Range Radiation Monitors, Rev. 1 JPM Page 9 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

  • Unit 3 is at 100% steady-state power.

Initiating Cue:

  • You are directed to complete Section 4.8 (Containment High Range Area Radiation Monitoring

[CHRRMS] Channel Check) of 3-OSP-204 (Accident Monitoring Instrumentation Channel Checks), beginning at Step 12.

  • Based on the test results, determine the appropriate operator response and complete any required procedure sections.

Operator Action(s): ____________________________________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

Procedure No.

T RKEY JNT UN 3-0SP-204 OPERATIONS SURVEILLANCE PROCEDURE Revision No.

FP SAFETY RELATED 9 CONTINUOUS USE

Title:

ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS Responsible Department: OPERATIONS Special Considerations:

FOR INFORMATION ONLY Before use, verify revision and change documentation (if applicable) with a controlled index or document.

DATE VERIFIED ~

INITIAL----"---

Revision Approved By Approval Date UNIT# UNIT3 DATE DOCT PROCEDURE 3 Brian Stamp 12/08/10 DOCN 3-0SP-204 SYS STATUS COMPLETED 9 Sam Shafer 10/13/14 REV 9

  1. OF PGS

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 2 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 REVISION

SUMMARY

Rev. No. Description 9 PCR 1996913, 10/13/14, Jonathan Lubert Revised to show #6 Unheated Thermocouple TE-3-6506 8 input to QSPDS failed. Input is bypassed.

PCR 1979583, 10/07/14, Michael Hargis Enhancement to correct mismatch between bypassed 38 RVLMS Thermocouples and the 38 QSPDS Flat Panel Display.

8 PCR 1973009, 07/08/14, Michael Hargis Edits shaded fields since instruments have been repaired and are no longer bypassed. TE-3-50E, TE-3-6507.

7 PCR 1883958, 06/24/13, Angel Ramirez Revise Acceptance Criteria for CET channel check AR 1878683.

6 PCR 1785540, 07/25/12, David Dagitz Revise procedure to reflect the completion of EC 246886, Reconfiguration of U3 Train "A" RVLMS to Restore System Health Capability. This EC reconfigured the RVLMS cables such that Sensor #6 Plenum will be out of se"(,ice for the post U326 cycle, but Sensor #3 will be in service for the cycle.

5 PCR 1783664, 07/23/12, Angel Ramirez

  • (EC 250002) Revised Note on Pages 17, 18, 19, and 23 in order to remove references from the superseded O-ADM-220, Equipment Abandonment Program.
  • Revised Section 4.11, Tables 7 and 8 (Pages 23 and 24) to reflect the correct inoperable and available sensors after refueling outage PT3-26.

(Ref. AR 1779281)

  • Revised Note 2 (Page 23) to accommodate for the available sensors following repairs during PT3-26 RFO.
  • Replaced Section 8.1.1.1 to reflect 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, as the implementing procedure for sensor bypass; thus, replacing O""ADM-220, Equipment Abandonment Program.
  • Revised Attachment 1, QSPDS CET/RVLIS Sensors Out-of-Service, item 3 bullets to reflect new process.
  • Revised Attachment 2, Temperature Element Descriptions and Locations A Train QSPDS and Attachment 3, Temperature Element Descriptions and Locations B Train QSPDS, Notes and updated sensor availability.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 3 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE AND SCOPE ............................................................................................. 4 2.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 4 3.0 PREREQUISITES ....................................................................................................... 4 4.0 INSTRUCTIONS .......................................................................................................... 5 4.1 General Requirements ................................................................................................ 5 4.2 PORV Position Indicator and PORV Block Valve Indicator Channel Check ................ 5 4.3 Safety Valve Position Indicator Channel Check ........................................................... 6 4.4 Containment Pressure (Wide Range) Channel Check ................................................ 7 4.5 Containment Pressure (Narrow Range) Channel Check ............................................. 8 4.6 Containment Water Level Monitor (Wide Range) Channel Check ............................... 9

4. 7 Containment Water Level Monitor (Narrow Range) Channel Check ......................... 1O 4.8 Containment High Range Area Radiation Monitoring (CHRRMS) Channel Check ........................................................................................................................ 11 4.9 Subcooling Margin Monitor Channel Checks ............................................................. 13 4.10 Core Exit Thermocouples Channel Check ................................................................. 15 4.11 Reactor Vessel Level Monitoring System Channel Checks ....................................... 22 4.12 QSPDS Health Check ............................................................................................... 24 5.0 RESTORATION AND DOCUMENTATION ............................................................... 25 6.0 ACCEPTANCE AND FUNCTIONAL CRITERIA ........................................................ 26 7.0 RECORDS ................................................................................................................. 26

8.0 REFERENCES

AND COMMITMENTS ..................................................................... 27 ATTACHMENTS ATTACHMENT 1 QSPDS CET/RVLIS Sensors Out-of-Service ............................... 29 ATTACHMENT 2 Temperature Element Descriptions and Locations A Train QSPDS ***********************************~****************************************************30 ATTACHMENT 3 Temperature Element Descriptions and Locations B Train QSPDS ........................................................................................ 32

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 4 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3

  • INITIAL 1.0 PURPOSE AND SCOPE 1.1 Purpose This procedure provides instructions for performing channel checks on the Accident Monitoring Instrumentation System. This surveillance procedure satisfies the requirements of T.S. 4.3.3.3 and Table 4.3-4 items 1-5, 8-14, and 16.

1.2 Scope 1.2.1 Frequency M

1.2.2 Applicability 1, 2, and 3 1.2.3 MODE Restrictions This surveillance may be performed in any MODE.

2.0 PRECAUTIONS AND LIMITATIONS 2.1 Precautions None 2.2 Limitations None

.@ PREREQUISITES

,@ OBTAIN Shift Manager's permission to perform this procedure. 7 End of Section 3.0

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 5 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.0 INSTRUCTIONS e General Requirements

'~ IF during performance of this procedure any of the following occur:

G Acceptance Criteria is UNSAT G) A malfunction occurs

{;) An abnormal condition is found THEN:

G NOTIFY the Unit Supervisor immediately.

Q DOCUMENT condition in Section 5.2.

End of Section 4.1 4.2 PORV Position Indicator and PORV Block Valve Indicator Channel Check

1. CHECK the following conditions:
  • PORV PCV-3-456, Green indicating light ON
  • PORV PCV-3-455C, Green indicating light ON
  • A4/1, PORV/SAFETYVALVE OPEN, CLEAR
  • A 712, PZR PORV HI TEMP, CLEAR
  • PZR Relief Stop Valve, MOV-3-536, Red indicating light ON
  • PZR Relief Stop Valve, MOV-3-535, Red indicating light ON Acceptance Criteria Results All conditions and indications checked above are D SAT D UNSAT satisfied.

End of Section 4.2

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 6 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.3 Safety Valve Position Indicator Channel Check

1. CHECKA4/1, PORV/SAFETYVALVE OPEN, CLEAR.
2. PRESS AND HOLD Test Switch on Unit 3 Pressurizer Safety Valve Acoustic Monitoring System.
3. CHECK Annunciator A 4/1, PORV/SAFETY VALVE OPEN, IN ALARM.
4. RELEASE Test Switch.
5. RESET Annunciator A 4/1, PORV/ SAFETY VALVE OPEN.
6. RECORD Pressurizer Safety Relief Tl-3-465 on VPA.

OF Acceptance Criteria Results Tl-3-465 indicates less than 200°F. D SAT D UNSAT End of Section 4.3

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 7 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.4 Containment Pressure (Wide Range) Channel Check

1. RECORD Channel A Pl-3-6306A.

psig

2. RECORD Channel 8 Pl-3-63068.

psig

3. CALCULATE the difference between Pl-3-6306A and Pl-3-63068.

psid Acceptance Criteria Results Wide Range Containment Pressure indication difference D SAT D UNSAT is less than 7.5 psid.

End of Section 4.4

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 8 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.5 Containment Pressure (Narrow Range) Channel Check

1. RECORD Channel A Pl-3-6425A.

psig

2. RECORD Channel 8 Pl-3-64258.

psig

3. CALCULATE the difference between Pl-3-6425A and Pl-3-64258.

psid Acceptance Criteria Results Narrow Range Containment Pressure indication D SAT D UNSAT difference is less than 1.0 psid.

End of Section 4.5

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 9 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.6 Containment Water Level Monitor (Wide Range) Channel Check

1. RECORD Channel A Ll-3-6309A.

inches

2. RECORD Channel 8 Ll-3-63098.

inches

3. CALCULATE the difference between Ll-3-6309A and Ll-3-63098.

inches Acceptance Criteria Results Wide Range Containment Water Level indication D SAT D UNSAT difference less than 3.0 inches.

End of Section 4.6

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 10 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.7 Containment Water Level Monitor (Narrow Range) Channel Check

1. RECORD Channel A Ll-3-6308A.

inches

2. RECORD Channel 8 Ll-3-63088.

inches

3. CALCULATE the difference between Ll-3-6308A and Ll-3-63088.

inches Acceptance Criteria Results Narrow Range Containment Water Level indication D SAT D UNSAT difference less than 15.0 inches.

End of Section 4. 7

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 11 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL Containment High Range Area Radiation Monitoring (CHRRMS)

Channel Check The following will cause the DCS DEGRADED CONTAINMENT (NOT EQ)

CONDITION to ALARM.

TURN AND HOLD Channel A Containment High Range Area Radiation Monitoring Function Selector switch in CHECK.

WHEN Trip 2 Alarm actuates AND indication stabilizes, THEN RECORD Channel A data in Table 1.

Table 1 CHRRMS Channel Check Channel A and B Channel A ChannelB Difference Rl-3-6311A (QR81) Rl-3-6311 B (QR82)

?JE1- ~E1-Rl-3-6311A (VPC) 3.5~-=t Rl-3-6311 B (VPC) Cf ET RAR-3-6311 A 5.-+E:r RAR-3-6311 B 4*3t"T

,@ RELEASE Function Selector switch. 7 JJ ENSURE Function Selector switch in OPERATE. '>

~ CHECK Operate Light ON. $

0 TURN AND HOLD Channel B Containment High Range Area Radiation Monitoring Function Selector switch in CHECK.

g WHEN Trip 2 Alarm actuates AND indication stabilizes, THEN RECORD Channel B data in Table 1. ~

f3J RELEASE Function Selector switch. s

g ENSURE Function Selector switch in OPERATE. 5

,@/ CHECK Operate Light ON. s ri9 PRESS Trip 1 and Trip 2 Alarm pushbuttons to RESET the alarms.

@ AND the difference between Channel Aand Channel B 1.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 12 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.8 Containment High Range Area Radiation Monitoring (CHRRMS)

Channel Check (continued)

12. (continued)

Acceptance Criteria Results Containment High Range Area Radiation Monitoring D SAT D UNSAT indication difference less than 0.5 decades at each location.

13. RESET DCS DEGRADED CONTAINMENT (NOT EQ) CONDITION by touching the Operator Reset of Degraded Containment - RESET button on the Degraded Containment screen on the DCS.
14. IF CHRRMS channel check is UNSAT, THEN GO TO 3-0NOP-094, Alternate Methods for Containment Post Accident Monitoring.

End of Section 4.8

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 13 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.9 Subcooling Margin Monitor Channel Checks NOTE

  • The minimum inputs for an OPERABLE Subcooling Margin Monitor (SMM)

Channel consists of one RCS pressure input, two hot leg temperature inputs, and two cold leg temperature inputs. The loss of any input will cause the SMM display to indicate POOR with black numbering on cyan background; however, the SMM calculated value is valid.

  • The Subcooling Margin Monitor remains OPERABLE with Temporary System Alteration installed on a temperature channel provided the minimum two of three leg temperature requirement is met.
1. CHECK Reactor Coolant Pumps in one of the following lineups:
  • Both RCP A and B running
  • Both RCP A and B shutdown
2. RECORD the Channel A and B Data from QSPDS in Table 2.

Table 2 Subcooling Margin Monitor Inputs Channel Parameter Channel A ChannelB Difference Reactor Coolant System Pressure (psig)

Hot Leg A Temperature (°F)

Hot Leg B Temperature (°F)

Hot Leg C Temperature (°F)

Cold Leg A Temperature (°F)

Cold Leg B Temperature (°F)

Cold Leg C Temperature (°F)

RCS Saturation Margin from the Core Summary page (°F)

Channel Difference for each set of parameters in Table 2.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 14 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 4.9 Subcooling Margin Monitor Channel Checks (continued)

3. (continued)

Acceptance Criteria Results RCS Pressure indication difference less than 42 psid. D SAT D UNSAT RCS Hot Leg Temperature indication difference less than D SAT D UNSAT 9.0°F.

RCS Cold Leg Temperature indication difference less D SAT D UNSAT than 9.0°F.

RCS Saturation Margin indication difference less than D SAT D UNSAT 5.0°F.

End of Section 4.9

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 15 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.10 Core Exit Thermocouples Channel Check

1. IF reactor power is less than or equal to 2%, THEN CALCULATE Avg T hot for Channels A and B as follows:

A. CALCULATE Channel A Average T hot*

+ + =

CH A Hot Leg CH A Hot Leg CH A Hot Leg Sum CH A A Temp B Temp CTemp Temp Table 2 Table 2 Table 2

=

Sum CH A Number of Avg CH A Thot Temp Loops Temp B. CALCULATE Channel B Average T hot*

+ + =

CH B Hot Leg CH B Hot Leg CH B Hot Leg Sum CH B A Temp B Temp C Temp Temp Table 2 Table 2 Table 2

=

Sum CH B Number of Avg CH B That Temp Loops Temp

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 16 of 33 PROCEDURE NO.;

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.10 Core Exit Thermocouples Channel Check (continued)

NOTE Those CETs that have XXXX entered into their data field below have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7.1.

2. RECORD Quadrant One CET Temperatures in Table 3:

Table 3 CET Quadrant 1 Temperatures Channel A ChannelB P7 R7 N10 P8 xx.xx N8 N6 L6 N4 K8 M11 xxxx M9 I xxxx L8 Functional Criteria Results

  • If Reactor Power is:::; 2%, then the CET reading is D SAT D UNSAT

+/- 15°F of its Channel Average T hot calculated in Section 4.10, Step 1.

  • If Reactor Power is > 2%, then the CET readings are 545°F to 650°F.

Acceptance Criteria Results At least 4 CETs in Quadrant 1 were determined to be D SAT D UNSAT functional.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 17 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.10 Core Exit Thermocouples Channel Check (continued)

NOTE Those CETs that have XXXX entered into their data field below have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7.1.

3. RECORD Quadrant Two CET Temperatures in Table 4:

Table 4 CET Quadrant 2 Temperatures Channel A ChannelB M3 KS HS K3 xxxx H3 J2 G2 G6 E4 G1 03 FS F3 Functional Criteria Results

  • If Reactor Power is~ 2%, then the CET reading is D SAT D UNSAT

+/- 1S°F of its Channel Average Thot calculated in Section 4.10, Step 1.

  • If Reactor Power is> 2%, then the CET readings are S4S°F to 6S0°F.

Acceptance Criteria Results At least 4 CETs in Quadrant 2 were determined to be D SAT D UNSAT functional.

REVISION NO.: PROCEDURE TITLE: PAGE:*

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 18 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.10 Core Exit Thermocouples Channel Check (continued)

NOTE Those CETs that have XXXX entered into their data field below have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7.1.

4. RECORD Quadrant Three CET Temperatures in Table 5:

Table 5 CET Quadrant 3 Temperatures Channel A ChannelB GB HB E10 F9 xxxx E7 EB 05 xxxx 810 C12 85 CB AB xxxx Functional Criteria Results

  • If Reactor Power is s 2%, then the CET reading is D SAT D UNSAT

+/- 15°F of its Channel Average T hot calculated in Section 4.10, Step 1.

  • If Reactor Power is> 2%, then the CET readings are 545°F to 650°F.

Acceptance Criteria Results At least 4 CETs in Quadrant 3 were determined to be D SAT D UNSAT functional.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 19 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.1 O Core Exit Thermocouples Channel Check (continued)

5. RECORD Quadrant Four CET Temperatures in Table 6:

Table 6 CET Quadrant 4 Temperatures Channel A ChannelB L14 K11 L12 H15 J12 H13 J10 H9 H11 E14 G15 E12 F13 F11 Functional Criteria Results

  • If Reactor Power is~ 2%, then the CET reading is D SAT D UNSAT

+/- 15°F of its Channel Average T hot calculated in '

Section 4.10, Step 1.

  • If Reactor Power is> 2%, then the CET readings are 545°F to 650°F.

Acceptance Criteria Results At least 4 CETs in Quadrant 4 were determined to be D SAT D UNSAT functional.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 20 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.10 Core Exit Thermocouples Channel Check (continued)

NOTE Attachment 2, Temperature Element Descriptions and Locations A Train QSPDS and Attachment 3, Temperature Element Descriptions and Locations B Train QSPDS provide a cross reference from temperature element core location to the specific temperature element designation. The temperature element designation is required to properly identify the component on any PWOs generated to address their failure.

6. ENSURE fill out of service CETs are documented per Attachment 1, QSPDS CET/RVLIS Sensors Out-of-Service.
7. IF Reactor Power is greater than 2%, THEN RECORD the following representative CET temperatures:

Channel A Channel B OF Acceptance Criteria Results If Reactor Power is greater than 2%, representative CET Channel A temperatures are between 545°F and 650°F. o SAT o UNSAT ChannelB D SAT D UNSAT

8. RECORD the following CET Subcooling temperatures:

Channel A OF ChannelB Reactor Power  %

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 21 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 4.10 Core Exit Thermocouples Channel Check (continued)

8. (continued)

Acceptance Criteria Results GET Subcooling Temperatures are within the Channel A temperature range below for the current reactor power:

D SAT D UNSAT 100% steady state 0°F - 30°F ChannelB D SAT D UNSAT

> 50% - < 100% 0°F - 60°F

>2 % - 50 % 45°F - 115°F

~2% N/A End of Section 4.10

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 22 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.11 Reactor Vessel Level Monitoring System Channel Checks NOTE Those sensors that have XXXX entered into their data field below have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7 .1.

1. RECORD the Channel A Thermocouple Temperatures in Table 7.

Table 7 Channel A Thermocouple Temperatures TH Heated Tu UnHeated AT Differential Location Thermocouple (°F) Thermocouple (°F) Temperature (°F)

NOTE 1: If heated thermocouple #1 is bypassed, AT value is calculated using the #2 heated thermocouple and the #1 unheated thermocouple. The #2 level sensor is OPERABLE and the #1 level sensor is inoperable.

(Section 8.1.2, Developmental 4.K)

(NOTE 1)

  1. 1 Head
  1. 2 Head NOTE 2: AT is calculated using the #5 heated thermocouple and the #5 unheated thermocouple. Following Refueling Outage PT3-26, Sensors #4, #6, and #7 will be INOPERABLE. (Section 8.1.2, Developmental 4.L)
  1. 3 Plenum
  1. 4 Plenum xxxx xxxx )()()()( (NOTE2)
  1. 5 Plenum
  1. 6 Plenum xxxx xxxx )()()()( NOTE2)
  1. 7 Plenum xxxx xx.xx )()()(X (NOTE 2)
  1. 8 Plenum

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 23 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.11 Reactor Vessel Level Monitoring System Channel Checks (continued)

2. RECORD Channel B Thermocouple temperatures in Table 8:

Table 8 Channel B Thermocouple Temperatures Location TH Heated Tu UnHeated L\ T Differential Thermocouple (°F) Thermocouple (°F) Temperature (°F)

  1. 1 Head
  1. 2 Head
  1. 3 Plenum
  1. 4 Plenum xxxx xxxx I
  1. 5 Plenum
  1. 6 Plenum
  1. 7 Plenum
  1. 8 Plenum xxxx xxxx NOTE Attachment 2, Temperature Element Descriptions and Locations A Train QSPDS and Attachment 3, Temperature Element Descriptions and Locations B Train QSPDS, provide a cross reference from temperature element core location to the specific temperature element designation. The temperature element designation is required to properly identify the component on any PWOs generated to address their failure.
3. ENSURE .§..U Out-of-Service RVLIS Sensors are documented per Attachment 1, QSPDS CET/RVLIS Sensors Out-of-Service.

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9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 24 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 INITIAL 4.11 Reactor Vessel Level Monitoring System Channel Checks (continued)

3. (continued)

Acceptance Criteria Results At least four Heated Junction Thermocouple (HJTC) Channel A sensor locations per channel display a differential D SAT D UNSAT temperature between 50°F and 120°F.

Channels D SAT D UNSAT

4. IF a QSPDS channel is UNSAT, THEN:

A. DECLARE the channel inoperable.

B. GO TO 3-0NOP-094, Alternate Methods for Containment Post Accident Monitoring.

End of Section 4.11 4.12 QSPDS Health Check

1. At each QSPDS panel, PERFORM the following:

A. CHECK date and time updating on the upper right of the screen.

8. CHECK System Health alarm green.

Acceptance Criteria Results All conditions checked above are satisfied. D SAT D UNSAT End of Section 4.12

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 25 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 5.0 RESTORATION AND DOCUMENTATION 5.1 Restoration None 5.2 Documentation

1. Acceptance Criteria D SAT D UNSAT
2. Functional Criteria D SAT D UNSAT Remarks:

Performed By:

(Signature) (Print) (I nit) (Date)

Reviewed By:

(Supervisor) (Print) (Date)

App(oved By:

(Shift Manager or SRO Designee) (Print) (Date)

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 26 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 6.0 ACCEPTANCE AND FUNCTIONAL CRITERIA 6.1 Acceptance Criteria All equipment has been demonstrated to be in the appropriate condition, to satisfy applicable channel checks and is in sufficient quantity that all channels are considered operable.

6.2 Functional Criteria None 7.0 RECORDS

1. The date, time, and section completed shall be entered in the Unit Narrative Log.
2. Problems encountered while performing the procedure (i.e., malfunctioning equipment, delays due to change in plant conditions, etc.) should be entered in the Unit Narrative Log.
3. Completed copies of the below listed items constitute Quality Assurance Records and shall be transmitted to QA Records.
  • Section 3.0
  • Section 4.0
  • Section 5.0

REVISION NO.: PROCEDURE TITLE:* PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 27 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3

8.0 REFERENCES

AND COMMITMENTS 8.1 References 8.1.1 Implementing

1. 3-0NOP-094, Alternate Methods for Post Accident Monitoring
2. 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations
3. ODl-C0-016, Operations Department Instruction 8.1.2 Developmental
1. Technical Specifications A. Section 3/4.3.3.3 Accident Monitoring Instrumentation and Table 4.3-4 Accident Monitoring Instrumentation Surveillance Requirements Items 1-5, 8-14, and 16.
2. FSAR A. Section 7.0, Instrumentation and Controls
3. Regulatory Guidelines A. Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Condition During and Following an Accident
8. NU REG 0737, Clarification of TMI Action Plan Requirements
4. Miscellaneous Documents (i.e., PC/Ms, ECs, Correspondence)

A. PTN-TECH-95-016

8. PTN-TECH-85-001 C. PTN-TECH-90-329 D. PTN-TECH 94-185 E. PC/M 92-031, Annunciator Window Engraving Drawings for Human Factors Engineering PC/M 95-139, RVLIS Sensor Abandonment

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 28 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 8.1.2 Developmental (continued)

4. (continued)

H. PC/M 00-016, Generic Package for Modifications I. PC/M 04-112, Emergency Response Data Acquisition and Display (ERDADS) Replacement J. PC/M 04-048, Qualified Safety Parameter Display System (QSPDS) Replacement K. CR 2008-38769 CA #1 L. CR 2008-38043 CA #6 M. SPEC-IC-007 8.1.3 Management Directives None 8.2 Commitments None

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 29 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 ATTACHMENT 1 QSPDS CET/RVLIS Sensors Out-of-Service (Page 1 of 1)

1. When a CET or RVLIS indication is identified as failed:
  • INITIATE a trouble and breakdown PWO assigned to l&C .
  • PLACE a Control Room green tag on the QSPDS display per ODl-C0-016, Operations Department Instruction.
2. IF the problem appears to be inside containment, THEN:
  • PLACE the PWO in the SNO file .
  • MOVE the green tag to the QSPDS cabinet in the Computer Room.
  • CHANGE the Control Room green tag designation per ODl-C0-016.
  • LIST the component in the Out-of-Service Logbook as an outage item.
3. IF it is determined that the point will NOT be repaired, THEN:
  • CLOSE the PWO .
  • PERFORM 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7.1 to bypass the point.
  • REMOVE the point from the OOS Logbook .
  • DOCUMENT bypassed points in 3-0SP-204, Accident Monitoring Instrumentation Channel Checks, under an AR-PCR.
4. WHEN it is determined that points need to be restored, THEN:
  • INITIATE a PWO .

SCHEDULE this PWO for the next refueling outage.

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 30 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 ATTACHMENT 2 Tem12erature Element Descri12tions and Locations A Train QSPDS (Page 1 of 2)

NOTE Instruments listed that are grayed out have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7 .1.

Location Temperature Element Description A-8 TE-3-1 E CET for A QSPDS LOC A-8 C-8 TE-3-27E CET for A QSPDS LOC C-8 C-12 TE-3-28E CET for A QSPDS LOC C-12 D-3 TE-3-29E CET for A QSPDS LOC D-3 D-5 TE-3-30E CET for A QSPDS LOC D-5 E-4 TE-3-4E CET for A QSPDS LOC E-4 E-10 TE-3-32E CET for A QSPDS LOC E-10 F-11 TE-3-11 E CET for A QSPDS LOC F-11 F-13 TE-3-33E CET for A QSPDS LOC F-13 G-2 TE-3-34E CET for A QSPDS LOC G-2 G-8 TE-3-35E CET for A QSPDS LOC G-8 G-15 TE-3-36E CET for A QSPDS LOC G-15 H-3 TE-3-37E CET for A QSPDS LOC H-3 H-5 TE-3-38E CET for A QSPDS LOC H-5 H-11 TE-3-40E CET for A QSPDS LOC H-11 J-10 TE-3-17E CET for A QSPDS LOC J-10 J-12 TE-3-18E CET for A QSPDS LOC J-12

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 31 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 ATTACHMENT 2 Tem12erature Element DescriQtions and Locations A Train QSPDS (Page 2 of 2)

Location Temperature Element Description K-8 TE-3-21 E CET for A QSPDS LOC K-8 L-6 TE-3-42E CET for A QSPDS LOC L-6 L-12 TE-3-44E CET for A QSPDS LOC L-12 L-14 TE-3-45E CET for A QSPDS LOC L-14 M-3 TE-3-46E CET for A QSPDS LOC M-3 N-8 TE-3-49E CET for A QSPDS LOC N-8 N-10 TE-3-50E CET for A QSPDS LOC N-10 E-7 TE-3-5E CET for A QSPDS LOC E-7 P-7 TE-3-51 E CET for A QSPDS LOC P-7

  1. 1 HD TE-3-6493 ICCS THMCPL for A QSPDS #1 HEAD
  1. 2 HD TE-3-6494 ICCS THMCPL for A QSPDS #2 HEAD
  1. 3 PL TE-3-6495 ICCS THMCPL for A QSPDS PLENUM #3
  1. 4 PL TE-3-6496 ICCS THMCPL for A QSPDS PLENUM #4
  1. 5 PL TE-3-6497 ICCS THMCPL for A QSPDS PLENUM #5
  1. 6 PL TE-3-6498 ICCS THMCPL for A QSPDS PLENUM #6
  1. 7 PL TE-3-6499 ICCS THMCPLforA QSPDSPLENUM #7
  1. 8 PL TE-3-6500 ICCS THMCPL for A QSPDS PLENUM #8

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 32 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 ATTACHMENT 3 Tem12erature Element Descri12tions and Locations B Train QSPDS (Page 1 of 2)

NOTE Instruments listed that are grayed out have been bypassed by the QSPDS software in accordance with procedure, 3-0P-042.1, QSPDS - Inadequate Core Cooling Monitor Infrequent Operations, Subsection 7.1.

Location Temperature Element Description B-5 TE-3-2E CET for B QSPDS LOC B-5 B-10 TE-3-3E CET for B QSPDS LOC B-10 E-8 TE-3-31 E CET for B QSPDS LOC E-8 E-12 TE-3-6E CET for B QSPDS LOC E-10 E-14 TE-3-7E CET for B QSPDS LOC E-14 F-3 TE-3-8E CET for B QSPDS LOC F-3 F-5 TE-3-9E CET for B QSPDS LOC F-5 F-9 TE-3-10E CET for B QSPDS LOC F-9 G-1 TE-3-12E CET for B QSPDS LOC G-1 G-6 TE-3-13E CET for B QSPDS LOC G-6 H-8 TE-3-14E CET for B QSPDS LOC H-8 H-9 TE-3-39E CET for B QSPDS LOC H-9 H-13 TE-3-41 E CET for B QSPDS LOC H-13 H-15 TE-3-15E CET for B QSPDS LOC H-15 J-2 TE-3-16E CET for B QSPDS LOC J-2 K-3 TE-3-19E CET for B QSPDS LOC K-3 K-5 TE-3-20E CET for B QSPDS LOC K-5

REVISION NO.: PROCEDURE TITLE: PAGE:

9 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECKS 33 of 33 PROCEDURE NO.:

3-0SP-204 TURKEY POINT UNIT 3 ATTACHMENT 3 Teml2erature Element DescriQtions and Locations 8 Train QSPDS (Page 2 of 2)

Location Temperature Element Description K-11 TE-3-22E CET for B QSPDS LOC K-11 L-8 TE-3-43E CET for B QSPDS LOC L-8 M-9 TE-3-23E CET for B QSPDS LOC M-9 M-11 TE-3-47E CET for B QSPDS LOC M-11 N-4 TE-3-48E CET for B QSPDS LOC N-4 N-6 TE-3-24E CET for B QSPDS LOC N-6 P-8 TE-3-25E CET for B QSPDS LOC P-8 R-7 TE-3-26E CET for B QSPDS LOC R-7

  1. 1 HD TE-3-6501 ICCS THMCPL for B QSPDS #1 HEAD
  1. 2 HD TE-3-6502 ICCS THMCPL for B QSPDS #2 HEAD
  1. 3 PL TE-3-6503 ICCS THMCPL for B QSPDS PLENUM #3
  1. 4 PL TE-3-6504 ICCS THMCPL for B QSPDS PLENUM #4
  1. 5 PL TE-3-6505 ICCS THMCPL for B QSPDS PLENUM #5
  1. 6 PL TE-3-6506 ICCSTHMCPL for B QSPDS PLENUM #6 I
  1. 7 PL TE-3-6507 ICCS THMCPL for B QSPDS PLENUM #7
  1. 8 PL TE-3-6508 ICCS THMCPL for B QSPDS PLENUM #8 I

JOB PERFORMANCE M JPM Page 1 of 1O JPM TITLE: Perform Reactor Coolant System Leak Rate Calculation - Manual Method JPM NUMBER: 02041036102 REV. 1-0 TASK NUMBER(S) I 02041036100/

TASK TITLE(S): Perform RCS Leak Rate Calculation KIA NUMBERS: 2.1.23 KIA VALUE: RO 4.3 I SRO 4.4 Justification (FOR KIA VALUES <3.0): N/A TASK APPLICABILITY:

~RO ~SRO ~ STA 0 Non-Lie ~SRO CERT 0 OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: Perform: [2J EVALUATION LOCATION: In-Plant: D Control Room: D Simulator: D Classroom: x I Lab: D Other: I Time for Completion: 30 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No Date Training Program Owner Date TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.b/RO,SRO

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 2 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.b/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 3 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE Updated to fleet template; Updated for L-15-1 See cover page N/A 1-0 01982463 text/grammar changes NRC Exam See cover page N/A 1-1 1-2 1-3 1-4 1-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.b/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 4 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION SIMULATOR SET-UP: N/A Required Materials:

  • Calculator General

References:

  • Demonstrate ability to perform specific system and integrated plant procedures during all modes of plant operation
  • Calculate RCS leak rate using 4-OSP-041.1, determine that the associated acceptance criteria are not met, and identify the required Technical Specification actions TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.b/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 5 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • Unit 4 is at 100% steady-state power and all prerequisites in 0-OSP-041.1 (Reactor Coolant System Leak Rate Calculation) have been met.
  • The RCS Leak Rate Calculation function in DCS is unavailable.
  • The combined Charging Pump primary seal leakage, as determined by Attachment 4 of 0-OSP-041.1, is 0.05 gpm; Non-Reactor Coolant Pressure Boundary (Non-RCPB) leakage, as determined by Attachment 6 of 0-OSP-041.1, is 0 gpm; primary-to-secondary leakage is 0 gpm.
  • CV-4-2821/2822 (Containment Sump Pump Discharge Valves) and LC-4-112A (VCT Level Controller) have been restored to normal, per 0-OSP-041.1.

Initiating Cue:

  • You are directed to perform a manual RCS leak rate calculation, using Attachment 3 of 0-OSP-041.1, given the following start/stop data:

Start Data Stop Data Parameter Value Parameter Value Time 1900 Time 2300 VCT Level 50% VCT Level 26%

Pressurizer Level 57% Pressurizer Level 57%

Primary Water Totalizer 0 Primary Water Totalizer 0 Boric Acid Totalizer 0 Boric Acid Totalizer 0 PRT Level 71% PRT Level 71%

RCDT Level 15% RCDT Level 28%

Containment Sump Level 120 gallons Containment Sump Level 120 gallons Tavg 580°F Tavg 580°F

  • Upon completing the calculation, determine whether the associated acceptance criteria are met

[RO,SRO] and identify the required Technical Specification actions, if any [SRO only].

Acceptance Criteria: Met ________ Not Met ________

Technical Specification action(s): ___________________________________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.b/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 6 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

Provide examinee with a copy of 4-OSP-041.1, Reactor Coolant System Evaluator Cue:

Leak Rate Calculation.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Performance Step: 2 Record leak rate start and stop data.

Critical: No Standard: Record leak rate start and stop data on table in Attachment 3 of 4-OSP-041.1.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 7 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Performance Step: 3 Determine RCS leak rate.

Critical: Yes Perform RCS leak rate calculation and determine reference leakage, per Attachment 3 of 4-OSP-041.1:

Standard:

  • Gross RCS leakage is 1.417 gpm (1.40 gpm to 1.43 gpm)
  • Identified leakage is 0.175 gpm (0.17 gpm to 0.18 gpm)

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Performance Step: 4 Determine whether or not leak rate acceptance criteria are met.

Critical: Yes Determine that Unidentified RCS Leakage criterion is not met (i.e., 1.192 Standard: gpm exceeds 1 gpm) and check Unsatisfactory box on Page 2 of Attachment 3.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 8 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Performance Step: 5

[SRO only] Determine any required Technical Specification actions.

Critical: Yes Determine that Technical Specification LCO 3.4.6.2.b applies and the required action is to reduce the leakage rate to within limits within four Standard:

hours or be in at least hot standby within the next six hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 5, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

01041036102, Perform RCS Leak Rate Calculation - Manual Method, Rev. 1 JPM Page 9 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

  • Unit 4 is at 100% steady-state power and all prerequisites in 0-OSP-041.1 (Reactor Coolant System Leak Rate Calculation) have been met.
  • The RCS Leak Rate Calculation function in DCS is unavailable.
  • The combined Charging Pump primary seal leakage, as determined by Attachment 4 of 0-OSP-041.1, is 0.05 gpm; Non-Reactor Coolant Pressure Boundary (Non-RCPB) leakage, as determined by Attachment 6 of 0-OSP-041.1, is 0 gpm; primary-to-secondary leakage is 0 gpm.
  • CV-4-2821/2822 (Containment Sump Pump Discharge Valves) and LC-4-112A (VCT Level Controller) have been restored to normal, per 0-OSP-041.1.

Initiating Cue:

  • You are directed to perform a manual RCS leak rate calculation, using Attachment 3 of 0-OSP-041.1, given the following start/stop data:

Start Data Stop Data Parameter Value Parameter Value Time 1900 Time 2300 VCT Level 50% VCT Level 26%

Pressurizer Level 57% Pressurizer Level 57%

Primary Water Totalizer 0 Primary Water Totalizer 0 Boric Acid Totalizer 0 Boric Acid Totalizer 0 PRT Level 71% PRT Level 71%

RCDT Level 15% RCDT Level 28%

Containment Sump Level 120 gallons Containment Sump Level 120 gallons Tavg 580°F Tavg 580°F

  • Upon completing the calculation, determine whether the associated acceptance criteria are met [RO,SRO] and identify the required Technical Specification actions, if any [SRO only].

Acceptance Criteria: Met ________ Not Met ________

Technical Specification action(s): __________________________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

Florida Power Light Company Turkey Point Nuclear Plant Unit 4 4-0SP-041.1

Title:

Reactor Coolant System Leak Rate Calculation Continuous Use.

Safety Related Procedure Responsible Department: Operations Revision Number 10A Issue Date 9/10/14 Revision Approval Date: 9/10/14 PCRs 08-1840, 08-2314, 08-4445, 08-6011, 08-5629, 09-3126, 10-0835, 569716, 1612830, 1666984, 1688327, 1680161, 1656821, 1697065, 1823275, 1920685, 1945723, 1952749, 1962952, 1965059, 1964686, 1969334, 1990194 PC/Ms90-024, 06-049,04-113, 06-031 W20 iO: JC/els/els/els

Procedure No.: Procedure

Title:

Page:

3 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 TABLE OF CONTENTS Section 1.0 PURPOSE..................................................................................................................... 4

2.0 REFERENCES

/RECORDS REQUIRED/

COMMITMENT DOCUMENTS............................................................................... 4 3.0 PREREQUISITES....................................................................................................... 6 4.0 PRECAUTIONS/LIMITATIONS.............................................................................. 7 5.0 SPECIAL TOOLS/EQUIPMENT.............................................................................. 10 6.0 ACCEPTANCE CRITERIA....................................................................................... 10 7.0 PROCEDURE 7 .1 RCS Leak Rate Calculation .... ..... .. ....... ..... .. ............. ...... .. ... .... ... ...... ........... ...... 11 ENCLOSURE/ATTACHMENTS Enclosure 1 TAvG Conversion Factor vs. TAvG...................................................................... 21 Attachment 1 Leak Rate Calculation Data Sheet (ERDADS Method).................................... 24 Attachment 2 Leak Rate Calculation Data Sheet (PC Leak Rate Method) .. . .. .. .. .. ... .. .. .. .. .. ... .. . 26 Attachment 3 Leak Rate Calculation Data Sheet (Manual Method) .. .. .. .. .. .. .. .. .. .. .. .. .. ... .. .. .. ... .. 28 Attachment 4 Determination of Charging Pump Primary and Secondary Packing Leakage....................................................................... 30 Attachment 5 Reactor Coolant System Leak Investigation Guideline..................................... 34 Attachment 6 Non-RCPB Leakage (Other Than Charging Pump Packing Leakage).............. 37 Attachment 7 RCS Leak Validity and Statistical Calculator Desktop Instructions ....... ..

W2010:J /els/els/els

Procedure No.: Procedure

Title:

Page:

4 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 1.0 PURPOSE 1.1 Provide instructions to calculate RCS leak rate by using the DCS, a PC computer, or manual method. This procedure meets the requirements of References 2.1.1 and 2.1.2.

2.0 REFERENCES

/RECORDS REQUIRED/COMMITMENT DOCUMENTS 2 .1 References 2.1.1 Technical Specifications

1. 3.4.6.2, RCS Operational Leakage
2. 4.4.6.2.1.c, RCS Water Inventory Balance
3. Table 3.3-4, Action 26
4. 3.4.6.1, Action a.3
5. 4.4.6.2.1.e, Primary to Secondary Leakage 2.1.2 FSAR
1. Section 4.1.3, Principal Design Criteria
2. Section 4.2.7, Leakage
3. Section 6.5, Leakage Detection and Provision for the Reactor and Auxiliary Coolant Loops
4. Section 6.6, Containment Isolation
5. Table 6.6-1, Containment Piping Penetrations and Valving
6. Section 9.8, Auxiliary Building Ventilation and Containment Purge Systems
7. Section 9.9, Control Room Ventilation System 01 O:JC/cls/cls/cls

Procedure No.: Procedure

Title:

Page:

5 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 8/20/14 2.1.3 Plant Procedures/Fleet Procedures

1. O-ADM-009, Containment Entries When Containment Integrity is Established
2. O-ADM-401, Engineering Process Monitoring Software Control
3. O-ADM-737, Post Maintenance Testing
4. 4-NOP-061.03, Reactor Coolant Drain Tank
5. 4-0SP-041.2, Reactor Coolant System Visual Leak Inspection and Leak Evaluation
6. 4-0SP-053 .4, Containment Integrity Penetration Alignment Verification
7. O-OSP-095.7, ERDADS Leak Rate Verification
8. O-OSP-102.9, Verification of PC Leak Rate Program
9. MA-AA-100-1011, Equipment Troubleshooting
10. OP-AA-100-1000, Conduct of Operations 2.1.4 Miscellaneous Documents (i.e., PC/M, ECs, Correspondence)
1. 10 CFR 50, Appendix A, Criterion 30
2. NUREG-1107, RCS Lkg: Reactor Coolant System Leak Rate Determination for PWRs
3. NRC Information Notice 94-46 Nonconservative RCS Leakage Calculation, Dated 6/20/94
4. PC/M 90-024, Containment Sump Level Setpoint Change
5. PC/M 06-031, Containment Recirculation Sump Debris GSI-191 Resolution
6. PC/M 04-113, Emergency Response Data Acquisition and Display System (ERDADS) Replacement
7. WCAP-16423-NP, Pressurized Water Reactor Owners Group Standard Process and Methods for Calculating RCS Leak Rate for Pressurized Water Reactors
8. WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors
9. CR 1963101, RCS Leak Rate Procedure Issues W2010:JC/cls/cls/ Is

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 2 .2 Records Required 2.2.1 The date, time, and section completed shall be entered in the Unit Narrative Log.

Also, problems encountered while performing the procedure should be entered; i.e., malfunctioning equipment, delays due to change in plant conditions, etc.

2.2.2 Completed copies of the QA Record Pages for the below listed items document compliance with Technical Specification surveillance requirements and shall be sent to QA Records for retention in accordance with Quality Assurance Records Program requirements:

1. Attachment 1, with DCS RCS Leak Rate Calculation printout attached
2.
  • Attachment 2, with Computer printout attached
3. Attachment 3
4. Attachment 4
5. Attachment 5
6. Attachment 6 2.3 Commitment Documents 2.3 .1 L-87-186, Instrumentation Port Column Assembly Leakage 2.3.2 PTN-SEG-87-41, Recommendations for Strengthening Leak Detection, Evaluation and Repair Process 2.3.3 PTN-PMN-87-545, Calculating RCS Leak Rate, (CAR-87-064) 2.3.4 PTN-PMM-88-351, Quality Assurance Audit QAO-PTN-88-909 2.3.5 PTN-ENG-LRAM-00-0028, Boric Acid Wastage Surveillance Program-License Renewal Basis Document 2.3.6 PTN-ENG-LRAM-00-0035, Steam Generator Integrity Program-License Renewal Basis Document 2.3.7 PTN-ENG-SESJ-05-041, Rev. 2, Primary to Secondary Leak Rate Tech Spec Changes 3.0 PREREQUISITES 3.1 The reactor is in Mode 1, 2, 3, or 4.

3.2 Reactor power, TAvG, and pressurizer level are stable.

3.3 No load adjustments are anticipated during the test. [Commitment - Step 2.3.3]

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 612114 4.0 PRECAUTIONS/LIMITATIONS 4.1 Changes made to parameters in Attachment 1, 2, or 3 may cause test inaccuracies.

4.2 Data shall be collected automatically by the DCS RCS Leak Rate Calculation Program or, if DCS is unavailable, manually from the corresponding listed meter.

4.3 The DCS RCS Leak Rate Calculation can be performed using Operator Workstations AW0401 or A W0402.

The DCS RCS Leak Rate Calculation requires selection of the calculation duration.

Available durations are 15 minutes, 30 minutes, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (normal). STOP time is the end time of the calculation. START time is STOP time minus the selected calculation duration.

4.4 An effective RCS leakage monitoring program should have at least three dissimilar, diverse and independent methods of monitoring leakage. (REF 2.1.4.8) 4.5 Programs used to calculate RCS leakage are maintained by Engineering Department using O-ADM-401, Engineering Process Monitoring Software Control.

4.6 Test completion is required:

  • Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady-state operation (average coolant temperature being changed less than 5°F/hour)
  • At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with RCS average coolant temperature being changed by less than 5°FI hour.

Testing may be extended up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> between successive test completion if changing plant conditions prevent steady-state operation.

[Commitment- Steps 2.3.5 and 2.3.6.]

4.7 Gross RCS leakage is the sum of Identified Leakage, Unidentified Leakage, and Non-RCPB Leakage. Negative values mask actual RCS leakage.

4.8 A low Unidentified RCS leak rate may be caused by leakage into the RCDT and PRT from sources other than the RCS (accumulator drains CV-852A, B, C; RV-859; RV-6511; MOV-744A and B stem leakoff [when RHR is isolated]; RV-706 [when RHR is isolated]).

4.9 The Identified leakage along with the total Non-RCPB Leakage are subtracted from Gross leakage to determine the unidentified leakage. Investigation should attempt to quantify and eliminate these Non-RCS sources from being included as RCS leakage (i.e., check accumulator levels, etc.) (NRC Information Notice 94-46).

4.10 Leakage results less than zero (negative) are recorded as a value of zero. (Negative leak rate results can be created when there are unaccounted for changes in the parameters used in the calculation. These may include temperature fluctuations of contained water volumes and in-leakage from unknown sources.)

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 4.11 RCS leakage investigation may stop when the source is found. For example, a Level 3 leak investigation requires a Level 2 investigation, which requires a Level 1 investigation.

If the source of the leakage is identified during the Level 1 investigation, the Level 2 and Level 3 investigations are not required. However, this does not excuse performance of a leak investigation required by a subsequent RCS leak rate calculation.

4.12 PRT level transmitter LT-4-470 has a high temperature rating of 200°F thus, due to the close proximity to the tank, any time PRT temperature exceeds 200°F PRT level indication may not be accurate.

4.13 Leak rate calculations should not be performed while refilling the RWST, and selection of leak rate durations should not include any time while the R WST is being refilled. The leak rate includes primary water flow rate and boric acid flow rate as part of the calculation.

Thus, the calculation assumes the primary water is being injected into the RCS, when it is actually flowing into the RWST. This will cause an erroneous leak rate calculation.

4.14 Non-Reactor Coolant Pressure Boundary (Non-RCPB) Leakage is the sum of leakage that impacts the Reactor Coolant System gross leakage calculation, but occurs outside the RCPB and therefore should not be included in Identified Leakage or Unidentified Leakage.

In order to account for Non-RCPB Leakage, it must be known to exist at the time of the leak rate calculation, and be identified, quantified, and documented. An example of Non-RCPB Leakage may include:

  • Charging pump packing leakage (known, quantified, and documented)
  • Charging pump relief valve leakage (known, quantified, and documented)
  • Seal Injection filter drain leakage (known, quantified, and documented) 4.15 In order to take credit for the Non-Reactor Coolant Pressure Boundary Leakage, the leakage from components associated with this category must be validated during the same time frame that the RCS leak rate calculation is being performed. Otherwise, any Gross Leakage that is not Identified Leakage must be considered to be Unidentified Leakage.

4.16 The Reactor Coolant Pressure Boundary (RCPB) consists of all those pressure-containing components which are part of the Reactor Coolant System or which are connected to the Reactor Coolant System, up to and including:

4.16.1 The outermost containment isolation valve in system piping which penetrates primary reactor containment.

4.16.2 The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment.

4.17 If a charging pump has primary packing leakage greater than 0.05 gpm, charging pump primary packing leakage shall be measured during all RCS leak rate determinations.

4.18 Gross Leakage is the sum of all leakage from the RCS control volume during the leak rate test time interval, ~ Time.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 4.19 The RCS Control Volume includes:

  • Pressurizer, and
  • Volume Control Tank Gross leakage is calculated by measuring the change in level, temperature, and pressure conditions in each segment of the RCS Control Volume.

4.20 The Reactor Coolant Water Volume (RCWV) is that portion of the Reactor Coolant System where the effective bulk temperature is represented by TAVE* This includes the reactor coolant liquid mass contained within:

  • Hot leg piping,
  • Cold leg piping,
  • Reactor Vessel, and

4.22 Calc-rev05c.xls requires StatAnalysis0.3B.xls open to properly update and calculate leak rate.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 5.0 SPECIAL TOOLS/EQUIPMENT 5 .1 Preprogrammed computer (PC Leak Rate Method) 5.2 Calculator (Manual Method) 5.3 Graduated measuring vessel (Attachment 6) 5.4 Timing device (Attachment 6) 5.5 RCS Leak Rate Validity and Statistical Calculation Desktop (Attachment 7) 6.0 ACCEPTANCE CRITERIA 6.1 RCS pressure boundary leakage shall not exist.

6.2 Unidentified RCS leakage is less than 1 gpm.

6.3 Identified leakage from the RCS is less than 10 gpm.

6.4 Primary to Secondary Leakage is limited to 0.1 gpm (150 gallons/day) through any one steam generator.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.0 PROCEDURE 7 .1 RCS Leak Rate Calculation Date/Time Started: ~~~~~~~~~~~~~~~

I 7 .1.1 Initial Conditions

1. Applicable prerequisites listed in Section 3.0 are satisfied.
2. The Chemistry Hot Lab is notified that an RCS Leak Rate will be performed and that all RCS sampling shall be terminated for the test duration.
3. Changes to parameters in the associated leak rate attachment are NOT anticipated for the duration of the leak rate test.

7 .1.2 Procedure Steps r=== NOTES ------------

I The order of preference for the leak rate calculation method used is DCS, PC Leak Rate, and then Manual. [Commitment - Steps 2. 3. 5 and 2. 3. 6]

  • If any OCS RCS leak rate calculation input is NOT operating correctly, the RCS leak rate calculation may be affected without any indication of the malfunction. Incorrect operation may be indicated by poor or bad status, no change in the readings or by a lack of 4 decimal places on the RCS, VCT, and RCOT parameters.

I . For the purpose of conducting the RCS water inventory balance, refer to the Precautions/Limitations and the Technical Specifications for definitions of leakage.

1. IF calculating an DCS RCS leak rate for a specific time period within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an End time prior to the current time, THEN perform the following:
a. Evaluate for leak rate duration, Start time, and End time by performing the following:
  • Check with Chemistry Department personnel if and when RCS sample valves were open.
  • Check if and when RCDT pumped down.
  • Check if and when Containment Sump pumped down.
  • Check if and when any makeup to VCT occurred.
  • Check for stable RCS TAVG*

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.1.2.1 (Cont'd)

b. Determine a leak rate duration to be used based upon the above information and the following step.
c. Ensure that there is a valid Green light on the Leak Rate Evaluation program as outlined in Steps 5 though 8 of Attachment 7 AND note the start time when a valid Green Light is obtained.

,.------- NOTE I1 The leak rate start time should be the same time that is selected in the Leak Rate Evaluation program and in accordance with the above step.

L-----------------------1

d. Determine leak rate start and end times to be used.

Start Time:- - - - - End Time- - - - -

e. On ERDADS Operator Workstation AW0401 or A W0402, select the

<UTILITIES MENU> button.

f. Select <RCS LEAK RATE CALCULATION>.
g. Select <CONFIGURE RCS LEAK RATE CALCULATION>
h. Select <OPERA TOR ENTERED TIME> for Select Leak Rate Calculation End Time.
i. Enter a value from 00 to 23 for End Time HH (hour).
j. Enter a value from 00 to 59 for End Time MM (minutes).
k. Select desired duration button for Select Leak Rate Calculation Duration.
1. Select <RUN CALCULATION> button.
m. Print the DCS leak rate calculation.
n. Record leak rate information on Attachment 1.
o. Mark Substep 7.1.2.2 through 7.1.2.15 NIA AND go to Substep 7.1.2.16.
2. Verify with Chemistry Department personnel that RCS sample valves are Closed.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7 .1.2 (Cont'd)

3. Direct the SNPO to perform the following to prevent pumpmg, system makeup, or parameter changes during test:
a. IF necessary, THEN manually pump the RCDT to between 15 and 25 percent using 4-NOP-061.03, Reactor Coolant Drain Tanlc CAUTION Maintaining sump level greater than 100 gallons ensures sump pumps will NOT cavitate.
b. IF necessary, and sump level is greater than 150 gallons, THEN manually pump down the Containment Sump as follows:

(1) Establish communication between the Unit RO and the operator at the appropriate breaker (40667 for 4A Containment Sump Pump or 40778 for 4B Containment Sump Pump).

(2) Begin pumping the Containment Sump by locally depressing and holding the Start pushbutton on the breaker faceplate.

(3) Monitor Containment Sump level on R-1418 (U4 VPA).

(4) WHEN the Containment Sump level reaches 100 to 125 gallons, THEN release the Start pushbutton.

4. IF the leak rate is expected to be greater than 0.25 gpm, THEN Close the following valves:
a. Containment Sump Pump Discharge, CV-4-2821
b. Containment Sump Pump Discharge, CV-4-2822

- - - -- ... - - - -- NOTE -- -- -- ... - .. -- - - .

VCT Level Controller LC-4-112A setpoint may be raised to avoid letdown diversion during the test.

5. IF necessary, THEN Adjust VCT Level Controller, LC-4-112A, up to 70%.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.1.2 (Cont'd)

.-----BlllllllBlllllllJl'l5filllii!lll!ll!!IBl!lllllllllllilllliill--lllllil!ll----~-----*

NOTE I Power level, current VCT level, and rate of makeup should be considered in determining a I 1

VCT level high enough to avoid makeup during the test. 1

6. Raise VCT level to between approximately 20 and 70 percent (high enough to avoid makeup during the test).
7. Record required data:
a. IF using the DCS method to determine RCS leak rate, THEN record the time and selected test duration on Attachment 1.
b. IF using the PC RCS leak rate method, THEN record required data under the Start column on Attachment 2.
c. IF using the manual RCS leak rate method, THEN record required data under the Start column on Attachment 3.

I . If the leak rate must be calculated over a short duration due to plant transients (i.e., xenon, startup, etc), a full length leak rate is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving steady-state operation.

I . Recommended minimum duration for the calculation is four hours.

L--- .... !

8. IF the surveillance can NOT be performed for the selected duration, THEN indicate the reason in the Remarks Section on Attachment 1, 2, or 3, as applicable.
9. IF Annunciator G 9/5, CNTMT SUMP HI LEVEL, or I 4/6, CNTMT SUMP HI LEVEL, actuates, THEN perform the following:
a. IF using the DCS RCS Leak Rate Calculation Program, THEN terminate the calculation.
b. IF using the PC method, THEN record Stop data in Attachment 2.
c. IF using the manual method, THEN record Stop data in Attachment 3.
d. Go to Substep 7.1.2.13.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.1.2 (Cont'd)

10. WHEN the RCS Leak Rate Calculation duration time has expired, THEN perform the following:
a. IF using the DCS RCS Leak Rate Calculation Program, THEN perform the following:
1. On DCS Operator Workstation A W0401 or A W0402, select the

<UTILITIES MENU> button.

2. Select <RCS LEAK RATE CALCULATION>.
3. Select <CONFIGURE RCS LEAK RATE CALCULATION>.
4. To calculate leak rate for the past 15 minutes from the Current time, select the <RUN CALCULATION> button AND go to Substep 7.1.2.1 O.d.(N/A if duration is other than 15 minutes)
5. To calculate leak rate from the Current time, perform the following:

(a) Select <CURRENT TIME> for Select Leak Rate Calculation End Time.

(b) Select applicable Select Leak Rate Calculation Duration button.

(c) Select <RUN CALCULATION> button.

(d) Go to Substep 7.1.2.10.d.

b. IF using the PC method, THEN record Stop data in Attachment 2.
c. IF using the manual method, THEN record Stop data in Attachment 3.
d. Print the DCS or PC RCS leak rate calculation, as applicable.
11. IF only one charging pump is affected by primary packing leakage that exceeds the capability of the leak collection method used to determine the packing leak rate, THEN perform the following:
a. Isolate the affected charging pump.
b. Re-perform the RCS leak rate calculation.
c. Expedite restoration of the isolated charging pump.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 INIT 7.1.2 (Cont'd)

12. IF two charging pumps concurrently have primary packing leakage that exceeds the capability of the leak collection method used to determine the packing leak rate, THEN perform the following:
a. Refer to Technical Specifications (Step 3.1.2)
b. Isolate the affected charging pumps.

NOTE -------l 1

The time period selected for leak rate calculation should NOT exceed two hours to limit the I amount of time both charging pumps are out of service.

c. Re-perform the RCS leak rate calculation.
d. Expedite restoration of the isolated charging pumps. ,..

NOTE*

Substeps 7.1.2.13 through 7.1.2.15 may be performed in any order.

L .....  !

13. Ensure open the following valves:
a. Containment Sump Pump Discharge, CV-4-2821
b. Containment Sump Pump Discharge, CV-4-2822
14. Restore VCT Level Controller, LC-4-112A, setpoint to 37 percent as follows:

(NIA if not applicable)

a. Place VCT Level Controller, LC-4-112A, in Manual.
b. Set VCT Level Controller to 37 percent.
c. Place VCT Level Controller, LC-4-l 12A, in Auto.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.1.2 (Cont'd)

Ill""""""------- _111111111 _ _ _ _ _ _ _ _ _ _ _ _ ,

NOTES When using the PC Leak Rate method, a choice of three options is presented to determine the density of the fluid.

Option 1 (preferred method) uses standard conditions, i.e., 70. 0 °F and 14. 7 psia. Option 1 determines the amount of fluid needed to replace the losses from the RCS.

Option 2 uses the average RCS conditions, i.e., 574°F and 2235 psig, to compute the leak rate. This option will give a higher leak rate than Option 1.

I Option 3 may be used to reflect the actual condition of the leaking fluid from the RCS if the leakage is known to be primarily from a location where temperature and/or pressure is 1

NOT the same as the RCS (examples leaking PORV, leak off the Cold legs).

L-----------------------1

15. Determine RCS Leak Rate using one or more of the following methods:
a. Determine the leak rate using the DCS method:

( 1) Compare start and stop parameters on DCS printout to verify that DCS responded to plant changes as follows:

(a) Start and stop parameters given to 2 decimal places except for time.

(b) Verify data changes from initial to final readings.

(2) IF DCS data is suspect with regard to accuracy or operability, THEN use a leak rate calculation method that is NOT dependent upon the suspect data.

(3) Attach DCS printout to Attachment 1.

b. Determine the leak rate using the PC Leak Rate method.

(I) Access Leakrate.bat in X:\PTN\Team\Ops\DeptShares\Leakrate Stat\Stat Leakrate \Leak\Leakrate (2) Perform program instructions.

(3) Steps may be repeated and information may be copied from the screen if no printer is configured.

(4) Attach personal computer printout to Attachment 2, if applicable.

c. Determine the leak rate using the manual method by performing the leakage calculations as required on Attachment 3.
16. Perform Attachment 7.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 9/10/14 7.1.2 (Cont'd)

NOTES

  • The below procedure step is the criteria by which the leak rate investigation Level or Action is based and the outcome can result from one of three types of data:
  • The calculated absolute value of RCS leakage
  • A statistical assessment using Attachment 7 that is based on the deviation between prior leak rates tests means and standard deviations Any time RCS Gross Leakage is greater than 1.0 gpm

_______ J L

17. Determine if a Reactor Coolant System leak investigation is required using the table below. Performance of Attachment 7 for the statistically derived data shall be performed every time, regardless of leak rate value.

Reactor Coolant System Leak Investigation Table Level or Action Required Gross 2: 0.05 gpm OR 1 Identified 2: 0.05 gpm OR Action required from Attachment 7 Identified > 0.2 gpm OR 2 Unidentified 2: 0.05 gpm OR Action required from Attachment 7 Identified> 0.5 gpm OR 3 Unidentified> 0.5 gpm OR Action required from Attachment 7

[Commitment - Steps 2.3.5 and 2.3.6]

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7.1.2 (Cont'd)

II""""""------- -------------lllllllll NOTES I

  • The performance of 4-0SP-041.2, Reactor Coolant System Visual Leak Inspection and Leak Evaluation, is only required for Gross RCS Leakage of> 1. 0 gpm.

I

  • Anytime the RCS leak rate is performed in Mode 3 or 4 when RCS Pressure less than 1 any SI Accumulator pressure, then perform a level 1 leak rate investigation to evaluate L _p:ib~e :_a~g~o: t:_si:c::~a: _ _ _ _ _ _ _ _ _ _ _ J
18. IF required, THEN perform a Reactor Coolant System Leak Investigation using Attachment 5 or 4-0SP-041.2, Reactor Coolant System Visual Leak Inspection and Leak Evaluation, as appropriate and in accordance with the above table.

NOTE In order to take credit for the Non-Reactor Coolant Pressure Boundary Leakage, the leakage from components associated with this category must be validated during the same time frame that the RCS leak rate calculation is being performed. Otherwise, any Gross Leakage that is NOT Identified Leakage must be considered to be Unidentified Leakage.

19. IF a Level 2 Investigation is performed, THEN subtract Attachments 4 and 6 results from the unidentified RCS leak rate values of Attachment 1, 2, and 3.
20. Record the RCS leak rates (gross, identified, and unidentified) obtained from Attachment 1, 2, or 3, as applicable, and charging pump packing leakage from Attachment 4, and Non-RCPB leakage (other than charging pump packing leakage) from Attachment 6, if performed, in the Unit Narrative Log.
21. IF a leak investigation is performed, THEN Record results in the Unit Narrative Log.

NOTE In consideration of the RCS leak rate surveillance frequency, any associated POD should be due in either 1 or 2 days.

22. IF SI Accumulator inleakage to the RCS are either unacceptable or indeterminate, THEN either establish conditions where RCS pressure is greater than all SI Accumulators AND re-perform the RCS inventory balance OR initiate a prompt operability assessment* to evaluate RCS leakage.

(NIA if not applicable.)

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 7 .1.2 (Cont'd)

23. IF Acceptance Criteria is Unsatisfactory, THEN perform the following:
a. Refer to applicable Tech Specs.
b. Go to 4-0NOP-041.3, Excessive Reactor Coolant System Leakage.

Date/Time Completed: _ _ _ _ _/_ _ _ _ __

REMARKS: ~------------------------------~

PERFORMED BY (Print) INITIALS VERIFIED BY (Print) INITIALS Reviewed By: Date: - - - - -

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ENCLOSURE 1 (Page 1 of 3)

TAVG CONVERSION FACTOR VS. TAVG FOR USE IN MODE 1 AT RCS PRESSURE OF 2240 PSIA TO 2260 PSIA (Gall"f W201 O:JC/cls/cls/cls

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22 Reactor Coolant System Approval Date Leak Rate Calculation 6/2/14 ENCLOSURE 1 (Page 2of3)

TAVG CONVERSION FACTOR VS TAVG FOR TAVG IN THE RANGE OF 320 TO 600 °F WITH RCS PRESSURE IN PSIAAS A PARAMETER 320

  • 360

" '\

\

'\.

~

400 "' ~

~

"'i~

~

440

~~

T-AVE ( ° F) ~ ~* *so psia

~~

480 ~~ 150- 000 osia

~~

~~

' ~~

~ ~ ~ 1000-125 psi a 520

~ ~~

~ r"5 r-....__ 1250

  • 1500 psia

~ ~-

560 r--.:::: I""--.~

!"-. 1500-1j50 ps1a bk0799 600

--- r...::::- ----*

r.:::::::-. -- 1750- 2000 psia

--- 2000- 2250 psia I I I

30 35 40 45 50 55 60 65 70 75 80 CONVERSION FACTOR (gall ° F)

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 612114 ENCLOSURE 1 (Page 3 of 3)

TAVG CONVERSION FACTOR VS TAVG FOR TAVG IN THE RANGE OF 200 TO 300 °F 32.7

/

32.0 31.3

[7 v

a::

0 f-u

<r:

30.6 29.8

/

v u... 29.0

z:

I/

v C)

U1 28.2 er::

w v

z 27.4 0

I-en (Gal/ F) 26.5 25.6 I/

v 24.7 IV 23.8 22.9 v

200 210 220 230 240 250 260 270 280 290 300 310 320 Tavg bk0656 W2010:J /els/els/els

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 3 (Page 1 of 2)

LEAK RATE CALCULATION DATA SHEET (MANUAL METHOD)

II QA RECORD PAGE II Date:

Start+ Stou *CF ERDADS Meter No. Start Stop 2 (Stop-Start) xCF

1. TIME NIA NIA 1111111111111 NIA min
2. VCTLevel Lll5 A LI-4-115  %  % 1111111111111 14.15 gal
3. PZRLevel L462 A LI-4-459A  %  % 1111111111111 42.1 gal
4. Primary Water Totalizer TOTPWCG V NIA gal gal 1111111111111 1 gal
5. Boric Acid Totalizer TOTBACG V NIA gal gal 1111111111111 1 gal
6. PRTLevel L470 A LI-4-470  %  % 1111111111111 100 gal
7. RCDT Level 411003 A LI-4-1003  %  % 1111111111111 3.2 gal
8. Cont. Sump Level Ll546 A LR-4-1418 gal gal 1111111111111 1 gal
9. TAVG AUCT_TAV_A **** op op op
    • gal TR-4-408
  • CF = Conversion Factor
    • Obtain CF for TAVG from Enclosure 1
        • If TAvG::; 540°F, use TR-4-410, TR-4-413, and average the temperatures CT~__+/-_Ltl of the operating loops. 2 RCS LEAK RATE CALCULATION
10. + _ _ _ gal 6-VCT Lvl Af>ZRLvl Primary Water Boric Acid 6-TAVG 6-Total gal Line 2 Line 3 Line 4 Line 5 Line 9
11. (-) _ _ _ _ _ gpm 6-Total gal 6-Time Gross RCS Line 10 unidentified Line 1 Leak Rate (Note 1)
12. IF RCS Leakage is greater than 0.1 gpm, THEN obtain primary to secondary leak rate from Chemistry AND record: gpm.

REFERENCE LEAKAGE (For Information)

13. (__ + ) _ _ _ gpm Af>RT Level 6-RCDT Level 6-Time Identified Line 6 Line 7 Line 1 Leakage
14. = _____ gpm Gross RCS Identified (Note 2) (Note 3) Unidentified Leak Rate Leakage Leakage Line 11 Line 13 Note 1: If the RCS gross leak rate is greater than 1 gpm, then immediately notify the Shift Manager.

Note 2: This is the combined Charging Pump primary leakage in gpm rounded to two decimal places. This value is determined using Attachment 4. However, the value from 4-0P-047, CVCS-Charging and Letdown, Subsection 7.14, Determination of Charging Pump Primary and Secondary Packing Leakage may be used if it is performed during the same time frame as the Leak Rate Calculation.

Note 3: This is the sum of all measured Non~ RCPB (other than Charging packing in gprn rounded to two decimal places. This value is detem1ined using Attachment 6.

W201 O:JC/cls/clslcls

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29 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 3 (Page 2of2)

LEAK RATE CALCULATION DATA SHEET (MANUAL METHOD)

II QA RECORD PAGE II Component Component Description Position Checked Verified Containment Sump Pump CV-4-2821 OPEN (Note 4)

Discharge Containment Sump Pump CV-4-2822 OPEN (Note 4)

Discharge LC-4-112A VCT Level Controller Set to 37%/AUTO Note 4: Control Switch Spring returns to Auto from Open.

ACCEPTANCE CRITERIA

1. RCS pressure boundary leakage shall not exist.
2. Unidentified RCS Leakage is less than 1 gpm.
3. Identified leakage from the RCS is less than 10 gpm.
4. Primary to Secondary Leak Rate is less than or equal to 0 .1 gpm.

D Satisfactory D Unsatisfactory Date _ _ _ _ _ _ __

Verified By: Date _ _ _ _ _ _ __

Reviewed By: Date _ _ _ _ _ _ __

Shift Manager or SRO Designee W201 O: JC/els/els/els

Procedure No.: Procedure

Title:

Page:

30 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 4 (Page 1 of 4)

DETERMINATION OF CHARGING PUMP PRIMARY AND SECONDARY PACKING LEAKAGE NOTES

" Charging pump primary packing leakage is determined by draining the seal pot below the level of the overflow drain and measuring the level increase with time.

  • A stopwatch should be used to ensure consistency of the measurements obtained.

" A conversion factor of 34.2 is used to convert inches per second to gallons per minute.

This factor uses a volume of 0. 57 gallons per inch of level change in the seal pot.

  • If secondary packing leakage is greater than primary packing leakage, the seal pot level will NOT increase and filling the seal pot may be required.
  • If this attachment is being performed while RCS leak rate data is being collected, the seal pot should NOT be filled using water from the RCS because this would cause an erroneous leak rate.
  • Charging pump packing leakage of greater than 0. 05 gpm requires a PWO to repack the pump. See 4-0P-047 PIL 4. 8.
  • If a charging pump has primary packing leakage greater than 0. 05 gpm, charging pump primary packing leakage shall be measured during all RCS leak rate determinations.

INITIALS CK'D VERIF

1. 4A Charging Pump Primary Seal Leakage Determination
a. Drain approximately one inch of water from the seal pot by opening 4A Chfg Pump Seal Water Head Tank Drain Isol Vlv, 4-1338, and lowering the level in the seal pot to approximately 7 inches on the level glass scale.
b. WHEN the seal pot level reaches approximately 7 inches of water, THEN Close 4A Chrg Pump Seal Water Head Tank Drain Isol Vlv, 4-1338.
c. Measure the time in seconds for the level in the seal pot sight glass to indicate an increase in level of at least one inch as follows:
1) IF after five minutes no visible increase in seal pot level is detected, THEN stop the measurement AND go to Substep 1.d AND record the leakage as zero.
2) WHEN the level in the seal pot sight glass has increased by one inch or five minutes have elapsed, THEN stop the measurement AND go to Substep 1.d AND record the level increase in inches and the elapsed time in seconds.
d. Calculate the primary packing leakage to the nearest one-hundredth of a gpm:

4A: _ _ _ inches (34.2 ) _ _ _ _ seconds = _ _ _ _ gal/min W201 O:JC/cls/cls/ Is

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 4 (Page 2 of 4)

DETERMINATION OF CHARGING PUMP PRIMARY AND SECONDARY PACKING LEAKAGE INITIALS CK'D VERIF

e. Observe secondary packing leakage for the 4A Charging Pump by viewing the cylinder end through the transparent cover. Leakage will appear as drops forming at the bottom of each cylinder.
f. Count AND record below the number of drops that form in one minute (dpm) at the South, Center, and North cylinders for the 4A Charging Pump:
1) 4A: South Cylinder secondary packing leakage ---- dpm
2) 4A: Center Cylinder secondary packing leakage _ _ _ dpm
3) 4A: North Cylinder secondary packing leakage _ _ _ dpm
2. 4B Charging Pump Primary Seal Leakage Determination
a. Drain approximately one inch of water from the seal pot by opening 4B Chrg Pump Seal Water Head Tank Drain Isol Vlv, 4-1339, and lowering the level in the seal pot to approximately 7 inches on the level glass scale.
b. WHEN the seal pot level reaches approximately 7 inches of water, THEN Close 4B Chrg Pump Seal Water Head Tank Drain Isol Vlv, 4-1339.
c. Measure the time in seconds for the level in the seal pot sight glass to indicate an increase in level of at least one inch as follows:
1) IF after five minutes no visible increase in seal pot level is detected, THEN stop the measurement AND go to Substep 2.d AND record the leakage as zero.
2) WHEN the level in the seal pot sight glass has increased by one inch or five minutes have elapsed, THEN stop the measurement AND go to Substep 2.d, AND record the level increase in inches and the elapsed time in seconds.
d. Calculate the primary packing leakage to the nearest one-hundredth of a gpm:

4B: inches X (34.2) seconds = gal/min

e. Observe secondary packing leakage for the 4B Charging Pump by viewing the cylinder end through the transparent cover. Leakage will appear as drops forming at the bottom of each cylinder.
f. Count AND record below the number of drops that form in one minute (dpm) at the South, Center, and North cylinders for the 4B Charging Pump.
1) 4B: South Cylinder secondary packing leakage - - - - dpm
2) 4B: Center Cylinder secondary packing leakage _ _ _ _ dprn
3) 4B: N01ih Cylinder secondary packing leakage _ _ _ dpm W201 O:JC/cls/cls/cls

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 4 (Page 3 of 4)

DETERMINATION OF CHARGING PUMP PRIMARY AND SECONDARY PACKING LEAKAGE INITIALS CK'D VERIF

3. 4C Charging Pump Primary Seal Leakage Determination
a. Drain approximately one inch of water from the seal pot by opening 4C Chrg Pump Seal Water Head Tank Drain Isol Vlv, 4-1340, and lowering the level in the seal pot to approximately 7 inches on the level glass scale.
b. WHEN the seal pot level reaches approximately 7 inches of water, THEN Close 4C Chrg Pump Seal Water Head Tank Drain Isol Vlv, 4-1340.
c. Measure the time in seconds for the level in the seal pot sight glass to indicate an increase in level of at least one inch as follows:
1) IF after five minutes no visible increase in seal pot level is detected, THEN stop the measurement AND go to Substep 3 .d AND record the leakage as zero.
2) WHEN the level in the seal pot sight glass has increased by one inch or five minutes have elapsed, THEN stop the measurement AND go to Substep 3 .d AND record the level increase in inches and the elapsed time in seconds.
d. Calculate the primary packing leakage to the nearest one-hundredth of a gpm:

4C: _ _ _ inches X (34.2) - - - - seconds = - - - - - gal/min

e. Observe secondary packing leakage for the 4C Charging Pump by viewing the cylinder end through the transparent cover. Leakage will appear as drops forming at the bottom of each cylinder.
f. Count AND record below the number of drops that form in one minute (dpm) at the South, Center, and North cylinders for the 4C Charging Pump:
1) 4C: South Cylinder secondary packing leakage _ _ _ dpm
2) 4C: Center Cylinder secondary packing leakage _ _ _ dpm
3) 4C: North Cylinder secondary packing leakage _ _ _ dpm
4. Record the results in the Unit Narrative Log.
5. Record the completion and results in STP.
6. Verify log entries specified in Subsection 2.2, Records Required, are completed.
7. Complete the QA Record Page for this attachment.

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 4 (Page 4 of 4)

DETERMINATION OF CHARGING PUMP PRIMARY AND SECONDARY PACKING LEAKAGE II QA RECORD PAGE II Determination of Charging Pump Primary and Secondary Packing Leakage Primary Packing Leakage l.d 4A Charging Pump gpm 2.d 4B Charging Pump gpm 3.d 4C Charging Pump gpm Secondary Packing Leakage South Center North Total l.f 4A Charging Pump _ _ _ dpm 2.f 4B Charging Pump _ _ _ dpm 3.f 4C Charging Pump _ _ _ dpm Date/Time Started / _ _ __ Date/Time Completed_ _ _ _ _/_ _ _ __

PERFORMED BY (Print) INITIALS VERIFIED BY (Print) INITIALS Reviewed By: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~_ _ _ _DA Shift Manager or SRO Designee W201 O:JC/ ls/els/els _

Procedure No.: Procedure

Title:

I Page:

34 Approval Date:

4-0SP-041.1 Reacto:r Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 5 (Page 1 of 3)

REACTOR COOLANT SYSTEM LEAK INVESTIGATION GUIDELINE A. Level 1 Reactor Coolant System Leak Investigation

1. Monitor trends of the following parameters for the previous 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
  • PORV Relief Temperature, TI-4-463*
  • PZR level
  • Reactor Head Vent Temperature, TI-4-6397*
  • Containment sump level
  • Reactor Head Vent to PRT Temperature, TI-4-6399*
  • PRT temperature and level
  • Letdown Relief to PRT Temperature, TI-4-141, OR review Narrative Logs for Annunciator A 516
  • PRMS Channels R-4-11,
  • Reactor Vessel Head Leakoff Temperature, R-4-12, and R-4-15 TI-4-401 *
  • PZR Relief temperatures
  • See RO Rounds (logs) for trend data which are found TI-4-465, 467, 469* in the Sequence Number range of 135 to 165.

.. - - - - - - - - - ... - - - - - .. - - - 11111111 -

NOTES fl When the source of the RCS leakage is identified, subsequent steps of the investigation are NOT required to be performed.

fl Filter vent and drain valves are a common source of RCS leakage.

L

2. Review the Unit Narrative Log for the previous 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
  • RCP Seal Water Injection Filters
  • Seal Water Injection Filters
  • Other activities that could impact leak rate calculation
3. IF necessary to bypass the CVCS demineralizers to aid in leak investigation, THEN perform the following:
a. Place Letdown Demineralizer Divert Valve, TCV-4-143, in the VCT Divert position.
b. WHEN the leak investigation of the CVCS demineralizers is complete, THEN place Letdown Demineralizer Divert Valve, TCV-4-143, is in the Auto position.
4. Initiate a Condition Report.
5. Verify log entries specified in Subsection 2.2 are completed.
6. Complete the QA Record Page for this attachment.

W201 O:JC/cls/cls/cls

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35 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 5 (Page 2 of 3)

REACTOR COOLANT SYSTEM LEAK INVESTIGATION GUIDELINE B. Level 2 Reactor Coolant System Leak Investigation r . . - . . - . . . . . . . . . . . . . . - . . . . . N"o"T'E .............. - ............. - ... - - ............... I I

If the source of the increased RCS leakage is identified while performing the step of a leak investigation, the subsequent investigational steps to locate the source of the leak are NOT required to be performed. A Condition Report should be generated to capture the event, I appropriate log entries made, and the appropriate QA documentation completed. I

- _. - ma - ... - mm - - - 118 - - - .. _, Elll - - - - - II& I

1. Perform a Level 1 Reactor Coolant System Leak Investigation.
2. Determine Charging Pump packing leakage using Attachment 4.
3. Determine Non-RCPB leakage (other than charging pump packing leakage) using Attachment 6.
4. Generate a troubleshooting plan using MA-AA-100-1011, Equipment Troubleshooting.
5. Generate a Condition Report.
6. Verify log entries specified in Subsection 2.2, Records Required, are completed.
7. Complete the QA Record Page for this attachment.

C.

.- Level 3 Reactor Coolant System Leak Investigation NOTE If the source of the increased RCS leakage is identified while performing the step of a leak investigation, the subsequent investigational steps to locate the source of the leak are NOT required to be performed. A Condition Report should be generated to capture the event, appropriate log entries made, and the appropriate QA documentation completed.

- am - am - Bll!I - - - - - .. - - - - - ... - -.. - mrll - ml I

1. Perform a Level 2 Reactor Coolant System Leak Investigation.
2. Perform 4-0SP-041.2, Reactor Coolant System Visual Leak Inspection and Leak Evaluation.

(!;[less than 1.0 gpm gross, THEN at AOM discretion, this may be NIA).

3. Verify log entries specified in Subsection 2.2, Records Required, are completed.
4. Complete the QA Record Page for this attachment.

W201 O:JC/cls/cls/cls

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4-0SP-041.1 Reactor Coolant System Leak Rate Calculation I 6/2/14 ATTACHMENT 5 (Page 3 of 3)

REACTOR COOLANT SYSTEM LEAK INVESTIGATION GUIDELINE QA RECORD PAGE II II Reactor Coolant System Leak Investigation

1. Indicate RCS leak investigation performed D Level 1 D Level 2 D Level 3 Date/Time Started / _ _ __ Date/Time Completed_ _ _ _ _/_ _ _ __

PERFORMED BY (Print) INITIALS VERIFIED BY (Print) INITIALS Reviewed W201 O:JC/cls/cls/cls

Procedure No.: Procedure

Title:

I Page:

37 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 6 (Page 1 of 1)

NON-RCPB LEAKAGE (OTHER THAN CHARGING PUMP PACKING LEAKAGE)

QA RECORD PAGE II II

1. Identify leaking non-RCPB components accounted for with this attachment.

Leak Rate*

Component Location (GPM rounded to 2 PWONo.

decimal places)

  • If possible, use graduated cylinder and stopwatch to determine leak rate. Record method used and leakage recorded and time interval over which it was recorded in the Remarks Section for each leak documented in the table above.
2. Total of Attachment 6 Non-RCPB Leakage (Sum of values above) = _______
3. Remarks (Attach additional sheets if necessary)
4. Ensure a PWO has been generated for each leak accounted for in this attachment and the PWO number is recorded in the Table in Step 1 above.

Performedby:~------------

Verified by: _ _ _ _ _ _ _ _ _ _ _ __

Reviewed by: -=--,,.-~~~~--:::=---::--=:--~~~-

Date:

Shift Manager or SRO Designee W201 O:JC/cls/cls/cls

Procedure No.: Procedure

Title:

Page:

28 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 612114 ATTACHMENT 3 (Page 1 of2)

KEY LEAK RATE CALCULATION DATA SHEET (MANUAL METHOD)

,, QA RECORD PAGE Date: 10~

Start+ Stor ERDADS Meter No. Start Stop 2

  • CF (Stop-Start) xCF
1. TIME NIA NIA /CJnn Z..300 /Ill/I/II/I/! NIA 2.40 min
2. VCTLeve! 1115 A LI-4-l15 i:*a  % 'l..6 % /////II/Ill/I 14.15 o gal
3. PZRLevel L462 A LI-4-459A ~c-~ % ~';J-  %  !/l/!llllllll 42.l 0 gal
4. Primary Water Totalizer TOTPWCG V NIA 0 gal 0 gal ///////////// 1 0 gal
5. Boric Acid Totalizer TOTBACG V NIA 0 gal () gal //I/Ill/II/II 1 0 gal
6. PRTLevel L470 A LI-4-470 -::J-1  % -:J. I  % II/I/Ill/Ill/ 100 O gal
7. RCDT Level 411003 A LI-4-1003 IC  % i.R % lll///l/lll// 3.2 +2 gal
8. Cont. Sump Level Ll546 A LR-4-1418 /?_ 0 gal 17..0 gal lll/!llll///I 1 O gal
9. TAVG AUCT_TAV_A ****

TR-4-408 6'"80 OF sao op £Bo OF ** O gal

  • CF= Conversion Factor
    • Obtain CF for TAVG from Enclosure 1 IfTAvG ~ 540°F, use TR-4-410, TR-4-413, and average the temperatures .CTQ__+/-_Ifil of the operating loops. 2 RCS LEAK RATE CALCULATION
10. -Yro + 0 - 0 - 0 0 ~34l> gal

~VCTLvl Af>ZRLvl Primary Water Boric Acid ATAVG ATotal gal Line 2 Line 3 Line4 Line 5 Line 9

11. (-) -3q-o .. 2'\-o I* 4-17. gpm (I
  • 4-o Tt> I
  • ts )

ATotal gal ATime Gross RCS Line 10 unidentified Line 1 Leak Rate (Note 1)

12. IF RCS Leakage is greater than 0.1 gpm, THEN obtain primary to secondary leak rate from Chemistry AND record: .RC gpm.

REFERENCE LEAKAGE (For Information)

13. ( 0 + 4- 2. ) '2-4-o = o * '~ gpm ( o* I t f-o o. 1 B )

APRTLevel ARCDT Level ATime Identified Line 6 Line 7 Line 1 Leakage

14. f*tf-17 - 0
  • 11-5 - 0 , () s *- _Q_ = /.I °I 2. gpm Gross RCS Identified (Note 2) (Note 3) Unidentified Leak Rate Leakage Leakage Line 11 Line 13 Note 1: If the RCS gross leak rate is greater than 1 gpm, then immediately notify the Shift Manager.

Note 2: This is the combined Charging Pump primary leakage in gpm rounded to two decimal places. This value *is determined using Attachlllent

  • 4. However, the value from 4-0P-047, CVCS-Charging and Letdown, I Subsection 7.14, Determination of Chargh1g Pump Primary and Secondary Packing Leakage may be used if it is performed during the same time frame as the Leak Rate Calculation.

3: This is the sum of all measured Non-RCPB leakage (other than Charging Pump packing leakage) in gpm rounded to two decimal places. This value is detem1ined using Attachment 6.

IJ\/2Q1n 1r 1 - 1 ~ 1 - 1 ~ 1 ::;Js

Procedure No.: Procedure

Title:

Page:

29 Approval Date:

4-0SP-041.1 Reactor Coolant System Leak Rate Calculation 6/2/14 ATTACHMENT 3 (Page 2of2)

LEAK RATE CALCULATION DATA SHEET (MANUAL METHOD)

II QA RECORD PAGE II Component Component Description Position Checked Verified Containment Sump Pump CV-4-2821 OPEN (Note 4)

Discharge

.*ccititainmenfSuml1.P~#1P
  • i.Pisctiarge
  • _:; * * *OPE}.t{tlQte:4)

LC-4-l 12A VCT Level Controller Set to 3 7%/AUTO Note 4: Control Switch Spring returns to Auto from Open.

ACCEPTANCE CRITERIA

1. RCS pressure boundary leakage shall not exist.

2 Unidentified RCS Leakage is less than I gpm.

3. Identified leakage from the RCS is less than 10 gpm.
4. Primary to Secondary Leak Rate is less than or equal to 0.1 gpm.

D Satisfactory 12(' Unsatisfactory Verified By: Date _ _ _ _ _ _ __

Reviewed By: Date _ _ _ _ _ _ __

Shift Manager or SRO Designee

JPM TITLE: Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications JPM NUMBER: 02068012301 REV. 0-0 TASK NUMBER(S) I 02068012300/

TASK TITLE(S): Evaluate And Direct Tech Specs Required Actions Due To Containment Spray System Out Of Spec/Service Conditions KIA NUMBERS: 2.2.37 KIA VALUE: RO 3.6 I SRO 4.6 Justification (FOR KIA VALUES <3.0): N/A TASK APPLICABILITY:

~RO ~SRO ~ STA D Non-Lie ~SRO CERT D OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: D Perform: [U EVALUATION LOCATION: In-Plant: D Control Room:

Simulator: D Classroom: x I Lab: D Other: I Time for Completion: 15 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No Instructor/Developer R~i~~~=--------~~~---~~--------- lns~al Revie Date SME (Technical Review) Date Approved by: ~~ ~~

---:----7--;;.~.,,.---r-:7~"-t':::>---r--::;::;r--=:::m-in~g-S"""'u-pe-rv~i-si-on_ _ _ _ _ _ __

Approved by:

Training Program Owner Date TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 2 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.2/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 3 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE New for L-15-1 NRC See cover page N/A 0-0 New JPM 01982463 Exam See cover page N/A 0-1 0-2 0-3 0-4 0-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.2/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 4 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION SIMULATOR SET-UP: N/A Required Materials:

  • 3-OSP-068.2, Containment Spray System Inservice Test - marked-up copy (front matter and Attachment 1)
  • Technical Specifications General

References:

  • Technical Specifications Task Standards:
  • Demonstrate ability to determine operability and/or availability of safety-related equipment
  • Determine that IST acceptance criteria for the 3A Containment Spray Pump and its discharge valve are not met, the system is inoperable, and Technical Specification LCO 3.6.2.1 Action a is required TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.2/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 5 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • Units 3 and 4 are both in Mode 1 at 100% power, with no equipment OOS, and in their normal electrical alignments.

Initiating Cue:

  • You are directed to review the data; perform Step 9 of Attachment 1 (3A Containment Spray Pump) of 3-OSP-068.2 and mark the surveillance SAT or UNSAT; list any parameters that are outside of their acceptable range in the Remarks section (Step 10 of Attachment 1); and [SRO only] identify any operability declarations and required Technical Specification actions, if applicable.

Operability Declarations/Required Actions applicable [SRO only]: _______________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.2/RO,SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 6 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

Provide examinee with marked-up copy of 3-OSP-068.2 (Containment Evaluator Cue: Spray System Inservice Test) - upfront material and Attachment 1 (3A Containment Spray Pump).

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 7 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Review the stroke-time data for MOV-3-880A, recognize that it is Performance Step: 2 outside of the required action range, and list the discrepancy in the Critical: Yes Remarks section.

Examinee reviews value for MOV-3-880A (3A Containment Spray Pump Standard: discharge valve), confirms that it is outside of the required action range, and lists the discrepancy in the Remarks section.

  • The stroke time for opening the 3A Containment Spray Pumps discharge valve (15.28 seconds) exceeds the Required Action time (i.e., 14.43 seconds).

Evaluator Note:

  • Operators performing surveillance failed to recognize that valve stroke was in required-action range and inappropriately indicated that valve performance was acceptable.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Review the 3A Containment Spray Pump bearing casing vibration Performance Step: 3 data (pump inboard horizontal vibration), recognize that it is Critical: Yes outside of the acceptable range, and list the discrepancy in the Remarks section.

Examinee reviews value for 3A Containment Spray Pump inboard Standard: horizontal vibration (coupling end), confirms that it is within the alert range, and lists the discrepancy in the Remarks section.

3A Containment Spray Pump inboard horizontal vibration (0.5614 in/sec)

Evaluator Note:

exceeds 0.325 in/sec, but not 0.700 in/sec.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 8 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Review the 3A Containment Spray Pump bearing casing vibration Performance Step: 4 data (pump inboard vertical vibration), recognize that it is outside Critical: Yes of the acceptable range, and list the discrepancy in the Remarks section.

Examinee reviews value for 3A Containment Spray Pump outboard Standard: vertical vibration (coupling end), confirms that it is within the alert range, and lists the discrepancy in the Remarks section.

3A Containment Spray Pump inboard horizontal vibration (0.6119 in/sec)

Evaluator Note:

exceeds 0.311 in/sec, but not 0.700 in/sec.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Performance Step: 5 Determine that the inservice test is unsatisfactory and complete Critical: Yes Step 9 of Attachment 1.

Determine that inservice test is unsatisfactory and check Unsatisfactory Standard:

in Step 9 of Attachment 1.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 9 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

[SRO only] Declare MOV-3-880A inoperable and identify the Performance Step: 6 appropriate Technical Specification LCO (3.6.2.1) and action Critical: Yes statement (a).

Declare MOV-3-880A inoperable and identify appropriate Technical Standard:

Specification LCO (3.6.2.1) and action statement (a).

  • Per Step 6.5, when a stroke time exceeds the required action time, the valve is declared inoperable, a WR is generated to correct the deficiency, and a CR is generated to determine Maintenance Rule implications.
  • Examinee may list CR/WR as a required action, but this is not a critical element of this step.
  • Applicability - Modes 1, 2, 3, and 4.
  • Action a - With one Containment Spray System inoperable restore the inoperable Spray System to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby within the next six hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 6, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

02068012301, Evaluate Containment Spray Pump Data Sheet and Apply Technical Specifications, Rev. 0 JPM Page 10 of 11 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 11 of 11 TURNOVER SHEET Initial Conditions:

  • Units 3 and 4 are both in Mode 1 at 100% power, with no equipment OOS, and in their normal electrical alignments.

Initiating Cue:

  • You are directed to review the data; perform Step 9 of Attachment 1 (3A Containment Spray Pump) of 3-OSP-068.2 and mark the surveillance SAT or UNSAT; list any parameters that are outside of their acceptable range in the Remarks section (Step 10 of Attachment 1); and [SRO only] identify any operability declarations and required Technical Specification actions, if applicable.

Operability Declarations/Required Actions applicable [SRO only]: _______________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

Florida P*ower & Light Company Turkey 'Point Nuclear Plant Unit 3 3-0SP-068.2

Title:

Containment Spray System Inservice Test Continuous Use Safety Related Procedure Responsible Department: Engineering Revision Number 6 Issue Date 8/27/12 Revision Approval Date: 8/16/12 ARs 1599262, 588220, 1644359, 1675276, 1659095, 1788790 PCRs 09-1687, 09-2895, 10-0787, 10-1298 PC/Ms84-144, 86-181,87-194, 08-026 ECs 247008 W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 8/16/12 LIST OF EFFECTIVE PAGES (Rev. 6)

Revision Revision Page Date Page Date 1 08/16/12 21 08/31/11 2 08/16/12 22 06/29/09 3 06/29/09 23 06129109 4 06129109 24 06/29/09 5 08/16/12 25 08/31/11 6 08/31/11 26 08/31/11 7 06/29/09 27 08/16/12 8 08/10/11 28 08/10/11 9 06/29/09 29 08/31/11 10 09/13/11 30 08/31/11 11 08/31/11 31 08/16/12 12 08/31/11 32 08/10/11 13 08/31/11 33 08/16/12 14 08/31/11 34 08/16/12 15 08/31/11 35 06/29/09 16 09/13/11 17 08/31/11 18 08/31/11 19 08/31/11 20 08/31/11 W97 :/J EE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 6/29/09 TABLE OF CONTENTS Section 1.0 PURPOSE..................................................................................................................... 4

2.0 REFERENCES

/RECORDS REQUIRED/COMMITMENT DOCUMENTS........ 4 3.0 PREREQUISITES....................................................................................................... 6 4.0 PRECAUTIONS/LIMITATIONS.............................................................................. 7 5.0 SPECIAL TOOLS/EQUIPMENT.............................................................................. 8 6.0 ACCEPTANCE CRITERIA....................................................................................... 8 7.0 PROCEDURE 7 .1 CS Gas Void Monitoring ................. .... ...... ........ .... .......... ... .. ... .... ............ ..... ..... 9 7.2 Containment Spray Pump 3A Group A Test..................................................... 10 7.3 Containment Spray Pump 3B Group A Test...................................................... 16 ENCLOSURES/ATTACHMENTS Enclosure 1 Containment Spray Pump Vibration Monitoring Locations.............................. 22 Attachment 1 3A Containment Spray Pump ...... .... ........ ......... ... .... ....... ... ... ... .. ....... .... .... ..... .... 25 Attachment 2 3B Containment Spray Pump............................................................................. 29 Attachment 3 Pump Reference Value Evaluation Sheet .. .. .. .. .... .. .. ..... .. ... .. .. .. .... .... .... .. .. ...... .... 33 Attachment 4 Unit 3 CS Gas Accumulation Monitoring Points .............................................. 35 W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 6/29/09 1.0 PURPOSE 1.1 This procedure provides the instructions for performing the inservice test of the Containment Spray pumps. This procedure ensures compliance with References Steps 2.1.1, 2.1.2, 2.1.4.5 and 2.1.5.

1.2 This procedure provides the method for obtaining data required by O-ADM-737, Post Maintenance Testing, for post maintenance pump data.

1.3 This procedure performs the required pump tests of the Containment Spray pumps m accordance with the ASME OM Code, Subsection ISTB.

1.4 This procedure performs the required valve exercise tests in accordance with the ASME OM Code, Subsection ISTC.

2.0 REFERENCES

/RECORDS REQUIRED/COMMITMENT DOCUMENTS 2.1 References 2.1.1 Technical Specifications

1. Section 3/4.6.2.1.b
2. Section 4.0.5 2.1.2 FSAR
1. Section 6.0, Engineered Safety Features
2. Section 14.0, Safety Analyses 2.1.3 Plant Drawings
1. 5613-M-3068, Sh 1, Containment Spray System 2.1.4 Plant Procedures
1. O-ADM-737, Post Maintenance Testing
2. 3-NOP-030, Component Cooling Water System
3. 3-NOP-068, Containment Spray System
4. 3-0SP-068.3, Containment Spray System Monthly Flowpath Verification
5. O-ADM-547, Gas Accumulation Management Program W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 8/16/12 2.1.5 Regulatory Guidelines

1. ASME OM Code 1998 Edition through 2000 Addenda, Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Nuclear Power Plants
2. ASME OM Code 1998 Edition through 2000 Addenda, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants 2.1.6 Miscellaneous Documents (i.e., PC/Ms, Correspondence)
1. PC/M 84-144, Replacement of 3-896T for Containment Spray Recirculation Test Line
2. PC/M 86-181, SI Pump Minimum Flow Recirculation Line Valve Actuator Replacement
3. Fourth Ten-Year Inservice Inspection Interval Inservice Testing Program for Pumps and Valves
4. Westinghouse System Description Auxiliary Coolant System, FPL-200/C/4 and FPL-200/D
5. JPN-PTN-SENP-95-026, Safety Evaluation for CCW Flow Balance and Post Accident Requirements to Support Current and Uprated Conditions (LER 250/95-006)
6. CR 02-0121, Incorporate Acceptance Criteria into all Procedures Used to Test Pumps in the IST Program
7. CR 03-0775, New 3B CSP Reference Values
8. 2004-1907-CR, Elevated Thrust Bearing Temperatures
9. OTSC 03-0425, Reduce Containment Spray Pump Required Suction Pressure to Allow Pump to Start with Reduced RWST Level
10. CR 2006-21547, CSP IST Procedure Does not Account for Discharge Orifice
11. CR 2006-17075, CSP Seal Water Heat Exchanger Monitoring Program Tracking CR
12. 5610-068-DR-002, Page 16, Section 2.3.3, Reference to Containment Spray System Design Temperature
13. PC/M 08-026, Containment Spray Pump Full Flow Recirculation Modification
14. PTN-ENG-SEMS-08-031 Rev. 2, Extended Power Uprate ECCS Fathom I Inputs to Westinghouse
15. EC 242393, PCM-08026 Containment Spray Pump Full Flow Recirculation I Modifications W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 8/31/11 2.2 Records Required 2.2.1 The date, time, and section completed shall be entered in the Unit Narrative Log.

Also, problems encountered while performing the procedure should be entered; i.e., malfunctioning equipment, delays due to changes in plant conditions, etc.

2.2.2 Prior to routing to QA Records, the completed QA Record Pages and Attachment 3 shall be routed to the IST Coordinator for analysis and review.

2.2.3 Completed copies of the below listed items document compliance with Technical Specification surveillance requirements and shall be sent to QA Records for retention in accordance with Quality Assurance Records Program requirements:

1. Attachment 1
2. Attachment 2
3. Attachment 3 2.3 Commitment Documents 2.3.1 L-88-400, Letter to NRC, Response to Inspection Report Nos. 50-250/86-06 and 50-251/86-06; Reconfirming or Establishing Reference Values 2.3.2 JPN-PTN-SENP-95-026, Safety Evaluation for CCW Flow Balance and Post Accident Requirements to Support Current and Uprated Conditions (LER 250/95-006) 2.3.3 JPN-PTN-SEMP-95-032, Engineering Evaluation for Pressure Locking and Thermal Binding Evaluation of Safety Related Power Operated Gate Valves 8' PREREQUISITES

,If}" The Containment Spray System is aligned per 3-NOP-068, Containment Spray System.

fi' The Containment Spray Pump bearing oiler bottle oil level must be visible prior to starting the pump and during the pump run.

g The Shift Manager has authorized performance of this test.

, ~The Component Cooling Water is aligned to the Containment Spray pump in accordance

~ with 3-NOP-030, Component Cooling Water System.

Instrumentation and test equipment shall be in calibration.

All personnel performing this test shall read and understand this procedure.

The RWST Purification Pump shall not be aligned to RWST during performance of I Subsections 7.2 and 7 .3.

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3-0SP-068.2 Containment Spray System Inservice Test 6/29/09

@ PRECAUTIONS/LIMITATIONS

@ When a pump is to be placed in service after maintenance, it shall be tested to verify pump operability.

@ In the event of unusual noise, overheating, excessive vibration, or other abnormal symptoms, discontinue test and notify the Shift Manager immediately.

fj Compare the test values with the ranges shown on appropriate acceptance criteria at all times through the test period. If values are not within the specified range, immediately notify the Shift Manager and the IST Coordinator (or designee) to evaluate the pump performance. If the results indicate that the pump is inoperable, remove the pump from service and do not return to service until cause of the deviation is determined and the condition has been corrected.

If the unit is in Cold Shutdown Mode 5 or Refueling Mode 6, Containment Spray Pump Discharge Isolation Valves 3-891A and 3-891B will be locked closed and remain locked closed during the test.

0 JC7 After a pump has been replaced, a new set or sets of reference values shall be determined from the results of the first inservice test. When a reference value or set of values may have been affected by repair or routine servicing of the pump, a new reference value or set of values shall be determined or the previous value reconfirmed by a Group A or Comprehensive inservice test run prior to returning the pumps to normal service.

[Commitment - Step 2.3.l]

)(Y a Valves with remote position indication are required to be observed at least once every two years to verify that valve operation is accurately indicated. When this is required, an observer will be positioned at the valve in communication with a second observer located at the remote position indicator. As the valve is operated, actual valve position will be compared to that indicated at the remote location.

d When performing valve manipulation to isolate the Containment Spray Header, the

.}t.J applicable Technical Specification action statement for one inoperable containment spray pump should be entered. Refer to Technical Specifications for appropriate LCO requirements and action statements.

ff Elevated pump coupling end (inboard) bearing housing temperatures may be experienced during extended pump runs. Pump operability limit is 210°F. Notify System Engineering of temperatures above 200°F for CS Pump 3B and 180°F for CS Pump 3A.

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3-0SP-068.2 Containment Spray System Inservice Test 8/10/11

@ SPECIAL TOOLS/EQUIPMENT

~Vibration Instrument and Probe Vstopwatch ACCEPTANCE CRITERIA

./'{)('"When new referern~e values are being determined, the test data shall be evaluated and

~ deviations between the previous and new set of reference values identified and verification that the new values represent acceptable pump operation shall be documented.

See Attachment 3)

The tested pump performance is satisfactory if all test values are within the pecified ranges.

~The Group A pump test is required to be performed once every 92 days.

~If the cooler outlet temperature taken at the pump seal gland exceeds 205°F, the pump P ~hall immediately be declared inoperable and the System Engineer shall be contacted.

~he required actions as a function of the measured stroke time for a valve are as follows:

Measured REQUIRED ACTIONS Stroke Time 1st Stroke 2nd Stroke Within

  • If 1st stroke time deviation is analyzed or Acceptable determined NOT to be due to a degraded Range
  • Mark 2nd Stroke N/A valve condition, then record deviation under Remarks. (See Note 1)

NOT within

  • Immediately retest valve
  • Generate a 3-day operability CR to analyze Acceptable data to determine if valve is showing Range but less OR acceptable operation and if a new than Required reference stroke time may be established
  • Declare valve Inoperable Action Time from this test.

Exceeds

Required Action

  • Generate WR to correct deficiency .

Time

  • Generate CR to determine maintenance rule implications.

Note 1: This allows not removing a valve from service for initial 1st stroke deviations caused by such things as failure to terminate the timing when the light goes out, burned-out light

_indication, power interruption, test equipment failure, etc.

f~ a valve fails the remote position indication verification test, the valve shall immediately

~ b.te declared inoperable and appropriate corrective action initiated.

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3-0SP-068.2 Containment Spray System Inservice Test 8/31/11 ATTACHMENT 1 (Page 1 of 4) 3A CONTAINMENT SPRAY PUMP QA RECORD PAGE II II INITIALS CK'D VERIF Date/Time Started: --~+---'--"""""'~--f----'--;_6_ha-'---'--u__,rs"--tt--=-~-.--o

.x::D Indicate below the reasons for testing the 3A Containment Spray Pump.

Reason for test: _ _ _ _/ _ _ _ _Quarterly Group A IST Test

_ _ _ _ _ _ _ _ Increased Surveillance for_ _ _ _ _ _ _ __

- - - - - - - -Reference Test

_ _ _ _ _ _ _ _Post Maintenance (w/o#)_ _ _ _ _ _ _ _ __

_ _ _ _ _ _ _ _ Other (Specify)_ _ _ _ _ _ _ _ _ _ __

~ Record the Test Equipment numbers and calibration due dates below:

Instrument Test Equipment No. Cal Due Date Pump Suction Press PI-3-1596C w..r~

Pump Disch Press PI-3-945A e.u..-rr.eri,t=

Flowmeter FI-3-945 CM. rr.,,.,,...r=

Vibration Instrument :r: 'I G:CS w vv- ~

Vibration Probe ,-=g S ::J::SA- u,t,vr.e..rvt=

Stopwatch :C. \~'2-\::. W-~

Contact Pyrometer :J: \ o -i.-z... -z.., CA.&~

e:?r: Pro~ .e.. ;I:. q i.'2.--"Z.-

~(Step 7 .2 .3) Verify calibrations of all listed instruments are current.

~(Step 7.2.19 and 7.2.20) Record MOV-3-880A Opening and Closing times:

Stroke 1st Stroke 2°d Stroke Acceptable Range Required Action Direction (Seconds) (Seconds)

OPEN IS'. ~B 9.82 to 13.28 rJ/k > 14.43 CLOSE (O*f'~ 9.86 to 13.34 N/flr-- > 14.50

_f_ v< ~ Verify stroke time falls within the acceptable range.

K Verify remote position indication AND locally observed position agree.

(NIA if not required)

$substep 7.2.21.2) Pre-start Suction Pressure, PI-3-1596C. f B. 5 psig I W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 8/31/11 ATTACHMENT 1 (Page 2 of 4) 3A CONTAINMENT SPRAY PUMP QA RECORD PAGE II II INIT r- - - - - - - -

1

~ounting pads

-y-----------,

installed at all pomts should be used with a magnet-mounted

.A{;/ :~celerometer.

~eps 6 and 7 may be performed concurrently or in either order.

S ~ (Step 7 .2.25) Record the following bearing casing vibration data below at the designated points as

-;/C}/ shown in Enclosure 1 AND compare the valves to the Acceptance Criteria.

(NIA if reference valve test)

% Pump Outboard Horizontal Vibration (Rotor End) 0. '1-( ab in/sec

§ Acceptable Range: Less than or equal to 0.325 in/sec Alert Range: Greater than 0.325 to 0.700 in/sec Required Action Range: Greater than 0.700 in/sec Pump Outboard Vertical Vibration (Rotor End) I f).fb3 7 in/sec

§ Acceptable Range: Less than or equal to 0.224 in/sec Alert Range: Greater than 0.224 to 0.538 in/sec Required Action Range: Greater than 0.538 in/sec ff' Pump Outboard Axial Vibration (Rotor End) I 0 dS-b I in/sec

~

Acceptable Range: Less than or equal to 0.325 in/sec Alert Range: Greater than 0.325 to 0.700 in/sec Required Action Range: Greater than 0.700 in/sec Pump Inboard Horizontal Vibration (Coupling End) IO*(°b/+inlsec

~

Acceptable Range: Less than or equal to 0.325 in/sec Alert Range: Greater than 0.325 to 0.700 in/sec Required Action Range: Greater than 0.700 in/sec

~ Pump Inboard Vertical Vibration (Coupling End) I f>* h/lf in/sec

§ Acceptable Range: Less than or equal to 0.311 in/sec Alert Range: Greater than 0.311 to 0.700 in/sec Required Action Range: Greater than 0.700 in/sec ff Motor Inboard Horizontal f). 2 Z 4-~ in/sec

¥Motor Inboard Vertical 'O * \ i,6( 1 in/sec W97:/JEE/cls/cls/fm

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3-0SP-068.2 Containment Spray System Inservice Test 8/16/12 ATTACHMENT 1 (Page 3 of 4) 3A CONTAINMENT SPRAY PUMP QA RECORD PAGE II II

6. (Cont'd)
fj)f'_ Motor Outboard Horizontal 0 * '2.<!l-31 in/sec

$Motor Outboard Vertical O*u5Cf t in/sec ff Motor Outboard Axial ()- OBot in/sec

..:+/-__ .fi}"' (Step 7.2.25) Record pump pressure parameters and perform calculations.

(ij- Discharge Pressure on PI-3-945A _~_q_'Z_psig

~ Suction Pressure, PI-3-1596C (Not used for dP calculation) _ ___._/""""-g_ psig

% Recirc Flow, FI-3-945 L}-Do gpm

% Calculate pump differential pressure (dP) as follows:

From Step 7.a From Step 5 Pump Pump Discharge Pre-Start Suction Orifice

- + = Differential Pressure Pressure Compensation

,, Pressure

?Av psig I

- r11> *'7 psig

~-

+ 2 psi = 2-1-'S* S psid Design Basis /JP limits at 400 gpm are> 250.55 and< 293.49 psid. I L-----------------------1 Compare the Pump Differential Pressure to the Acceptance Criteria.

(Tech Spec 4.6.2.1.b)

H-A-cceptable Range: 250.55 psid to 293.49 psid D Required Action Range: Less than 250.55 psid or greater than 293.49 psid

~(Step 7.2.25) Record the following pump operating data:

~ Seal leakage: ~ drops/min for one minute.

9. 3A Containment Spray Pump and related valves (if required) test is:

Satisfactory _ _ __ Unsatisfactory _ _ __

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3-0SP-068.2 Containment Spray System Inservice Test 8/10/11 ATTACHMENT 1 (Page 4 of 4) 3A CONTAINMENT SPRAY PUMP QA RECORD PAGE II II Record Seal Water Heat Exchanger Cooler outlet temperature using contact pyrometer on pump seal gland (Acceptance Criteria is T ~ 205°F):

~'1* 5 op If Cooler outlet temperature is ~190°F, notify System Engineer.

Date/Time Completed: _ _ f"D_!J_¥------/r-r1_o_fl0 PERFORMED BY (Print) INITIALS nFv.i:t s O~.ex0vhY~ k REVIEWED BY: ~~~~~~~~~~~~~~~~~~~-

Shift Manager or SRO designee Date REVIEWED BY: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

IST Coordinator or designee Date W97:/JEE/cls/cls/fm

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one Containment Spray System inoperable restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With two Containment Spray Systems inoperable restore at least one Spray System to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Restore both Spray Systems to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position and that power is available to flow path components that require power for operation;
b. By verifying that on recirculation flow, each pump develops the indicated differential pressure, when tested pursuant to Specification 4.0.5:

Containment Spray Pump 241.6 psid while aligned in recirculation mode.

TURKEY POINT - UNITS 3 & 4 3/4 6-12 AMENDMENT NOS. 137 AND 132

EN~.BQ~~~--- M u J NUCLEAR Page 1 of 9 FLEET JPM TITLE: Review Primary and Secondary Sample Results and Determine Applicable Actions JPM NUMBER: 02200063301 REV. 0-0 TASK NUMBER(S) I 02200063300/

TASK TITLE(S): Respond to Deviation from Secondary Chemistry Limits KIA NUMBERS: 2.1.34 KIA VALUE: SRO 3.5 Justification (FOR KIA VALUES <3.0): N/A TASK APPLICABILITY:

D RO ~SRO D STA D Non-Lie ~SRO CERT D OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: D Perform: ~

EVALUATION LOCATION: In-Plant D Control Room:

Simulator: D Classroom: x I Lab: Other: I Time for Completion: 10 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No 11 I8 /tq-Date Date Training Program Owner Date TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/SRO

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 2 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 3 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE New for L-15-1 NRC See cover page N/A 0-0 New JPM 01982463 Exam See cover page N/A 0-1 0-2 0-3 0-4 0-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 4 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION SIMULATOR SET-UP: N/A Required Materials:

  • 3-ARP-097.CR.D, Control Room Response - Panel D
  • 3-ONOP-071.1, Secondary Chemistry Deviation from Limits
  • 0-ADM-651, Nuclear Chemistry Parameters Manual
  • 0-ADM-115, Notification of Plant Events
  • AD-AA-07, Notification of Chief Nuclear Officer General

References:

  • 3-ARP-097.CR.D, Control Room Response - Panel D
  • 0-ADM-651, Nuclear Chemistry Parameters Manual
  • 3-ONOP-071.1, Secondary Chemistry Deviation from Limits
  • 0-ADM-115, Notification of Plant Events
  • AD-AA-07, Notification of Chief Nuclear Officer
  • Technical Specifications Task Standards:
  • Demonstrate knowledge of primary and secondary plant chemistry limits
  • For the given conditions, determine the secondary-chemistry action level (level 3), minimum required plant response (quickly/safely reduce power to <5%), and required notifications (Chief Nuclear Officer, Site VP, Plant General Manager, Operations Manager, Nuclear Duty Officer, Duty/Shift Maintenance Supervisor, Duty Engineering Manager, Work Management Director, Work Week Manager, Senior Resident Inspector, Site Oversight Manager, and Corporate Communications Manager)

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 5 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • Unit 3 is holding at 55% power, while repairs are being performed on the 3A Feedwater Pump.
  • Annunciator D-4/3 (SECONDARY SAMPLE SYSTEM TROUBLE) is currently in alarm.
  • The Chemistry Department reports the following concentrations for the most recent 3A Steam Generator blowdown sample:

o Chloride - 140 ppb o Sodium - 300 ppb o Sulfate - 80 ppb Initiating Cue:

  • Based on the given conditions, determine the appropriate secondary-chemistry action level; the minimum required plant response; and any required notifications (list below):

o Action Level: __________

o Plant Response:

o Notifications:

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.1.a/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 6 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

If requested, provide examinee with a copy of window D-4/3 of 3-ARP-097.CR.D (Control Room Response - Panel D), 0-ADM-651 (Nuclear Evaluator Cue:

Chemistry Parameters Manual), and 3-ONOP-071.1, Secondary Chemistry Deviation from Limits.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 7 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Performance Step: 2 Determine the appropriate secondary-chemistry Action Level and Critical: Yes minimum required plant response.

Determine that the appropriate secondary-chemistry Action Level is 3 Standard: (based on sodium concentration) and the minimum required plant response is to reduce power below 5% as quickly as safe plant operations permit.

Evaluator Note: Appropriate response is from RNO for Step 9.b of 3-ONOP-071.1.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 8 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Performance Step: 3 For the given conditions, obtain the required materials and determine Critical: Yes the necessary notifications.

For the given conditions, obtain the required materials and determine that notifications must be made to the Chief Nuclear Officer, Site VP, Plant General Manager, Operations Manager, Nuclear Duty Officer, Duty/Shift Standard:

Maintenance Supervisor, Duty Engineering Manager, Work Management Director, Work Week Manager, Senior Resident Inspector, Site Oversight Manager, and Corporate Communications Manager.

If requested, provide examinee with a copy of 0-ADM-115 (Notification of Evaluator Cue: Plant Events), OP-AA-100-1000 (Conduct of Operations), and/or AD-AA-07 (Notification of Chief Nuclear Officer).

  • 0-ADM-115, Enclosure 2 (Plant Management and NRC Resident Notification Table) calls for notifications for the following conditions/

events:

o Unplanned shutdown or load reduction of greater than 5%

o Chemistry action level 2 or 3 Evaluator Note:

  • AD-AA-07 stipulates that the CNO will be notified for unplanned reductions in power (greater than 5%).
  • OP-AA-100-1000, Attachment 12 (Significant Event Reporting) calls for notifications for the following conditions/events:

o Action level 2 or greater chemistry parameters o Unplanned shutdown or load reduction of greater than 5%

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 3, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

02200063300, Review Primary and Secondary Sample Results and Determine Applicable Actions, Rev. 0 JPM Page 9 of 10 DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

  • Unit 3 is holding at 55% power, while repairs are being performed on the 3A Feedwater Pump.
  • Annunciator D-4/3 (SECONDARY SAMPLE SYSTEM TROUBLE) is currently in alarm.
  • The Chemistry Department reports the following concentrations for the most recent 3A Steam Generator blowdown sample:

o Chloride - 140 ppb o Sodium - 300 ppb o Sulfate - 80 ppb Initiating Cue:

  • Based on the given conditions, determine the appropriate secondary-chemistry action level; the minimum required plant response; and any required notifications (list below):

o Action Level: __________

o Plant Response:

o Notifications:

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

REVISION NO.: PROCEDURE TITLE: PAGE:

9B 24 CONTROL ROOM RESPONSE - PANEL D PROCEDURE NO.: WINDOW:

3-ARP-097.CR.D TURKEY POINT UNIT 3 4/3 (Page 1 of 1)

CAUSES: 1. Condenser tube leak

2. S/G carryover D4/3
3. Condensate and Feedwater System non-soluble particulates
4. Spent resin column or air inleakage SECONDARY SAMPLE SYSTEM TROUBLE DEVICE: SETPOINT: LOCATION:

MICROMAX 2 FFOS As determined by Secondary Chemistry N/A ALARM CONFIRMATION

1. IDENTIFY alarming channel on MICROMAX 2 FFOS.

OPERATOR ACTIONS

1. IF Condensate OR Feedwater dissolved oxygen indications are alarming, THEN DIRECT Unit 3 NPO to report Air-In-Leakage per 3-NOP-073.03, Steam Jet Air Ejector In-Leakage.
2. IF Air-In-Leakage is greater than 15 SCFM, THEN NOTIFY Chemistry to perform compensatory sampling per 0-NCCP-210, SPING and DAM Monitor Channel Checks.
3. DIRECT Secondary Nuclear Chemistry to sample affected component to validate alarm.
4. ENTER 3-ONOP-071.1, Secondary Chemistry Deviation from Limits, AND CONSULT Secondary Nuclear Chemistry for actions while awaiting sample confirmation.
5. IF high conductivity is confirmed, THEN GO TO appropriate section of 3-ONOP-071.1, Secondary Chemistry Deviation from Limits.
6. IF high conductivity can NOT be confirmed, THEN:

A. CONTINUE monitoring MICRMAX 2 FFOS.

B. CONSULT Secondary Nuclear Chemistry for further recommendations.

REFERENCES:

1. FPL Drwg 5613-M-3032, Sh 1
2. FPL Drwg 5613-M-3032, Sh 4
3. FPL Drwg 5610-E-26, Sh 39
4. FPL Drwg 5610-M-2
5. PC/M 93-205

Procedure No.: Procedure

Title:

Page:

37 Approval Date:

0-ADM-651 Nuclear Chemistry Parameters Manual 4/26/13 5.7 Secondary Systems - Power Operation 5.7.1 Steam Generators Blowdown ( 50% Power)

CONTROL PARAMETER NORMAL VALUE Cation Conductivity S/cm < 0.8 Chloride, ppb(c) < 1.0 Sodium, ppb < 0.5 (d)

Sulfate, ppb < 1.0 CONTROL Parameters Action Level Parameter Frequency 1 2 3 Cation Conductivity increase continuous - >1 (a) >4 (a) above baseline, µS/cm @ 25°C Sodium, ppb continuous >5 (b) >50 >250 Chloride, ppb (c) daily >10 (b)

>50 >250 Sulfate, ppb (d) daily >10 (b)

>50 >250 Diagnostic Parameters Parameter Consideration Continuously monitored. Refer to references [9, 10, and 11]

pH @ 25°C for pH Optimization, listed in Chapter 5 of Miscellaneous Document 2.1.5.2.

Refer to Chapter 7 - Reasonable consistency between values Specific Conductivity and pH of pH, ammonia, amine, conductivity, etc. should be Agent achieved.

Perform each planned shutdown. See Chapter 7 of Hideout Return Evaluation Miscellaneous Document 2.1.5.2.

Crevice chemistry, steam quality, and impurity source Silica consideration.

Notes:

a. At plants operating with organic amines, cation conductivity values will be elevated as a result of organic acids. In such cases, identification of a possible Action Level 2 or 3 excursion can be based on increases of 1 or 4 µS/cm, respectively, above the normal baseline value. The 1 µS/cm value corresponds to a maximum chloride concentration of 80 ppb or a maximum sulfate concentration of 110 ppb. The 4 µS/cm value corresponds to a maximum chloride concentration of 330 ppb or a maximum sulfate concentration of 445 ppb. If these Action Level 2 or Action Level 3 threshold values for cation conductivity increase above baseline are exceeded, the site has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to obtain a blowdown sample and analyze it to confirm both chloride and sulfate concentrations before being required to enter the appropriate Action Level based solely on cation conductivity increase above baseline. Once the blowdown sample is obtained and analyzed for both chloride and sulfate concentrations, action can be based on the measured chloride and sulfate concentrations and not on the cation conductivity increase above baseline.
b. Per Section 5.6.1 for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after escalation above the 50% (MPV) power level, the Action Level 1 threshold for sodium, chloride, and sulfate will remain 10 ppb, 20 ppb, and 20 ppb, respectively, before reverting to the Action Level 1 values in this Section 5.7.1.
c. Limit per Commitment - Step 2.3.1 is < 150 ppb.
d. Limit per Commitment - Step 2.3.1 is < 100 ppb.

W97:/mr/cls/cls

Procedure No.: Procedure

Title:

Page:

12 Approval Date:

3-ONOP-071.1 Secondary Chemistry Deviation from Limits 3/14/14 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES Requirements are based on 0-ADM-651, NUCLEAR CHEMISTRY PARAMETERS MANUAL.

Power reduction may be performed using 3-GOP-103, POWER OPERATION TO HOT STANDBY, or 3-GOP-100, FAST LOAD REDUCTION.

9 Monitor Action Level 2 OR 3 Reached Return to Step 8.

a. Check Step 8 (Action Level 1 response) a. Observe NOTE prior to Step 8 AND go to complete or in progress Step 8.
b. Check SG sodium, chloride and sulfate less than b. IF sodium, chloride, or sulfate exceed 250 ppb, 50 ppb OR exceed 50 ppb for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, THEN reduce power below 5% as quickly as safe plant operations permit.
c. Check SG Iodine <0.1 Ci/gm c. Be in at least Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Mode 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(Tech Spec 3/4.7.1.4).

d. Check hydrazine chemical injection available d. Restore hydrazine injection to service using 0-NCOP-076, CHEMICAL INJECTION SYSTEM OPERATION, OR IF the ratio of feedwater hydrazine to feedwater oxygen decreases to a value< 2 and is NOT restored to a value >=2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, determine shutdown rate after a discussion with Chemistry and Operations management. A shutdown to Mode 2 or lower may NOT be necessary. If this ratio is restored to a value of >=2, the plant may return to full power as consistent with 0-ADM-651, Section 5.6.2, and Action Level 1 criteria on Foldout Page.
e. Check SG cation conductivity <1.4 S/cm e. IF SG cation conductivity is:
  • <4.0 S/cm, AND Foldout Page NOTE 2 does NOT apply, THEN reduce power below 50%

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

  • >4.0 S/cm, AND Foldout Page NOTE 2 does NOT apply, THEN reduce power as recommended by Chemistry within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
f. Check feedwater dissolved oxygen <10 ppb f. Consult with Chemistry for the following:
1. Identify AND correct source of Oxygen.
2. Determine if a power reduction to below 50%

is appropriate.

g. Return to Step 8 W97:PEB/cls/fm/ab

Procedure No.: Procedure

Title:

Page:

23 Approval Date:

0-ADM-115 Notification of Plant Events 8/1/12 ENCLOSURE 2 (Page 1 of 3)

PLANT MANAGEMENT AND NRC RESIDENT NOTIFICATION TABLE Condition or Event Notes Significant plant configuration control issues. OP-AA-100-1000 Reactivity event, including any mis-positioned control OP-AA-100-1000 rod events.

Unplanned entry into a shutdown technical specification action statement 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in duration or OP-AA-100-1000 less.

Technical specification action statement that will not be OP-AA-100-1000 met within allowed time requirement.

Loss of offsite power. OP-AA-100-1000 Reactor or Turbine trip. OP-AA-100-1000 and AD-AA-07 Unplanned shutdown or load reduction of greater than OP-AA-100-1000 and AD-AA-07 5%.

Major equipment failure or malfunction (including OP-AA-100-1000 safeguards equipment).

Any unexpected, significant plant transient. OP-AA-100-1000 Load restriction or inability to meet load dispatcher OP-AA-100-1000 requirements.

Chemistry Action Level 2 or 3. OP-AA-100-1000 Unexplained reactivity change. OP-AA-100-1000 Any event or operating condition that occurs that is not OP-AA-100-1000 enveloped in the plant design basis.

Significant plant equipment damage (>$100,000). OP-AA-100-1000 and AD-AA-07 Shutdown risk classification > scheduled. OP-AA-100-1000 Orange or Red on line risk classification. OP-AA-100-1000 OP-AA-100-1000 and AD-AA-07 A leak or spill from the SFP or other contaminated source with potential to contaminate groundwater is not an Enclosure 2 event. Notify Chemistry to evaluate and report using Attachment 5. Chemistry is Hazardous, environmental, or radioactive material responsible for notifications involving event requiring EPA notification. radioactive releases. The hazardous material coordinator (HMC) is responsible for Federal, State, and Local agency reports for non-nuclear hazardous material releases through internal company departments. If internal company departments cannot be contacted, reports will be made directly to the agency.

W2002/AWT/cls/mr/mr

Policy No.:

FPL NUCLEAR POLICY AD-AA-07 Revision No.:

0 NOTIFICATION OF Effective Date:

CHIEF NUCLEAR OFFICER 01/04/10 Page No.:

1 of 2 POLICY STATEMENT The Chief Nuclear Officer (CNO) shall be notified of problems or events that can affect the operations of the FPL Nuclear Fleet.

POLICY INTENT Timely notifications are made to the Chief Nuclear Officer on problems or events that effect the proper operation of the nuclear sites.

  • Ensures that timely reports are provided to the Chief Nuclear Officer regarding reactor trips and OSHA Recordable Injuries.
  • Ensures refueling outage durations are approved by CNO.

APPLICABILITY CNO Direct Reports, Regional VP Direct Reports, and Nuclear Division Duty Officers IMPLEMENTATION The following information, important to the proper operation of the FPL Nuclear Fleet shall be brought to the attention of the CNO by the cognizant site and/or corporate management during normal business hours. The Nuclear Division Duty Officer shall be informed of this information by cognizant site and/or corporate contacts both during business and non-business hours; however, it is important that site / corporate management contact the CNO during normal business hours.

Problems / Events:

  • Problems or potential problems involving the Nuclear Regulatory Commission (NRC) (e.g. notifications, enforcement actions, etc.)
  • Injury of a serious nature or fatality of any employee or contractor.
  • A request to Access Control for an unfavorable termination of access that involves an issue of trustworthiness and reliability.
  • Acts of known or suspected sabotage.
  • External threats to generation (e.g. fires, accidents, system dispatch information).
  • Hazardous weather warnings (hurricanes, tornadoes, blizzards, or cold weather) which could affect normal plant operations.
  • Significant labor issues.

Policy No.:

FPL NUCLEAR POLICY AD-AA-07 Revision No.:

0 NOTIFICATION OF Effective Date:

CHIEF NUCLEAR OFFICER 01/04/10 Page No.:

2 of 2 Problems / Events (continued):

  • Significant quality issues - examples of such issues would include:
  • Any and all breakdowns in material control at FPL or any of its suppliers.
  • Systematic weaknesses in either programs or procedures being utilized by FPL.
  • Media interest or events likely to result in media interest.
  • Unplanned reductions in power (greater than 5%).
  • Spills or releases of radioactive material requiring immediate notification of state or federal agencies, or a leak or spill of radioactive material that meets the Fleet Ground Water Protection Program criteria to be communicated to state and local officials.
  • A non-radiological environmental event or occurrence for which immediate notification is required to any Local, State or Federal environmental authority.
  • A matter judged to be provocative and/or significant relating to the nuclear plants or staffs.

OSHA Recordable Injuries:

An Accident Investigation Summary Report, including corrective actions, is to be submitted by the Site Vice President to the CNO within 10 working days of the accident for any utility or contractor OSHA recordable injury.

Reactor Trips:

Following any reactor trip, the sites shall provide the CNO with a brief of the event. The elements of the report should include:

  • Cause / apparent cause of trip
  • Circumstances surrounding trip (ongoing maintenance, load threats, etc.)
  • Response of operating crew to event, including any human performance issues noted
  • Equipment malfunctions / anomalies noted
  • Any other items deemed significant The report is due within eight (8) hours of the trip and must be signed by the site Vice President.

Refueling Outages:

Information on refueling outage durations shall be approved by the CNO prior to the information being provided to any external organization.

REFERENCE None Approved: SIGNATURE ON FILE Chief Nuclear Officer

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 76 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 1 of 7) 1.0 PRINCIPLE This procedure describes the protocol to be used for reporting occurrences and significant events to ensure proper response is initiated both onsite and offsite, and to ensure that appropriate management is promptly informed of the event or occurrence.

2.0 STANDARD The Operations Director is responsible for notifying the Plant Manager or equivalent of any reported event, as appropriate. The events considered appropriate should be contained in the situations listed in the main body of this attachment. This attachment is not all-inclusive. Site specific procedures shall be referenced to ensure that all required notifications and reports are completed in a timely manner.

3.0 EXPECTATIONS 3.1 Shift Manager

1. The Shift Manager is the designated representative of station senior management and has command and control authority for the site.
2. The Shift Manager is responsible for directing appropriate immediate actions to place the unit(s) in a safe and stable condition following any transient / event, maintaining safe and conservative operation of the facility as the highest priority.
3. The Shift Manager is responsible for initial determination of all reportable or potentially reportable items, and is responsible to ensure that appropriate notifications are performed.

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 77 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 2 of 7) 4.0 MAIN BODY 4.1 Significant Event Reporting

1. For declarations of any EP classification, notifications shall be made in accordance with applicable site emergency plan procedures.

The Shift Manager is responsible for notification of all appropriate offsite personnel. The Unit Supervisor is responsible for the notification of the Shift Manager and necessary onsite personnel. The following situations require timely verbal notification:

A. Reactor trip, inadvertent radioactive liquid or gaseous releases, B. Major equipment failure or malfunction (includes all safeguards equipment),

C. Unexplained reactivity changes, D. Loss of offsite power, E. Major personnel injury or radiation overexposure, F. Accidents occurring on Station property (except minor injuries),

G. All regulatory reportable occurrences, H. Events requiring (red phone) NRC notification, I. Violations of local, state or federal pollution regulations, J. Turbine trip, or K. Load restrictions or inability to meet load dispatcher requirements (example: rod control problems).

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 78 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 3 of 7) 4.2 NRC Resident Inspector Notifications NOTE Consideration should be given to informing the NRC resident inspector of an issue anytime that the OPS manager is informed.

The Shift Manager or designee is responsible for notifying the resident NRC inspector in accordance with approved procedures to include but not limited to:

  • NRC immediate reports
  • Significant plant configuration control issues 4.3 Engineered Safety Feature (ESF) Actuation The Shift Manager is responsible to notify appropriate personnel and agencies of any engineered safety feature actuation, as required by the regulatory notification guidance.

4.4 Documentation

1. The Unit Supervisor should document in the Unit Journal the time of Shift Manager notification.
2. The Shift Manager shall document, in the Unit Journal, the time the Operations Director or Assistant Operations Manager is notified (when outside notification is required) or the time any outside agency is notified.

Refer to site specific reporting requirement procedures for detailed list of required notifications. The following table is an example of required notifications.

5.0 DOCUMENTATION None

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 79 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 4 of 7)

6.0 REFERENCES

1. AD-AA-07, Notification of Chief Nuclear Officer
2. NP-606, Reporting Environmental Events and Non-Compliance

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 80 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 5 of 7)

6.0 REFERENCES

(continued)

2. (continued)

EVENT NOTIFY

  • Reactivity event, including any mis-positioned control rod
  • Site VP events.
  • Plant Manager
  • Hazardous material incident.
  • Operations Director
  • Shutdown risk classification> scheduled or online risk
  • Nuclear Duty Officer classification - Orange
  • Senior Resident Inspector
  • Site Oversight Manager
  • Injury requiring offsite medical attention.
  • Corporate Communications
  • Significant plant equipment damage in excess of $100,000. Manager
  • Major enforcement actions, fines or other sanctions or a serious operating event that could lead to this action, including events which have been, or may be, brought to the attention of NRC upper management.
  • Action level II or greater chemistry parameters.
  • Any event or operating condition that occurs that is not enveloped in the plant design basis.
  • Significant labor issues.
  • Any event that proceeds in a way significantly different than expected; for example:
  • LCO action that will not be met within allowed time requirement;
  • Initiative of a prompt investigation or similar.
  • Plant Manager
  • Security threats of any nature against plant or personnel.
  • Operations Director
  • Acts of known or suspected sabotage
  • Nuclear Duty Officer
  • Hazardous weather warnings
  • Senior Resident Inspector
  • Corporate Communications Manager
  • System Controller

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 81 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 6 of 7)

6.0 REFERENCES

(continued)

2. (continued)

EVENT NOTIFY

  • Shutdown LCO action statement entry.
  • Site VP Unplanned entry into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or less shutdown LCO action statement.
  • Plant Manager
  • Operations Director
  • Nuclear Duty Officer
  • Duty Maintenance Supervisor
  • Work Week Manager
  • Senior Resident Inspector
  • Unplanned shutdown or load reduction of
  • Site VP greater than 5%.
  • Plant Manager
  • Operations Director
  • Nuclear Duty Officer
  • Duty / Shift Maintenance Supervisor
  • Duty Engineering Manager
  • Work Management Director
  • Work Week Manager
  • Senior Resident Inspector
  • Significant breakdown of plant radiological
  • Site VP or environmental controls.
  • Plant Manager
  • Any serious personnel radioactive contamination requiring extensive on-site
  • Sr. Manager Operations / Operations Director decontamination or outside assistance.
  • Nuclear Duty Officer
  • Duty / Shift Health Physics Supervisor
  • Senior Resident Inspector
  • Corporate Communications Manager

REVISION NO.: PROCEDURE TITLE: PAGE:

14 CONDUCT OF OPERATIONS 82 of 99 PROCEDURE NO.:

OP-AA-100-1000 NUCLEAR FLEET ADMINISTRATIVE ATTACHMENT 12 SIGNIFICANT EVENT REPORTING (Page 7 of 7)

6.0 REFERENCES

(continued)

2. (continued)

EVENT NOTIFY

  • A status or capability change on any
  • Site VP generator Reactive Power resource, including the status of each automatic
  • Plant Manager voltage regulator and power system
  • Operations Director stabilizer and the expected duration of the change in status or capability.
  • Nuclear Duty Officer
  • A status or capability change on any
  • Transmission System Operator (within 30 other Reactive Power resources under minutes) the Generator Operators control and the expected duration of the change in status or capability.

JPM TITLE: Approve Liquid Waste Release Permits JPM NUMBER: 02061051101 REV. 0-0 TASK NUMBER(S) I 02061051100 I TASK TITLE(S): Approve Liquid Waste Release Permits KIA NUMBERS: 2.3.6 KIA VALUE: SRO 3.8 Justification (FOR KIA VALUES <3.0): N/A TASK APPLICABILITY:

D RO ~SRO D STA D Non-Lie ~SRO CERT D OTHER: _ _

APPLICABLE METHOD OF TESTING: Simulate/Walkthrough: D Perform: ~

EVALUATION LOCATION: In-Plant: D Control Room: D Simulator: D Classroom: x I Lab: D Other: I Time for Completion: 20 Minutes Time Critical: No Alternate Path [NRC]: No Alternate Path [INPO]: No Developed by: _ _ _ _ _ ___,

11 2-t \ Lf Date Validated by:

Date Approved by:

Training Program Owner TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/SRO

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 2 of 10 JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 3 of 10 UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE New for L-15-1 NRC See cover page N/A 0-0 New JPM 01982463 Exam See cover page N/A 0-1 0-2 0-3 0-4 0-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 4 of 10 SIMULATOR SET-UP: N/A Required Materials:

  • 0-NOP-061.11A - marked-up copy
  • 0-NCOP-003, Attachments 1 & 5 - marked-up copies
  • 0-NCOP-003, Preparation of Liquid Release Permits

References:

  • 0-NOP-061.11A, Controlled Liquid Release from Recycle Monitor Tank A
  • 0-NCOP-003, Preparation of Liquid Release Permits
  • Demonstrate ability to approve release permits
  • Identify the requirements that must be met to perform a liquid release, when Process Radiation Monitor R-18 is out of service; given a completed Liquid Release Permit, recognize various conditions that invalidate the permit TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 5 of 10 I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • A liquid release is to be initiated from Recycle Monitor Tank A, with Process Radiation Monitor R-18 out of service.
  • Recycle Monitor Tank A has been recirculated, sampled, and analyzed, and a Radioactive Liquid Release Permit has been generated.
  • 0-NOP-061.11A, Controlled Liquid Release from Recycle Monitor Tank A, is complete through Step 4.1.1.1.

Initiating Cue:

  • The Shift Manager directs you to identify any procedural requirements that must be met prior to commencing the release and review the Radioactive Liquid Release Permit for completeness and accuracy.

Procedural Requirements: ______________________________________________________

Radioactive Liquid Release Permit issues, if any: ____________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.3/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 6 of 10 JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

  • Provide examinee with marked-up copies of 0-NCOP-003, Attachments 1 & 5, and 0-NOP-061.11A.

Evaluator Cue:

  • If requested, provide examinee with the following:

o Offsite Dose Calculation Manual (ODCM) o 0-NCOP-003, Preparation of Liquid Release Permits Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 7 of 10 Performance Step: 2 Identify any procedural requirements that must be met prior to Critical: Yes commencing the release, with R-18 out of service.

Identify the following requirements:

  • At least two independent samples must be analyzed, per the ODCM sampling/analysis program
  • At least two technically qualified members of the facility staff must independently verify the release rate calculations Standard:
  • At least two technically qualified members of the facility staff must independently verify the discharge valve lineup
  • A jumper must be installed from terminal 20 to terminal 16 on K850-QR-66 to temporarily defeat the R-18/RCV-018 interlock and permit operation of RCV-018
  • Shift Manager permission must be obtained prior to initiating the unmonitored release These requirements are stipulated in:
  • Section 2 (Radioactive Liquid Effluents) of the ODCM, in Control 2.1 (Radioactive Liquid Effluent Monitoring Instrumentation, Evaluator Note:

Functionality, and Alarm/Trip Setpoints) and Control 2.2 (Concentrations in Radioactive Liquid Effluents)

  • Section 2.2 (Limitations) of 0-NOP-061.11A Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 8 of 10 Performance Step: 3 Review the Radioactive Liquid Release Permit for completeness Critical: Yes and accuracy.

Recognize the following issues on the Radioactive Liquid Release Permit and determine that it is invalid/incomplete:

  • The value for Radioactive Analysis - Specific Activity (Liquid) in Part I exceeds 1 x 10-4 µCi/ml and the requisite analysis approval by the Standard: Radiochemist (or designee) in Part III is missing.
  • The value for Total Estimated Dose after this Release exceeds 0.25 mR/month, which is the Administrative Release Limit.
  • Two signatures are required in the Permit Prepared by block in Part III, but only one is present.
  • The Radiochemist (or designee) approval in Part III is stipulated in the permit and Step 2.1.2 of 0-NOP-061.11A.

Evaluator Note:

  • The requirement for two signatures in the Permit Prepared by block in Part III is stipulated in Step 7.3.7 of 0-NCOP-003.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee completes Step 3, state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

02061051101, Approve Liquid Waste Release Permits, Rev. 0 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 9 of 10 Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

  • A liquid release is to be initiated from Recycle Monitor Tank A, with Process Radiation Monitor R-18 out of service.
  • Recycle Monitor Tank A has been recirculated, sampled, and analyzed, and a Radioactive Liquid Release Permit has been generated.
  • 0-NOP-061.11A, Controlled Liquid Release from Recycle Monitor Tank A, is complete through Step 4.1.1.1.

Initiating Cue:

  • The Shift Manager directs you to identify any procedural requirements that must be met prior to commencing the release and review the Radioactive Liquid Release Permit for completeness and accuracy.

Procedural Requirements: ______________________________________________________

Radioactive Liquid Release Permit issues, if any: ____________________________________

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

Procedure No.

RKEY I O-NOP-061.11A NORMAL OPERA TING PROCEDURE ...,.____ Re-v-is-io-n-N-o.------i SAFETY RELATED 18 CONTINUOUS USE

Title:

CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A Responsible Department: OPERATIONS Special Considerations:

This is an Upgraded Procedure. Initial use should include increased awareness because of potential technical and/or sequential changes to the procedure. After initial use of this procedure, provide comments back to the Procedure Upgrade Project.

FOR INFORMATION ONLY Before use, verify revision and change documentation (if applicable) with a controlled index or document.

DATE VERIFIED INITIAL_--t:.____ _

Revision Approved By Approval Date UNIT#

DATE DOCT PROCEDURE 0 Steve Murano 03/09/10 DOCN O-NOP-061.11A SYS STATUS COMPLETED 18 Sam Shafer 06/03/14 REV 1B

  1. OF PGS

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 2 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT REVISION

SUMMARY

Rev. No. Description 1B PCR 1967308, 06/03/14, Adam Lefcourt Swap order of Steps 31 and 32 on Attachment 1 Page 10 of 10 and update procedure references.

1A PCR 1885460, 07/03/13, Shaun Matthews Editorial to correct nomenclature inconsistencies.

1 PCR 10-1859, 05/28/10, J. Mesa Added Precaution, Notes and Steps in order to address and implement changes incorp~rated in O-EPIP-20101, Duties of Emergency Coordinator.

0 PCR 09-2498, 03/09/10, GT Slaby New upgraded procedure replacing Sections 1.0 thru 4.0, 5.1, 5.2, 6.0 and, 7.0 of O-OP-061.11, revised 4/2/04 C1. As a group with O-NOP-061.11 A thru O-NOP-061.11 E, will supersede O-OP-061.11 in its entirety.

Deleted QA Pages and references to them throughout. Logging the activity satisfies atl QA record requirements. Deleted logging statements throughout.

The requirements on what to log are covered in Conduct of Operations, O-ADM-200 and Operations Narrative Logbooks, O-ADM-204. Deleted Precautions and Limitations that were generic in nature and could be considered as training aids. Moved Precautions, Limitations, and Prerequisites that did not apply to every section to the individual sections that they do apply.

Split procedure into attachments for field use and body for Control Room use.

Split procedure into two sections, placing a release with R-18 inoperable as an Infrequent Operation. Also split attachments for R-18 operable/inoperable to simplify IV requirements.

Expanded instructions for Fl-1064, WST COND PMP FLOW IND, inoperable to prevent reoccurrence similar to one described in CR# 2009-7590.

Replaced example thumbwheel settings with values typically found in O-PMl-067.5 and the Liquid Release Permit.

PCR 09-2508 Cancels O-OP-061.11.

REVISION NO.: PROCEDURE TITLE: PAGE:

1B CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 3 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ................................................................................................................... 4 2.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 4 3.0 PREREQUISITES ....................................................................................................... 6 4.0 NORMAL OPERATIONS ............................................................................................. 7 4.1 Startup ......................................................................................................................... 7 4.1.1 Unit 3 RCO Actions for Initiating a Controlled Liquid Release ..................................... 7 4.2 Operation ..................................................................................................................... 9 4.3 Shutdown .................................................................................................................. 10 5.0 INFREQUENT OPERATIONS ................................................................................... 11 5.1 Controlled Liquid Release with R-18 Not Operable ................................................... 11 5.1.1 Unit 3 RCO Actions for Initiating Controlled Liquid Release, with R-18 Not Operable .................................................................................................................... 11 5.1.2 Unit 3 RCO Actions for Terminating a Controlled Liquid Release, with R-18 Not Operable ............................................................................................................. 12 5.2 Flushing R-18 with Service Water ............................................................................. 13 6.0 RECORDS ................................................................................................................. 15

7.0 REFERENCES

AND COMMITMENTS ..................................................................... 16 ATTACHMENTS ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable .......... ._ ........................................................... 18 ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable ................................................................ 28

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 4 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 1.0 PURPOSE ,

This procedure provides instructions for making a controlled release of the contents of Recycle Monitor Tank A to Circulating Water.

PRECAUTIONS AND LIMITATIONS Precautions The SNPO shall perform the release only upon receipt of an approved Radioactive Liquid Release Permit, and a completed Recirculation and Sampling Verification Sheet from O-NCOP-003, Preparation of Liquid Release Permits, that is reviewed and signed by the Shift Manager.

Approval by the Radiochemist or designee is required for controlled discharge of waste tanks whose contents contain a total non-gas isotopic activity of greater than or equal to 1.0 E-4 microCuries/ml.

Process Radiation Monitor R-18 should be in service and frequently observed during the release to assure count rate is NOT approaching the R-18 Warning Limit as stated on the Liquid Release Permit.

If R-18 count rate exceeds the expected R-18 WARN limit, the release shall be terminated until the cause determined.

If R-18 count rate exceeds R-18 ALARM setpoint, release shall be terminated and Chemistry shall verify ODCM Limits NOT exceeded.

If there are NO Circulating Water Pumps operating, liquid effluent shall NOT be released from the plant.

A release that exceeds the R-18 Warn OR HIGH ALARM setpoints requires the SM to be notified to consult O-EPIP-20101, Duties of Emergency Coordinator.

REVASION NO.: PROCEDURE TITLE: PAGE:

1B CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 5 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT f85' Limitations

,J At least one Circulating Water Pump on Unit 3 or Unit 4 shall be operating. The Liquid Release Permit may require more than one Circulating Water Pump operating.

J() If PRMS R-18 is NOT operable, effluent releases may continue provided that prior to initiating a release:

15 At least two independent samples are analyzed per ODCM 2.1.1.

~ At least two technically qualified members of the Facility Staff independently verify the release rate calculations per ODCM 2.1.1.

Rf At least two technically qualified members of the Facility Staff independently verify discharge line valve alignment per ODCM 2.1.1.

J¥' A jumper installed from Terminal 20 to Terminal 16 on K850-QR-66, to temporarily defeat the R-18/RCV-018 interlock and permit operation of RCV-018.

!& SM permission is obtained prior to initiating the unmonitored release.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 6 of 37 PROCEDURE NO.:

O-NOP-061.11 A TURKEY POINT PLANT INITIAL

,ff PREREQUISITES

~ ENSURE Process Monitor R-18 is operable or equivalent sampling ~\

per Offsite Dose Calculation Manual 2.1 is available.

ENSURE Circulating Water System is in service per 3(4)-NOP-010, Circulating Water System.

ENSURE Laundry Drain System is aligned per O-NOP-061.10, Waste Disposal System Laundry Drain System.

ENSURE Waste Disposal System is aligned per O-NOP-061.13, Waste Disposal System - Transferring Water to Portable Demineralizer Skid for Processing.

ENSURE Fl-1064, WST COND PMP FLOW IND, is operable or flow estimation per Offsite Dose Calculation Manual 2.1 is available.

ENSURE power available to the applicable Radwaste Discharge to Seal Well solenoid valves:

J!j' 3P09-15 for SV-3-1413 and SV-3-1414 ff 4P09-15 for SV-4-1413 and SV-4-1414 End of Section 3.0

REVISION NO.: PROCEDURE TITLE: PAGE:

1B CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 7 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 4.0 NORMAL OPERATIONS e

@ Startup Unit 3 RCO Actions for Initiating a Controlled Liquid Release (1)( IF PRMS channel R-18 is out of service OR NOT operable, THEN

)(._) GO TO Section 5.1, Controlled Liquid Release with R-18 Not Operable.

2-. ENSURE receipt of approved Radioactive Liquid Release Permit and Tank Recirculation and Sampling Verification attachment of O-NCOP-003, Preparation of Liquid Release Permits.

A. REVIEW Radioactive Liquid Release Permit.

B. CHECK Recycle Monitor Tank A specified on Release Permit.

C. RECORD Radioactive Liquid Release Permit Number and Date/Time started on Attachment 1, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable.

3. ADJUST PRMS R-18 HIGH ALARM and WARN setpoints as follows:

A. SLIDE R-18 drawer out to gain access to thumbwheels.

NOTE

  • Thumbwheel switches used to select setpoints are configured as follows:
  • Forward Switch Setting = digit
  • Middle Switch Setting = decimal and digit
  • Rear Switch Setting = exponent
  • Examples:
  • Forward 4, Middle 0, Rear 4 equals to 4.0 X 104 or 40 Kcpm
  • Forward 2, Middle 5, Rear 4 equals to 2.5 X 104 or 25 Kcpm
  • Thumbwheel switch settings allow for only two significant digits. Values on the Radioactive Liquid Release Permit may have more significant digits.

Rounding down provides conservative setpoints.

  • A release that exceeds the R-18 Warn OR HIGH ALARM setpoints requires the SM to be notified to consult O-EPIP-20101, Duties of the Emergency Coordinator.

POSITION HIGH ALARM setpoint thumbwheels to R-18 Limit Setpoint specified on Radioactive Liquid Release Permit.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 8 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL 4.1.1 Unit 3 RCO Actions for Initiating a Controlled Liquid Release (continued)

3. (continued)

C. POSITION WARN setpoint thumbwheels to R-18 Warning Limit specified on Radioactive Liquid Release Permit.

D. CLOSE R-18 drawer.

E. Simultaneously PRESS HIGH ALARM AND CHECK digital display CPM indication equals R-18 Limit Setpoint specified on Radioactive Liquid Release Permit.

IV F. Simultaneously PRESS WARN AND CHECK digital display CPM indication equals Warning Limit specified on Radioactive Liquid Release Permit.

IV G. Simultaneously PRESS SOURCE CHECK AND CHECK digital display CPM indication responds to source.

4. ENSURE number of operating Circulating Water Pumps is equal to or greater than minimum specified on Radioactive Liquid Release Permit AND RECORD total number operating on Permit.
5. ENSURE Independent Verifications complete.

End of Section 4.1.1

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 9 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 4.2 Operation

1. PROVIDE the following to SNPO:
  • Radioactive Liquid Release Permit
  • Attachment 1, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable
2. REQUEST SNPO commence controlled release from Recycle Monitor Tank A.
3. WHEN notified by SNPO release is in progress, THEN PERFORM the following every 15 minutes until release is terminated:

A. RECORD R-18 reading.

8. CHECK R-18 count rate has NOT exceeded Warning Limit.

C. IF indication approaches Warning Limit, THEN NOTIFY SNPO AND SM.

4. WHEN notified by SNPO that release is complete, THEN:

A. OBTAIN completed Attachment 1, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable, from SNPO.

B. PERFORM Section 4.3.

End of Section 4.2

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 10 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL 4.3 Shutdown

1. IF Process Radiation Monitor R-18 HIGH ALARM setpoint was adjusted in Section 4.1.1, Step 3, THEN RESET R-18 HIGH ALARM setpoint by performing the following:

A. SLIDE R-18 drawer forward to gain access to thumbwheel switches.

NOTE

  • Thumbwheel switches used to select setpoints are configured as follows:
  • Forward switch setting = digit
  • Middle switch setting = decimal and digit
  • =

Rear switch setting exponent

  • Examples:

4

  • Forward 2, Middle 5, Rear 4 equals 2.5 X 10 or 25 Kcpm
  • Forward 4, Middle 0, Rear 4 equals 4.0 X 104 or 40 Kcpm
  • The R-18 HIGH ALARM setpoint is the Unusual Event EAL threshold value in O-EPIP-20101, Duties of the Emergency Coordinator.

B. POSITION HIGH ALARM setpoint thumbwheels to setting of 1.6 X 104 cpm.

IV C. POSITION WARN ALARM setpoint thumbwheels to setting of 1.0 X 104 cpm.

IV D. CLOSE R-18 drawer.

E. Simultaneously PRESS HIGH ALARM AND CHECK digital display CPM indication equals HIGH ALARM setpoint of 1.6 X 104 cpm.

IF R-18 HIGH ALARM OR WARN LIGHT is ON, THEN NOTIFY the SM to consult O-EPIP-20101, Duties of the Emergency Coordinator.

of~: :t!c:1 4.3

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 11 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL 5.0 INFREQUENT OPERATIONS 5.1 Controlled Liquid Release with R-18 Not Operable 5.1.1 Unit 3 RCO Actions for Initiating Controlled Liquid Release, with R-18 Not Operable

1. Obtain SM permission.
2. ENSURE receipt of approved Radioactive Liquid Release Permit and Tank Recirculation and Sampling Verification Sheet Attachment of O-NCOP-003, Preparation of Liquid Release Permits.

A. REVIEW Radioactive Liquid Release Permit.

8. CHECK Recycle Monitor Tank 6 specified on Radioactive Liquid Release Permit.

C. RECORD Radioactive Liquid Release Permit Number and Date/Time started on Attachment 2, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable.

IV

3. IF Process Radiation Monitor Channel R-18 is out of service due to failure or maintenance, causing a trip signal to be generated, THEN:

A. REQUEST l&C Maintenance install jumper from Terminal 20 to Terminal 16 on K850-R-18 in QR-66.

l&C IV

8. HANG Caution Tag is hung on Drawer R-18 to alert the operator(s) of the installed jumper.

IV

4. ENSURE number of operating Circulating Water Pumps is equal to or greater than minimum specified on Radioactive Liquid Release Permit AND RECORD total number operating on Permit.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 12 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL 5.1.1 Unit 3 RCO Actions for Initiating Controlled Liquid Release, with R-18 Not Operable (continued)

NOTE ODCM requirements for independent samples and release rate calculation verifications are confirmed by two Chemistry Department signatures in the PREPARED BY block of the Radioactive Liquid Release Permit.

5. CHECK ODCM requirements complete.

IV

6. ENSURE Independent Verifications complete.
7. PROVIDE the following to the SNPO:
  • Radioactive Liquid Release Permit
  • Attachment 2, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable
8. REQUEST SNPO commence controlled release from Recycle Monitor Tank A.

End of Section 5.1.1 5.1.2 Unit 3 RCO Actions for Terminating a Controlled Liquid Release, with R-18 Not Operable

1. WHEN notified liquid release is complete, THEN OBTAIN completed attachment from SNPO.
2. REQUEST l&C Maintenance remove jumper from Terminal 20 to Terminal 16 on K850-R-18 in QR-66.

l&C IV

3. REMOVE Caution Tag hung on Drawer R-18.

IV S~d!o.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 13 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 5.2 Flushing R-18 with Service Water NOTE This section is performed by request of Chemistry, RCO, or SM.

1. ENSURE the following valves CLOSED:
  • 4743, WCT PUMP* DISCH SAMPLE VALVE
  • 4761, WHT/LWT TO DISCH CANAL
  • 1805, LIQUID RELEASE MANUAL RECIRC TO RWB
  • 4745, WCT PUMP RECIRC
2. ENSURE 4749, LIQUID RELEASE STOP VALVE, LOCKED CLOSED.
3. ATTACH temporary water hose from 70-189, SERVICE WATER FLUSH VALVE, to hose connection at 4738, WASTE COND TKS DRAIN.
4. UNLOCK AND OPEN 4738, WASTE COND TKS DRAIN.
5. OPEN 47428, B WASTE CONDENSATE PUMP B DISCHARGE VALVE.
6. OPEN 4748, WCT PUMP RELEASE VALVE.
7. OPEN 4747, WASTE COND PUMPS AND MONITOR TANKS PUMP DISCH TO WHT.
8. OPEN 70-189, SERVICE WATER FLUSH VALVE.

NOTE R-18 should NOT be flushed for longer than 7 minutes.

9. FLUSH R-18 with Service Water until any of the following conditions are met:
  • Background count rate returns to normal Maximum of 7 minutes elapse

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 14 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL 5.2 Flushing R-18 with Service Water (continued)

10. WHEN R-18 flush is complete, THEN:

A. CLOSE 70-189.

8. CLOSE AND LOCK 4738, WASTE COND TKS DRAIN.

IV

c. CLOSE 47428.

D. CLOSE 4748.

E. CLOSE 4747.

11. REMOVE temporary hose installed in Section 5.2, Step 3 AND STORE in assigned area.

IV End of Section 5.2

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 15 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 6.0 RECORDS

1. Completed pages of the below listed items constitute Quality Assurance Records and shall be transmitted to QA Records for retention in accordance with Quality Assurance Records Program:
  • Attachment 1, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable
  • Attachment 2, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable
2. Completed copies of the below listed items, shall be retained in the Shift Manager file until the next performance of that section or attachment.
  • Attachment 1, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable
  • Attachment 2, Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable
3. The date, time, and section completed shall be logged in the Unit Narrative Log.
4. Any problems encountered while performing the procedure should be logged in the Unit Narrative Log (i.e., malfunctioning equipment, delays due to changes in plant conditions, etc.).

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 16 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT

7.0 REFERENCES

AND COMMITMENTS 7 .1 References 7.1.1 Implementing

1. O-ADM-213, Technical Specification Related Equipment and Risk Significant SSC Out-Of Service Logbook
2. O-NCOP-003, Preparation of Liquid Release Permits
3. 3(4)-NOP-010, Circulating Water System
4. O-NOP-061.10, Waste Disposal System - Laundry Drain System
5. O-NOP-061.13, Waste Disposal System - Transferring Water to Portable Demineralizer Skid for Processing
6. Offsite Dose Calculation Manual 7.1.2 Developmental
1. 561 O-M-3046, Sheet 4 - CVCS, Boron Recycle System
2. 561 O-M-3061, Sheet 4 - Waste Disposal System Liquid, Polishing Demineralizer
3. 561 O-M-3061, Sheet 8 - Waste Disposal System Liquid, Waste Monitor Tanks
4. 5613-M-3010, Sheet 1 - Circulating Water
5. O-PMl-067.5, Process Radiation Monitoring System Channel R-18 Calibration Procedure
6. Memorandum JPE-PTP0-85-1209, Dated Nov. 7, 1985, Item 27
7. PC/M 88-264, PRMS Rate Meter Output
8. PC/M 96-047, Replacement of FT-1064 with Addition of an Isolation and Drain Valve
9. DRM-200 (V-6) Digital Ratemeter Operation and Maintenance Manual, Z0889

REVISION NO.: PROCEDURE TITLE: PAGE:

1B CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 17 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT 7.1.3 Management Directives None 7.2 Commitments None

REVISION NO.: PROCEDURE TITLE: PAGE:

18 .CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 18 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 1 Controlled liguid Release from Rec~cle Monitor Tank A with R-18 OQerable (Page 1 of 10)

LIQUID RELEASE PERMIT NUMBER: DATE/TIME: I REMARKS:

PERSONNEL PERFORMING MANIPULATIONS PRINTED NAME INITIALS COMMENCEMENT OF RELEASE DATE RELEASE (MM/DD/YY): TIME RELEASE (HH:MM, 24 HR CLOCK):

STARTED: I I STARTED:

RELEASE PRINTED NAME I SIGNATURE:

STARTED BY:

COMPLETION OF RELEASE DATE RELEASE (MM/DD/YY): TIME RELEASE (HH:MM, 24 HR CLOCK):

COMPLETED: I I COMPLETED:

RELEASE PRINTED NAME I SIGNATURE:

COMPLETED BY:

REVIEW OF RELEASE RELEASE PRINTED NAME I SIGNATURE: DATE: (MM/DD/YY):

REVIEWED BY: I I

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 19 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liguid Release from Rec~cle Monitor Tank A with R-18 Operable (Page 2 of 10)

1. REVIEW Radioactive Liquid Release Permit.
2. CHECK Recycle Monitor Tank 6 specified on Release Permit.
3. IF Recycle Monitor Tank A ORB is on recirculation, THEN STOP Monitor Tank Pump per O-NOP-061.10, Waste Disposal System -

Laundry Drain System.

4. COMPLETE Table 1, Recycle Monitor Tank A Pre-Release Alignment:

Table 1 Recycle Monitor Tank A Pre-Release Alignment COMPONENT POSITION ALIGNED COMPONENT DESCRIPTION NUMBER REQUIRED BY MONITOR TANK ROOM LAUNDRY WASTE WATER FILTER TO BA 1305 CLOSED MONITOR TANKA ISOL VALVE 1306 MONITOR TANK PUMP A OUTLET TO RAD CLOSED WASTE BUILDING 1307 MONITOR TANK PUMP B OUTLET TO RAD CLOSED WASTE BUILDING 1249B MONITOR TANK PUMPS DISCH LOCAL CLOSED SAMPLE 1804 MONITOR TANK PUMP DISCH TO WHT CLOSED DISCH HOR 4748 WCT PUMP RELEASE VALVE CLOSED WASTE COND PUMPS AND MONITOR 4747 CLOSED TANKS PUMP DISCH TO WHT 4745 WCT PUMP RECIRC CLOSED 4761 WHT/LWT TO DISCH CANAL CLOSED

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 20 of 37 PROCEDURE NO.:

O-NOP-061 . 11 A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liguid Release from Recycle Monitor Tank A with R-18 Operable (Page 3 of 10)

4. (continued)

Table 1 Recycle Monitor Tank A Pre-Release Alignment COMPONENT POSITION ALIGNED COMPONENT DESCRIPTION NUMBER REQUIRED BY MONITOR TANK ROOM 1805 LIQUID RELEASE MANUAL RECIRC TO RWB CLOSED 4743 WCT PUMP DISCH SAMPLE VALVE CLOSED 1279 MONITOR TANK PUMP RECIRC TO CLOSED MONITOR TANKA STOP 1283 MONITOR TANK PUMP RECIRC TO CLOSED MONITOR TANK B STOP 4749 STOP VALVE BEFORE RCV-018 WASTE TO LOCKED DISCHARGE CLOSED 4798 ISOLATION VALVE FOR FT-1064 OPEN 4799 DRAIN VALVE FOR FT-1064 CLOSED

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 21 of 37 PROCEDURE NO.:

O-NOP-061 . 11 A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liguid Release from Recy:cle Monitor Tank A with R-18 012erable (Page 4 of 10)

5. IF using P206A Monitor Tank Pump A for release, THEN COMPLETE Table 2.

Table 2 Recycle Monitor Tank A Release Via P206A Monitor Tank Pump A COMPONENT POSITION ALIGNED COMPONENT DESCRIPTION NUMBER REQUIRED BY MONITOR TANK ROOM 1282 MONITOR TANKA OUTLET VALVE OPEN 1288 MONITOR TANK PUMP A SUCTION OPEN

. 1290 MONITOR TANK PUMP A DISCHARGE OPEN 1303 MONITOR TANK CROSS CONNECTION CLOSED 1281 MONITOR TANK B OUTLET CLOSED THROTTLE MONITOR TANK PUMP RECIRC TO 1279  % turn MONITOR TANK A STOP OPEN WASTE BORON PANEL - SOUTH MONITOR TANK PUMP A TANK SELECTOR P206A MT-A SWITCH

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 22 of 37 PROCEDURE NO.:

O-NOP-061 . 11 A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liguid Release from Rec~cle Monitor Tank A with R-18 O~erable (Page 5 of 10)

6. IF using P206B Monitor Tank Pump B for release, THEN COMPLETE Table 3:

Table 3 Recycle Monitor Tank A Release Via P206B Monitor Tank Pump B COMPONENT POSITION ALIGNED COMPONENT DESCRIPTION NUMBER REQUIRED BY MONITOR TANK ROOM 1281 MONITOR TANK B OUTLET CLOSED 1282 MONITOR TANKA OUTLET VALVE OPEN 1303 MONITOR TANK CROSS CONNECTION OPEN 1292 MONITOR TANK PUMP B SUCTION OPEN 1294 MONITOR TANK PUMP B DISCHARGE OPEN 1290 MONITOR TANK PUMP A DISCHARGE CLOSED THROTTLE MONITOR TANK PUMP RECIRC TO 1279  % turn MONITOR TANK A STOP OPEN WASTE BORON PANEL - SOUTH MONITOR TANK PUMP B TANK SELECTOR P206B MT-A SWITCH

7. START Monitor Tank Pump aligned in Attachment 1, Step 5 or Attachment 1, Step 6.

RCV-018, WASTE DISCHARGE LINE VALVE.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 23 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable (Page 6 of 10)

9. UNLOCK AND OPEN 4749, STOP VALVE BEFORE RCV-018 WASTE TO DISCHARGE.
10. Slowly OPEN 1296, MONITOR TANK PUMP TO WASTE DISPOSAL SYSTEM, until Fl-1064, WST COND PMP FLOW IND, indicates release flow rate specified on Radioactive Liquid Release Permit.
11. NOTIFY U3 RCO Recycle Monitor Tank A release has commenced.

NOTE If Fl-1064, WST COND PMP FLOW IND, is out of service, the release may continue provided an estimate of the release flow rate is made every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the release, per ODCM Action 2.1.3.

12. IF at any time, Fl-1064, WST COND PMP FLOW IND, is out of service, THEN:

A. NOTIFY Chemistry estimates of release flow rate are required every four hours during the release per ODCM Action 2.1.3.

B. NOTIFY RCO to log entry in Equipment Out-Of-Service Logbook per O-ADM-213, Technical Specification Related Equipment and Risk Significant SSC Out-Of-Service Logbook.

13. RECORD the following on Radioactive Liquid Release Permit and SNPO Log Book:
  • Start time
  • Tank level
  • Release flow rate

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 24 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable (Page 7 of 10)

NOTE An R-18 Warning Limit may be exceeded due to a miscalculation on the Liquid Release Permit, an incorrect thumbwheel adjustment, rapid manipulation of Valve 1296, or other factors associated with high activity. All data should be evaluated prior to the re-issuance of the Liquid Release Permit.

14. REQUEST RCO monitor R-18 to ensure both the following:
  • R-18 responds to the liquid release.
  • R-18 count rate does NOT exceed the Warning Limit.

NOTE Release time for Recycle Monitor Tanks should be approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

15. MONITOR waste tank levels to ensure Recycle Monitor Tank A level lowers, and all other waste tank levels NOT changing unexpectedly.

NOTE Valve 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANK A STOP, throttled ~turn OPEN, should prevent Recycle Monitor Tank Pump cavitation, however, if the pump exhibits cavitation characteristics, 1279 should be throttled while maintaining a recirculation flowpath.

16. MONITOR operating Monitor Tank Pump for cavitation.

A. IF cavitation is evident, THEN ADJUST 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANKA STOP position, while maintaining a recirculation flowpath.

17. IF at any time, R-18 count rate exceeds the Warning Limit, THEN:

A. STOP Monitor Tank Pump used for release.

8. GO TO Attachment 1, Step 19 to terminate the release.

1 WHEN Recycle Monitor Tank A level lowers to between 15% and 10%, THEN Monitor Tank Pump used for i-:'.- --:

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 25 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable (Page 8 of 10)

19. RECORD the following on the Radioactive Liquid Release Permit:
  • Stop time
  • Tank level
  • Required R-18 count rate data
20. CLOSE AND LOCK 4749, STOP VALVE BEFORE RCV-018 WASTE TO DISCHARGE.

IV

21. CLOSE RCV-018, WASTE DISCHARGE LINE VALVE.

IV

22. CLOSE 1296, MONITOR TANK PUMP TO WASTE DISPOSAL SYSTEM.

IV

23. ENSURE Monitor Tank Pump used for release, STOPPED.
24. CLOSE 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANK A STOP.
25. ENSURE 1282, MONITOR TANK A OUTLET VALVE, CLOSED.
26. ENSURE 1303, MONITOR TANK CROSS CONNECTION, CLOSED.
27. ENSURE 1290, MONITOR TANK PUMP A DISCHARGE, OPEN.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 26 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable (Page 9 of 10)

28. FLUSH R-18 with Service Water as follows:

A. ATTACH temporary hose from 70-189, SERVICE WATER FLUSH VALVE, to hose connection at 4738, WASTE COND TKS DRAIN.

8. UNLOCK AND OPEN 4738, WASTE COND TKS DRAIN.

C. OPEN 47428, B WASTE CONDENSATE PUMP B DISCHARGE VALVE.

D. OPEN 4748, WCT PUMP RELEASE VALVE.

E. OPEN 4747, WASTE COND PUMPS AND MONITOR TANKS PUMP DISCH TO WHT.

F. OPEN 70-189, SERVICE WATER FLUSH VALVE.

NOTE R-18 should NOT be flushed for longer than 7 minutes.

G. CONTINUE flushing R-18 with Service Water until any of the following conditions are met:

  • Background count rate returns to normal
  • Maximum of 7 minutes elapse H. WHEN R-18 flush is complete, THEN:

(1) CLOSE 70-189.

(2) CLOSE AND LOCK 4738, WASTE COND TKS DRAIN.

IV (3) CLOSE 47428.

4748.

REVISION NO.: PROCEDURE TITLE: PAGE:

1B CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 27 of 37 PROCEDURE NO.:

O-NOP-061 . 11 A TURKEY POINT PLANT INITIAL ATTACHMENT 1 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Operable (Page 10 of 10)

28. (continued)

I. REMOVE temporary hose installed in Attachment 1, Step 28.A AND STORE in assigned area.

IV

29. RETURN completed Radioactive Liquid Release Permit to Hot Chemistry Lab.
30. INFORM U3 RCO of completion of the liquid release.
31. LOG completion of the release in the Narrative Logs.
32. RETURN this procedure to the Control Room.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 28 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liguid Release from Rec~cle Monitor Tank A with R-18 Not 012erable (Page 1 of 10)

LIQUID RELEASE PERMIT NUMBER: DATE/TIME: I REMARKS:

PERSONNEL PERFORMING MANIPULATIONS PRINTED NAME INITIALS DATE RELEASE STARTED:

(MM/DD/YY):

I COMMENCEMENT OF RELEASE I

TIME RELEASE STARTED:

l (HH:MM, 24 H~ CLOCK)

RELEASE PRINTED NAME I SIGNATURE:

STARTED BY:

COMPLETION OF RELEASE DATE RELEASE COMPLETED:

(MM/DD/YY):

I I TIME RELEASE COMPLETED:

I(HH:MM, 24 H~ CLOCK):

RELEASE PRINTED NAME I SIGNATURE:

COMPLETED BY:

l I REVIEW OF RELEASE RELEASE PRINTED NAME I SIGNATURE: DATE (MM/DD~)

REVIEWED BY: I

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 29 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 2 of 10)

1. REVIEW Radioactive Liquid Release Permit.
2. CHECK Recycle Monitor Tank~ specified on Release Permit.
3. IF Recycle Monitor Tank A OR B is on recirculation, THEN STOP Monitor Tank Pump per O-NOP-061.10, Waste Disposal System -

Laundry Drain System.

4. COMPLETE Table 4, Recycle Monitor Tank A Pre-Release Alignment:

Table 4 Recycle Monitor Tank A Pre-Release Alignment COMPONENT POSITION ALIGNED VERIFIED COMPONENT DESCRIPTION NUMBER REQUIRED BY BY MONITOR TANK ROOM LAUNDRY WASTE WATER FILTER 1305 TO BA MONITOR TANK A ISOL CLOSED VALVE MONITOR TANK PUMP A OUTLET 1306 CLOSED TO RAD WASTE BUILDING MONITOR TANK PUMP B OUTLET 1307 CLOSED TO RAD WASTE BUILDING MONITOR TANK PUMPS DISCH 12498 CLOSED LOCAL SAMPLE MONITOR TANK PUMP DISCH TO 1804 CLOSED WHT DISCH HOR 4748 WCT PUMP RELEASE VALVE CLOSED WASTE COND PUMPS AND 4747 MONITOR TANKS PUMP DISCH CLOSED TOWHT

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 30 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liguid Release from Rec~cle Monitor Tank A with R-18 Not OQerable (Page 3 of 10)

4. (continued)

Table 4 Recycle Monitor Tank A Pre-Release Alignment COMPONENT POSITION ALIGNED VERIFIED COMPONENT DESCRIPTION NUMBER REQUIRED BY BY MONITOR TANK ROOM 4745 WCT PUMP RECIRC CLOSED 4761 WHT/LWT TO DISCH CANAL CLOSED LIQUID RELEASE MANUAL 1805 CLOSED RECIRC TO RWB WCT PUMP DISCH SAMPLE 4743 CLOSED VALVE MONITOR TANK PUMP RECIRC 1279 CLOSED TO MONITOR TANK A STOP MONITOR TANK PUMP RECIRC 1283 CLOSED TO MONITOR TANK B STOP STOP VALVE BEFORE RCV-018 LOCKED 4749 WASTE TO DISCHARGE CLOSED 4798 ISOLATION VALVE FOR FT-1064 OPEN 4799 DRAIN VALVE FOR FT-1064 CLOSED

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 31 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 4 of 10)

5. IF using P206A Monitor Tank Pump A, THEN COMPLETE Table 5:

Table 5 Recycle Monitor Tank A Release Via P206A Monitor Tank Pump A COMPONENT POSITION ALIGNED VERIFIED COMPONENT DESCRIPTION NUMBER REQUIRED BY BY MONITOR TANK ROOM 1282 MONITOR TANKA OUTLET VALVE OPEN MONITOR TANK CROSS 1303 CLOSED CONNECTION MONITOR TANK PUMP A 1288 OPEN SUCTION MONITOR TANK PUMP A 1290 OPEN DISCHARGE 1281 MONITOR TANK B OUTLET CLOSED THROTTLE MONITOR TANK PUMP RECIRC 1279  % turn TO MONITOR TANK A STOP OPEN WASTE BORON PANEL - SOUTH MONITOR TANK PUMP A TANK P206A MT-A SELECTOR SWITCH

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 32 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liguid Release from Rec~cle Monitor Tank A with R-18 Not OQerable (Page 5 of 10)

6. IF using P2068 Monitor Tank Pump B, THEN COMPLETE Table 6:

Table 6 Recycle Monitor Tank A Release Via P2068 Monitor Tank Pump B COMPONENT POSITION ALIGNED VERIFIED COMPONENT DESCRIPTION NUMBER REQUIRED BY BY MONITOR TANK ROOM 1281 MONITOR TANK B OUTLET CLOSED 1282 MONITOR TANK A OUTLET VALVE OPEN MONITOR TANK CROSS 1303 OPEN CONNECTION MONITOR TANK PUMP B 1292 OPEN SUCTION MONITOR TANK PUMP B 1294 OPEN DISCHARGE MONITOR TANK PUMP A 1290 CLOSED DISCHARGE THROTTLE MONITOR TANK PUMP RECIRC 1279  % turn TO MONITOR TANK A STOP OPEN i

WASTE' BORON PANEL - SOUTH MONITOR TANK PUMP B TANK P2068 MT-A SELECTOR SWITCH

7. START Monitor Tank Pump aligned in Attachment 2, Step 5 or Attachment 2, Step 6.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 33 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 6 of 10)

NOTE ODCM requirements for independent samples and release rate calculation verifications are confirmed by two Chemistry Department signatures in PREPARED BY block of Liquid Release Permit.

8. CHECK ODCM requirements complete.

IV

9. OPEN RCV-018, WASTE DISCHARGE LINE VALVE.

IV

10. UNLOCK AND OPEN 4749, STOP VALVE BEFORE RCV-018 WASTE TO DISCHARGE.

IV

11. ENSURE Verifications complete.
12. Slowly OPEN 1296, MONITOR TANK PUMP TO WASTE DISPOSAL SYSTEM, until Fl-1064, WST COND PMP FLOW IND, indicates release flow rate specified on Radioactive Liquid Release Permit.
13. NOTIFY U3 RCO Recycle Monitor Tank A release has commenced.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 34 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 7 of 10)

NOTE If Fl-1064, WST COND PMP FLOW IND, is out of service, the release may continue provided an estimate of the release flow rate is made every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the release, per ODCM Action 2.1.3.

14. IF at any time Fl-1064, WST COND PMP FLOW IND, is out of service, THEN:

A. NOTIFY Chemistry estimates of release flow rate are required every four hours during the release per ODCM Action 2.1.3.

B. NOTIFY RCO to log entry in Equipment Out-Of-Service Logbook per O-ADM-213, Technical Specification Related Equipment and Risk Significant SSC Out-Of-Service Logbook.

15. RECORD the following on Radioactive Liquid Release Permit and SNPO Log:
  • Start time
  • Tank level
  • Release flow rate NOTE Release time for Recycle Monitor Tanks should be approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
16. MONITOR waste tank levels to ensure Recycle Monitor Tank A level lowers, and all other waste tank levels NOT changing unexpectedly.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 35 of 37 PROCEDURE NO.:

O-NOP-061.11 A TURKEY POINT PLANT INITIAL ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 8 of 10)

NOTE Valve 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANKA STOP, throttled% turn OPEN should prevent Recycle Monitor Tank Pump cavitation, however, if pump exhibits cavitation characteristics, 1279 should be throttled while maintaining a recirculation flowpath.

17. MONITOR Monitor Tank Pump for cavitation.

A. IF cavitation is evident, THEN ADJUST 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANKA STOP position, while maintaining a recirculation flowpath.

18. WHEN Recycle Monitor Tank A level between 15% and 10%, THEN STOP Monitor Tank Pump, used for release.
19. RECORD the following on Radioactive Liquid Release Permit:
  • Stop time
  • Tank level
20. CLOSE AND LOCK 4749, STOP VALVE BEFORE RCV-018 WASTE TO DISCHARGE.

IV

21. CLOSE RCV-018, WASTE DISCHARGE LINE VALVE.

IV

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 36 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 9 of 10)

22. CLOSE 1296, MONITOR TANK PUMP TO WASTE DISPOSAL SYSTEM.

IV

23. ENSURE Monitor Tank Pump used for release, STOPPED.
24. CLOSE 1279, MONITOR TANK PUMP RECIRC TO MONITOR TANK A STOP.
25. ENSURE 1282, MONITOR TANK A OUTLET VALVE, CLOSED.
26. ENSURE 1303, MONITOR TANK CROSS CONNECTION, CLOSED.
27. ENSURE 1290, MONITOR TANK PUMP A DISCHARGE, OPEN.
28. FLUSH R-18 with Service Water as follows:

A. ATTACH temporary hose from 70-189, SERVICE WATER FLUSH VALVE, to hose connection at 4738, WASTE COND TKS DRAIN.

8. UNLOCK AND OPEN 4738, WASTE COND TKS DRAIN.

C. OPEN 47428, B WASTE CONDENSATE PUMP B DISCHARGE VALVE.

D. OPEN 4748, WCT PUMP RELEASE VALVE.

E. OPEN 4747, WASTE COND PUMPS AND MONITOR TANKS PUMP DISCH TO WHT.

F. OPEN 70-189, SERVICE WATER FLUSH VALVE.

NOTE R-18 should NOT be flushed longer than 7 minutes.

G. FLUSH R-18 with Service Water for a maximum of 7 minutes.

REVISION NO.: PROCEDURE TITLE: PAGE:

18 CONTROLLED LIQUID RELEASE FROM RECYCLE MONITOR TANK A 37 of 37 PROCEDURE NO.:

O-NOP-061.11A TURKEY POINT PLANT INITIAL ATTACHMENT 2 Controlled Liquid Release from Recycle Monitor Tank A with R-18 Not Operable (Page 10 of 10)

28. (continued)

H. WHEN R-18 flush is complete, THEN:

(1) CLOSE 70-189.

(2) CLOSE AND LOCK 4738, WASTE COND TKS DRAIN.

IV (3) CLOSE 47428.

(4) CLOSE 4748.

(5) CLOSE 4747.

I. REMOVE temporary hose installed in Attachment 2, Step 28.A AND STORE in assigned area.

IV

29. RETURN completed Radioactive Liquid Release Permit to Hot Chemistry Lab.
30. INFORM U3 RCO of completion of liquid release.
31. RETURN this attachment to the Control Room.
32. LOG completion of release in SNPO Log Book.

Procedure No.: Procedure

Title:

Page:

25 Approval Date:

O-NCOP-003 Preparation of Liquid Release Permits 11/1/10 ATTACHMENT 1 (Page 1 of 1)

RADIOACTIVE LIQUID RELEASE PERMIT FLORIDA POWER AND LIGHT CO LRP No. L- 2 oJ x -0/ 5" TURKEY POINT PLANT RADIOACTIVE LIQUID RELEASE PERMIT DATE: -tvol~

Monitor Tank Waste Monitor Tank Volume to be Released /01 ooo Gals.

I

)(A DB DA DB DC Part I - Pre-Release Data and Calculations Radiochemical Analysis - Specific Activity (Liquid) 4. 33CJ e-3 µCi/ml Calculated Activity to be Released I* "42 e S µCi Estimated Dose for this Release "

  • cnei /,, E - 3 mR Month-to-date dose prior to this Release 2.4boE- I mR Total Estimated Dose after this Release '2.*52..G}E-\ mR Administrative Release Limit 0 .25 mR/month LC/EC Dissolved Gas Activity after dilution Expected R-18/19 Countrate LC/EC :s; 1.0

<2 X 10-4 !JCi/ml ,

'1* S"t,be-2.

'2.*~S-4 E4 CPM

!JCi/ml Part II - Limits R-18/19 Background =CPM ( Z A2oe~ ) R-18/19 Setpoint = CPM ('4* ooo e+ )

R-18/19 Warning . =CPM (S-. lo&E~ )

Max. Release Flow Rate 100 GPM Min. No. of CW Pumps ohe. ( 1)

~ h.o~ A.t\ D Recirc. Start Time Recirc. Pump Disch Press

'~ ho&A~ 1--o 0...1:'\0 psig Sample Time Recirculation Pump Flowrate: ., ,.;)/~ GPM Notes: )

Minimum 2 hr recirc. time when using 1 inch mini-recirc. on WMTs.

Part III - Authorization and A rovals:

The approval of the analysis by the Radiochemist (or designee) shall be obtained if the Specific Activity in Part 1 is greater than or equal to 1 x 10-4 µCi/ml. The Shift Manager shall review and sign Attachment 5 ensuring that the tank recirculated was the same tank that was sampled and that the ermit was enerated for the correct tank.

Technician Radiochemist SM Part IV - Release Data Release Performed By No. of Circ. Water pumps in service , Units 1 and 2 1 Units 3 and 4 Release Date:  : Release Start Time:  : Release Stop Time:

R-18 (or R-19) Readings every I I I I I I I I I I I 15 min from the start of the release I I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

Recorder/Meter Readings (CPM)  : Maximum  : Average Flow Rate (Estimate) GPM , le-Ve1 bcl'ore - - - - - - o;; - 1 Level after  %

Procedure No.: Procedure

Title:

Page:

31 O-NCOP-003 Preparation of Liquid Release. Permits 6/24/12 ATTACHMENT 5 (Page 1 of 1)

TANK RECIRCULATION AND SAMPLING VERIFICATION SHEET I have verified that the _ _A_-_/vl_I ___ tank was placed on recirculation at /2.. hourrs a..,,, on TIME to~ with a flowrate of __N~/_A ___ gpm [for WMT] OR a recirculation pump DATE discharge pressure of ___ To___ psig [for MT].

1----------~--------- l If the WMT is on mini recirc, a minimum 2 hr recirc is required.

.J This was verified by all of the following methods:

J5 Valve lineup

,)' Logbook entries jfj' Review of applicable procedure Senior Nuclear Plant Operator: --~~-------

Date/Time: fo~ I I 2. houyJ "Jo I have sampled the A- /v1T tank in accordance with (circle one):

CQY-TP-104-ooID CY-TP-104-004 7 This tank was sampled at '"1oc.urs 4jv on TIME

-fvftJ DA and was verified to be recirculated One tank volume prior to sampling.

Chemistry Technician: ---~......__,_

Shift Manager or designee:

FINAL PAGE

TURKEY POINT UNITS 3 & 4 OFFSITE DOSE CALCULATION MANUAL 2.0 RADIOACTIVE LIQUID EFFLUENTS CONTROL 2.1: RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION, FUNCTIONALITY AND ALARM/TRIP SETPOINTS The radioactive liquid effluent monitoring instrumentation channels shown in Table 2.1-1 shall be FUNCTIONAL with their Alarm/Trip Setpoints set to ensure that the limits of Control 2.2 are not exceeded. The Alarm/ Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in this OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY : At all times, except as indicated in Table 2.1-1 ACTION :

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above Control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel non-functional, or change the setpoint so it is acceptably conservative.
b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels FUNCTIONAL, take the ACTION shown in Table 2.1-1. Restore the non-functional instrumentation to FUNCTIONAL status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to Administrative Control 1.3, why this non-functionality was not corrected in a timely manner.
c. The provisions of Administrative Control section 1.6.3 are not applicable.

SURVEILLANCE REQUIREMENTS 2.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated FUNCTIONAL by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 2.1-2.

2-5 REV. 20

TURKEY POINT UNITS 3 & 4 OFFSITE DOSE CALCULATION MANUAL 2.0 RADIOACTIVE LIQUID EFFLUENTS CONTROL 2.1 : RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION, FUNCTIONALITY AND ALARM/TRIP SETPOINTS (continued)

TABLE 2.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT FUNCTIONAL ACTION

1. Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release
a. Liquid Radwaste Effluent Line 1* 2.1.1
b. Steam Generator Blowdown Effluent Line 1 ** 2.1.2
2. Flow Rate Measurement Devices
a. Liquid Radwaste Effluent Line 1* 2.1.3
b. Steam Generator Blowdown Effluent Line 1 *** 2.1.3 per Steam generator
  • Applicable during liquid effluent releases.
    • Applicable during blow down operations.
      • Applicable during blow down operations when Primary to Secondary Leakage is detected.

2-6 REV. 20

TURKEY POINT UNITS 3 & 4 OFFSITE DOSE CALCULATION MANUAL 2.0 RADIOACTIVE LIQUID EFFLUENTS CONTROL 2.1 : RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION, FUNCTIONALITY AND ALARM/TRIP SETPOINTS, (continued)

TABLE 2.1-1, (Continued)

TABLE NOTATION ACTION 2.1.1 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with the surveillance requirement of Control 2.2.1, and
b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve line-up; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2.1.2 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross (beta or gamma) radioactivity at a lower limit of detection of no more than 1 X 10-7 microcuries/ml or analyzed isotopically (Gamma) at a lower limit of detection of at least 5 x 10-7 microcuries/ml :

a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT I-131, or
b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/gram DOSE EQUIVALENT I-131.

ACTION 2.1.3 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow.

2-7 REV. 20

Procedure No.: Procedure

Title:

Page:

15 Approval Date:

0-NCOP-003 Preparation of Liquid Release Permits 2/9/11 7.3 Operation with R-18 Out-Of-Service 7.3.1 Have two qualified Chemistry technicians obtain independent samples.

7.3.2 Perform isotopic analysis using 0-NCOP-310, Calibration and Operation of the Gamma Spectroscopy System on each sample.

7.3.3 IF the Total Specific Activity of both samples is greater than or equal to 1.0E-5 Ci/ml, THEN perform the following:

1. Sum the total isotopic activity in each sample for nuclides with a counting error of less than or equal to 10 percent.
2. Calculate the percent variation and enter into the Chemistry Data Management System.
3. Verify the result meets the statistically calculated control limits for duplicates on liquid releases (see 0-NCAP-216, Radiochemistry Quality Control Samples).
4. IF the percent variation is within limits, THEN continue preparing the release permit.
5. IF the percent variation exceeds the control limits, THEN repeat Subsection 7.4 7.3.4 IF the Total Specific Activity of both samples is less than 1.0E-05 Ci/ml, THEN perform the following:
1. Calculate the percent variation for both samples.
2. IF the percent variation is less than or equal to 20%, THEN continue preparing the release permit.
3. IF the percent variation is greater than 20%, THEN repeat Subsection 7.4 7.3.5 Use the original/first isotopic analysis to complete the permit.

7.3.6 WHEN the permit has been prepared, THEN a second technically qualified person shall review and verify the calculations.

7.3.7 Both the chemistry technician preparing the permit and the person verifying the calculations shall sign Part III, Attachment 1, Permit Prepared By.

W2003:WRS/cls/mr/cls

Classify Events and Fill Out SNF 02201052311 1-0 I 02201052300/

TASK TITLE(S): Classify Significant Events KIA Justification VALUES <3.0): N/A D Non-Uc !ZS] D Simulate/Walkthrough: Perform: [i]

In-Plant D Control Room:

Simulator: c Classroom: x I Lab: c Other: I Time for Completion: 30 Minutes Time Critical: Yes Alternate Path [NRC]: No Alternate Path [INPO]: No Developed by: ,,,, /4-

~~~~~~~~~~~-=~~~~~~~~~~

Date lfff2.. )J Date Date Training Program Owner 003-F10, Revision 2 L-15-1 Adrnin

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 2 of 10 JOB PERFORMANCE MEASURE VALIDATION CHECKLIST ALL STEPS IN THIS CHECKLIST ARE TO BE PERFORMED PRIOR TO USE.

REVIEW STATEMENTS YES NO N/A

1. Are all items on the signature page filled in correctly?
2. Has the JPM been reviewed and validated by SMEs?
3. Can the required conditions for the JPM be appropriately established in the simulator if required?
4. Do the performance steps accurately reflect trainees actions in accordance with plant procedures?
5. Is the standard for each performance item specific as to what controls, indications and ranges are required to evaluate if the trainee properly performed the step?
6. Has the completion time been established based on validation data or incumbent experience?
7. If the task is time critical, is the time critical portion based upon actual task performance requirements?
8. Is the job level appropriate for the task being evaluated if required?
9. Is the K/A appropriate to the task and to the licensee level if required?
10. Is justification provided for tasks with K/A values less than 3.0?
11. Have the performance steps been identified and classified (Critical /

Sequence / Time Critical) appropriately?

12. Have all special tools and equipment needed to perform the task been identified and made available to the trainee?
13. Are all references identified, current, accurate, and available to the trainee?
14. Have all required cues (as anticipated) been identified for the evaluator to assist task completion?
15. Are all critical steps supported by procedural guidance? (e.g., if licensing, EP or other groups were needed to determine correct actions, then the answer should be NO.)
16. If the JPM is to be administered to an LOIT student, has the required knowledge been taught to the individual prior to administering the JPM?

TPE does not have to be completed, but the JPM evaluation may not be valid if they have not been taught the required knowledge.

All questions/statements must be answered YES or N/A or the JPM is not valid for use. If all questions/statements are answered YES or N/A, then the JPM is considered valid and can be performed as written. The individual(s) performing the initial validation shall sign and date the cover sheet.

Protected Content: None TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.4/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 3 of 10 UPDATE LOG: Indicate in the following table any minor changes or major revisions (as defined in TR-AA-230-1003) made to the material after initial approval. Or use separate Update Log form TR-AA-230-1003-F16.

PREPARER DATE

  1. DESCRIPTION OF CHANGE REASON FOR CHANGE AR/TWR#

SUPERVISOR DATE Updated to fleet template; Updated for L-15-1 See cover page N/A 1-0 01982463 text/grammar changes NRC Exam See cover page N/A 1-1 1-2 1-3 1-4 1-5 TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.4/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 4 of 10 SIMULATOR SET-UP: N/A Required Materials:

  • 0-EPIP-20101
  • 0-EPIP-20134
  • F439, Florida Nuclear Plant Emergency Notification Form
  • F444, Guidance For Determining Protective Action Recommendations (PARS)
  • F668, Turkey Point EAL Classification Tables (Hot)
  • F669, Turkey Point EAL Classification Tables (Cold)

General

References:

  • 0-EPIP-20101, Duties of Emergency Coordinator
  • 0-EPIP-20134, Offsite Notifications and Protective Action Recommendations Task Standards:
  • Demonstrate knowledge of emergency action level thresholds and classifications
  • Within 15 minutes, declare a Site Area Emergency (CS1); within 15 minutes of declaration, complete a Florida Nuclear Plant Emergency Notification Form per 0-EPIP-20134 (Offsite Notifications and Protective Action Recommendations) with no errors on required items that are marked with an asterisk (with the exception of Item 2B)

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.4/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 5 of 10 I will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues. When you complete the task successfully, the objective for this job performance measure will be satisfied.

DURING THE JPM, ENSURE PROPER SAFETY PRECAUTIONS, FME, AND/OR RADIOLOGICAL CONCERNS AS APPLICABLE ARE FOLLOWED.

Initial Conditions:

  • Unit 4 is at 100% power.
  • Unit 3 is in Mode 5, with the following conditions:

o RCS temperature is at 170°F.

o The core has been reloaded.

o Containment closure has been established.

Subsequent Conditions:

  • Pressurizer level is found to be off-scale low and the RCS draindown level indicators are unavailable.

3 RAD-3-6311B (CHRRM) has increased to 1.1 x 10 R/hr.

  • After 35 minutes, RCS draindown level indication is restored.
  • Wind speed and direction are 5 mph and 287 degrees, respectively.

Initiating Cue:

  • You are the Emergency Coordinator in the Control Room. Given the initial conditions, classify the event using 0-EPIP-20101 (Duties of Emergency Coordinator) and, if necessary, issue protective action recommendations using 0-EPIP-20134 (Offsite Notifications and Protective Action Recommendations).
  • By raising your hand, you signify that you have completed the event declaration. At that time, the Examiner will provide you with a Florida Nuclear Plant Emergency Notification Form, which you will complete using 0-EPIP-20134 (Offsite Notifications and Protective Action Recommendations).
  • When you have completed the Florida Nuclear Plant Emergency Notification Form, raise your hand to inform the Examiner that you are done.
  • There is an element of this task that is Time Critical.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2 L-15-1 NRC Admin JPM A.4/SRO DRAFT - NRC L-15-1 EXAM SECURE INFORMATION

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 6 of 10 JPM PERFORMANCE INFORMATION Start Time:

NOTE: When providing Evaluator Cues to the examinee, care must be exercised to avoid prompting the examinee. Typically cues are only provided when the examinees actions warrant receiving the information (i.e., the examinee looks or asks for the indication).

NOTE: Critical steps are marked with a Y below the performance step number. Failure to meet the standard for any critical step shall result in failure of this JPM.

Performance Step: 1 Obtain required materials.

Critical: No Standard: Obtain required materials.

Provide examinee with the following:

  • 0-EPIP-20101, Duties of Emergency Coordinator
  • 0-EPIP-20134, Offsite Notifications and Protective Action Evaluator Cue:

Recommendations

  • F668, Turkey Point EAL Classification Tables (Hot)
  • F669, Turkey Point EAL Classification Tables (Cold)

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Review Turkey Point EAL Classification Tables (F668 and F669) and Performance Step: 2 0-EPIP-20101 (Duties of Emergency Coordinator), based on the Critical: No given conditions.

Standard: Review F668, F669, and 0-EPIP-20101, based on the given conditions.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 7 of 10 Performance Step: 3 Determine the highest emergency classification level and EAL Critical: Yes number using F669 (Turkey Point EAL Classification Tables [Cold]).

  • Determine that highest emergency classification level is a Site Area Emergency and EAL number is CS1.

Standard:

  • Examinee raises his/her hand, within 15 minutes, to signify completion of event declaration.
  • Log event declaration time: _____________.

Evaluator Note:

  • Declaration time is the start time for completion of the Florida Nuclear Plant Emergency Notification Form.

Upon receiving the event declaration, provide examinee with the following:

  • F439, Florida Nuclear Plant Emergency Notification Form Evaluator Cue:
  • F444, Guidance For Determining Protective Action Recommendations (PARS)

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

TR-AA-230-1003-F10, Revision 2

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 8 of 10 Complete a Florida Nuclear Plant Emergency Notification Form Performance Step: 4 (F439), in accordance with 0-EPIP-20134 (Offsite Notifications and Critical: Yes Protective Action Recommendations).

  • Within 15 minutes of event declaration, a Florida Nuclear Plant Emergency Notification Form (F439) is completed, in accordance 0-EPIP-20134 (Offsite Notifications and Protective Action Standard: Recommendations), with no errors on required items identified with an asterisk.
  • Item 2B is N/A, until offsite agencies are contacted.

Evaluator Note: Log form completion time: _____________.

Performance: SATISFACTORY _______ UNSATISFACTORY Comments:

Terminating Cues: When the examinee submits the Florida Nuclear Plant Emergency Notification Form (F439), state This completes the JPM.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

Stop Time:

TR-AA-230-1003-F10, Revision 2

02201052311, Classify Events and Fill Out SNF, Rev. 1 JPM DRAFT - NRC L-15-1 EXAM SECURE INFORMATION Page 9 of 10 Examinee: Evaluator:

RO SRO STA Non-Lic SRO CERT Date:

LOIT RO LOIT SRO PERFORMANCE RESULTS: SAT: UNSAT:

Remediation required: YES NO COMMENTS/FEEDBACK: (Comments shall be made for any steps graded unsatisfactory).

EXAMINER NOTE: ENSURE ALL EXAM MATERIAL IS COLLECTED AND PROCEDURES CLEANED, AS APPROPRIATE.

EVALUATORS SIGNATURE:

NOTE: Only this page needs to be retained in examinees record if completed satisfactorily. If unsatisfactory performance is demonstrated, the entire JPM should be retained.

TR-AA-230-1003-F10, Revision 2

JPM Page 10 of 10 TURNOVER SHEET Initial Conditions:

  • Unit 4 is at 100% power.
  • Unit 3 is in Mode 5, with the following conditions:

o RCS temperature is at 170°F.

o The core has been reloaded.

  • Containment closure has been established.

Subsequent Conditions:

  • Pressurizer level is found to be off-scale low and the RCS draindown level indicators are unavailable.

3 RAD-3-6311B (CHRRM) has increased to 1.1 x 10 R/hr.

  • After 35 minutes, RCS draindown level indication is restored.
  • Wind speed and direction are 5 mph and 287 degrees, respectively.

Initiating Cue:

  • You are the Emergency Coordinator in the Control Room. Given the initial conditions, classify the event using 0-EPIP-20101 (Duties of Emergency Coordinator) and, if necessary, issue protective action recommendations using 0-EPIP-20134 (Offsite Notifications and Protective Action Recommendations).
  • By raising your hand, you signify that you have completed the event declaration. At that time, the Examiner will provide you with a Florida Nuclear Plant Emergency Notification Form, which you will complete using 0-EPIP-20134 (Offsite Notifications and Protective Action Recommendations).
  • When you have completed the Florida Nuclear Plant Emergency Notification Form, raise your hand to inform the Examiner that you are done.
  • There is an element of this task that is Time Critical.

NOTE: Ensure the turnover sheet that was given to the examinee is returned to the evaluator.

TR-AA-230-1003-F10, Revision 2

KEY FLORIDA NUCLEAR PLANT EMERGENCY NOTIFICATION FORM Online Verification: STATE MIAMI-DADE COUNTY MONROE COUNTY

  • 1. A. This Is A Drill B. This Is An Actual Event
2. A. Date XX/XX/XX *B. Contact Time: C. Reported by: Name Communinicator's Name D. Message Number: 1 E. Reported From: Control Room TSC EOF F. Initial/New Classification OR Update Notification
  • 3. SITE A. Crystal River UNIT 3 B. St. Lucie UNIT 1 C. St. Lucie UNIT 2 D. Turkey Point UNIT 3 E. Turkey Point UNIT 4
  • 4. EMERGENCY CLASSIFICATION: A. Notification Of Unusual Event B. Alert C. Site Area Emergency D. General Emergency
  • 5. A. EMERGENCY DECLARATION: B. EMERGENCY TERMINATION Date: XX /XX /XX Time: XX:XX
  • 6. REASON FOR EMERGENCY DECLARATION:** A. EAL Number: CS1 OR B. Description
7. ADDITIONAL INFORMATION OR UPDATE: A. None OR B. Description
  • 8. WEATHER DATA: A. Wind direction from 287 degrees. B. Downwind Sectors Affected EFG
  • 9. RELEASE STATUS: A. None (Go to Item 11) B. In Progress C. Has occurred, but stopped (go to Item 11)
10. RELEASE SIGNIFICANCE CATEGORY (at the Site Boundary)

A. Under evaluation B. Release within Normal Operating Limits (Tech Specs)

C. Non-Significant (Fraction of PAG Range) D. PAG Range (Protective Actions required)

E. Liquid release (no actions required)

  • 11. UTILITY RECOMMENDED PROTECTIVE ACTIONS FOR THE PUBLIC:

A. No recommended actions at this time. B. The utility recommends the following protective actions:

EVACUATE ZONES:NOT APPLICABLE OR Miles Evacuate Sectors Shelter Sectors No Action Sectors SHELTER ZONES:NOT APPLICABLE 0-2 2-5 5 - 10 AND consider issuance of potassium iodide (KI)

If form is completed in the Control Room, go to item 15. If completed in the TSC or EOF, continue with item 12.

12. PLANT CONDITIONS:

A. Reactor Shutdown? YES NO B. Core Adequately Cooled? YES NO C. Containment Intact? YES NO D. Core Condition: Stable Degrading

13. WEATHER DATA: A. Wind Speed mph B. Stability Class
14. ADDITIONAL RELEASE INFORMATION: A. Not applicable (Go to Item 15)

Distance Projected Thyroid Dose (CDE) for 1 Hour Projected Total Dose (TEDE) for 1 Hour 1 Mile (Site Boundary) B. mrem C. mrem 2 Miles D. mrem E. mrem 5 Miles F. mrem G. mrem 10 Miles H. mrem I. mrem

15. (Do not read to State) EC or RM Approval Signature(EC - Shift Manager) DateXX/XX/XX TimeXX:XX MESSAGE RECEIVED BY: Name( Date / / Time
    • IF EMERGENCY CLASS ESCALATION IS KNOWN TO BE NECESSARY AND A NEW NOTIFICATION FORM WILL BE TRANSMITTED WITHIN 15 MINUTES, THEN YOU MAY GO TO EC/RM APPROVAL SIGNATURE LINE.
  • ITEMS ARE EVALUATED FOR NRC PERFORMANCE INDICATORS (PIs)

F439 / Page 1 of 2 - Rev 10

FLORIDA NUCLEAR PLANT EMERGENCY NOTIFICATION FORM METEOROLOGICAL WORKSHEET SECTOR

REFERENCE:

The chart below can be used to determine sectors affected by a radiological release, through comparison with wind direction from the meteorological recorders in the Control Room.

If the wind direction is directly on the edge of two sectors (e.g., 11°, 33°, 56°, etc.), an additional sector should be added to the protective action recommendations. For example, if the wind direction is from 78°,

then the affected sectors for PARs should be L, M, N and P.

SECTOR INFORMATION:

WIND SECTOR WIND FROM DEGREES WIND TOWARD SECTORS AFFECTED

[A] N 348-11 S HJK

[B] NNE 11-33 SSW JKL

[C] NE 33-56 SW KLM

[D] ENE 56-78 WSW LMN

[E] E 78-101 W MNP

[F] ESE 101-123 WNW NPQ

[G] SE 123-146 NW PQR

[H] SSE 146-168 NNW QRA

[J] S 168-191 N RAB

[K] SSW 191-213 NNE ABC

[L] SW 213-236 NE BCD

[M] WSW 236-258 ENE CDE

[N] W 258-281 E DEF

[P] WNW 281-303 ESE EFG

[Q] NW 303-326 SE FGH

[R] NNW 326-348 SSE GHJ STABILITY CLASSIFICATION

REFERENCE:

Either ERDADS or the below chart can be used to determine atmospheric stability classification for notification to the State of Florida. Primary method is from T via the South Dade (60 meter) tower. Backup method is from Sigma Theta via the Ten Meter Tower. If neither meteorological tower is available, Stability Classification shall be determined using data from National Weather Service (See 0-EPIP-20126, Off-site Dose Calculations).

CLASSIFICATION OF ATMOSPHERIC STABILITY:

Primary Backup Stability Pasquill Delta T Sigma Theta Classification Categories

(°F) Range (Degrees)

Extremely unstable A T -1.7 ST 22.5 Moderately unstable B -1.7 <T -1.5 22.5 > ST 17.5 Slightly unstable C -1.5 <T -1.4 17.5 > ST 12.5 Neutral D -1.4 <T -0.5 12.5 > ST 7.5 Slightly stable E -0.5 <T +1.4 7.5 > ST 3.8 Moderately stable F +1.4 <T +3.6 3.8 > ST 2.1 Extremely stable G +3.6 <T 2.1 > ST Meteorological information needed to fill out the Florida Nuclear Plant Emergency Notification Form is available from the Dose Calculation Worksheet (0-EPIP-20126). The Worksheet shall be filled out by Chemistry and given to the Emergency Coordinator.

F439 / Page 2 of 2 - Rev 10

GUIDANCE FOR DETERMINING PROTECTIVE ACTION RECOMMENDATIONS (PARS)

KEY BASED ON PLANT CONDITIONS No recommended General NO actions at this time Emergency?

(Note 3)

YES Miles Evacuate Sectors Shelter Sectors No Action Sectors Severe Loss of Core Damage Physical 0-2 NONE ALL NONE NO NO (Actual or Projected) Control of Plant 2-5 NONE (sectors affected) ALL REMAINING (Note 1) (Note 2) 5 - 10 NONE NONE ALL (Note 3)

YES YES Continue to Assess Miles Evacuate Sectors Shelter Sectors No Action Sectors Conditions and Evaluate 0-2 ALL NONE NONE further Protective Action 2-5 (sectors affected) ALL REMAINING NONE Recommendations based on 5 - 10 NONE ALL NONE Off-Site Dose Projections (Note 3)

NOTES:

(1) Severe core damage is indicated by any of the following:

- Loss of critical functions required for core protection (e.g. loss of injection with LOCA)

- High Core temperatures (Valid CET > 700°F)

- CHRRM Reading of greater than or equal to 1.3E4 R/hr (2) Loss of physical control of Control Room or reactor operating areas required for continued safe plant operation to intruders.

(3) See additional Guidance for Determining PARs in Emergency Plan Implementing Procedures.

F444/1:5 - Rev. 4 (0-EPIP-20134)

BASED ON MANUAL DOSE CALCULATIONS RELEASE DURATION LESS THAN 2 HOURS (PUFF RELEASE)

Beyond 10 miles use this column and the 10 mile dose value.

Total Dose Thyroid Dose 0-2 Miles 2-5 Miles 5-10 Miles TEDE OR CDE Dose (mRem) (mRem) Use 1 Mi. value Use 2 Mi. Value Use 5 Mi. Value

< 500 mRem <1000 mRem None None None 500 mRem 1000 mRem S(ALL) but but S(DW) S(DW)

<1000 mRem <5000 mRem 1000 mRem 5000 mRem S(ALL) but but S(ALL) S(ALL)

<5000 mRem < 25000 mRem 5000 mRem 25000 mRem E(ALL) E(DW)+S(AR) E(DW)+S(AR)

RELEASE DURATION GREATER THAN OR EQUAL TO 2 HOURS Beyond 10 miles use this column and the 10 mile dose value.

Total Dose Thyroid Dose 0-2 Miles 2-5 Miles 5-10 Miles TEDE OR CDE Dose (mRem) (mRem) Use 1 Mi. value Use 2 Mi. Value Use 5 Mi. Value

< 500 mRem <1000 mRem None None None 500 mRem 1000 mRem but but S(ALL) S(DW) S(DW)

<1000 mRem <5000 mRem 1000 mRem 5000 mRem but but E(ALL) E(DW)+S(AR) E(DW)+S(AR)

<5000 mRem < 25000 mRem 5000 mRem 25000 mRem E(ALL) E(ALL) E(DW)+S(AR)

SUMMARY

0 - 2 MI. 2 - 5 MI. 5 - 10 MI.

PARs based on - Plant Conditions PARs based on - Total Dose (TEDE)

PARs based on - Thyroid Dose (CDE)

Most Conservative PARs based on Plant Conditions and Dose Projections LEGEND OF ABBREVIATIONS S - Sheltering recommended E - Evacuation recommended DW - Downwind plus 2 adjoining sectors AR - All Remaining sectors ALL - All Sectors F444/2:5 - Rev. 4 (0-EPIP-20134)

CAUTION Previously issued PARs, unless found to be less conservative, are to remain in effect until the source of the threat is clearly under control.

FPL is required to provide county and state governmental authorities with recommendations for protective action to be taken by the public during radiological emergencies at the Turkey Point Nuclear Plant. The responsible authorities are the State Division of Emergency Management (DEM), Miami-Dade County Office of Emergency Management and Monroe County Office of Emergency Management.

Protective Action Recommendations (PARs) should be made utilizing all of the available data. This includes plant status, off-site dose projections, and/or field monitoring data. The more conservative recommendations should be made.

Beginning at the top left side, answer the General Emergency question. If yes, continue on, following the arrows, and answering the other question blocks. Record the PARs based on Plant Condition (A) in the Summary Block at the bottom of the page. From the PAR based on Plant Condition's block continue following arrow to next box, and determine PARs based on Off-site Dose Projections (B) Total Dose (TEDE) and Thyroid Dose (CDE). In determining PARs, both plant conditions AND off-site doses must be considered for all PARs. If a release has not occurred, then proceed with issuance of PARs from the plant condition determination.

To determine PARS from off-site doses, find the blocks that correspond with the Total Dose (TEDE) and Thyroid Dose (CDE) at 1, 2 and 5 miles from the Dose Calculation Worksheet (0-EPIP-20126). Follow across to the column that indicates the distance where that dose was found i.e., first block for 1 mile, second block for 2 miles, or third block for 5 miles. (B) Record the PARs based on Off-site Doses in the Summary Block. Once PARs are determined for all mile sectors for both Total Dose (TEDE) and Thyroid Dose (CDE)

(B), then a comparison with the Plant Condition PARs (A) is performed, and the most conservative PARs for each mile sector is selected for issuance to off-site agencies.

The following example is provided:

EXAMPLE A release has occurred at the Turkey Point Plant. The wind direction is from the SSE and the projected off-site accumulated Thyroid Dose (CDE) is 5,000 mRem at 1 mile, 1,000 mRem at 2 miles, and less than 1,000 mRem at 5 miles. The plant is in a General Emergency with CHRRM at 100 R/hr, no core damage indicators, and no loss of physical control of the plant.

F444/3:5 - Rev. 4 (0-EPIP-20134)

Using the PAR Worksheet, the following recommendations should be made:

Based on our current assessment of all the information now available to use, Florida Power & Light Company recommends that you consider taking the following protective actions.

A. EVACUATE all people between 0 and 2 miles from the plant.

B. SHELTER all people between a 2 and 5 mile radius form the plant who are in Sectors Q, R, and A (refer to Attachment 1).

C. No protective actions is recommended between a 5 and 10 mile radius from the plant.

Due to the large political and legal ramifications of these recommendations and the potential impact on FPL, the following guidelines, format, and content should be used.

(1) If the emergency has not been classified as a GENERAL EMERGENCY and the off-site doses are LESS THAN 500 mRem Total Dose (TEDE) or 1,000 mRem Thyroid Dose (CDE) at 1 mile over the projected duration of the release, no protective action is recommended. When reporting to DEM and other off-site agencies who inquire, this should be reported in a manner similar to the following:

Based on our urgent assessment of all the information now available to us, Florida Power & Light Company recommends that you consider taking the following protective actions - NONE. This recommendation may change in the future, but we cannot now say when it may change or what the change may be.

(2) When available, both plume calculation and off-site monitoring results should be evaluated when making protective action recommendations. If significant discrepancies exist between field monitoring results and plume dispersion calculations, then the discrepancy should be reviewed, and the appropriate value should be selected in the determination of protective action recommendations.

(3) Thyroid Dose (CDE) Limits for PARs are based on adult thyroid. These limits are consistent with EPA Guidelines based on the following criteria:

a. Uncertainty and potential errors associated with age specific parameters, and
b. Level of conservatism in the adult values.

(4) Loss of physical control of the plant to intruders shall be determined by the Emergency Coordinator based on the current operating mode requirements of the unit / plant, and the availability of equipment required for continued safe operation.

F444/4:5 - Rev. 4 (0-EPIP-20134)

GUIDANCE FOR THE USE OF POTASSIUM IODIDE (KI) - A THYROID BLOCKING AGENT

1. The EOF RP Manager in consultation with the TSC RP Supervisor will determine the need to dispense Potassium Iodide (KI) based upon a projected or actual thyroid Committed Dose Equivalent (CDE) of greater than or equal to 5 rem. (The thyroid CDE of greater than or equal to 5 rem is based on the FDA recommended threshold for ingestion of KI by pregnant and lactating women).
2. The TSC RP Supervisor and the OSC RP Supervisor will coordinate KI distribution once a decision for use has been determined.
3. The TSC RP Supervisor is responsible for KI distribution to personnel in the Unit 3 and 4 Control Room and the TSC and Field Monitoring Teams and to Security personnel not assigned to the OSC.
4. The OSC RP Supervisor is responsible for distribution in the OSC.
5. KI should be administered and ingested within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the determination is made that thyroid CDE is greater than or equal to 5 rem.
6. When KI is issued, thyroid intakes will be estimated by whole body counts.
7. Administering KI after an uptake may limit thyroid CDE depending on time after exposure.
8. Caution emergency response personnel of potential KI side effects if they are allergic to shellfish or iodide. Emergency response personnel who know they have such allergies should be replaced in lieu of directing them to ingest KI.
9. All KI tablets are stored in the RP kits in the Unit 3 and 4 Control Room, TSC, OSC, and Field Monitoring Team Kits.

F444/5:5 - Rev. 4 (0-EPIP-20134)