ML20199F145
| ML20199F145 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/31/1997 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F1-98-0017, W3F1-98-17, NUDOCS 9802030070 | |
| Download: ML20199F145 (268) | |
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W3F198 0017 A4.05 PR January 29,1998 U.S. Nuclear Regulatory Commission Attn:- Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF 38 Report of Facihty Changes, Tests, and Experiments Gentlemen:
Enclosed is the Report of Facility Changes, Tests, and Experiments for Waterford 3 which is submitted pursuant to 10CFR50.59. This report covers the period from December 1,1995, through May 31,1997.
If you have any questions regarding this report, please contact me at (504) 739-6242 or T.J. Gaudet at (504) 739-6666.
Very truly yours, 7
w E.C. Ewing ,
Director Nuclear Safety & Regulatory Affairs -
g q) ; g ECE/ELUssf
Enclosures:
50.59 Summary Report 50.59 Report
- cc: E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRR),
J. Smith, N.S. Reynolds, NRC Resident inspectors Office 9802030070 970531 PDR ADOCK 0500o382:
R PDR
. .______ A
SjJMMARY This report provides the Waterford 3 Facility Changes made pursuant to 10CFR50.59(a)(1). The report covers the period from December 1,1995, through May 31,1997. None of the items in the report were found to involve an unreviewed safety question.
Section I of the report identifies 179 Facility Changes which consist of: 44 Design Changes (DCs),37 Condition identification / Work Authorizations (Cl/WAs),15 Temporary Alteration Requests (TARS),4 Document Revision Notices (DRNs),29 License Document Change Requests (LDCRs),31 Miscellaneous Evaluations, and 19 Commitment Changes.
Section ll of the report identifies 50 Procedure Changes which consist of: 25 Plant Procedures and 25 Special Test Procedures (STPs).
QUALITY REVIEW OF 50.59
SUMMARY
REPORT DESIGN CHANGES - Reviewed the Design Change Information table on the Planning
& Scheduling / Site Scheduling home page. Those DCs identified as either ' Closure' or
' Approved / Testing' were included.
Cl/WAs - Reviewed Cl/WA information in SIMS. Those Cl/WAs shown as either
' Complete' or ' Closed' were included.
TARS - Reviewed the Control Room TAR Leg.
DRNs - spoke with Ron Cummins who verified DRNs had been issued / incorporated.
LDCRs - verified LDCRs against the FSAR Coordinator's log to ensure agreement between FSAR and 50.59 submittals. Any items shown as ' Forwarded'(to FSAR revision 10) or ' Canceled' in the log were not included in this submittal.
MISCELLANEOUS - Setpoint changes were confirmed with Michelle Grooms as having been installed. TRM changes were reviewed against TRM amendments. Calculations were reviewed against calculations in the library. SPEERs were reviewed against hard copies in Programs Engr. Some items were accepted on the basis of the 50.59 itself -
Cycle 9 Core Reloads, W4.101s, CRs, PElR, ERs.
CCEFs - reviewed against CCEF log.
PLANT PROCEDURES - verified against the available procedure information in IDEAS to ensure procedures had been issued.
STPs - verified against available information in SIMS. Only those STPs shown as having been implemented were included.
ENTERGY OPERATIONS, INC.
WATERFORD 3 SES DOCKET NO 50-382 LICENSE NO. NPF-38 REPORT OF FACILITY CHANGES, TESTS, AND EXPERIMENTS PER 10CFR50.59 DECEMBER 1,1995 THROUGH MAY 31,1997
~
WATERFORD 3 4 10CFR50.59 REPORT ENTERGY OPERATIONS, INC.
DECEMBER 1,1995 THROUGH MAY 31,1997 TABLE OF CONTENTS Item Page No, _ No.
- 1. FACILITY CHANGES A. DESIGN CHANGES 1, DC-3311, Corrosion Rcte Monitoring of Component Cooling Water 1 (CCW) and Aux;llary Component Cooling Water (ACCW) Heat Exchangers A&B (Revision 0, Revision 1, and Revision 2)
- 2. DC-3322, Permanent Facility for Makeup Water Treatment Service 2 Vendor (Revision 0 and Revision 4)
- 3. DC-3388, Corrosion Rate Monitoring of Various Cooling Water 3 Systems
- 4. DC-3390, Instrument Air / Station Air Enhancements 4
- 5. DC-3427, Condenser Air Removal Vacuum Flow Measurement 5
- 6. DC-3429, instrument Air Supply to Fail Open Containment isolation 6 Valves (Revision 0 and Revision 3)
- 7. DC-3433, Enhancement to EFW Turbine Oil System (Revision 2) 7
- 8. DC-3434, EH System Constant Pressure Pumps (Revision 2) 8
- 9. DC-3435, Replacement of Fuel and CEA Hoists on Dual-Masted 9 Refueling Machine
- 10. DC-3440, HPSI Flow Control Valve Replacement 10
- 11. DC-3448, Annunciator Inter!ock for Monitoring a Loss of Bus 11 Transfer Capability (Revision 0 and Revision 1) i
- 12. DC 3462, Control Room Noise Reductions 12
- 13. DC-3468, Essential Ch!lled Water Chemistry improvement / 13 NNS Train Relocation to the Supplemental Chilled Water Sys'em 14 DC 3470, ACCW System Waterhammer 14 15 DC-3472. Hot and Cold Leg RTD Noise Abatement 15
- 16. DC-3476: Additional Source of Containment Building Temporary 16 Power for Outsges
- 18. DC-3478, Safety injection Nitrogen Valve Replacement 18
- 19. DC-3480, Safety injection Tank Lt /el and Pressure Alarm 19 Relocation
- 21. DC-3488, Reactor Coolant Pump Controlled Bleed-Off Flow 21 Transmitter Replacement
- 22. DC-3489, Addition of Maintenance Capability in CVC Letdown 22 to the VCT
- 24. DC-3492, Addition of Filters on the Component Cooling 24 Water System
- 25. DC-3493, Stroke Time Speed Reduction of CCW Header 25 Isolation Valves ,
- 26. DC-3494, Steam Generator Blowdown Flow Restoration 26
- 27. DC-3495, Boric Acid and Primary Makeup Flow improvements 27
- 28. DC-3498, HPSI Flow Control Valve Position Indicators (Revision 0 28 1 and Revision 1) il
- 29. DC-3502, Reduce Bypass Leakage from Penetrations 53 and 65 29
- 30. DC-350S, Emergency Diesel Generator Miscellaneous Vent, Drain, 30 and Sample Lines
- 31. DC 3506, Auxiliary Steam Test 31
- 32. DC-3508, MSR Shell Drain Tank Normal Level Control Valve 32 Fall Open Modification
- 33. DC-3513, CCW/ACCW Pump Improvements 33
- 34. DC-3518, Condenser Air Evacuation System improvements 34
- 36. DC-3522, Removal of Steam Generator Feed Pump Suction 36 Relief Valves
- Isolation Valves
- 38. DC-3530, Emergency Feedwater Pump Turbine Steam 38 Supply Drains
- 39. DC-3533, Post-Accident Sampling System Sampling Capacity 39 Improvements
- 40. DC-8006, Maii1 Turbine Turning Gear Control Switch Modification 40
- 41. DC-8016, Dry Cooling Tower Sump Pump Discharge Check Valve 41-42, DC-8025,- Seal Oil Vapor Extractor Drain Line Improvements 42
- 43. DC-8026, High Pressure Turbine Gland Steam Spillover Capacity 43 Improvement
- 44. - DC-8028, E22 Fan Shut Down Interlock with the Toxic Chemical 44 Monitors lil
B. - CONDITION IDENTIFICATION / WORK AUTHORIZATION (Cl/WA)
- 1. Cl 300334/WA-01142862, Warte Tanks As built Strainer 45 Size Discrepancy
- 2. Cl 301153/WA-01144191, NI Log Power Channel Calibration 46 Safety Channel A, B, C, or D [ Evaluation is also for Ml-003-102 (Rev 6) and OP 903-102 - see Item II.A.6)
- 3. Cl-301410/WA-01144762, Unit 3 Metal Waste Pond Transfer 47 Pump Recirculation Line Orifice Deletion j
- 4. Cl-301785/WA+1145481, Rearrange Instrumentation for 48 Component Cooling Water Flow from Essential Chillers
- 5. Cl-302111/WA-01146073,1/2" Heater Core Tubing for EDG 'D' 49
- 6. Cl-302337/WA-01146881, Replacement of Door D 085 50
- 7. Cl-302375/WA-01146537, CC-963B Needle Valve Replacement 51
- 8. Cl-302702/WA 01147033, TCCW Pumps 'A' and 'B' Air Eliminator 52 Additions
- 9. Cl 302960, Elevator Machine Room, RAB -35 53
- 11. Cl-303093/WA-01147774, CVCS Charging Pumps Pulsation 55 Dampeners: Schrader Valve Replacement
- 12. Cl-303246/WA-01148080, Safety injection Vent Line Addition 56
- 13. Cl-303338, Handra:Is in RAB and Cooling Tower Area 57
- 14. Cl-304444/WA-01150338, Ductwork Access Panels for Valves 58 HVC-101 and HVC-102 15i Cl-304502/WA-01153320, Fire Water Storage Tanks A & B 59
- 16. Cl-304506/WA-01150447, Installation of Vent on Charging Pump 60 Discharge Pipe iv
f
- 17. Cl 305472, Repair of Fire Detector 278-19 61
- 18. Cl 305473/WA 99100019, Terminate New Safety Related 62 Cables for the Ultimate Heat Sink Tornado Missile Protection Project (Revision 0 and Revision 1)
- 19. Cl-305849/WA-01152287, Evaluation of Unprotected Opening 63 Between Fire Areas RAB 2 and Roof E
- 20. Cl 306068/WA-99100019, Terminate New Safety Related Cables 64 for the Ultimate Heat Sink Tornado Missile Protection Project (Revision 0 and Revision 1)
- 21. Cl-306751/WA-01153561, LPSI Vent Additions at Containment 65 Penetrations
- 22. Cl 306764/WA-01153606, Addition of Pressure Equalization Lines 66 to Sl 125A(B) and SI-412A(B)
- 23. WA 01142925, Main Steam Trap Drain Line Replacement 67
- 24. WA-01144346, Removal of Vibration Trip Switch from Emergency 68 Diessl Generator 'A'
- 25. WA-01144462, Boric Acid Makeup Thermal Relief Valve Removal 69
- 26. WA-01146547, CC 963A Needle Valve Replacement 70
- 27. WA-01148519, Reactor Drain Tank Outlet Check Valve BM-107, 71 Internal Components Removal
- 28. WA 01148765 and WA-01148764 EDG Lube Oil Filter Internal 72 Relief Valves Replacement with a Valve Port Closure Assembly
- 29. Cl-303903/WA-01149350, Modification of RCP Seal Heat Exchanger 73 Baffle Bolting (Revision 0 and Revision 1)
- 30. WA-01155142, Plug Tubes in Spent Fuel Pool Heat Exchanger 74
- 31. WA-01155273, Reactor Closure Head Exhaust Manifold Bolt Hole 75 Modification
- 32. WA-01156373, Installation of Vent / Test Connection on Gland 76 Steam Condenser Exhaust v
- 33. WA-01157203, Rowiro Ground Detection Relay in Emergency 77 Diesel Generator 'A' 34.- WA-01157227, Rowire Ground Detection Relay in Emergency 78 Diesel Generator 'B'
- 35. W . i158027, HPSI Vent Additions at Penetrations 79 55, v6,57, & 58
- 36. WA-01158036, Shield Building Maintenance Hatch Seal 80 System Regulator
' 37, WA 01158198, EH Fuller Earth and Contaminant Filters - 81 Siphon Breaker Addition vi
C. TEMPORARY ALTERATION REQUEST (TAR)
- 1. TAR 96-003, Mux Site RA460, ' Annunciation Temporary 82 Disconnection
- 3. TAR-96-009, Compliance with TS 3.7.6.5 due to Installation 84 of Blank Off Plate on Control Room Normal Outside Air Intake Duct (Rovision 1 and Revision 1)
- 4. TAR 96-012, Gag Closed Valves CC-807A and CC-823A 85
- 5. TAR-96-014, RWSP Low Level Alarm 86
- 6. TAR-96-015, Determing AH-25 Inlet Damper Operators, 87 SVS-103A(B)
- 7. TAR-96-016, Steam Generator Feed Pump Lube Oil 88 Temporary Purifier
- 11. TAR-97-007, Containment Sump Pump 'A' Bypass 92
- 13. TAR "7-009, Blank Plate Installation at Orifice SI-lFI-0321 in 94 Support of DC-3440 (Revision 0 and Revision 1)
- 14. TAR-97-010, Blank Plate Installation at Orifice SI-IFl-0331 in 95 Support of DC-3440 (Revision 0 and Revision 1)
- 15. TAR-97-011, Blank Plate Installation at Orifice SI-IFl-0341 in 96 Support of DC-3440 vii
D. DOCUMENT REVISION NOTICES (DRN)
- 2. DRN E 9600940, Paging System Wiring 98
- 3. DRN's M 9501315, M 9501325, M 9501336, M 9501337, M 9501338, 99 and M 9501339, Revise Instrument Air Flow Diagrams
- 4. DRN-M 9701151, Lower Emergency Feedwater Temperature from 70 Dog. F to 40 Dog. F 100 viii
E. LICENSE DOCUMENT CHANGE REQUESTS (LDCR)
- 1. LDCR 95-0078, Revises FSAR Section 8.3.1 101
- 2. LDCR 95-0117, FSAR Changes to Fuel Pool Ccoling System 102
- 3. LDCR 96-0151 Exception for CIAS Channel Assignment to 103 HRA Valves
- 4. LDCR-96-0161, FSAR Changes on Ultimate Heat Sink 104
- 5. LDCR-96-0170, Revises FSAR Table 6.2-32, Sheets 75 & 76 105
- 6. LDCR-96-0171, Revises FSAR Section 13.7 106
- 7. LDCR-97-0001, Revises FSAR Figures 9.5.1-16 end 9.5.1-20 107 .
- 8. LDCR-97-0005, Revises FSAR Figure 1.219 108
- 9. LDCR 97-0015, Revises FSAR Section 6.4.4.2 109
- 10. LDCR 97-0023, Maintenance Requirements: Rotating 110 Equipment in Storav
- 12. LDCR-97-0042, Breathing Air Requirements for Control 112 Room and Fire Brigado Personnel
- 14. LDCR-97 0047, Location of Cabinets C3A(B) and C-4 Outside 114 the CVAS Boundary (Revision 0 and Revision 1)
- 15. LDCR 97-0061, Ultimate Heat Sink Design Basis 115
- Meteorological Conditions
- 16. LDCR 97-0069, Resetting the High Log Power Trip Setpoint 116 to 0.257% and Changes to FSAR Section 15.4.1.1
- 17. LDCR-97-0104, Revises FSAR Section 9.51.2 and 117 Figure 9.5-6 ix
- 18. LDCR 97-0110, Technical Requirements Manual Table 3.6-2 118 and FSAR Table 6.2 32
- 19. LDCR 97-0123, Revises FSAR Table 6.2-32 119
- 20. LDCR-97-0123, Revises FSAR Sections 12.3.2.3 and 12.3A.1 120
- 21. LDCR 97-0131, FSAR 10.3.2 and 10.4.0, and EC-M96-028, 121 EFW Heat Trace Temperature Requirements .
- 23. LDCR-97-0138, Revises FSAR Section 9.5.1 123
- 24. LDCR-97-0140, Revises FSAR Sections 13.1 and 13.5, 124 Table 13.1-2, and Figure 13.1-6
- 25. LDCR 97-0141, Revises FSAR Section 6.2.1, Containment 125 Functional Design
- 27. LDCR-97-0170, Revises FSAR Section 9.5.1 127
- 28. LDCR-97-0176, Revises FSAR Sections 8.1 and 8.2 128
- 29. LDCR-97-0201, Revises FSAR Section 9.2.5.3.2, Tables 9.2-1, 129 9.23,9.2-9,9.210, and Figures 9.2-4, 9.2-4a, and 9.2-Sa
,a t
- - - = = - -
X
l F. MISCELLAN20US EVALUATIONS
- 1. SPC-95-010-0, Wet Cooling Tower Basin Temperature Setpoint .130
- 2. SPC-94-005-0, Pressurizer Pressure Hi/Lo Alarm Annunciator 131 H0531 RC (PAC 0100-X,Y
- 3. Condition Report 96-0543, CCW/ACCW Throttle Valve Positions 132
- 4. CR-96-0543, Essential Chille 'B' Operability without Restrictions 133
- 5. SPC-96-003-0, Condensate Storage Pool Level 134
- 6. SPC-96-004-0, Loss of instrument Air Accumulator CVR-101, 135 CVR-201 Alarm Setpoints
- 7. W4.101, EFW Heat Trecing Evaluation (Revision 0 and Revision 1) 136
- 8. SPC-93-017-0, CCW and Shutdown Cooling Room Temperature 137 Setpoints
- 9. TRM-004, Table 3.3-11 and Section 3.7.10.4 138
- 10. Closure of isolation Valves CVR-401 A(B) (Revision 0 and Revision 1) 139 1 'i . Control of Safety and Reliof Valve Blowdowrs Ring Settings, 140 Root Cause Analysis 96-0463
- 12. Post-Accident Sampling System (PASS) Heat Exchanger Operation 141 and Chilled Water (CHW) Chemical Addition (PElR OM-112)
- 13. SPC-96-011-0, CC Accumulators for CC-807A(B), CC-808A(B), 142 CC-822A(B), and CC-823A(B)
- 14. SPC-94-017-0, RAB Ambient Negative Pressure, 143 HVRIPAC5275 A and B
- 15. W3 Pump and Valve Inservice Test Plan, Revision 8, Change 2 144
- 16. Calculation EC-M95-012, Minimum Submergence to Prevent 145 Vortexing Calculation (Revision 1)
- 17. TRM Section 4.7.11.2 146 18.- SPC-97-001-0, Emergency Breathing Air Low Pressure Alarm 147 xi
- 19. Drawing G-587, Common Foundation Structure Masonry, Sheet 1 148 ,
- 20. Instrument Calibration Calculations and Instrument 149 Calibration Data for Refueling Water Storage Pool Level Loops, SI ILO305 A, B, C, and D
- 21. - Burnup Extension Program 150
- 22. Calculation EC-M92-015 151-
- 23. Cycle 9 Core Reload (Mode 6 Only) 152
- 24. Calculation MN(Q)-6-45 (Revision 1, Change 1) . 153
- 25. Cycle 9 Core Reload 154
- 26. Engineering Request ER-W3-97-0064-00, Evaluction of 171 Plant Fire Doors j
- 28. SPEER 9501423, Replacement of EGA-13GA(B) and EGA-137A(B) 173
- 29. SPEER 9501467, Replacement of EDG Cool Down Trip 174 Circuit Check Valves
- 30. SPEER 9701667, NUKON Blanket Insulation - Reactor 175 Vessel Closure Head
- 31. SPEER 9701684, Wet Cooling Tower Nozzle Replacement 176 xii 3
G. COMMITMENT CHANGES
- 1. - Delete Safety Evaluations included in PORC Meeting Minutes 177
- 3. Testing the Post-Accident Sampling System 179
- 4. Evaluation of Emergency Classification Procedure 180
- 5. Incorporate Firewatch into the Managemer.t Observation Program 181
- 6. Makeup and Function of the Condition Review Board 182
- 7. Retrieve Missing Steam Generator Plugs 183
- 8. Completion of EQ Data Record Form 184
- 9. Use of " Auto Sequential" Control Rod insertion 185
- 10. Training and Qualification Requirements for Health Physics 186 Contract Personnel
- 11. Departmental Self Assessments to include Design Engineering 187
- 12. Shelf Life of Okonite Electrical Tape and Nuclear Splice Cement 188
- 13. LER 91-006 Commitments 189
- 14. Review of Permanent Control Room Modifications 190
- 15. Component Cooling Water Mekeup Commitments 191 i
- 16. Flant Operations Review Committee Review of Licensee 192 Event Reports
- 17. Breakdown in the Plant Overtime Policy 193 l
- 18. Use of Portable Radios and Telephones 194 l
- 19. Use of Portable Radios and Telephones 195 xiii l
l l I
ll. PROCEDURES A. PLANT PROCEDURES
- 1. CE-002-001, Maintaining Steam Generator Chemistry (Revision 12) 196
- 2. CE-002-002, Maintaining Condensate and Feedwater 198 Chemistry (Revision 8)
- 3. CE-002-013, Maintaining Essential Services Chill Water 199 Chemistry (Revision 10)
- 4. CE-002-036, Chemical Control of Zebra Mussels in 200 Circulating Water System (Revision 0)
- 5. HP-001-220, Bioassay Program (Revision 8) 201
- 6. Ml-003-102 (Revision 6) and OP-903-102 (Evaluation is also for 202 Cl-301153/WA-01144191, NI Log Power Channel Calibration Safety Channel A, B, C, or D - see item I.B.2 )
- 7. Mi-005-415, Main Turbine Eccentricity / Key-Phasor/Zero 203 Speed Tachometer / Acceleration Monitoring System (Revision 4, Change 2)
- 8. Ml-005-463, Plant Protection System Bistable Calibration 204 (Revision 1, Change B), OP-004-004, Centrol Element Drive (Revision 7, Change A), OP-009-007, Plant Protection System (Revision 4, Change A), OP-010-001, General Plant Operations (Revision 17, Change D), OP-903-107, Plant Protection System Channel A,B,C,D Function Test (Revision 12, Change B)
- 9. MVI-008-002, Containment Penetration Modification for 205 Refueling (Revision 1)
- 10. MM-TEM-043, Installation and Removal of the Temporary <
Reactor Coolant Pump Seal (Revision 0)
- 11. NOECP-001, Development, Revision, and Deletion of Procedures, 207 Standard and Guides (Revision 4)
- 12. NOECP-402, NPIS Common Foundation Basemat Integrity 208 Check (Revision 1) .
xiv
._ _ . . _ __ ._. . _ ________ _ _ _ . . . . . .._m
- 13. NOECP-405, installation of Pre-Engineered Access Platforms 209 and Ladders, (Revision 0)
- 14. OP-002-006, Fuel Pool Cooling and Purification 210
- 15. OP-003-014, Control Room Heating and Ventilating 211 (Revision 5, Change A)
- 16. OP-010-001, General Plant Operations (Revision 18) 212
- 17. OP-100-009, Control of Valves and Breakers (Revision 13, 213 Change 1)-
- 18. OP-901-520, Toxic Chemical Release (Revision 2, Change A) 214
- 19. PE-005-040, Diagnostic Differential Pressure Testing of Motor 215 Operated Valves (Revision 1)
- 20. RF-003-003, Steam Generator Sludge Removal (Revision 4) 216
- 21. RF-TEM-001, Dropped Fuel Assembly Stabilization 217
- 22. _RF-TEM-002, Recovery of Fuel Assembly LAR338 218
- 23. STA-001-004, Local Leak Rate Test (LLRT) (Revision 1, Change 3) 219
- 24. UNT-005-014, Offsite Dose Calculation Manual (Revision 5, 220 Change 1 and Change 3)
- 25. UNT-006-019, Control of Local Leak Rate Testing (Revision 1, 222 Change 3) xv
B. SPECIAL TEST PROCEDURES (STP)
- 1. STP-01135315, On-line Leak Test of SI-120A(B) and SI-121 A(B) 223 2.- STP-01145052, Nitrogen Injection to Condenser 224
- 3. STP-01147312, CCW Heat Exchanger Performance 225 Test (Modes 1-4)
- 4. STP-01149859, Control Room Carbon Dioxide Collection 226
- 5. STP-d1150154, CCW System Flow Balance Test (Revision 0 227 and Revision 1)
- 6. STP-01150615, CS-118A Flow Test 228
- 7. STP-01151879, CC-181 A(B) and CC-135A(B) Leak Test 229
- 8. STP-01156079, Shutdown Cooling Heat Exchanger "A" 230 Performance Test
- 9. STP-01156080, Shutdown Cooling Heat Exchanger "B" 231 Performance Test
- 10. STP-01156545, Vacuum Fill of LPSI Penetration No. 39 232
- 11. STP-01156593, Vacuum Fill of LPSI Penetrations No. 36 and 37 233
- 12. STP-01157063, Pressure Test of Class 1 RCS Vent 234 and Drain Lines
- 13. STP-01157743, Change in RWSP Level with One CVAS Makeup 235 Damper Failed Closed
- 14. STP-01158622, CVAS Boundary Test 236
- 15. STP-01158897, Shutdown Cooling Flow Control Valves 237
- 16. STP-301702, Throttling Condenser Outlet isolation Valves 238
- 17. STP-307622, DC-3435 Acceptance Test 239
- 18. STP-99100019A, Ultimate Heat Sink Cable Re-route 240 Train 'A' Retest xvi
.19 - STP-99003468A, Essential Chilled Water Flow Balance, -241
. Minimum Room Temperature Test, and DC-3468 Acceptance Test for Train A
- 20. STP-99003468B, Essential Chilled Water Flow Balance, 242
- Minimum Room Temperature Test, and DC-3468 Acceptance Test for Train B -
21, STP-99003470, Acceptance Test for DC-3470 (Revision 0 243 and Revision 1) -
- 22. STP-99003478, SIT Nitrogen Supply PCV Leak Test. 244
- 23. STP-99003492, DC-3492 Acceptance Test 245
- 24. STP-99003523, Acceptance Test for DC-3523-246 25.- STP-99100019-B, Ultimate Heat Sink Cable Reroute 247 Train B Retest xvii
- l. FACILITY CHANGES A. DESIGN CHANGES
- 1. DC-3311. Corrosion Rate Monitorina of Component Coolina Water (CCW) and Auxiliary Component Coolina Water (ACCW) Heat Exchanoers A&B (Revision O.
Revision 1. and Revision 2)
DESCRIPTION OF CHANGE Revision 0 installed coupon racks on both trains of the ACCW system in the CCW Heat Exchanger rooms. Initial testing revealed there was insufficient differential pressure across the CCW Heat Exchanger tubes to provide adequate flow through the new coupon racks. Revision 1 increased the supply and return tubing on the ACCW system coupon racks and added a third rack in the non-essential portion of the CCW system.
Revision 2 removes the coupon racks on both A and B trains of the ACCW system because they do not have adequate flow through the coupons.
REASON FOR CHANGE There were originally no provisions for monitoring corrosion or erosion in either the CCW or the ACCW systems which are constructed primarily of carbon steel materials.
Subsequent problems with the modification have been identified.
SAFETY EVALUATION According to the safety evaluation, the proposed changes to DC-3311 will not reduce the margin of safety as defined in the basis for any TS or safety analysis and no unreviewed safety questions are created. After implementation, a single coupon rack will have been installed on the non-safety portions of the CCW system and two 1" globe valves and two 1" check valves will remain on the ACCW system where the coupon racks were removed. These connections were previously vent and drain connections and they can still be used as originally intended. The new valves will remain because they are welded connections and they have potential for future use:
1 L . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - _ _
- 2. DC-3322. Permanent Facility for Makeuo Water Treatment Service Vendor (Revision 0 and Revision 4)
DESCRIPTION OF CHANGE
~
The DC removes all obsolete equipment associated with the domineralizer system and -
provides services for the permanent use of contractor domineralizer equipment. The DC also installs instrumentation to monitor the water purity at the discharge from the vendor.domineralizer equipment.
- i REASON FOR CHANGE Eliminates maintenance of unused equipment, simplifies the flow path and reduces the system pressure drop, and provides more flexibility for arrangement of the vendor furnished domineralizer equipment, and provides more space for locating domineralizer equipment inside the Water Treatment Building (WTB).
SAFETY EVALUATION According to the safety evaluation there are no FSAR postulated accidents affected by
- the DC. There are no new system interactions introduced and affected systems are not ;
-safety related. The systems affected serve no safety function and they are not required for operation during the safe shutdown of the plant following an accident or to mitigate c the consequences of an accident.-
2 l
- 3. DC-3388. Corrosion Rate Monitorina of Various Coolina Water Systems DESCRIPTION OF CHANGE The DC will install a corrosion test loop in each of the following systems: ,
One (1) in each Emergency Diesel Generator (EDG) Jacket Water System, ;
One (1) in Essential Chilled Water piping Trains A and B, One (1) in the supplemental Chilled Water piping, and, One (1) for each of the supplemental chiller's Condensing Water Systems.
The test loop will consist of 3/4 inch supply and return tubing, a coupon rack fabricated from 3/4 inch tubing, and a corrator probe. Each test loop will have isolation valves for the coupon rack and a chemistry sample point / drain valve.
REASON FOR CHANGE The installation of corrosion test loops to the listed systems will provide chemistry with a quantitative method to assess the effectiveness of the chemical treatment program used in preventing corrosion related piping and component failures.
SAFETY EVALUATION The safety evaluation determined that the DC will not reduce the margin of safety as defined in the SAR or Technical Specification bases, nor result in any unreviewed safety questions. The evaluation states that the pressure boundary of the affected systems is maintained because the design of the test loop is in accordance with the ASME code requirements for the systems. The possibility of a failure in a pressure boundary is not increased by this DC, nor is any new method of failure introduced. The DC does not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the SAR. .
3
' 4c ' DC-3390; instrument Air / Station Air Enhancements (Revision 0)
DESCRIPTION OF CHAT 1GE The DC will install isolation valves; connections for a temporary _ outside air source, and the interconnection from the Station Air (SA) to instrument Air (IA) will be modified to tie ,
in downstream of the lA receiver.
REASON FOR CHANGE
-The existing compressed air systems are not designed for system outages in order to perform maintenance tasks on main components.1 Also, there are no provisions for-temporary a_ir enmpressors to be connected to the compressed air systems in case a
- overall system outage is required. The lack of flexibility resulted in the compressed air systems being named as a high reliability risk.
The DC will result in increased reliability and maintainability of the compressed air systems. q
- SAFETY EVALUATION iAccording to the safety _ evaluation a complete loss of IA could cause a " Loss of Normal
- Feedwater Flow" but this modification will not increase the probability of occurrence of this accident. The evaluation notes that the lA and SA systems are non-safety and non-seismic and according to the safety evaluation in the FSAR, "The complete loss of IA or SA during full power operation or under accident conditions does not reduce the ability of the reactor protective system or the engineered safety features and their.
supporting systems to safely shutdown the reactor or to mitigate the consequences of an accident." ;
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- 5. DC-3427. Condenser Air Removal Vacuum Flow Measurement DESCRIPTION The proposed changa adds flow measurement capability to the air evacuation lines which run _from the condenser shells to the air evacuation equipment. These flow loops will indicate the amount of non-condensable gases which are flowing from each condenser shell, which is a measure of how much air is being liberated in each condenser shell, by inteakage or outgassing of introduced liquid. The output data will be routed to a personal computer for collection of the data, and from there to the PMC where it will be archived and displayed.
REASON FOR CHANGE The changes from this modification will provide diagnostic information to aid in reducing the condenser pressure during normal operation, and reducing the dissolved oxygen in the condensate /feedwater. These improvements are especially needed during winter months when the dissolved oxygen value is elevated.
SAFETY EVALUATION The modification of the condenser air evacuation system will make it possible to obtain diagnostic data which may be used to improve steam cycle efficiency and improve secondary water chemistry by reducing dissolved oxygen, thus affording additional protection to the steam generator tubes, which are a pressure-boundary element for preventing the spread of radioactivity. This modification does not result in an unreviewed safety question.
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- 6. DC-3429. Instrument Air Supply to Fail Open Containment isolation Valves (Revision 0 and Revision 3)
DESCRIPTION The proposed change provides a means to remotely recharge the existing air accumulators for valves CC-641, CC-710, CC-713, CC-807A(B), CC-808A(B), CC-822A(B), CC-823A(B), and CS-125A(B).
REASON FOR CHANGE Containment isolation valves CC-641, CC-710, CC-713, CC-807A(B), CC-808A(B), CC-822A(B), CC-823A(B), and CS-125A(B) are air operated, designed to fail open, and they are required to be capable of closing and staying closed in accordance with GDC 56 and 57. Although each of these valves is connected to safety related air accumulatorr., the accumulators are not adequately sized to provide long term closure of the valves if necessary, SAFETY EVALUATION Neither the failure of any of the affected RCB isolation valves or failure of the Instrument Air (IA) system are postulated to initiate any accidents identified in the FSAR. To preclude an incraase in the consequences of any accident identified in the FSAR, the new piping, tubing, and associated fittings and components are designed in accordance with appropriate ASME, regulatory, and industry standards. Appropriate administrative controls will also be implemented to ensure equipment operates as required anri containment penetration integrity is maintained. Potential events with the proposed change (e.g., cylinder or tubing rupture) have also been evaluated to ensure no new accident or equipment malfunction not previously evaluated in the FSAR is created. TS bases for containment systems, containment isolation valves, and containment building penetrations were evaluated to ensure there is no reduction in any margin of safety.
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- 7. DC-3433. Enhancement to EFW Turbine Oil System (Revision 2)
DESCRIPTION The proposed change will rotate the tee and_a valve and associated fittings on the drain plug of the EFW Turbine Lube Oil system. This will be accomplished by drilling and tapping the drain plug so the valve and associated fittings can be installed on to it.
REASON FOR CHANGE The present method for removing oil and taking samples is difficult. _The location of the drain plug does not lend itself to easy access. The location of conduit and the cooling system for the turbine are obstructions that make it difficult to remove the drain plug for the lube oil.
SAFETY EVALUATION The FSAR accidents affected by this change are those initiated by EFW start.
However, this chance has no potential to start EFW so no accidents are affected. All components will be installed to ASME Class 3 requirements to preclude any adverse affect on the EFW turbine. No new system interactions are created and no new methods of failure are introduced by this change. No margin of safety is reduced and no protective boundanes are affected.
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- 8. DC-3434. EH System Constant Pressure Pumps (Revision 2F DESCRIPTION The proposed change installs hoses on the EH Pump discharge lines to eliminate or reduce pump induced vibration.
REASON FOR CHANGE Since the new constant pressure pumps were installed by Revision 1, pump induced vibration has been experienced.
SAFETY EVALUATION I
According to the safety evaluation, there are no unreviewed safety questions associated with this change. Events listed in the FSAR which may be affected by the proposed change include turbine trip event and turbine trip with a single active failure.
The probability or consequences of these accidents will not be increased. The integrity of the EH system will be maintained and reliability improved by reducing system vibration caused by the EH pumps. No new interconnections to any important-to-safety equipment will be created and no new failure methods introduced. No margin of safety associated with a structure, system, or component will be reduced.
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- 9. DC-3435. Replacement of Fuel and CEA Hoists on Dual-Masted Refuelina Machine DESCRIPTION The proposed change : eplaces the CEA and Fuel Hoists on the dual-masted Refueling Machine and removes the dry-sipping hoses attached to the festoon for the kefueling Machine, it adds permanent communications cable to the Refueling Machine trolley to _
replace the temporary communications used previously. It also deletes the FSAR requirement for the CEA hoist to stall at a load below the tensile capacity of a CEA.
REASON FOR CHANGE The CEA hoist drive has previousi, failed and has also locked up. The CEA hoist is required to stall below the hoist tensile capacity of 600 pounds. This is not possible since the CEA hoist alone weighs 1000 pounds. The temporary communication setup has failed in the past and requires replacement to ensure reliable communication between the Control Room and the refueling station. The dry-sipping system was removed earlier but the hoses were left in place. The hoses will be removed to allow installation of the communication cable.
SAFETY EVALUATION According to the safety evaluation, the replacement of the CEA and Fuel hoists on the Refueling Machine, installing a permanent communications cable, removing the dry-sipping hoses, and deleting the FSAR requirement for the CEA hoist will not reduce any safety margins for the Refueling Machine or the Fuel Handling System. No accidents or important-to-safety equipment are affected by these changes and no unreviewed safety question is created 9
- 10. DC-3440. HPSI Flow Control Valve Replacement -
DESCRIPTION The existing HPSI flow control valves'have seats that are subject to cracking. Due to l the design of these valves, the cracked seats cannot be replaced without a high probability of damaging the valve body. In addition, the valve design has free play between the stem and disc which can cause the disc to become cocked or could ultimately result in stem disc separation due to wear.
REASON FOR CHANGE The new valve bodies have an Acme screw threaded and tack welded seat design which can be replaced, if necessary, without' damaging the valve body ' Also, the 1 increased strength of the new stem / disc assembly will eliminate the existing stem disc wear / separation problem.
i SAFETY EVALUATION
' According to the safety evaluation, the new valve bodies and stem / discs supplied by the original equipment manufacturer are essentially like-for-like replacements with a few minor differences; The repiccoment valves have been specified to allow the same -
- flow as the existing valve, and have been specified to be able to pass a large enough -
particle size to preclude flow blockage in the recirculation mode of operation. These requirements will ensure that the necessary flow can be achieved through the new valves, and that the valves will properly perform their accident mitigation function. The installation sequence was developed ensuring that sufficient boration flow paths are available throughout the installation activity. No unreviewed safety question is involved w;th this modification.
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- 11. DC-3448. Annunciator Interlock for Monitorina a Loss of Bus Transfer,Qaoability (Revision 0 and Revision 1)
DESCRIPTION OF CHANGE The proposed changes for the new automatic bus transfer control circuit are to: 1) prevent a slow opening Unit Auxiliary Transformer (UAT) supply breaker to still be closed when a close signal to the Startup Transformer (SUT) supply breaker is present;
- 2) prevent an unacceptably slow closing SUT supply breaker from closing on the bus beyond a predetermined time, or exceeding the acceptable phase angle difference between the SUT feeder and the bus being transferred; 3) monitor the automatic bus transfer control circuit for electrical component failures.
REASON FOR CHANGE The existing bus transfer scheme resulted in catastrophic equipment failure due to parallel closure of SUT and UAT breakers. The proposed change alters the control logic of the automatic bus transfer scheme to prevent such eouipment failures due to bus transfers that are outside certain parameters. In addition, the modification provides a means to monitor the automatic bus transfer control circuit.
SAFETY EVALUATION Failure of the automatic bus transfer scheme would result in a loss of Offsite Power (LOOP). The proposed change decreases the probability of a LOOP by improving the reliability of the automatic bus transfer scheme. The new system interactions with existing safety related equipment involve an increase in bus ' dead time' imposed by the introduction of an 'early b' contact. The existing scheme has a dead bus duration of approximately 1 to 2 cycles. For the new scheme, any equipment that was operating prior to transfer will see approximately 3 to 5 cycles of bus dead time. The proposed change was analyzed up to 14 cycles v/hich is comparable to an additional start type of stress on the motors. The increase in dead time is insignificant in the long term performance characteristics of the large motors (safety and non-safety). An evaluation of the increased dead time rc:;uiting in mechanical stresses on the shafts of large motors concluded there is no thermal or mec",anical damage to the motors or the driven equipment. There are no unreviewed safety questions associated with this change.
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' 12. DC-3462c Control Room Noise Reductions DESCRIPTION The purpose of this change is to further reduce the ambient background noise levels in
- the main operating area of the Control Room by installing divider panels in the walkways on the east and west sides of the main control panel, it also performs a post installation air flow test on all diffusers to ensure that proper air flow throughout the Control Room is maintained. It also adds six access ceiling panels in the proper area of the Control Room in order to maintain temperatures inside the Control Room within the set limits defined in NUMARC 87-00 and FSAR Appendix 8.1 A.
REASON FOR CHANGE '
Upon comp!stion of Phase 1.of DC-4362, operations personnel identified that although 4 the Control Rom Noise Reduction modification resulted in improvement / reduction in ambient noist levels within the proper area, reduction in noise levels along the walkway aisles on both sides of the main control paneI would further improve the communication ability of the control room operators / staff. Adding the six access ceiling panels will ensure the temperature in the proper area of the Control Room will remain within an acceptable range during/ following a Station Blackout SAFETY EVALUATION The safety evaluation has confirmed that the proposed changes will not reduce the margin of safety as defined in the basis of any TS or safety analysis and no unreviewed safety questions are created.. There is no impact on the function of the main control panels due to installation of the divider panels. Calculation EC-C96-003 verified that'-
the divider panels will remain in place during a design basis earthquake and will not affect the function of nearby safety related control panels. Proper air flow throughout the Control Room will be maintained and demonstrated by air flow testing as part of the acceptance test. The overall function of the Control Room HVAC will not be impacted.
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,. 13. DC-3468. Essential Chilleg Water Chemistry improvement / NNS Trai_! r
! Relocation to the SupM inental Chilled Water Systent DESCRIPTION The proposed change will: 1) disconnect the NNS chilled water train from essential chilled water and connect it to the supplemental chilled water system; 2) remove essential chilled water bypass lines and replace with valved and flanged test connections; 3) disable chilled water temperature control valves in full open position;
- 4) modify control logic for AH-25 and AH-30; 5) abandon air handling units AH-31 and AH-27 in place REASON FOR CHANGE l During chilled water flow testing in 1995, the air handling units were found to have little or no flow with the downstream temperature control valves fully open. Subsequent chemistry analysis showed excessive corrosion products in these lines, which was attributed to stagnant or low flow conditions in various parts of the chilled water system.
These stagnant conditions are the result of temperature control valves remaining closed when the associated air handling unit is secured. The essential chilled water bypass valves are fall-open valves. The concern is this feature could leave one train of
' chilled water with diminished flow in the event a bypass valve failure occurred. The failure is currently mitigated by the redundancy of the essential chillers.
SAFETY EVALUATION Implementation of the proposed changes to the essential chilled water system as described do not result in an unreviewed safety question both during installation and in the final modified configuration. The removal of the NNS train from essential chilled water does not have an adverse affect on the safety-related function of essential chilled water since after implementation, no automatic function is required to isolate nonessential loads. The removal of the chiller bypass valves and disabling temperature control valves to a passive full open position provides the design flow to room coolers at all times. The modification of the control logic of the control valves at the cable vault and switchgear area air handling units to modulate when the fans are idle will avoid stagnant lines and allow loads to be accepted rapidly when fan trains are switched. The resulting flow rates have been resolved with respect to minimum required flow for adequate heat transfer within the chiller evaporators as well as the minimum flow requirement for the chilled water pumps. Abandoning the unused -
blowdown filter room air handling units and the unused computer room air handling unit has no safety significance.
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E14; DC-3470. ACCW System Waterhammer DESCRIPTION The proposed change will add the following design features to the ACCW system to eliminate and/or mitigate hydraulic transients in the system: 1) a pressure transmitter that will monitor high point pressure and automatically start the ACCW pump on low -
pressure; 2) a jockey pump that will operate continuously when the ACCW pump is .
- idle._--Valve ACC-110 will be modified to become an MOV in order to mitigate the -
- hydraulic transient effects due to either air intrusion or column separation which could occur upon an ACCW pump start following a Loss of Offsite Power (LOOP). The EDG sequence for a LOOP without an accident will be changed to automatically. start the ACCW pump on the 17 second load block. _ Interlocks will be installed on valves ACC-
- 112A(B) and ACC-139A(B), tubing changed on ACC-139A(B), and a time delay added to ACC-112A(B) to eliminate a hydraulic transient in the Essential Chill portion of the system. Four new pipe hangers will be added, seven modified, and two deleted to enhance ACCW piping structural integrity, REASON FOR CHANGE The ACCW system has been found to have had either air intrusion or column separation occur in the system. The absence of water in the ACCW system'is --
considered significant because of the potential for a hydraulic tra elsnt in the eve at of an automatic ACCW pump start. The original design of the ACCW system did not contain specific design features to prevent its susceptibility to air intrusion or column -
- separation during idle periods. The essential chiller portion of the system is also
' susceptible to column separation if the essential chillers are aligned to the Wet Cooling Towers (WCT) and the ACCW pump is not running, or if the essential chillers are switched from the WCT to the Dry Cooling Towers (DCT) and ACC-112A(B) closes -
before ACC-139A(B).
SAFETY EVALUATION The ACCW system is not considered an initiator for any accident described in the SAR, The safety function of the ultimate heat sink, including the ACCW system, to mitigate the consequences of an accident by dissipating the heat removed from the reactor and
- its auxiliaries after a design basis accident, is enhanced by the design change of the ACCW system. The proposed change also enhances the design of the ACCW system
- and helps ensure that ACCW piping, support, and equipment loads remain within their design allowables during normal and accident conditions.
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15.--- DC-3472. Hot and Cold Leo RTD Noisa Abatement DESCRIPTION The proposed change modifies the shield grounding scheme for the Hot and Cold Leg -
RTD cables, it also improves cable and shielding configurations to reduce electro-tmagnetic interference (EMI) and separates power conductors from RTD circuit conductors.
- REASON FOR CHANGE This change is being made to reduce the various anomalies associated with the Hot I and Cold Leg RTD circuits.
SAFETY EVALUATION The safety evaluation has concluded that all of the proposed changes have no adverse impact on safe operation of the plant. The proposed changes do not modify electrical circuit interconnections that would alter the operation of the plant. The proposed EMI improvements are all passive in nature and do not add any active components that would potentially alter the intended function of the systems. The proposed changes will not cause nor do they affect any accidents described in the FSAR. No margin of safety is affected by the changes and there are no unreviewed safety questions associated with these changes.
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- 16. DC-3476. Additional Source of Containment Buildina Temporary Power for Outaaes DESCRIPTION The Hydrogen Recombiner System is part of the Combustible Gas Control Sye.em.
The function of this dual-train system is to limit the concentration of hydrogen in the containment (following a LOCA) to a 4% volume. Each recombiner is equipped with electric heating coils which are powered from independent safety buses. Thase components are s. tarted manually by the operator upon detection of a high concentration of hydrogen (there are no automatic actuation signals). The recombiners, which are designed to process combustible gas and air mixtures to remove hydrogen by reacting with oxygen to form water, are located such that uniform mixing of the atmosphere occurs b/ natural circulation. The A (West) and B (North) recombiners will be revised to include transfer capability from the normal heater load to new distribution panels. These panels will have individual breakers feeding 480v receptacles to power the required refueling loads on elevation +46.
REASON FOR CHANGE This change will provide additional power source in containment for use during refuel outages.
SAFETY EVALUATION s
The Hydrogen Recombiners will be energized from their respective power sources during all plant operating cycles. They will be fully operational and capable of performing their design basis function prior to closing containment. These components are not required during Modes 3 through 6, per TS 4.6.4.2. Each Hydrogen Recombiner will have an additional molded case switch installed in the supply circuit but these switches will be qualified for 1E operation. Administrative controls will ensure that the power supply is functional and is returned to the normal position for the subsequent operating cycle.
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17,- DC-3477. Feedwater Pump Vibration Mon;torina DESCRIPTION The proposed change will replace the existing single-axis vibre'aon monitoring equipment and thrust bearing wear detectors on the turbine driven steam generator feed pumps with 2-axis vibration monitoring equipment and new thrust bearing wear detectors. This includes a change to the Control Room display and recording equipment to accept the new data generated.
REASON FOR CHANGE
- The new vibration monitoring system is more reliable, easier to malatain, and provides more complete information.
SAFETY EVALUATION
- A feedwater pump turbine trip, which may be caused by the vibration monitoring equipment (thrust bearing wear channel), is enveloped by the Loss of Feedwater event (which is classified as an Infrequent incident in the Accident Analysis Section of FSAR Chapter 15). There are no Unreviewed Safety Questions involved in this change.
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- 18. DC-3478. Safety inlection Nitronen Valve Replacement 1
DESCRIPTION The proposed change will replace the existing Safety injection Tank (SIT) Nitrogen Fill valves with 1" diameter, soft-seated, air operated ball valves.
REASON FOR CHANGE The existing hard-seated valved leaked by excessively when the valve was closed.
The replacement valve is soft-seated, which is the correct design for preventing leakage in gas service.
SAFETY EVALUATION According to the safety evaluation, replacement of the hard seated globe valves with soft seated ball valves will provide a more positive shutoff for leakage in gas service.
This allows the valves to adequately perform their safety function of maintaining Safety injection Tank pressure. Since the same type of valve operator is used (fail-closed, air- .,
- operator), no new failure modes are introduced. The valve is sized to limit flow capacity to the rated capacity of the SIT safety relief valves, thus ensuring that the SITS '
cannot be overpressurized. There are no new system interactions or connections created by this modification.
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- 19. DC-3480. Safety iniection Tank Level and Pressure Alarm Relocation DESCRIPTION The water level and nitrogen pressure operability limits for the Safety injection Tanks (SITS) have been broadened to require less operator action. As a result of this change to TS 3.5.1, SIT pressure and level instrumentation loops are being modified to enable the instruments to monitor the increased pressure operating ranges and to provide redundant wide range level indication loops to facilitate channel checking of the instruments. The number of SIT level loops is changed from two narrow range and one wide range to two wide range level loops only. The narrow range pressure loops are modified so their upper ranges are expanded from 650 to 750 psig and the wide range pressure loops will have their upper range expanded from 700 to 750 psig. Level alarms, which are presently on the narrow range loops, will be transferred to the wide range loops.
REASON FOR CHANGE The revised level and pressure instrument ranges will more effectively monitor the increased parameter ranges afforded by the change to TS 3.5.1. The revised TS provides wider operating ranges for level and pressure so less attention to SIT servicing will be required while still ensuring preservation of the design bases.
Extension of the ranges of the pressure instrumentation will also satisfy the requirements of RG 1.97 without the need for the previously approved exception.
SAFETY EVALUATION The accidents which are mitigated by the SITS are the large and small break LOCK The proposed changes to the SIT level and pressure instrumentation will have no impact on operation of the SITS. No system functions or interactions are affected by the change and no margin of safety is reduced. Therefore, there is no unreviewed safety question.
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- 20. DC-3483. LPSI Pumo Minimum Flow Recirculation Line: Check Valve Replacement DESCRIPTION The proposed change replaces the existing LPSI pump minimum recirculation 2" check valves with 2" stop check valves to provide a positive means to isolate the LPSI pumps ,
from the common recirculation / test header.
.R_EASON FOR CHANGE Each of the HPSI and CS pump minimum flow recirculation lines are presently equipped with stop check valves to allow for maintenance on the downsteam instruments and valves. The LPSI pump minimum flow recirculation lines are presently equipped with only check valves and the pumps can only be isolated by closing " fail open" solenoid valves which could leak by if power were lost.
- SAFETY EVALUATION The safety evaluation states that the function of the LPSI pump minimum recirculation valves is to protect the LPSI system from overpressurization condition from the HPSI and CS systems during testing. Tha stop check valve which will replace existing piston check valves will not affect the function or operation of the Si system because the replacement valves function as check valves when in the open position. This replacement of a check valve with a stop check valve will not impact either the Si system function or integrity, except to permit maintenance to be done outside the SDC outage windows and to eliminate the risk of the downstream solenoid valve failing open <
during maintenance of the recirculation headers. This valve will have the T-handle locked open during normal plart operating conditions to ensure that the minimum flow for pump protection is maintained. The proposed change does not reduca the margin of safety and does not create an unreviewed safety question.
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- 21. DC-3488. Reactor Coolant Pumo Controlled Bleed-Off Flow Transmitter Replacement DESCRIPTION Replace the RCP Seal Bleed-Off flow transmitters, which are magnetically-coupled rotameters, with coriolis flow element / transmitters.
REASON FOR CHANGE The existing flow elements are cbsolete.
SAFETY EVALUATION The worse case accident associated with this change is a small break LOCA. The replacement flo el elements and new piping meet or exceed the original component ratings for pressure boundary retention and seismic qualification. The piping is qualified as ASME Class 2, the flow element is not required to be ASME qualified, but will be (nird-party fetted for pressure retention and seismic qualificetion. The safety evaluation has determined there are no unreviewed safety questions.
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- 22. DC-3489. Addition of Maintenance Capability in CVC Letdown to the VCT QESCRIPTION The proposed change installs a 3" manual ball valve in the Letdown line to the Volume Control Tank (VCT);
REASON FOR CHANGE This valve can be utilized during plant shutdown conditions to isolate the VCT from Letdown piping Closing this valve will allow shutdown cooling to be aligned to the letdown purification lon exchangers while maintenance is performed downstream of the new valve.
gAffTY EVALUATION According to the safety evaluation, the new valve will replace the existing CVC-166 valve as the boundary between SDC flow and the VCT. The same administrative controls which ensure CVC-166 is not inadvertently opened during SDC purification will do so for the new valve. Failure of the new valve in the open position would have the same consequences as failure of CVC-166 when it is used to isolate SDC flow from the VCT. The valve wi'l be rarmally locked open to prevent inadvertent manual valve operailon, in addition, the letdown portion of CVCS is not required for safe shutdown or to maintain the plant in safe chutdown condition. No new system connections are required and no protective boundaries are affected.
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- 23. DC-3491. Additional Trisodium Phosphate (TSP) Baskets for the Safety iniection System Sump DESCRIPTION DC 3491 willincrease the storage capacity of TSP to support raising the post-accident Safety Injection system sump pH to >= 7.0 under limiting conditions for maximum boric acid concentrations based on next and future fuel design changes.
REASON FOR CHANGE Fuel design changes will result in higher boron concentrations in the Refueling Water Storage Pool (RWSP) and Safety injection Tanks (SIT). The SI system injects the borated water from the SIT and the RWSP into the Reactor Coolant System (RCS) for post accident cooling to limit core damage and fission products release and to ensure adequate shutdown margins. This design change increases the capacity of TSP that can be stored to achieve the required pH with higher boron concentrations in the RWSP.
SAFETY EVALUATION According to the safety evaluation, the individual baskets of TSP are seismically designed and will not impact any system or equipment in the event of an earthquake.
The baskets will have negligible effects on Containment free volume or flood levels ard they do not affect any other systems and do not require new interconnections. The additional TSP maintains the margin of safety by ensuring the post- accident Containment sump pH is >= 7.0.
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- 24. DC-3492. Addition of Filters on the Component Coolina Water System Q_ElgRIPTION E
The proposed change installs a filtration system on the non-essential related portion of the CCW system.
REASON FOR CHANGE The CCW system is provided with facilities for chemical addition for control of bacteria but no provicions are made for removal of the suspended solids which build up in the system as the chemicals kill the bacteria.
SAFETY EVALUATION A Loss of Coolant Accident (LOCA) relies partly on the CCW system to dissipate heat removed from the reactor and its auxiliaries during a design basis accident. Failure of the non essential portions of the CCW cannot initiate a LOCA or any other previously analyzed accident. Therefore, no accident or its consequences are affected by the proposed change. To preclude draining of the CCW Surge Tank and subsequent tripping of both CCW tanks on low NPSH, a calculation was performed which demonstrated that the isolation valves would isolate the non-essential portion of the system before the Surge Tank would be drained. To prevent air entering the safety related portions of CCW, routine maintenance of the filters would be proceduralized, including fill and venting of the filter vessels. Thus, no important-to safety equipment 6
will be affected by the change. No accident different than previously ovaluated in the FSAR will be created, no new failure methods are created, and no matgin of safety is reduced, j
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- 25. DC-3493. Stroke Time Speed Reduction of CCW Header Isolation Valves DESCRIPTION The proposed change modifies the pneumatic actuation time of certain isolation valves ,
which have caused undesirable pressure transients during past valve surveillance. ;
REASON FOR CHANGE i During surveillance testing of valves CC-200A(B), CC 563, CC-727, CC 641, CC 713, and CC 710, a pressure spike causes some thermal relief valves to momentarily open l with resultant loss of a small amount of CCW inventory, SAFETY EVALUATION i
There are no analyzed accidents initiated by any of the valves affected by this change and no accident consequences are Increased The safety function of neither these-valves nor other important-to safety equipment is adversely affected by the proposed change and no new fallare modes are created No protective boundaries are affected and no margin of safety is reduced, i
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- 26. DC 3494. Steam Generator Blowdown Flow Restoration l
DESCRIPTION !
1 The proposed change replaces the four Anchor Darling SG Blowdown Containment I lsolation valves with two Anchor Darling double-disc,4" gate valves, and with two Masonellan double seated,4" globe valves.
REASON FOR CHANGE I l
The existing gate valves use a flexible wedge disc that exhibits thermal binding when the valve has been closed for long periods of time. As a result, the Emergency Operating Procedure function to control steam generator level has been negatively affected and procedural " work arounds" had to be implemented requiring an operator to manually open the blowdown valves.
SAFETY EVALUATION The proposed changes have been evaluated for their impact on safety-related systems and components. The review has determined there is no impact on the operation and performance of the SG Blowdown System (SGBS); there are no changes to the interactions between SGBS and other safety-related systems or components; there is no reduction in established safety margins; and no unreviewed safety question is created Since the proposed changes retain the function and control philosophy of the existing system and the design criteria of the new valves is very similar to the existing valves, there is no increase in the probability of occurrence of an accident previously evaluated in the FSAR. Replacing the inside and outside Containment isolation valves does not lead to increased consequences of an accident previously evaluated in the FSAR.
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- 27. DC-3495. Boric Acid and Primary Makeuo Flow Imorovements DESCRIPTION This design change improves the accuracy and operating flow range of the Boric Acid blending system by establishing limits on the boric acid blend flow control valves and installing a new flowmeter.
REASON FOR CHANGE Since the end of refueling outage 7, the amount of boric acid being added to the volume control tank has been increasing and experiences some variations. Two possible scenarios which lead to boron / water ratio inaccuracies are: 1) higher actual flows which exceed the measurement capability of flow instrumentation with the BAM/PMU valves full open; 2) high inaccuracy at low flow rates due to the square root function of the orifice plate for the PMU flow measurement device.
SAFETY EVALUATION According to the safety evaluation, the proposed chango will have no affect on the inadvertent dilution event described in the FSAR. The proposed change improves the indication of flow in the full range of the valve opening which enhances the means of detecting abnormal leakage from RCS. The opening of the flow control valves will 4 continue to be limited by the existing mechanical limiters per the original design of the valves. No new system interactions are created by this change. No protective boundaries are affected and no margin of safety is reduced.
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I DESCRIPTION
_ The proposed change provides regulated power a*Jpply to the valve position indicating .
- circuits and replaces the existing analog indicators with more accurate digital indicators, i
l REASON FOR CHANGE
, Engineering evaluation indicated that the bus voltage variance and tight tolerance contributed to HPSI and LPSI flow control valve indication problems. The evaluation
- i. recommended regulating the voltage source to the position indication circuit to preclude i errors due to voltage variations from the feed bus.
SAFETY EVALUATION The HPSI and LPSI flow control valve position indicators are provided for monitoring anticipated operational occurrences and of accident conditions by the Operator. These ,
- valve position indication circuits are part of safety-related display instrumentation of !
! Plant Process Display and ESF monitoring system but have no sense, command, or '
control function for accident mitigation, Providing a regulated power supply and
- replacing the indicators does not represent a change to the facility described in the FSAR and does not result in an unreviewed safety question.
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- 29. DC-3502. Reduce Bypass Leakaae from Penetrations 53 and 65 DESCRIPTION DC-3502 provides a perrnanent solution to reduce potential bypass leakage from the non-essential lines associated with penetrations 53 and 65. CVR-401B will be removed and placed upstream of CVR-401 A. CVR-401B wiil be powered from an independent train of electrical power, making it redundant to CVR-401 A. Excess flow check valves CVR-402A(B) will be removed. Valves will be added on the Penetration 53 non-essential line to facilitate Local Leak Rate Testing.
REASON FOR CHANGE The non-essential instrument lines from Penetrations 53 and 65 communicate directly with the containment atmosphere. Each line routes from containment to a solenoid globe valve that closes automatically on a CIAS. An excess flow check valve is located downstream of the automatic valves. The tubing for these non-essentialinstrument lines up to and including the excess flow check valves is ASME Section Ill, Class 2, seismic Category 1. The remaining portion of the lines are non safety and although seismically supported, these lines downstream of the isolation valves are not classified as seismic Category 1. The linns terminate at cabinet C-4 which is located outside the area exhausted by the Controlled Ventilation Area System (CVAS). Postulating a single active failure of CVR-401 A or B and a tube rupture on the non-safety part of the non-essenti J instrument lines, bypass leakage would be limited by an excess check flow valve, CVR-402A or B.
SAFETY EVALUATION The proposed change enhances plant safety by reducing the potential bypass leakage through penetrations 53 and 65. This proposed modification of Penetrations 53 and 65 non-essential lines with both containment isolation valves CVR-400 (formerly CVR-4018) and CVR-401 (formerly CVR-401 A) outside containment has been evaluated per Licensing Amendment Request NPF-38181. According to the safety evaluation, the ,
change in power supplies for CVR-400 and CVR-401, the implementation activities associated with DC-3502, and the effect of DC-3502 as it relates to GL 96-06 do not :
create an unreviewed safety question. I 29 l
- 30. DC-3503. Emeraency Diesel Generator Miscellaneous Vent. Drain. and Samol t Lines DESCRIPTION This design change modifies EDG A and B by: 1) adding a sample line and valve to the
- Lubricating Oil system; 2) adding a vent line and valve to the Jacket Water Piping system; 3) adding a vent line and valve to the Fuel Oil Piping system; 4) adding a drain line and valve to the lube oil cooler water boxes; 5) seismically qualifying the Klene valve extension tubes for seismic ll over I mounting; 6) removing previously abandoned tubing on the sides of the engines; 7) permanently installing a hydraulic Jack on the side of the lube oil filter housings; and 8) removing Jacket water control valves EGC-107 and EGC-108 from the inlet line to the combustion air heaters.
REASON FOR CHANGE The addition of vents, drains, and sample line to the Jacket Water, Fuel Oil, and Lube Oil systems is intended to help service the systems during maintenance and surveillance activities. Removing the two abandoned lines eliminates comments and questions made about these lines during walkdowns and surveillances. Permanently installing the jack will preclude it from being misplaced or lost. Removing the control valves (and their thermostatic controller) is based on a recommendation from the supplier and the problems with obtaining fully qualified parts.
SAFETY EVALUATION The review of the proposed changes has determined there is no impact on the operation and performance of the EDGs, there is no reduction in safety margins established for any safety-related system or component,~ and no unreviewed safety questions. The changes do not affect the performance of the EDGs or their ability to provide redundant electric power to ESF buses 3A3 S and 3B3-S.
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- 31. DC-3506. Auxiliary Steam Test DESCRIPTION This design change will install a new 6" pipe tee with a blind flanged connection on line 3MS6-90B to allow for connection of a temporary boiler to supply steam to turbine-driven EFW Pump NB for post-maintenance testing during plant outages. The existing non-safety heat tracing circuit for the piping modified will be replaced with a longer cable of a different type, due to the added heat sink of the new pipe fittings. A second identical" installed spare" heat tracing cable will ensure that periods of heat tracing unavailability are minimized.
REASON FOR CHANGE Technical Specification 3.7.1.2 requires that all three EFW pumps be operable in Modes 1,2, and 3. If only two are operable and the third cannot be returned to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the plant must be placed in 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 />. Because of these requirements, routine maintenance and periodic overhaul of steam turbine-driven EFW Pump NB occurs during refuel outages. Testing of the turbine and pump can then only be performed during plant startup when any problems discovered could extend the plant outage.
SAFETY EVALUATION The proposed changes will not reduce the margin of safety as defined in the basis of ,
any TS or safety analysis and no unreviewed safety questions are created. The proposed added test connection maintains the integrity of the Main Steam pressure boundary by using materials and methods identical to those used in the existing system. In addition, the blind flange will only be removed during plant outages when the MS system is not in service.
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- 32. DC-3508. MSR Shell Drain Tank Normal Level Control Valve Fail Open Modification DESCRIPTION Temporary Alteration TA-95-005 disabled the automatic level control valves which are in the drains of the Moisture Separator / Reheater Shell Drain Tanks (SDT) and discharge to the #2 Feedwater Heater shell sides. These control valves are air-operated globe valves which fail open on loss of air. The valves are blocked in the failed-open position to reduce the possibility of a SDT Level Hi Hi condition, which will result in a turbine trip. This change will incorporate the changes installed by TA 005 Into the permanent plant design base configuration.
REASON FOR CHANGE The original design of the MSR SDT Normal Level Control Valves (NLCV)was to modulate and maintain a minimum level in the tank. In the event the valve failed closed, level modulation for that specific SDT would be shifted to the Alternate Level Control Valve (ALCV). During routine maintenance that required closing one NLCV, it was identified that the ALCV could not maintain level, it was necessary to reopen the NLCV to prevent level from tripping the turbine on Hi-Hi level. Tne ALCVs were undersized by 7.8% and could not handle full design flow, installation of TA-95-005 demonstrated that the NLCVs being blocked in the full open position allowed the plant to function satisfactorily and did not adversely affect plant operation.
SAFETY EVALUATION The proposed change involves secondary steam cycle components which are not '
required for safe operation of the plant, and are not required for plant shutdown or accident mitigation. The change does not affect any radioactive material creation, storage, processing, or discharge, and does not create an unreviewed safety question.
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- 33. DC 3513 CCW/ACCW Pump Improvements DESCRIPTION The proposed CCWiACCW pump improvements include utilization of an internal vent process in place of the sight and vent fitting currently installed on the thrust bearing, replacing the existing oil supply and return piping lines to minimize oil leakage, and installing manual vent valves on the pump casing to allow adequate venting following maintenance.
REASON FOR CHANGE These improvements are required because the sight and vent fittings and the existing supply and return lines have been a chronic source of oil leakage, and there is concern that the pumps may be susceptible to air entrainment in the pump casing.
SAFETY EVALUATION According to the safety evaluation, the installation of manual vent valv9s on the pump casings will not impact either the CCW or ACCW system function or integrity. The valves permit manual venting of the pump casings prior to placing the system in service without removing the 3/4" vent plug from the casing. The modifications made to the oil lubrication systems represent design improvements intended to minimize oil leakage and improve pump reliability. These modifications do not affect operation of the pumps and have been reviewed for impact on the seismic qualification of the pumps. The
, changes do not reduce the margin of safety and will not create an unreviewed safety l question.
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- 34. DC 3518. Condenser Air Evacuation System Imorovements I
DESCRIPTION l
The proposed change adds Air Evacuation (AE) volumetric capacity by modifying the existing evacuation skids, reroutes cooling water to all three AE skids in order to supply colder water which will increase their performance and efficiency, and removes the requirement for the Main Condenser being evacuated down to 2.5" Hg within 103 minutes in section 10.4.2.
REASON FOR CHANGE With six air cooler sections all connected, any differences in pressure will result in unequal flow The waterboxes are cooled by the CW system which contains small debris and sitt that tends to get caught at the inlet tubesheets. This contributes to unequal pressures and flows from the air cooler vents and in turn causes unequal venting of oxygen and non-condensable gases from the waterboxes. During winter months when river water is cold and condenser pressure is low, the AE system lacks the volumetric capacity to adequately remove in leaking air and non-condensables to control the condensate dissolved oxygen. High levels of dissolved oxygen can be detrimental to the SGs and other secondary components. Maintaining TCCW temperature at the high end of the acceptable range decreases performance of the evacuation pump. There is no design basis or system requirement for the condenser draw down time; therefore, it is being removed from the FSAR.
SAFETY EVALUATION Loss of Condenser Vacuum is an accident which could be initiated by the proposed change; however, the proposed change lessens that chance by increasing the efficiency and capacity of the AE system. SG Tube Rupture could cause the secondary systems to become radioactive; however, the leakage paths would be redirected automatically upon detection of high radiation so the consequences would not be increased, in addition, the three vacuum pumps are sized to account for the increased volume of air that must be removed in the winter versus a lower amount in the summer.
The volume increase is caused by a decrease in condenser pressure in the winter.
Therefore, this change will not change the maximum mass flow of air from the
- condenser. Because the proposed change more efficiently enables the AE system to maintain condenser a
- r arid non-condensables at a low level, the TCCW system, Main Condenser, and AE system are not adversely affected by the change. No new system interfaces ar6 created by the change No protective boundaries are affected and no margin of safety is reduced.
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- 35. DC-3521. Route DCT Sumps Discharoe to Circulatina Water System DESCRIPTION The proposed change will install a new 8" isolation valve on line 7CW16-55 to allow future connection of the discharge of the Dry Cooling Tower (DCT) sumps. The reroute of the sump discharge is required to add dilution to the sump effluent and effectively reduce offsite dose.
REASON FOR CHANGS When rain fills the DCT sumps, the discharge radiation monitors alarm as the sumps are being pumped out, stopping the discharge. Naturally occurring activity in the surrounding concrete is causing the alarm. By routing the DCT sump discharge to CW, the radiation monitor setpoint can be changed permanently to prevent naturally occurring activity from setting off the alarm. This will greatly dilute any activity released and reduce the dose reported at the discharge point.
SAFETY EVALUATION The Circulating Water system is a non-safety, non-seismic and is not postulated to initiate any accident analyzed in Chapter 15 of the FSAR. The proposed change does not increase the probability or consequences of an accident or failure of equiprnent important-to safety. CW system pressure boundary integrity will be maintained by using materials and construction methods identical to the existing system. No margin of safety is reduced and there is no unreaeym! safety question associated with this change.
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- 36. DC-3522. Removal of Steam Generator Feed Pumo Suction Relief Valves DESCRIPTION This change removes the existing Lonergan relief valves and also removes three supports on each of the SGFP suction trim piping to prevent thermal binding of the branch lines.
REASON FOR CHANGE Numerous work packages have been written to document that these valves have had problems with seat leakage and/or lifting and not reseating.
SAFEW EVALUATION The proposed change does not involve an unreviewed safety question. The effect of the valve removal on the SGFP suction operation is considered a design enhancement t since an inadvertent actuation of the valve is eliminated while the potential overpressure flows can be accommodated through the pump shaft seal leakoff. The function, operation, and piping boundaries of the blind flange installed in place of these valves does not affect any safety related structure, system, or component important-to-safety or needed for accident mitigation.
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-The proposed change installs keylock switches which can be used to overrido the emergency actuation signals to valves CS 125A(b,, CC 807A(B), CC-8MA(B), CC-822A(B), and CC-823A(B), closing or allowing them to be closed. The new switches are located in the Auxiliary Building Relay room and produce Control Room annunciation when they are moved from the normal position. The CS switches return normal valve control to the Control Room; the Containment Cooler switches unconditionally close the inlet and discharge velves for the CCW flow to the individual cooler.
REASON FOR CHANGE The lack of control of the CS valves may be undesirable if a further system failure is
, experienced during a CSAS initiation while the containment is pressurized due to the accident - containment isolation is maintained by a single check valve under these circumstances. Similarly, if a containment fan cooler should experience a leak or break under post - LOCA conditions, it should be possible to close the CCW isolation valves to the cooler for protection of the CCW water inventory, and containment isolation.
Both of these scenarios present a potentially unmonitored release of post accident radioacJvity and will be terminated by remote manual 2 %sure of the valves by switches provided by this change.
SAFETY EVALUATION According to the safety evaluation, the keylock switches which will be Installed by this plant modification are designed to mitigate the consequences of equipment failure 1
subsequent to an accident. The switches are normally open, are alarmed in the off.
normal position, and are keylocked in the normal position under administrative control to minimize the possibility of the switches blocking the designed operation of the associated valves. No unreviewed safety question is created by this modification.
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- 38. DC-3530. Emoroency Feedwater Pump Turbine Steam Supply Drains DESCRIPTION Install drain lines at low points in Emergency Feedwater Pump Turbine steam supply lines to prevent condensate accumulation.
REASON FOR CHANGE Concern that excessive condensation could be carried to the EFWPT governor valve during a startup transient, valve response time could be slowed due to water impingement, causing a turbine overspeed trip. There is also a potential for waterhammer from accumulated condensate when steam is admitted to the lines.
SAFETY EVALUATION Addition of the condensate drains in the steam supply lines of the EFWPT will not increase the probability or consequences of a feedwater flow accident. Analysis has shown that with previously evaluated drain isolation valves open plus all six trap lines failed, there will still be sufficient steam to operate the EFWPT. The design change meets the system pressure, temperature, Code, and seismic criteria. Therefore, the function of the EFV,ST will not be affected by this change. No new system interconnections are created and no new accidents are created.
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- 39. DC 3533. Post Accident Samplina System Samplino Capacity Improvements QESCRIPTION The proposed change will connect the two existing Pnmary Sample system lines from the HPSI pump recirculation lines to the PASS SIS sump sample pump line. The existing lines will be isolated from the new PASS lines by two, normally closed manual isolation valves. New check valves will be added to prevent recirculation between the sample lines. The PASS process control panel will also be modified.
REASON FOR CHANGE Under certoln accident or post accident conditions, there may be insufficient pressure in the RCS to push a sample from the RCS hot leg to the PASS sample stations. Under these conditions, the PASS SIS sample pump would be the only means of taking an RCS liquid sample. In addition, there is currently no means to perform a functional check of the PASS SIS sump sample pump.
SAFETY EVALUATION The proposed changes modify the Primary Sample and PASS systems outside containment, are not connected to the RCS, and do not affect any accident in the FSAR. The new and existing tubing and components are all qualified for the sample point design pressure and temperature. Therefore, the proposed chtnge will not affect any accident that involves a Primary Sample or PASS line break or leak. The effect of operation of the PASS on the NPSH available for the CS and HPSI pumps was evaluated and determined to have no affect on pump operation. The Si recirculation header sample isolation valves will be replaced with environmentally qualified components to ensure these valves remain functional in the post accident environment.
No protective boundary is affected by these changes and there are no unreviewed safety questions.
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- 40. DC 8006. Main Turbine Turnino Gear Control Switch Modification DESCRIPTION An additional interlock to the Main Turbine Zero Speed circuit will be added as a confirmatory contact of an existing tachometer under speed switch. The currently existing and the proposed switches are located at different positions on the Main Turbine-Generator shaft. Nine direct digital control switches will be removed from CP.
- 1. These DDS loops were installed as a means for providing automatic control for cooling water temperature control valves for FW pump turbine and Main turbine lobe oil, generator stator, and chemical feed.
REASON FOR CHANGE During the turbine roll up at the completion of Refuel Outage 5, the main turbine turning gear automatically engaged while the rotor speed was approximately 50 rpm causing significant damage to the turning gear mechanism. Turning gear engagement at higher speeds could cause considerable damage to bearings, bull gear, and may even result in rotor blade failure.
SAFETY EVALUATION FSAR Sections 10.0,10.1, and 10.2 describe the Main Turbine-Generator System. The main turbine turning gear is not described in these sections. This proposed modification improves the reliability of the turning gear engagement logic. The Chemical Feed meter pumps are components of the Domineralized Water System.
FSAR Section 9.2.3.2 describes the DWS. DWS serves no safety function since it is not required to achieve safe shutdown or initigate the consequences of an accident.
The FSAR does not describe the automatic operation of any of the loops for which the control / selector switches are being removed. This SE has determined that an Unreviewed Safety Question does not exist.
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- 41. DC-8016. Drv Coolina Tower Sumo Pumo Discharae Check Valve DESCRIPTION The proposed change installs a check valve in the DCT sump pump discharge piping.
REASON FOR CHANGE The DCT pumps have check valves installed to prevent back flow through the pumps and back into the sump. However, the radiation monitor sample piping is installed downstream of the sump pump check valves and the monitor has no provision for preventing backflow to the sump, SAFETY EVALUATION There are no accidents in the FSAR that are affected by this change; therefore, there is no increase in accident probability or consequences. The DCT sump pumps are required to remove water from the DCT areas during Probable Maximum Precipitation.
To ensure adequate water removal capacity, only one pump will be taken out of service at a time. The o_ther sump pumps will remain operable. In addition, a portable 100 gpm sump pump will be staged at the sump to be worked. Weather conditions will also be considered prior to breaching the system boundaries. No new system connections are created and no new failure methods are introduced. No margin of safety is affected.
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- 42. 0C 8025. Seal Oil Vapor Extractor Drain Line Improvements DESCRIPTION The drain lines for the Seal Oil vapor extractors will be reconfigured so they no longer drain into a common loop seal trap but drain into their own individual loop seal trap.
This new arrangement will eliminate the recirculation flow path that now exists between the two vapor extractors.
REASON FOR CHANGE The proposed change will eliminate startup problems encountered by the standby vapor extractor when it is started. The reconfigured lines will reduce the buildup of oil and water deposite in the standby vapor extractor housing and will eliminate binding problems caused by these deposits.
SAFETY EVALUATION The Seal Oil vapor extraction piping loop is a non-safety, non-seismic pipe line located within ths Turbine Building. The function, operation, and piping boundaries of the Seal Oil vapor extraction piping loop do not affect ot communicate with any safety-related structure, system, or component important to plant safety or accident mitigation. The proposed change will not affect any operating parameter or system function that would initiate a turbine trip such as low turbine bearing oil pressure, low seal oil differential pressure, or low turbine oil tank level.
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- 43. DC-8026. Hiah Pressure Turbine Gland Steam Spillover Capacity improvement DESCRIPTION The size and routing configuration of the high pressure (HP) turbine gland steam spillover piping loop and control valve station will be revised in order to increase the steam flow rates from the HP turbine "X" chambers. The new pip!ng design will encompass new design information provided by Westinghouse. Due to potentially high steam flow rates with worn seals, piping materials will be changed from carbon steel to stainless steel. This will eliminate flow-accelerated corrosion problems over the life of the plant.
REASON FOR CHANGE The gland seal spillover piping loop (including the spillover control valve station) for the HP turbine is not properly sized to handle steam leakage rates from worn gland seals.
This results in excess steam pressure in the "X" chambers on each end of the turbine.
The excess steam pressure causes additional steam to flow from the "X" chambers to the "Y" chambers. This additional steam flow overwhelms the leak off lines from the 'Y" chambers and causes steam to leak along the shaft and out into the environment. This results in lower plant efficiency, loss of condensate inventory, condensation in the turbine deck house, and personnel safety problems.
SAFETY EVALUATION The HP turbine spillover piping loop is non safety, non-seismic pipe line located within the physical boundaries of the Turbine Building. The function, operation, and piping boundaries of this piping loop do not affect or communicate with any safety related structure, system, or component important to plant safety or accident mitigation. Since the proposed changes retain the functionality and control philosophy of the existing system and the design criteria of the new pipe is adequate for all conceivable flow conditions, there is no increase in the probability of occurrence of an accident previously evaluated in the FSAR. Increasing the size and routing of the pipe will not lead to increased consequences of an accident previously evaluated in the FSAR. The modifications do not reduce the margin of safety as defined in the bases for any technical specification or safety analysis.
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- 44. DC-8028. E22 Fan Shut Down Interlock with the Toxic Chemical Monitors DESCRIPTION The proposed change installs an interlock from the Toxic Chemical Monitors to trip the E22 Normal RAB Exhaust Fans. Either Toxic Chemical Monitor will trip both the E22 fans. A low flow signal then ' 'os the operating supply fan and closes the intake / exhaust dampers to isolate the RAB.
REASON FOR CHANGE The change enhances integrity of the Control Room Envelope during a toxic chemical event. If a toxic chemical monitor trip of the E22 fans occurs, the operator will be able to override the trip and restart the fan. This override, with the trip present, will actuate an alarm in the Control Room to remind the operator this override condition is present.
This override feature allows the operator to restart the fans if the concentration or type of toxic chemical does not represent a hazard and allows fan restart due to spurious trips or trips during testing.
SAFETY EVALUATION The E22 exhaust fans a e part of the non-safety RAB Normal Ventilation System which is not required to operate for any design basis accident but is required to trip on a Safety injection Actuation Signal (SlAS). A new separation method for an associated information circuit is created with this plant change. However, the low voltage of the circuit, coupled with the fusing protection of the power supply, ensures there is no increase in the probability of a malfunction of important to-safety equipment. A new system interaction is created between the Toxic Cl.amical Monitors and the RAB Normal Ventilation System. Both of these systems are non safety and are not required for any accident and no new accidents are created and the change will enhance Control Room Envelope integrity. A new system interaction is also created by the new separation method. However, the fusing protection ense es no new accident is created. The proposed change increases the protective boundary for the Control Room by eliminating the negative pressure effect upon the enve; ope by the RAB Normal Ventilation System.
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B, CONDITION IDENTIFICATION / WORK AUTHORIZATION (Cl/WA)
- 1. Cl 300334/WA 01142862. Waste Tanks As-built Strainer Size Discrepancy DESCRIPTION The proposed change revises the size of strainer LWM-MSTRN-0001 A, from i 1/2" to 2".
REASON FOR CHANGE This resolves a discrepancy between the Liquid Waste Management system drawings and the as-built configuration.
SAFETY EVALUATION According to the safety evaluation, the strainer size revision to the flow diagram, isometric drawings, specification, and SIMS to reflect the as-built configurat'on will not impact the subject strainers or Liquid Waste Management system ability to perform its intended function. The proposed change does not affect the approved design basis -
- and there is no unreviewed safety question.
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- 2. pl 301153/WA-01144191. NI Loa Power Channel Calibration Safety Channel A.
{LC. or D lEvaluation is also for Ml-003102 (Rev 6) and OP-903-102 - see j,em II. A.61 DESCR,PTION The procedure revisions and Cl/WA reflect a conservative calibration of excore log power to 100% to match 100% reactor thermal power. Performance of the revised calibration wil! allow the exiting of the Excore Nuclear Instrumentation (ENI) Log Power Indicator LCO.
REtSON FOR CHANGE The revision enhances performability of the procedure. The revised tolerances and values for the " Log Calibrate" switch ensure that the ENI log power indicators used are calibrated to reactor thermal power.
Change in tolerance and values for the " Log Calibrate" switch positions are due to a change in the general b!as required by the changes in neutron flux at the log power neutron detector (described in the Engineering Input for the Cl/WA),
SAFETY EVALUATION According to the safety evaluation the only accident which has radiological release consequances for the log power trip is the Uncontrolled CEA Withdrawal from Subcritical Conditions.
The log power trip remains Out of Service until it can be returned to service. Since the equipment is Out of Service no credit is taken for its safety function and all Technical Specification LCOs will be met. It is not Applicable in Mode 1. The change in the general bias does not create any new system interactions or connections that did not previously exist, Indicated log power is altcred to better and more conservatively match actual power, The adjustment will improve the eccuracy of all log power indicators and the indications will remain within the loop accuracles.
The change in the general bias will maintain the margin of safety and assumptions used for the safety analysis by ensuring that the log power signal used for the log power trip correctly represents reactor power.
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- 3. Cl 301410/WA-01144762. Unit 3 Metal Waste Pond Transfer Pumo Recirculation Line Orifice Deletion DESCRIPTION Delete Restrictive Orifice LWM MFE 0297 from design drawings.
REASON FOR CHANGE Deletion will allow drawings to reflect the as built configuration.
SAFETY EVALUATION According to the safety evaluation, deletion of Restrictive Orifico LWM 627 from the flow diagram and piping Isometric drawing to reflect as-built conditions will have no impact on the Unit 3 Low Volume Metal Waste Pond Recirculation lines ability to perform its intencied function. FSAR Figure 9.3 3 will require revision to reflect the orifice deletion. No unreviewed safety question is associated witn this change, i
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_ -. _ . , . . , - _ _ . ._ - _ _ . , ~ . . _ . . _ _ _ _ _ _ _ _ . - . _ _ . ., -- .
- 4. Cl 301785/WA-01145481 Rearrance Instrumentation for Component Coolina Water Flow from Essential Chillers 4 DESCRIPTION This proposed change will replace the flow indicating switches and change the range of CCW flow from esseritial chillers to O to 1500 GPM. It will also change the range of the flow indication loop to O to 1500 GPM.
REASON FOR CHANGE Flcw indicators for CCW to the Essential Chillers are pegged high SAFETY EVALUATION This proposed change does not increase the probability of occurrence or consequences of an accident. This WA Repair Package does not increase the probability of occurrence or consequences of a malfunction of equipment important to safety. The presently installed flow indicating switches are used to trip the chillers on low CCW flow (510 GPM). The replacement switches will have the same setpoint (510 GPM). The tolerance will be changed to 125 GPM and the reset limit will be changed to 700 GPM due to the increased range of the flow switches. The CCW flow for the chillers will not bo affected. The change in reset limit and tolerance is acceptable fo-the flow required to operate the chillers. This change will enhance the operation of the chillers. No margin of safety is reduced and no unroviewed safety question is created.
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- 5. Cl-302111/WA-01146073.1/2" Heater Core Tubino for EDG 'B' DESCRIPTION The Wt package installs two,2-directional tubing suppcit clamps on the Heater cue vent te 'ng for EDG 'B',
REASON FOR CHANGE A routine walkdown of the plant identified excessive vibration of the 1/2" Heater core nf vent tubing during operation. The tubing is not supported per the manufacturer's recommendation.
SAFETY EVALUATION According to the safety evaluation, installation of the clamps minimizes line vibration and maintains the stability of the 1/2" vent line. The clamps do not affect the ability of the Heater core to perform its function of heating the inlet air as designed.
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- 6. ' Cl-302337/WA-01146881. Replauement of Door DiO85 DESCRIPTlQhl The Cl installs replacement doors ar;d door frame for Door D-085. Replacements will be installed such that the door swings into the Control Room envelope and not out of:
the Control Room envelope as currently exist.Ein addition the card readers for the door will be deactivated and the door will be administratively controlled as locked and alarmed.
RE.ASON FOR CHANGE' D-085 acts as a barrier in the opening between the Control Room HVAC Room and the
- RAB Normal HVAC area. _ lt is a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Appendix R fire barrier, a security barrier, and
--seismically mounted in place. The change of the door swing will result in better sealing of the door with a positive pressure in the Control Room envelope.
SAFETY EVALUATION .
According to the safety evaluation D-085 is not described in the FSAR, however, the swing of the door is shown on FSAR figures. The replacement door and change of
- door swing will not impact the function of the door in any way. During installation of the replacement provisions will be in place to ssal the envelope boundary in the event a condition occurs that will require isolating the control room envelope. The replacement door and hardware will also be UL rated for a 3-hour fire barrier. Replacement of D-085 will eliminate a problem area that has allowed air leakage when the Control Room has been isolated in the past. Reversing the swing will result in better sealing of the door.
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7.- Cl-302375/WA-01146f.37. CC-963B Needle Valve Replacement DESCRIPTION
--This Cl/WA will replace the current needie valve for CC 963B with a metering valve.
The metering valve has a much lower flow coefficient thsn the existing needle valve and is available with a vernier handle. Additionally, the time' delay setpoints for the -
- SDC HX A&B CCW VLV_ TROUBLE (Shutdown Cooling Heat Exchanger A&B,
' Component Cooling Water Valve) alarms will be increased from 5 seconds to 15 seconds.
- REASON FOR CHANGE in an effort to reduce pressure transients in the Component Cooling Water (CCW) system, the stroke times for CC-963A&B were increased to approximately thirty _
- seconds. This was accomplished by throttling a needle valve in the instrument air line supply for the actuator to 1/16 turn open. An attempt to stroke _ CC-903A closed failed
- when the velve v/ent 40% open and slowed to a stop. The needle valve was adjusted and CC-963A stroked closed in 31 seconds (Corrective Action Document CR-96-0430 discusses). Use of a metering valve with vernier handle will provide Operations with a .
means to accurately establish and verify the valve settings.
~ SAFETY EVALUATION According to the safety evaluation CC-963B is a normally closed, fail open valve.
-Instrument Air (IA) is supplied to provide the motive force to close CC-963B, Upon an SIAS, CC-963B is required to automatically open to pass flow through Shutdown -
. Cooling Heat Exchanger B. The fail open position assures.that CC-963B is capable of performing its safety function regardless of the status of the lA system. Since there are no safety functions associated with the closed stroke for the valve, there are no accidents affected that may have radiological release consequences altered by the-Cl/WA.
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- 8. Cl-302702/WA-01147033. TCCW Pumps 'A' and 'B' Air Eliminator Additions DESCRIPTION Air eliminators will be added to the casing vents for the Turbine Cooling Water Pumps
'A' and 'B',
REASON FOR CHANGE To prc; vide a continuous and automatic method for entrapped air removal.
SAFETY EVALUATION The proposed air eliminator additions are confined to the Turbine Closed Cooling Water System, a non-safety, non- radioactive system which has no affect on any FSAR accident scenarios. The air eliminator vents will not impact any other equipment or systems therefore an increase in the probability or consequences of a malfunction of equipment important to safety is not valid. There are no margins of safety associated with the TCCW System or TCCW Pumps and no unreviewed safety question is associated with this change.
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- 9. Cl-302960. Elevator Machine Room. RAB -35 DESCRIPTION This work package redefines certain fire barriers around the Elevator Machine room along with repairing a fire door. This change allows the Elevator Machine Room to be combinsd with the Elevator 'C' shaft in order to be treated as one fire enclosure.
REASON FOR CHANGE The Elevator Machine Room is open to the Elevator 'C' shaft by a 4'x6' opening for equipment access. The Elevator Machine Room is part of the shaft rated wall system and must be maintained for room integrity. Fire Door 173 will require repair to maintain room and shaft ir'egrity.
SAFETY EVALUATION I
According to the safety evaluation, per GL 86-10, this change meets the guidelines found in 10CFR50, Appendix R. Therefore, this change does not have any affect on '
any accident previously evaluated in the UFSAR, on any important to safety equipment, and does not create a new accident.
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- 10. Cl-302997/WA-01147507. EFW 'AB' Pump Security Cane Removal DESCRIPTION This WA package removes the expanded metal screen walls and ceiling in the EFW
'AB' Pump area and removes or relocates security related conduit, card readers, and other electrical boxes attached to the cage walls.
REASON FOR CHANGE The expanded metal ceiling of the cage has been identified to have at least three broken welds on ceiling members. A vertical post was found detached from the ceiling support members and a few of the ceiling panels were found sagging over the pump area.
SAFETY EVALUAYlON According to the safety evaluation, the removal of the expanded metal cage will not affect the pump or its components. The cage was installed for security reasons only; however, the area has now been decontrolled and the cage is no longer required.
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- 11. Cl-303093/WA-01147774. CVCS Charaina Pumos Pulsation Dampeners:
Schrader Valve Replacement DESCRIPTION The proposed change replaces the nitrogen fill valve (Schrader Valve) on all three trains of the CVCS Charging Pumps pulsation dampeners with a valve / adapter -
assembly.
REASON FOR CHANGE Corrective action document CR-95-0077 identified a reliability problem with the Schrader valve which allows nitrogen fill for the CVCS Charging Pump discharge pulsation dampener.
SAFETY EVALUATION The proposed change does not increase the probability or consequences of any accident or malfunction of equipment important to safety and the margin of safety will not be reduced.- The new components will be seismically qualified and designed to
- ASME Class 2 requirements which are the requirements of the CVC system. The design pressure and temperature of the new component is consistent with the existing CVCS Charging Pumps pulsation dampener design. The change does not require the system to be operated outside its design test limits. The new component is not part of the CVCS pressure boundary since the pressure boundary is maintained within the stem / bladder shell assembly.
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- 12. Cl-303246/WA-01148080. Safety Iniection Vent Line Addition DESCRIPTION The proposed change will add a vent downstream of check valve SI-216 which will include a 1 inch line and associated manual valve that will normally be closed.
REASON FOR CHANGE The change will provide adequate venting capability for the downstream header.
SAFETY EVALUATION ,
The addition of the vent line to the SI system will not increase the probability of occurrence or consequences of an accident previously evaluated in the SAR. The SI 1 system functions to mitigate the consequences of an accident (i.e. LOCA and MSLB) and is not considered to be the initiator of any accident. Additionally, the function of the Si system will remain unchanged as a result of the installation of the new vent line.
The piping, fittings, and associated velve will be designated Safety Related and Seismic and designed to AMSE Class 2 requirements. The vent valve will be normally .
closed and verified closed periodically. The vent line will provide added assurance that all air is removed during venting activities which is important to ensuring that the SI system performs as expected. No new system interactions have been created, no new _
methods of failure are created for the SI system, and no margin of safety is reduced.
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- 13. Cl-303338. Handrails in RAB and Coolina Tower Area DESCRIPTION The proposed change installs handrails on the RAB roof and the Cooling Tower Q-deck area.
REASON FOR CHANGE Personnel safety and to reduce the possibility of items falling into the Wet Cooling Towers (WCT).
SAFETY EVALUATION According to the safety evaluation, there is no unreviewed safety question. The handrails will become an integral part of the RAB roof and Q-deck structures. They are bounded by the current tornado missiles identified in FSAR Table 3.5-10. The plywood installed to prevent weld slag from falling into the WCT will cover only 48 of
- approximately 456 square feet of grating and open area and will be removed in the event of a tornado warning. '
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- 14. Cl-304444/WA-01150338. Ductwork Access Panels for Valves HVC-101 and HVC-102 DESCRIPTION The proposed change installs a 12"x20" access panel upstream of valve HVC-101 and a 10"x14" access panel upstream of HVC-102.
REASON FOR CHANGE Valves HVC-101 and HVC-102 require periodic maintenance to maintain positive closure and ensure leak tightness. These valves are located in seismic category I ductwork in a very congested area of the H&V Control Room Equipment Room.
Removal of the valves is time consuming and requires removal of portions of adjacent seismic supports. The access panels are needed to allow for minor adjustments to the valve seats and to inspect and/or clean the valve seats.
SAFETY EVALUATION According to the safety evaluation, installation of the access panels will not increase the probability or consequences of an accident. The Control Room HVAC system is not postulated to initiate any accidents and the access panels will not degrade the seismic category I HVAC system. The only important to safety equipment that could be affected is the valves themselves. This change only provides for easier access to them but does not affect their function. The only significant risk associated with the change is the risk of a fire during welding activities. To ensure a fire would be contained and the effects minimized, the downstream dampers will be gagged closed and the seats of downstream valve HVC-101 will be reworked and tested prior to welding. To maintain the margin of safety, the modified ductwork will ba smoke tested to ensure no leakage exists.
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- 15. Cl-304502/WA-01153320. Fire Water Storaae Tanks A & B DESCRIPTION The proposed change deletes the ball drip valves FP-211 A(B) and FP-212A(B) located on the Fire Water Storage Tanks A and B fill lines.
REASON FOR CHANGE
~
During implementation of work packages to replace Fire Water Storage Tank A and B
. fill lines and ball drip valves, it was determined that the potential existed to drain the tanks below the minimum level specified in the FSAR. Properly functioning ball drip valves could drain tanks to the level of the inlet nozzles.
SAFETY EVALUATION The proposed change does not impact equipment important-to-safety. The proposed change results in an increase in usable' volume in the fire water storage tanks. FSAR Figure 9.5.1-1 is revised to delete the ball drip valves. The change does not constitute an unreviewed safety question.
- 16. Cl-304506/WA-01150447. Installation of Vent on Charaina Pump Discharae Pipe DESCRIPTION The proposed change will install a vent upstream of relief valve CVC-192, at the high point, in order to provide adequate venting capability for the upstream piping. The vent will consist of a short (approximately 10 inches) of 1/2" tubing and an associated manual valve that will normally be closed, and a cap.
REASON FOR CHANGE CR-96-0893 identified that adequate vents are not available on the charging pumps discharge piping for removing air following routine maintenance on the pumps. This inability to effectively fill and vent the charging pump and its associcte piping could permit air accumulation in the piping with no method for removal.
SAFETY EVALUATION According to the safety evaluation, the proposed change will not increase the probability or consequences of an accident. The function of CVCS will not be changed and the vent valve will be normally closed and administratively controlled. The new equipment will be designed consistent with the existing system design to preclude any effect on equipment important to safety. No new system interconnections or new failure modes will be created. The margin of safety will not be reduced since the additional vents will not affect the capability of the CVCS to perform its safety function.
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- 17. Cl-305472. Repair of Fire Detector 27B-19 DESCRIPTION This package is to revise the CSGM Software to remove the preaction sprinkler actuation logic from Detector 278-19.
REASON FOR CHANGE Resolution for Condition Report CR-96-1131 - Smoke Detector 27B-19, located in the
+7 communication room is arranged to actuate suppression system FPM-27. . This room does not have suppression and the detector should not activate FPM-27.
SAFETY EVALUATION The safety evaluation states that this is a software change to eliminate any maloperation actions from undesired activation of preaction sprinklers. The detector ,
- only provides indication and does not actuate any sprinklers since there are none in the room. No new accidents or system interactions are created and no margin of safety is reduced.
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- 18. Cl-305473/WA-99100019. Terminate New Safety Related Cables for the Ultimate Heat Sink Tornado Missile Protection Project (Revision 0 and Revision 1)
QESCRIPTION Electrical circuits essential to safe shutdown after a design basis tornado on the safety Train 'B' east side c.,ooling area are being relocated to cable in raceways protected from tornado generated missiles. Revision 1 is an enhancement to the previous revision.
References to additional Technical Specifications were added to help facilitate the correct TS Action Statement entries. Two attachments referencing Licensing's position on the PMP event were also added.
REASON FOR CHANGE The conduits / cables that are not protected from missiles generated by a design basis tornado and that are required for safe shutdown following a design basis tornado will be re-routed to bring the plant into compliance with its licensing basis documents with respect to a design basis tornado.
SAFETY EVALUATION The safety evaluation considered Fuel Handling Accidents, Probable Maximum Precipitation, and any other accident which required the UHS. The chan0es do not affect the probability or consequences of any of these accidents nor do they alter the function of any equipment important-to-safety. The functions of circuits essential for -
mitigating the consequences of these accidents have not been affected and the associated equipment will be functionally tested for proper operation after changes have been implemented. No protective boundaries are affected and no margins of safety are reduced. Therefore, there are no unreviewed safety questions associated with these changes.
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- 19. Cl-305849/WA-01152287. Evaluation of Unprotected Openino Between Fire Areas RAB 2 and Roof E DESCRIPTION This is an evaluation of the acceptability of an opening between Fire Areas RAB-2 and Roof E.
REASON FOR CHANGE While performing a Quality Assurance audit, it was identified that there is no fire-related damper in the 36"x24" discharge opening separating Fire Area RAB-2 from Fire Area Roof E.
SAFETY EVALUATION According to the safety evaluation, the minimal amount of combustibles on either side of the opening do not increase the probability or consequences of an accident. As evaluated per GL 86-10, the unsealed penetration does not increase the consequences of a malfunction of equipment important-to-sefety, does not create the possibility of a new accident, and does not reduce the margin of safety, i
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- 20. Cl-306068/WA-99100019. Terminate New Safety Related Cables for the Ultimate Heat Sink Tornado Missile Protection Project (Revision 0 and Revision
.D DESCRIPTION Electrical circuits essential to safe shutdown after a design basis tornado on the safety Train 'B' east side cooling area are being relocated to cable in raceways protected from tornado generated missiles. NOTE: Revision 1 is an enhancement to the previous revision. References to additional Technical Specifications were added to help facilitate the correct TS Action Statement entries. Two attachments referencing Licensing's position on the PMP event were also added.
REASON FOR CHANGE The conduits / cables that are not protected from missiles generated by a design basis tornado and that are required for safe shutdown following a design basis tornado will be re-routed to bring the plant into compliance with its licensing basis documents with respect to a design basis tornado.
SAFETY EVALUATION The safety evaluation considered Fuel Handling Accidents, Probable Maximum Precipitation, and any other accident which required the UHS. The changes do not affact the probability or consequences of any of these accidents nor do they alter the function of any equipment importani-to-safety. The functions of circuits essential for mitigating the consequences of these accidents have not been affected and the associated equipment will be functionally tested for proper operation after changes have been implemented. No protective boundaries are affected and no margins of safety are reduced. Therefore, there are no unreviewed safety questions associated with these changes.
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- 21. Cl-306751/WA-01153561. LPSI Vent Additions at Containment Penetrations DESCRIPTION Nitrogen gas bubbles in the LPSI headers were identified by ultrasonic examination.
The examinations were conducted at the highest point of the outside containment LPSI piping near the penetration. Cl-306751 was initiated to add vent lines downstream of the outside containment isolation valves on the LPSI headers in order to provide adequate venting capability The vents will include a one inch line and associated manual valve that will normally be closed. 4 REASON FOR CHANGE The vents are required to provide the capability to remove nitrogen from the LPSI headers. A hydraulic transient could result if the LPSI pumps are started with the ,
nitrogen in the piping.
SAFETY EVALUATION The safety evaluation determined that no unreviewed safety questions exist as a result of the vent line additions and that the margin of safety as defined in the SAR and Technical Specification bases will not be reduced. The installation will meet the requirements of GDC 55 for Containment isolation Valves. The affected penetrations do not require type 'C' testing per FSAR 6.2-43 since the LPSI system does not
. constituta a potential containment atmosphere leak path following a LOCA.
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- 22. Cl-306764/WA-01153606. Addition of Pressure Eaualization Lines to SI-125A(B) and SI-412A(B)
DESCRIPTION Pressure equalization lines are being added to connect the leak off nipples on the bonnets of valves Sl-125A(B) and SI-412A(B) to the lines adjacent to the valves. This will prevent pressure locking of the valves if a hydraulic transient occurs.
REASON FOR CHANGE Hydraulic transients in the Shutdown Cooling Heat Exchanger lines could cause the inlet and outlet isolation valves (SI-125A(B) and SI-412A(B)] to become inoperable due to pressure locking of the valves.
SAFETY EVALUATION The proposed channes to SI-125A(B) and SI-412A(B) will not increase the probability of either a LOCA or v *c1SLB, The operability of the valves will be improved by precluding pressure e,&, .7 due to a hydraulic transient. Thus the likelihood of their malfunction will not be increased. The new connections meet all the design criteria required for the Si lines; therefore, the margin of safety is not reduced.
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- 23. WA-01142925. Main Steam Trao Drain Line Replacement DESCRIPTION This change replaces existing carbon steel pipe that shows signs of external corrosion with like-size stainless steel pipe on the Main Steam (MS) Trap Drain Line.
REASON FOR CHANGE Existing carbon steel pipe has leaked and shown signs of corrosion due to exposure to outside atmosphere.
SAFETY EVALUATION There are no accidents affected by the MS piping to be changed; therefore the probability or consequences of an accident are not increased. These drain lines do not interact with any important-to-safety equipment. No new system connections are created as this is replacement of the existing piping. No margin of safety is reduced and there is no unreviewed safety question created.
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- 24. WA-01144346. Removal of Vibration Trio Switch from Emeraency Diesel Generator 'A'
- DESCRIPTION -
The proposed change removes the vibration trip switch from EDG 'A'.
REASON FOR CHANGE The vibration trip switch function is to shut down the engine in the event of a catastrophic failure, it senses increasing velocity, or acceleration, on the diesel and provides a trip signal to the engine control panel in Test Mode only. The trip is -
bypassed in Emergency Mode. The trip is intended to mitigate component damage but has not reliably performed its intended design function. The Cooper-Bessemer Owner's Group has recommended utilities remove the svatch.
SAFETY EVALUATION There is no unreviewed safety question as a result of this change. The EDG is required to operate during a LOCA/ LOOP but does not initiate any accident. Spurious trips have occurred in Test Mode because of the unreliability of the vibration switch. ,
However, this protection is not enabled in Emergency Mode; therefore, no accident or consequences will be affected by removal of the switch. No new system connections or interactions are created by the change neither is there any effect on a protective boundary or margin of safety.
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- 25. WA-01144462. Boric Acid Makeup Thermal Relief Valve Removal DESCRIPTION This WA repair package removes the thermal relief valves installeri in the BAM and CVC system which were installed for thermal ovorpressure protection for the associated piping.
REASON FOR CHANGE A problem was identified with the relief valve installation. Back pressure on the discharge side of the valve allowed body to bonnet joint leakage causing contamination problems.
SAFETY EVAL.UATION The safety evaluation states that three accidents involve CVCS: boron dilution, increased inventory, and CVCS malfunction. None of these are affected by this change; therefore, ne radiological release is affected by the change. The valves were installed to prevent overpressurization that might be caused by the heat tracing installed to prevent boric acid precipitation. The heat trace was removed by SM-1904, so no overpressure condition sill be created and the valves are not needed. No important-to-safety equipment is affected and the margin of safety is not reduced by this change.
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- 26. WA-01146547. CC-963A Needle Valve Replacemen.1 DESCRIPTION This work package will replace the existing needle valve in the Instrument Air supply to valve CC-963A with a metering valve.
REASON FOR CHANGE in an effort to reduce pressure transients in the CCW system, the stroke times for CC-963A(B) were increased to approximately 30 seconds. This change will result in better control of the lA flow rate to the valve actuators for CC-963A(B).
SAFETY EVALUATION According to the safety evaluation. there are no accidents whose probability would be increased by this change. On a SIAS, CC-963A must remain closed bacause the CCW pump does not have the capacity to meet the additional flow demand of the Shu'down HX; Replacement of the needle valve will ensure there is sufficient air pressure to maintain the valve in the closed position. Once a CSAS is ganerated, CC-963A is required to open automatically to pass CCW flow through SDCHX 'A', This fail epen position ensures CC-963A is capable of performing this safety function regardless of the status of IA, The replacement valve will be procured as safety-related, seismic I to ensure the safety-related portion of the lA supply is not compromised. No new system interactions are created by the change and no margin of safety is reduced.
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27, WA-01148519. Reactor Drain Tank Outlet Check Valve BM-107. Internal Components Removal DESCRIPTION This repair package will remove the internal components of the Reactor Drain Tank (RDT) outlet check valve, BM-107.
REASON FOR CHANGE The corrective action program identified a potential to form a hydraulic lock which could cause containment isolation valve BM-109 to fall to close upon demand. This condition exists if BM-110 is closed prior to BM-109
- SAFETY EVALUATION According to the safety evaluation, BM-107 has no safety function, and does not serve a pressure class break or a containment isolation function. The change will improve the ability of the Boron Management (BM) system to isolate during a CIAS but will not increase the prcbability or consequences of any accident. Administrative controls are in place to prevent reverse flow to the EDT so there is no increase in the likelihood of a malfunction o' BM equipment. No new system interactions are created and no margin of safety is reduced by this change.
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- 28. WA-01148765 and WA-01148764. EDG Lube Oil Filter Internal Relief Valves Replacement with a Valve Port Closure Assembly DESCRIPTION '
The proposed change replaces the relief valves on the EDG lube oil filters with a seismically qualified Cooper Bessemer valve port closure assembly.
REASON FOR CHANGE During a review of EDG "B" trend data, it was noted that lube oil strainer differential pressure was showing an adverse trend. The lube oil filter internal relief valves are susceptible to leaking, allowing debris to bypass the filter and accumulate in the strainer. .
SAFETY EVALUATION <
Replacement with a valve port closure assembly does not alter the function or operation of the diesel generator lube oil system and does not create any new system
. interfaces.-- No unreviewed safety questions exist and the margin of safety will not be reduced because: 1) the new component is seismically qualified; 2) the change does not require the EDG lube oil system to be operated outside its design test limits; and 3) -
the new component is not part of the EDG lube oil filter pressure boundary.
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- 29. Cl-303303/WA-01149350. Modification of RCP Seal Heat Exchanaer Baffle Boltina (Revision 0 and Revision 1)
QESCRIPTION This work package documents acceptance of the thread damage to the baffle bolt holes
- in the RCP shaft. It also adds a Belleville spring washer between the locking cup and the bottom of the counter bore in the baffle for the Reactor Coolant Pump baffle bolts.
REASON FOR CHANGE Preload !s lost on the RCP baffle bolts mainly through thread embedment which is the plastic debrmation of localized high points on the bolt and/or threaded con'ponent. To reduce this loss, a Belleville spring washer has been added bot.veen the locking cup' and the bottom of the counterbore in the baffle. This will improve the bolted joint resistance to loosening by maintaining a preload on the baffle bolt.
SAFETY EVALUATION According to the safety evaluation, the four accidents initiated by a RCP malfunction are not increased by adding the Belleville spring washers. The loss of flow events are initiated by electrical power failures and are not affected by the spring washers nor is the seized / sheared shaft event. Loss of preload may allow the baffle to vibrate which -
could result in higher pump vibration and/or damage to the RCP shaft seal. This change helps maintain preload, thus there is no increase in likelihood of a malfunction.
There are no new system interactions or connections as a result of this change to create a new accident.
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- 30. WA-01155142. Plua Tubes in Spent Fuel Pool Heat Exchanaer DESCRIPTION The proposed change will revise the Spent Fuel Pool Cooling system thermal performance given in the FSAR as a result of Spent Fuel Pool Heat Exchanger (SPFHX) tube plugging.
REASON FOR CHANGE The Fuel Pool Cooling system removes decay heat generated in the spent fuel pool and maintains the spent fuel pool temperature below SRP requirements. Calculation 2 EC-S96-003 evaluated the normal, partir' core fuel offload plus eleven refueling batches, and maximum, full core offload p a ten refueling batches, fuel pool decay heat loads at various times after reactor shL down. Calculation MN(Q)-9-7, Change 2, establishes the fuel pool cooling system operating parameters to maintain the fuel pool temperature according to the SRP when 6 heat exchanger tubes are plugged in accordance with WA-01155142.
SAFETY EVALUATION According to the safety evaluation, there are no unreviewed safety questions and no margin of safety is reduced as a result of the proposed change. The Design Basis Fuel Handing Accident (Limiting Fault) was reviewed for possible adverse affects of the proposed change. No other design basis accidents are affected. The proposed change does not alter the existing operation or design function of the FP Cooling system. For possible cross-leakage between the CCWS and SFP system, CCWS is at a higher pressure so water would not leak into it during normal operations. If CCWS pressure were to fail and a leak were to occur from SFP cooling, radiation monitors in the CCWS would identify the leakage mitigating the consequences. Tube flow
. velocities will remain below the HX design and failure of tube plugs is bounded by the existing FSAR failure and effects analysis. System pressure boundary integrity will be maintained by installing the plugs under the ASME Section XI Repair program.
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- 31. WA-01155273. Reactor Closure Head Exhaust Manifold Bolt Ho!e Modification DESCRIPTION i he proposed change increases the bolt hole size for the reactor vessel closure head exhaust manifold (CEDM clamshell) ducts from 7/8" diameter to 1 1/2" diameter to provide improved installation capabilities.
REASON FOR CHANGE The present arrangement requires precise alignment between the clamshell duct boit holes (7/8") and the tapped holes (3/4") on the reactor vessel skirt assembly. The ductwork configuration and weight (approximmely 2000 pounds) tend to deform creating misalignment during installation.
SAFETY EVALUATION The proposed bolt hole enlargement will be negated by the use of washers around the bolt heads. This change has been evaluated and designed to maintain the non-safety related seismic qualifications presently identified in the FSAR, No accidents or important-to-safety equipment are affected by this change and no unreviewed safety question is created.
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- 32. WA-01156373. Installation of Ventfrest Connection on Gfand Steam Condenser Exhaust DESCRIPTION-This repair package will install a vent / test connection on the Gland Steam Condenser (GSC) Exhaust in order to connect instrumentation and obtain system operating information.
REASON FOR CHANGE Engineering could not determine the cause of condensate problems with the GSC exhausters because there is not enough information to evaluate the systems. There is no instrumentation available for indication of flow, pressure, temperature, etc.
SAFETY EVALUATION The proposed change will not increase the probability of a loss of condenser vacuum, The vent / test connection will be designe.d and installed to the same standards as the existing piping. The Air Evacuation (AE) system will continue to perform its intended function; therefore, there will be no increase in the consequences of a steam generator tube rupture. No new system interfaces are created, no important-to-safety equipment will be affected, no protective boundaries are affected, and no margin of safety is reduced as a result of this change. 9 76
- 33. WA-01157203. Rewire Ground Detection Relay in Emeraer,3/ Diesel Generator DESCRIPTION The proposed change disconnects the 64F, field ground detection relay, from the generator field circuit while field flashing is in progress. This will be accomplished by connecting a normally open contact from ths 52C1 starting sequenco rolay in series with the 64F relay. The 52C1 is designed to actuate when the EDG achieves approximately 78% rated ouput voltage and 570 rpm (when field flashing is no lor @r needed).
REASON FOR CHANGE During EDG field flashing, the station battery system, which has its own ground detection system separate from the EDG field ground detection system, is connected to the generator field. As a result of the EDG field protecUun relay being connected to ground, the corresponding DC bus is experiencing a ground during field flashing.
SAFETY EVALUATION The proposed change does not add or delete any components exclusive of an electrical cable used in the re-wiring. There is no potential introduced by this change that will modify the operation or expectations of the EDG or the field flashing circuit as described by design basis document W3-DBD-002. No new changes to procedures or testing are introduced. The new contact does introduce a potential failure point -
should the contact fail open, the field ground detection relay will be isolated from the circuit and, consequently, there will be no field ground annunciation should a fault occur on the generator field. However, this change removes the single negative to ground fault that normally exists during field flashing; therefore, overall, this is an improvement to the present configuration. The proposed change has no impact on the operation of the EDG or the ability to start the generator field flashing and the ability to complete a successful safe shutdown is not % graded.
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i
- 34. WA-01157227. Rewire Ground Detection Relav in Emeroency Diesel Generator 1
DESCRIPTION The proposed change disconnects the 64F, field ground detection relay, from the generator field circuit while field flashing is in progress. This will be accomplished by connecting a normally open contact from the 52C1 starting sequence relay in series with the 64F relay. The 52C1 is designed to actuate when the EDG achieves approximately 78% rated ouput voltage and 570 rpm (when field flashing is no longer needed),
REASON FOR CHANGE During EDG field flashing, the station battery system, which has its own ground detection system separate from the EDG field ground detection system, is connected to the generator field. As a result of the EDG field protection reby being connected to ground, the corresponding DC bus is experiencing a ground during field flashing.
SAFETY EVALUATION The proposed change does not add or delete any components exclusive of an electrical cable used in the re wiring. There is no potential introduced by this change that will modify the operatior or expectations of the EDG or the field flashing circuit as described by design basis decument W3-DBD-002. No new changes to procedures or testing are introduced. The new contact does introduce a potential failure point -
- should the contact fail open, tha field ground detection relay will be isolated from the l circuit and, consequently, there will be no field ground annunciation should a fault occur on the generator field. However, this change removes the single negative to ground fault that normally exists during field flashing; therefore, overall, this is an I improvement to the present configuration. The proposed change has no impact on the l operation of the EDG or the ability to start the generator field flashing and the ability to complete e successful safe shutdown is not degraded.
l 78 l
l
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- 35. WA-01158027; HPSI Vent Additions at Penetrations 55. 56. 57. & 58 DESCRIPTION The proposed change will int, tall new ASME Class 2, Seismic 1, vent assemblies to HPSI piping outside containment penetrations 55,56,57, & 58.
REASON FOR CHANGE The vents are required to provide the capability to remove gas from the HPSI headers.
A hydraulle transient could result if the HPSI pumps are started with gas in the piping.
SAFETY EVAL.UATION The safety evaluation has determined no unreviewed safety questions exist as a result of the proposed change and the mergin of safety defined in the FSAR and Technical Specification Bases will not be reduced. The function of the Safety injection system will not be affected since the vent lines have a passive function of maintaining the HPSI pressum boundary. The materials to be used will be ASME Class 2, Seismic 1.
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- 36. WA-01158036. Shield Buildina Maintenance Hatch Seal System Reaulator DESCRIPTION This package will allow an existing regulator and drain to remain in place.
REASON FOR CHANGE A regulator and drain valve were found between Station Air valve SA-7021 and the Maintenance Hatch Seal control panel, The regulator and valve were not tagged or shown on any drawings.
SAFETY EVALUATION The critical characteristic of the regulator is the pressure setting. The regulator is currently fully open, not decreasing the supply pressure, and therefore not adversely affecting any equipment. The seal panel is designed to receive Station Air pressure and has inte.nal regulators which decrease the incoming pressure. The proposed function of the drain valve is to expel any condensation from the Station Air system before it reaches the regulator. The existing drain valve is capable of this function.
These components will not adversely affect the maintenance hatch seal and do not create an unreviewed safety question.
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_ _ _ _ = _ _ _ _ _ _ _ _ _ _ _
- 37. WA 01158198. EH Fuller Earth and Contaminant Filters - Slohon Breaker Addition DESCRIPTION This change adds a siphon breaker inside the Electro Hydraulic (EH) reservoir discharge line in order to prevent siphoning of systern fluid when the filters are being replaced during maintenance. The siphon breaker will consist of one 1/8" diameter through hole in the horizontal portion of the discharge line.
REASON FOR CHANGR Inlet isolation valves are provided to the Fuller Earth and Contaminant filters but no outlet isolation valves exist. The filters discharge directly into the EH reservoir at atmospheric pressure. The discharge line inlet is below the fluid level to preclude aeration of the fluid. However, the system is siphoning back through the discharge line whenever the reservoir fluid level is above the discharge line inlet into the reservoir.
SAFETY EVALUATION The safety evaluation determined that no unreviewed safety questions exist as a result of this change. The change will not affect operation of the main turbine and will not increase the probability or consequences of any accident. No new system interactions are created and no important to-safety equipment is affected. Maintenance capability of the EH system will be improved with no adverse affect on margin of safety or accident response.
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l C. TEMPORARY ALTERATION REQUEST (TAR)
- 1. TAR 96-003. Mux Site RA4602/ Annunciation Temporary Disconnection DESCRIPTION The TAR will temporarily disconnect digital inputr, to the Safety Parameter Display System (SPDS) associated with the Plant Monitoring Computer (PMC) Mux Site RA4602. It will also ccuse the loss of Sequence of Events (SOE) points and disconnect the remaining annunciator inputs to the mux site.
REASON FOR CHANGE DC 3374 (ltem 17 of 1996 report) installed optical isolation devices to separate the annunciator circuitry from the PMC multiplexer interface to res, !ve a common ground problem that existed with the annunciator ground detector circuliry. Based upon four failures experienced to date with the optical isolators in mux site RA4602 Design Engineering recommended disconnecting the mux site from the Annunciator System.
This will ensure operation of the Annunciatcr System.
SAFETY EVALUATION According to the safety evaluation, the TAR will not affect any accidents previously analyzed, nor will it create any new accidents. The information used for SPDS and SOE is used as an Operctor/ Emergency Plan personnei aid. The SPDS points are ESFAS status points which are available from the Plant Protections System (PPS) indicators. The SOE points are available from the long term archival file of the PMC.
The PMC is not considered available during or after an accident. The points do not provide any control functions and will not affect equipment important-to-safety.
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- 2. TAR-96-004. Goa CC-958A(B) and CC 190A(B) Closed DESCRIPTION The Shutdown Heat Exchanger shell side thermal relief valves, CC 958A(B), and the CCW Heat Exchanger tube side thermal relief valves, CC-190A(B), will be gagged closed.
BEASON FOR CHANGE i
P w wives do not resent after lifting during system surveillances because system opma%9 pressure is too close to system design pressure.
SAFETY EVALUATION l According to the safety evaluation, gagging CC 958A(B) and CC-190A(B) will not affect the overall system performance or reliability in a way which could lead to an accident occurring. The relief valves are required only as thermal reliefs in the event the shell side of the Shutdown Heat Exchangers or the tube side of the CCW Heat Exchangers are isolated. During operation, these heat exchangers are not isolated and they do not receive an automatic isolation signal during a Design Basis Accident.
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- 3. TAR-96-009. Compliance with TS 3.7.6.5 due to Installation of Blank Off Plate on Control Room Normal Outside Air intake Duct (Revision 0 and Revision 1)
DESCRIPTION The proposed change will block the outside air intake to HVC-101 and HVC-102, which are the normal outside air intake valves for the control room HVAC. This will be done by putting a blind flange upstream of fire damper FD 36 in the HVAC shaft (OAl plenum).
REASON FOR CHANGE Testing performed on HVC-101 and HVC-102 identified that both valves are leaking.
Leakage past those valves, which are in series, could result in unfiltered air entering the control room during the recirculation mode of operation.
SAFETY EVAL llATION According to the safety evaluation, the accidents which could be potentially affected are those requiring isolation of the control room HVAC. During normal operation, HVC-101 and HVC-102 provide 2200 cfm of outside air. The loss of outside air will not affect safe operation of the plant. These valves close during an accident and are not required to re-open. Since this change blanks off outside air, the intent of the valves' safety function will be maintained under the proposed change. The change will minimize the consequences of a radiological release on the control room since no active function will be required to iso l ate the control room during an accident. Since the seismically supported blank off plate is passive, there are no new methods of failure.
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- 4. TAR-96-012. Gao Closed Valves CC-807A and CC-823A DESCRIPTION The proposed temporary alteration will gag valves CC-807A (CFC 'C' inlet valve) and CC-823A (CFC 'C' outlet valve) closed.
REASON FOR CHANGE Gag valves closed in order to perform maintenance on the valve operators, which were discovered leaking.
SAFETY EVALUATION According to the safety evaluation, no unreviewed safety question exists for the proposed change: it does not impact the intended functions of the CCW or CCS systems; valves CC-807A and CC-823A will be gagged closed while maintenance is performed to ensure containment integrity is maintained; thermal relief for CFC 'C' ensures components and struc' 'es between the inlet and outlet isolation valves to CFC 'C' will not exceed their d< gn limits; CCW piping and valves maximum internal pressures will not excoed allowable stress limits; containment design limits are not exceeded post-accident assuming one CFC is available for containment heat removal; the installed gags do not impact the seismic qualification of the valves; and, short term operation with increased CCW flow to CFC 'A' will have no affect on the cooling coils.
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- 5. TAR-96-014. RWSP Low Level Alarm DESCRIPTION The proposed change raises the low level alarm of the RWSP to 17'5" and the reset to 17'8". It also revises procedure OP 901-522 to add a step to isolate the RWSP purification line by closing the manual isolation valve FS-423 within 30 minutes of the onset of a design basis earthquake.
REASON FOR CHANGE A portion of the RWSP purification line is not seismically analyzed. During a design basis earthquake, the line could potentially break and divert water from the suctions of the ECCS pumps resulting in an insufficient volume of water. This RWSP line is not automatically isolated in a DBE event.
, SAFETY EVALUATION According to the safety evaluation, the RWSP will lose 7,200 gallons in inventory based on leakage from a pipe break downstream of the RWSP purification pump. This is based on the maximum pump capacity of 240 gpm running for 30 minutes. The procedure revision ensures operator action is taken within this 30 minutes. The TS required minimum borated water volume of 475,500 gallons available for the RWSP to maintain its safety function is maintained by the change to the RWSP setpoint. Raising the low level setpoint compensates for the 7,200 gallon inventory loss in the event of RWSP purification line break during an earthquake. The proposed TAR and procedure revision do not affect the probability or consequences of an accident and do not adversely affect any important-to-safety equipment. Neither is any margin of safety reduced.
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- 6. TAR 96-015. Determina AH 25 Inlet Damoer Operators. SVS-103A(B)
DESCRIPTION This TAR will fall dampers SVS-103A(B) to their safe open position by determing the valve operator. '
REASON FOR CHANGE This change will ensure the inlet dampers remain in their fall safe position SAFETY EVALUATION According to the safety evaluation, the inlet dampers are required to open to allow sufficient cooling air flow to various Switchgear and Cable Vault areas. Failing the dampers open will ensure these components are capable of performing their safety relatod function during normal and accident conditions. This proposed change will reduce the likelihood that the dampers would not open as required during an accident.-
In addition, the gravity damper on the standby unit will be verified closed whenever a unit is secured to preclude the possibility of short circuiting air through the standby unit.
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- 7. TAR-96-010. Steam Generator Feed Pumo Lube Oil Temporary Purifier DESCRIPTION The proposed change installs a temporary lube oil purifier for the Steam Generator Feed Pumps.
REASON FOR CHANGE The Lube oil for the feed pumps shows an increasing trend in water content. The temporary purifier will reduce the water content in the lube oil.
' SAFETY EVALUATION Installation and operation of this lube oil purifier will not impact the ability nf the 1 Feedwater pumps to supply water to the Steam Generators. The probability and consequences of a loss of feedwater accident are not affected by this change. The parameters of the temporary purifier are similar or more conservative than those of the permanent centrifuge; therefore, installation of the purifier skid will not increase the probability of loss of feedwater due to tripping the feed pump from loss of lube oil.
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- 8. TAR-97-002. Determinate AH-25 Inlet Damper Operators DESCRIPTION The purpose of this temporary alteration is to fall damper SVS-103B to its safe open position by determinating the valve operator.
REASON FOR CHANGE This change is being implemented to ensure the inlet damper remains in its fall safe position.
- SAFETY EVALUATION According to the safety evaluation, the inlet damper is required to open to allow sufficient cooling air flow to various switchgear and cable vault areas. Falling the damper open will provide assurance that the component is capable of performing its safety-related function during normal and accident conditions. Furthermore, the gravity dampers provided on the discharge of each unit will prevent reverse air flow through the standby unit. The TAR contains explicit instructions to verify the gravity damper is closed whenever the unit is secure. This will preclude the possibility of short circuiting air through the standby unit. No protective boundary is affected by this . change. ,
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- 9. TAR 97-004. Addition of Portable lon Exchance Vessels to the ACCW Filtration Skids DESCRIPTION The proposed change involves placement of portable ion exchange units on the ACCW
'A' and ACCW 'B' filtration / chemical addition skids.
REASON FOR CHANGE Zine must be removed to less than 1.0 ppm prior to discharge of basins to the Circulating Water system, SAFETY EVAL.UATION The safety evaluation concludes that the addition of temporary ? -a exchange vessels to each ACCW basin will not reduce the level of performance of the ACCW system. The ACCW system will be able to perfoim its design function as described in the FSAR, during normal operation, and during design basis accidents No unreviewed safety question is created.
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- 10. TAR 97-005. Disable Condenser Air Evacuation System Exhaust to RAB Normaj Ventilation System DESCRIPTION The proposed change disables the diversion of the Condenser Vacuum Pump Exhaust Header to the RAB normal ventilation duct which provides a filtered release through the plant stack and eliminates a plant protection system instrument problem resulting from a positive pressure condition in the RWSP during an accident. The OP 901-202 deviation is to remove immediate operator action to go into Main Condenser.
Evacuation System (MCES) divert. In addition, it adds into the subsequent operator action section, steps so that aft- shutdown starts, operators manually divert the off-gas to the plant stack.
REASON FOR CHANGE The licensed design of the Recirculation Actuation Signal (RAS) involved in LOCAs does not work correctly under some circumstances and the reference leg of the four RWSP level instruments does not account for various HVAC lineups and condenser air evacuation flow paths. The proposed change prevents those circumstances and ensures RAS functions as designed.
SAFETY EVALUATION Rerouting condenser off-gas so that it cannot go to the plent stack for the remainder of Cycle 8 neither causes events previously analyzed to exceed off-site dose limits, nor causes the amount of activity released to exceed the amount giver, by the NRC as acceptable. Therefore, there is no unreviewed safety question associated with the proposed change.
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- 11. TAR 97-007. Containment Sumo Pumo 'A' Bvoass DESCRIPTION This temporary alteration will install a temporary sump and pump beneath the Containment sump weir box to collect Containment sump influent and route it, via the normel system flowpath, to the Waste Tanks. Should influent exceed the capacity of-the tempora:y pump, the 'B' Containment sump pump can be used.
REASON FOR CHANG 4 Use of this temporary alteration will facilitate calibration of the sump level instruments.
4 SAFETY EVALUATION There are no accidents or equipment important to safety listed in the FSAR affected by this change and the TAR will only be in place during Modes 5 and 6. Therefore, no accident or equipment malfunction probabilities or consequences are increased as a result of the change, The temporary power to the pump has adequate protective devices to preclude any effect on equipment important-to-safety. In addition, no margin i of safety is reduced due to this change.
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- 12. TAR 97-008. Blank Plate Installation at Orifice Sl lFl 0311 in Support of DC-3440 DESCRIPTION The proposed change consists of the temporary replacement of flow measuring orifice SI lF10311 with a blank plate. The orifice normally provides means to obtsin the flow rate for the HPSI 1 A flow loop.
REASON FOR CHANGE The blank plate will serve to isolate the HPSI 1 A cold leg injection line and allow drainage of the upstream piping surrounding valves SI 225A(B). Once isolated, these valves will be removed and replaced under DC 3440.
SAFETY EVALUATION According to the safety evaluation, the HPSI system is used primarily for post-accident mitigation and is not a credible initiator of any accident. During the installation of the blank plate, the requisite number of shutdown cooling and boration flowpaths will be maintained without the need for the HPSI 1 A injection flowpath to be in service.
Therefore, none of the consequences of accidents listed in the FSAR will be affected.
No other equipment important-to-safety will be affected and RCS temperature will be below 200 deg F so one injection system is acceptable without single failure consideration on the basis of stable reactivity condition and restrictions prohibiting cora alterations and positive reactivity changes. No new system interactions will result from the proposed change and no new accidents or malfunctions of equipment will be created. No margin of safety defined in the TS Bases will be reduced and there are no unreviewed safety questions associated with this change.
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3440 (Revision 0 and Revision 1) l DESCRIPTION 1
The proposed change consists of the temporary replacement of flow measuring orifice SI IFl-0321 with a blank plate. The orifice normally provides means to obtain the flow rate for the HPSI 1B flow loop.
REASON FOR CHANGE The blank plate will serve to isolate the HPSI 18 cold leg injection line and allow drainage of the upstream piping surrounding valves SI-226A(B). Once isolated, these valves will be removed and replaced under DC-3440.
SAFETY EVALUATION According to the safety evaluation, the HPSI system is used primarity for post accident mitigation and is not a credible initiator of any accident. During the installation of the blank plate, the requisite number of shutdown cooling and boration flowpaths will be maintained without the need for the HPSI 1B injection flowpath to be in service.
Operability of two trains of SDC will be maintained throughout the activity. With one flowpath of a SDC loop isolated, total SDC flow for the loop will be through the remaining flowpath. This will be controlled administratively by the applicable reduced inventory procedure. Valve positions listed in this procedure have been established by testing to prevent vortexing or pump run out. Therefore, none of the consequences of accidents listed in the FSAR will be affected. No other equipment important to safety will be affected and RCS temperature will be below 200 deg. F so one injection system is acceptable without single failure consideration on the basis of stable reactivity condition and restrictions prohibiting core alterations and positive reactivity changes.
No new system interactions will result from the proposed change and no new accidents or malfunctions of equipment will be created. No margin of safety defined in the TS Bases will be reduced and there are no unreviewed safety questions associated with this change.
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- 14. TAR-97-010. Blank Plate Installation at Orifice SI IFI-0331 in Support of DC-3440 (Revision 0 and Revision 1)
DESCRIPTION The proposed change consists of the temporary replacement of flow measuring orifice SI IFl.0331 with a blank plate. The orifice normally provides means to obtain the flow rate for the HPSI 2A flow loop.
REASON FOR CHANGE The blank plate will serve to isolate the HPSI 2A cold leg injection line and allow drainage of the upstream piping surrounding valves SI 227A(B). Once isolated, these valves will be removed and replaced under DC 3440.
SAFETY EVALUATION According to the safety evaluation, the HPSI system is used primarily for post-accident mitigation and is not a credible initiator of any accident. During the installation of the blank plate, the requisite number of shutdown cooling and boration flowpaths will be maintained without the need for the HPSI 2A injection flowpath to be in service.
Therefore, none of the consequences of accidents listed in the FSAR will be affected.
No other equipment important-to-safety will be affected and RCS temperature will be below 200 deg. F so one injection system is acceptable without single failure consideration on the basis of stable reactivity condition and restrictions prohibiting core alterations and positive reactivity changes. No new system interactions will result from the proposed change and no new accidents or malfunctions of equipment will be created. No margin of safety defined in the TS Bases will be reduced and there are no unseviewed safety questions associated with this change.
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- 15. TAR 97-011. Blank Plate Installation at Orifice SI IFl-0341 in Support of DC-3440 DESCRIPTION The proposed change consists of the temporary replacement of flow measuring orifice
- SI lFl.0341 with a blank plate. The orifice normally provides means to obtain the flow rate for the HPSI 28 flow loop.
REASON FOR CHANGE-The blank plate will serve to isolate the HPSI 2B cold leg injection line and allow drainage of the upstream piping surrounding valves SI 228A(B), Once isolated, these valves will be removed and replaced under DC-3440.
SAFETY EVALUATION According to the safety evaluation, the HPSI system is used primarily for post-accident mitigation and is not a credible initiator of any accident. During the installation of the blank plate, the requisite number of shutdown cooling and boration flowpaths will be maintained without the need for the HPSI 2B injection flowpath to be in service.
Therefore, none of the consequences of accidents listed in the FSAR will be affected.
No other equipment important to safety will be affected and RCS temperature will be below 200 deg. F so one injection system is acceptable without single failure consideration on the basis of stable reactivity condition and restrictions prohibiting core alterations and positive reactivity changes. No new system interactions will result from the proposed change and no new accidents or malfunctions of equipment will be created. No margin of safety defined in the TS Bases will be reduced and there are no unreviewed safety questions associated with this change.
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D. DOCUMENT REVISION NOTICES
- 1. DRN-C-9600048. Resolution of W3-DBD-027 Open Item 1.
3.4 DESCRIPTION
-The DRN provides the resolution of the Open item associated with Design Basis Document (DBD)-027 concerning the weight of major equipment I
REASON FOR CHANGE The DRN will achieve consistency between the DBD, FSAR, and Design Specification SW12 DI.001(O)in documerG g the weight of major equipment (Reactor vessel, Steam Generator, Reactor Coolant Pump and the Pressurizer).
SAFEW EVALUATION According to the safety evaluation, the DRN revises the weight of the Reactor Coolant Pump in FSAR Table 3.8-20_to coincide with the weight shown on the vendor drawing, determined to contain the more accurate weight. This DRN does not increase the probability of occurrence of an accident previously evaluated in the FSAR. The DRN does not reduce the margin of safety as defined in the bases for any technical specification or appropriate safety analysis.
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- 2. DRN E 9600940. Paolna System Wirino DESCRIPTION This change allows paging system wiring to be run in electric metallic tubing (EMT) from the Service Building Warehouse to the extended portion of the Service Building. It also allows the wiring, once entering the extended portion of the Service Building extension, to be run in free air above the suspended ceiling.
REASON FOR CHANGE Provide paging capability to new portion of Service Building.
SAFETY EVALUATION l
This change allows wiring for the non safety paging system to be run in electric metallic <
tubing (EMT) from the Service Building Warehoust 'o the extended portion of the Service Building. It also allows the wiring, once er, ering the extended portion of the Service Building extension, to be run in free air above the suspended ceiling. No accidents or equipment important to-safety are affected by the paging communication system and no margin of safety is reduced.
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- 3. DRN's M 9501315. M 9501325. M 9501336. M 9501337. M-9501338. and M-9501339. Revise Instrument Air Flow Disarams DESCRIPTION The Instrument Air flow diagrams G-152, Sheets 1 through 6, are being revised to reflect as-built conditions. The as built changes entall valve position and tag number revisions which are for reference only and are not part of the plant design basis. The drawings are also revised to reflect removal of temporary Instrument Air connections, as directed by drawing notes.
REASON FOR CHANGE This is an administrative change only as the drawings are being changed to reflect the current plant as built condition. No physical changes are being made to the plant.
SAFF,TY EVAL.UATION According to the safety evaluation, the change is purely administrative. No physical change to the plant is proposed. Therefore, the change will have no impact on any accident or important to-safety equipm6nt identified in the FSAR. The change does not create any new accident and does not reduce the margin of safety.
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- 4. DRN M-9701151 Lower Emeroency Feedwater Temperature from 70 Dea. F to 40 Dec. F DESCRIPTION The proposed change will allow for cold water feed to the sicam generators (SG) at no less than 40 degrees F and to raise the number of cycles from 8 to 20.
REASON FOR CHANGE The change is being made to comply with the overall design of the plant which allows for feeding of the SGs from alternate water sources. The number of cycles is being changed to provide more margin to the life of the SGs. Under various operating and emergency conditions, water from the WCT basins or from the Condensate Storage Tank (either of which may be as low as 40 degrees F) may be injected into the SGs.
SAFETY EVALUATION Based on analysis by the SG vendor, this change will have no impact on the structural integrity of the SGs or the safety analyses. There are no accidents in the FSAR affected by this change since there are no accidents in which the EFW system is initially aligned to accept water from the WCT basins or from the auxiliary feedwater system. By analysis of the temperature change and the number of cycles, there is no adverse effect on the SGs and no other important-to-safety equipment is affected by this change. No new system interconnections are created by this change and there is no reduction in a margin of safety.
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E. LICENSING DOCUMENT CHANGE REQUESTS (LDCR)
- 1. LDCR 95-0078. Revises FSAR Section 8
3.1 DESCRIPTION
The proposed change revises tables and figures in the FSAR required to conform with the revised loading of the Emergency Diesel Generators as indicated in calculation EC-E90-006, revision 2.
REASON FOR CHANGE The EDG's are required to power Engineered Safety Features (ESF) loads during a Loss of Offsite Power (LOOP). The change is the amount of loads being supplied by the EDG's. The new loadings are still within the ratings of the EDG's.
SAFETY EVALUATION According to the safety evaluation, the function, ratings, and capacities of the EDG's have been previously evaluated and reviewed. The net changes (combination of all load revisions) are decreases in the loadings of the EDG's. The probability or consequences of an accident are not increased by this change. Since the EDG's total load remains less than their rating, no important-to-safety equipment is affected by the change. The margin of safety is not reduced since there is no change to a protective boundary by this revision.
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- 2. LDCR 95-0117. FSAR Chanaes to Fuel Pool Coolina System DESCRIPTION The proposed change will revise the fuel pool cooling heat loads and fuel pool cooling system thermal performance given in the FSAR.
REASON FOR CHANGE Calculation EC-S96-003 evaluated the normal, partial core fuel offload plus eleven refueling batches, and maximum full core offload plus ten refueling batches and fuel pool decay heat loads at various times after reactor shutdown. Calculation MN(Q)-9-7 establishes the fuel pool cooling system operating parameters to maintain the fuel pool temperature in accordance with the Standard Review Plan SAFETY EVALUATION According to the safety evaluation, the proposed change does not impact any spent fuel or fuel handlirig accident and does not alter any equipment. Therefore, no accidents or impor' ant to-safety equipment are affected. Revision of the heat loads and system thenrel performance does not require any new system interactions, therefore no new accident possibilities are created. The proposed change maintains maximum fuel pool temperature below 140 degrees F, therefore no margin of safety is reduced.
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- 3. LDCMG-0151. Exception for CIAS Channel Assionment to HRA Valves DESCRIPTION This LDCR adds an exception to FSAR 6.2.4.3 2 to identify that Containment isolation Actuation Signal (CIAS) Channel A is exclusively used in Hydrogen Analyzer (HRA)
Train A containment isolation valve control circuits. The same applies to ClAS Channel B.
REASON FOR CHANGE The current FSAR 6.2.4.3.2 states that "wherever two automatic isolation valves are in series, each valve operator is actuated from a separate and redundant CIAS channel, A or B, ..." Waterford 3 design is such that CIAS Channel A is exclusively used in HRA Train A containment isolation valve control circuits. Similarly, CIAS Channel B is exclusively used in HRA Train B containment isolation valve circuits.
SAFETY EVALUATION HRA is a measuring system that samples gas from various parts of containment. it is not possible for HRA to initiate a reactor trip, turbine trip, or loss of off site power. It is cet part of the primary system pressure boundary.
Accident analyses assume containment is isolated and leakage is no more than an assumed value. During performance of surveillance procedures OP-903-120 and OP-903-094, one or more of six HRA valves are open. While open, a single failure of a CIAS relay contact to open can prevent automatic closure of the open HRA valves.
However, during each procedure an operator is in active control of the valve position.-
The operator presence and the procedure itself constitute administrative control. Thus, as specified under Technical Specification (TS) 3.6.3 the lack of an automatic closure signal does not in . Nase the likelihood that the penetration is open during the early stages of an accident. In addition, HRA is one of the systems in the TS 6.8.4 program to ensure leak-tightness of equipment connected to contsinment. Therefore by design and testing, gas samples going through HRA do not leak into the Reactor Auxiliary Building. Operators are directed by emergency operating procedures to put HRA on-line as committed to in FSAR 6.2.5.3 in accordance with Regulatory Guide 1.7, 103
- 4. LDCR-96-0161. FSAR Chances on Ultimate Heat Sink DESCRIPTION The proposed change revises the Ultimate Heat Sink (UHS) design basis based on analyses from calcellations EC M95-008, MN(Q)-9-9, MN(Q)-9-10, and MN(Q)-9-17.
REASON FOR CHANGE As a result of testing on the CCW Heat Exchanger, the UHS was evaluated in order to reduce the design heat duty on the CCW Heat Exchanger.
SAFETY EVALUATION According to the safety evaluation, the proposed change revises certala UHS design parameters but does not alter the operation or function of UHS. Thus no accidents are affected. The calculations upon which the change is based demonstrate tnat there wil!
be no incrcase in the probability or consequences nf a malfunction of equipment important-to-safety. There are no new system interconnections and UHS function is not altered; therefore, there will be no new accidents or equipment malfunctions created by this change, The analyrzie also shows that 110 Dog, F cooling water to the -
Essential Chiller will not impact the, Essential Chiller or its safety function; therefore, the margin of safety is not reduced.
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- 5. LDCR-96-0170. Revises FSAR Table 6.2-32. Sheets 75 & 76 DESCRIPTION Table 6.2-32 is revised to reflect the correct classification of the Hydrogen Analyzer (HRA) Containment isolation valves.
REASON FOR CHANGE Table 6.2-32 of the FSAR currently shows the valves as classification A2, the correct classification of the valves is Clacs D.
SAFETY EVALUATION The safety evaluation states that HRA is a measuring system that samples gas from various parts of primary containment.11 is not possible for HRA to initiate a reactor trip, turbine trip, or loss of off-site power, it is not part of the primary system pressure boundary. The Hydrogen Analyzers are closed systems and are located outside of containment. All tubing is Seismic Class I and Safety Related. The analyzer skids are qualified for containment accident pressure. The isolation va%s are normally locked closed and receive a CIAS (See LDCR 90-0151, Exceotion k, CIAS Channal Assignment to HRA Valves, Item 4 of this report.) if isolation is required during system calibration and the valves are open.
Accident analyses assume that containment leakage is below a particular level. The leak-tight design of the whole HRA process stree.n means that a failure to automatically isolate any or all of the HRA containment isolation valves does not release post-accident contaminants into the Reactor Auxiliary Building. During performance of surveillance procedures one or more of six HRA valves are open. While open, a single failure of a CIAS relay contact to open can prevent automatic closure of the open HRA valves. However, during each procedura an operator is in active control of the valve position. HRA is one of the systems in the Technical Specification 6.8.4 program to ensure leak tightness of equipment connected to containment.
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- 6. LDCR-96-0171. Revises FSAR Section 13.7 l QESCRIPTION The LDCR incorporates information into the FSAR which describes the attributes associated with the Technical Requirements Manual (TRM) and includes the TRM as an extension of the FSAR.
REASON FOR CHANGE This is an administrative change only and has no impact on plant equipment or procedures. In accordance with the Technical Specification improvements Program and NRC criteria, specific requirements in the Technical Specifications have been relocated to a controlled document - the Technical Requirements Manual.
SAFETY EVALUATION According to the safety evaluation the change is purely administrative. No requirements associated with systems, structures, or components will be changed.
Therefore, the change will have no impact on any accident or anticipated operational
- occurrence identified in the FSAR.
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- 7. LDCR-97-0001. Revises F.*j.f Ciaures 9.5.1-16 and 9.5.1-20 DESCRIPTION The LDCR revises the FSAR Figures to downgrade the exterior east wall of the Turbine Generator Building (TGB) Switchgear Room. It resu :s in the downgrading of two fire doors (D62, D190) and no penetration seats.
REASON FOR CHANGE The barrier is currently rated and maintained at two hours as an insurance barrier.
Exterior plant walls are neither rated nor maintained as fire walls unless there exist in-situ hazards as determined by a fire protection engineer.
SAFETY EVALUATION According to the safety evaluation, the TGB Switchgear Room contains non-safety plant equipment. As a result, TGB fires are not analyzed in the FSAR Fire Hazards Analysis. No in-situ fire or explosion hazards have been identified in the vicinity of the east exterio, wall of the TGB Switchgear Room. The usn of ignition sources and the staging of ccmbustibles will continue to require evaluation by the Fire Protection Permit process.
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- 8. LDCR-97-0005. Revises FSAR Flaure 1;2-19 DESCRIPTION The proposed change is a revision to FSAR figure 1,2-19 to delete the ultrasonic tank, ultrasonic generator, and rinse tanks.
-REASON FOR CHANGE -
This equipment was removed as identified in condition report CR-94-0823.
SAFETY EVALUATION The proposed change only revises a figure in the FSAR to reflect the as-built condition of the plantc It does not affect any accident, equipment important-to-safety, or reduce the margin of safety.-
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- 9. LDCR-97-0015. Re"ises FSAR Section 0.4.
4.2 DESCRIPTION
The proposed change revises the assumed number of people in the Control Room during a toxic gas accident given in FSAR Section 6.4.4.2, paragraph e, from 7 to 16.
REASON FOR CHANGE
. Calculation EC-M5002 provides a basis for carbon dioxide production and oxygen depletion assuming 16 people in the control room during a toxic gas accident, which better reflects conditions that may exist in the control room.
SAFETY EVALUATION According to the safety evaluation, the proposed change only revises information in the FSAR to agree with design basis calculations. The control room HVAC system and important-to-safety equipment are not affected by the change. Therefore, there is no increased probability or consequence of an accident or squipment malfunction. Since -
the change reflects the design basis calculation, no margin of safety is reduced.
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- 10. LDCR 97-0023. Maintenance Reautrements: Rotatina Eauipment in Storaae l DESCRIPTION The proposed cha ige revises the FSAR and QA Program Manual to take exception to the provisions of RG 1.38 and ANSI N45.2.2 for motors in storage, eliminates the requirement for insulation resistance testing of all motors in storage, and revises shaft rotation requirements for motors in storage.
REASON FOR CHANGE The proposed change eliminates unnecessary insulation resistance tests and allows shaft rotation frequencies to be adjusted according to experience and type of motor. In addition, some motor shafts cannot or should not be rotated (e.g., computer fan motors, strip chart recorder motors).
SAFETY EVALUATION According to the safety evaluation, each motor is thoroughly tested before being placed in service. Elimination or adjustment of the in-storage tests will not affect the reliability of the motors in service. Therefore, no accidents or important-to-safety equipment would be adversely affected by the change and the margin of safety is not reduced.
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DESCRIPTION The proposed change will revise the EFW design basis operating point given in the FSAR and the EFW TS Bases 3/4.7.1.2.
REASON FOR CHANGE Calculation EC-M96-004 evaluated the EFW demand events and determined the bounding event for the EFW system is the Feedwater Line Break (FWLB), For a FWLB, the EFW is required to deliver 575 gpm at 1100 psia (1085 psig) to the intact Steam Generator to remove decay heat from the Reactor Coolant System. Calculation !
MN(Q)-10-1 demonstrates that the two electric driven EFW pumps or the one turbine driven pump can provide 575 gpm to a single SG at a pressure of 1102 psig, the first main steam safety valve setpoint plus 3% operating tolerance. Other changes revise EFW information in the FSAR and TS Bases to be consistent with the SER and safety evaluations for approved design changes.
SAFETY EVALUATION According to the safety evaluation, the proposed change does not alter the operation or function of the EFW system, nor does it create new system connections. Therefore, it does not increase the probability of an accident, it does not create the possibility of a new accident, nor does it affect important to safety equipment. Calculation EC-M96-004 determined that an EFW flow of 575 gpm at 1100 psia to the intact steam generator provides adequate margin to remove decay heat from the RCS. Calculation MN(Q)-10-1 determined that two electric driven EFW pumps or one turbine driven EFW pump can provide 575 gpm at a pressure of 1102 psig. Therefore, this change does not increase the consequences of an accident nor does it reduce the margin of safety.
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12, LDCR-97-0042. Breathino Air Reauirements for Control Room and Fire Briaade Personnel DESCRIPTION The proposed change revises the FSAR to accurately reflect the location ani Othod of providing and replenishing breathing air for control room and fire brigade p innel.
REASON FOR CHANGE Condition Report CR-96-0591 identified discrepancies with FSAR section 9.5.1.3.1 in its comparison of Waterford 3 to the guidance of Appendix 'A' to BTP APCSB 9.5-1, SAFETY EVALUATION According to the safety evaluation, the proposed change revises the FSAR to accure ely describe how Waterford 3 satisfies BTP APCSB 9.5-1. The change does not compromise the ability of the fire brigade to _ respond to a fire or of control room personnel to bring the plant to a safe thutdown if n3cessary. No equipment important-to-safety is affected by this change and the margin of safety is not reduced.
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- 13. LDCR-97-0046. Eliminate Contents of FSAR Table 9.3-14. CVCS Component Diesel Loadina DESCRIPTION The proposed change eliminates the contents of FSAR Table 9.3-14, CVCS Component Diesel Loading, which is redundant to Table 8.3-1.
REASON FOR CHANGE Table 9.3-14 is not typically referred to when changing the load blocks for the emergency diesel generators. All information concerning emergency diesel assignment of loads shown on Table 9.3-14 is part of the current Tsble 8.3-1.
SAFETY EVALUATION According to the safety evaluation, the proposed change only relocates information from one FSAR table to another FSAR table. This change does not affect any design basis accident, has no affect on any equipment, does not physically change the plant, and does not reduce the margin of safety.
113
- 14. LDCR-97-00 /. Location of Cabinets C3A(B) and C-4 Outside the CVAS Boundarv (Revision 0 and Revision 1)
DESCRIPTION -
The proposed change will prescribe Local Leak Rate testing requirements for Penetrations 53 and 65 to address these potential bypass leakage paths created by the location of cabinets C-3A(B) and C-4.
REASON FOR CHANGE -
In resporise to NRC Qut sJon 480.36, LP&L stated that instrumentation lines through Penetration 53 and 65 form e closed system outside containment, are seismically
_ qualified, and terminate in an .rea exhausted through the filters of Controlled -
Ventilation Area System (CVAS). Therefore, they did not need to be bypass leakage tested.- Contrary to this, these lines terminate in cabinets C-3A(B) and C-4 which are :
outside the CVAS boundary. In addition, the CVR sensing lines constitute a closed
- system for instrument lines outside Containment in accordance with ANSI /ANS-
- 56.2/N271-1976 and ISA 67.02-1980.
SAFETY EVALUATION According tc the safety evaluation, the proposed change corrects the FSAR by -
incorporating appropriate leak testing requirements for identified potential bypass leakage paths associated with Penetrations 53 and 65, There are no desig . changes and the function or operation of equipment important to safety will not be affected as a result of this change. The evaluation shows that the integrity of the containment barrier-is maintained within specified acceptance criteria and radiological releases remain
.within acceptance limits for offsite and control room dose. Thus the margin of safety is preserved.-
114 1
15.- LDCR 97-0061. Ultimate Heat Sink Desian Basis Meteoroloaical Conditions DESCRIPTION The design basis of the UHS is based on the most limiting one hour historical meteorological condition of 102 degrees F wet bulb / 78 degrees F dry bulb. The proposed change will add equivalent meteorological conditions that the UHS can reject its design basis heat load.
RgASON FOR CHANGE The proposed change identifies the bounding design meteorological conditions in order
. to maintain a 115 degree F CCW outlet temperature to the plant auxiliaries during a design basis accident.
SAFETY EVALUATION According to the safety evaluation, the proposed change does not alter the operation or function of the UHS or its components. It enhances plant safety by identifying the bounding meteorological conditions in order to maintain a 115 degree F CCW outlet
~
temperature to the plant auxiliaries during a design basis accident. The proposed change does not alter the function of the UHS and it will continue to operate within its design capabilities. Therefore, there is no adverse affect on important-to-safety equipment. No accidents are affected since no new system interconnections are created. Because the analyzed meteorological conditions are equivalent to the most limiting one hour historical meteorological conditions, the margin of safety is not reduced.
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- 16. LDCR 97-0069 Resettina the Hiah Loa Power Trio Setpoint to 0.257% and Chanaes to FSAR Section 1Q4M DESCRIPTION The proposed change resets the HLP1 setpoint from 0.0257% to 0.257%, removes the CEA movement restriction at RCS temperatures below 520 degrees F, and revises FSAR Section 15.4.1.1.
REASON FOR CHANGE Reanalysis of the CEA withdrawal from subcritical conditions indicates that the HLPT setpoint of 2.2%, which corresponds to an equipment setpoint of 0.257%, has to be increased by a factor of 2 (i.e.,2.2
- 2 = 4.4% power) to account for the effect of additional decalibration of the log power channel.
SAFETY EVALUATION According to the safety evaluation, the rqalysis indicates that the DNBR remains bounded by the DNBR SAFDL value. I addition, the CPC's will trip the reactor at a -
point much iower than the HLPT setpoint. Therefore, there are no increased consequences of a previously evaluated accident. Changing the HLPT setpoint does not affect the operation of any equipment important-to-safety and no new system interactions are created. The analysis irdicates that an uncontrolled CEA withdrs;.al will be terminated before exceeding the UNBR SAFDL'or fuel centerline melt limit, 4 therefore the margin of safety is not reduced.
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! - 17. LDCR-97-0104. Revises FSAR Section 9.5.7.2 and Fiaure 9.5-6 i
DESCRIPTION
- The proposed change revises FSAR Section 9.5.7.2 and Figure 9.5-6 to show that ths
. non-safety, non-seismic Emergency Diesel Generator Lube Oil Storage Tank is not used.-
- REASON FOR CHANGE Engineering determined that the tank could not be used for_its intended purpose because of its close proximity to EDG 'B' intake air piping.
. SAFETY EVALUATION
.The safety evaluation states that the scope of the proposed change is limited to -
changing the FSAR to show that the non-safety, non-seismic lube oil storage tank is not used to store EDG lube oil as currently described in the FSAR. There will be no physical change to the plant and existing plant equipment will remain intact. This -
change does not create an unreviewed safety question.
4 117
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- 18. LDCR-97-0110. Technical Reauiret >nts Manual Yable 3.6-2 and FSAR Table 6.2-32 DESCRIPTION The proposed change removes the (*) note for the EFW Flow Control Valves from TRM Table 3.6-2 and changes the note on FSAR Table 6.2-32 for EFW Flow Control Valves to read "No credit is takea for these isolation valves in meeting the requirements of ,
GDC 55 through 57."
REASON FOR CHANGE The EFW FCVs, EFW-228A(B) and EFW-229A(B), are not required to be containment isolution valves per GDC 57. These valves do not have NRC-approved " direct" position indication in the control room and thus should not be credited for containment isolation. This change restores the original licensing basis which stated these valves are not credited for containment isolation per the GDC.
SAFETY EVALUATION According to the safety evaluation, the proposed change will have no impact on the
. function of the EFW system. EFW FCVs will continue to receive the ESFAS Main Steam isolation System (MSIS) signal which will close the valves unless an EFW Actuation Signal (EFAS) is present. FSAR Table 6.2-32 originally indicated that EFW FCVs were not required to met GDC 55 through 57. A subsequent licensing document change revised the FSAR and TRM Table 3.6-2 by indicating that these valves could be credited as providing an isolation function pursuant to GDC 55-57. However, these valves do not have " direct" position indication per RG 1.97 and thus should not be credited for containment isolation. Containment isolation will consist of at least one -
remote operated valve and a closed system inside containment. This meets the requirements of GDC-57 and is defined in the FSAR as a Class C penetration. This is an administrative change only and no physical change to the plant will be made and no method c., peration will be affected.
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- 19. LDCR 97-0123. Revises FSAR Table 6.2-32 DESCRIPTION The proposed change to FSAh Table 6.2-32 deletes the closure time of the Atmospheric Dump Valves, MS-116A(B).
REASON FOR CHANGE ADV closure time has no relevance to any safety analysis conclusion. Closure times are requested by RG 1.70, but FSAR 6.2.4.2.3, Item I, says that the time assigned to the ADVs was conservatively selected and provides no other basis. FSAR 15.1.1.4 and 15.1.2.4 both assume an ADV remains open without crediting any automatic function.
SAFETY EVALUATION The safety analysis relevant to the ADVs assumes that one ADV remains open for the duration of the accident. In this accident, releases from the affected steam generator are from a primary-to-secondary leak of one gallon per minute which is assumed to exist in the affected steam generator for the duration of the accident. Most accidents assume that allowed primary-to-secondary leakage goes into the faulted steam generator. Thirty minutes after the inadvertent opening of the ADV, operators begin to cool down the plant. They achieve shutdown cooling conditions at approximately 11, 300 seconds. That is the presumed end of the event because it is reasonable to assume that operators have manually closed the ADV by the time shutdown cooling begins.Section XI trending of ADV performance willidentify the adequacy of ADV operation independent of any values posted in FSAR Table 6.2-32.
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- 20. LDCR-97-0126. Revises FSAR Sections 12.3.2.3 and 12.3A.1
- DESCRIPTION The proposeo change updates FSAR sections 12.3.2.3 and 12.3A.1 with a description of the use of other computer codes in radiation dose rate and shielding calculations.
One of the computer codes utilized is MicroShield.
REASON FOR CHAN,f,g .
The MicroShield computer code was used in the calculation of radiation dose rates in the Reactor Auxiliary Building, elevation -15', valve gallery near CS-117A and CS-117B valves. The calculation was performed and the results subsequently incorporated into the FSAR in May 1991. However, there is no reference to the MicroShield computer code in the FSAR.
SAFETY EVALUATION The use of a properly verified and validated computer code in radiation dose rate and shielding calculations does not adversely affect any accidents, equipment, or margin of safety factors. By applying standardized and accepted reference problems (i.e., ANSI, ANS, ASME, etc.), a proper QC lest verification assures that other computer codes, when utilized by a properly trained individual, will accurately and conservatively calculate dose rates and shielding problems at various points of interest. The MicroShield computer code has been successfully verified and validated by industry and W3SES to yield accurate and conservative results for radiation dose rate and shielding calculations.
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- 21. LDCR-97-0131. FSAR 10.3.2 and 10.4 9. and EC-M96-028. EFW Heat Trace Temperature Reauirements DESCRIPTION Calculation EC-M96-028 establishes minimum temperature requirements for the EFW heat trace system to ensure there is no standing water and to prevent the formation of ,
water slugs during startup of the Terry Turbine. The calculation provides the basis for heat trace temperature alarm setpoints and administrative requirements.
REASON FOR CHANGE i
The minimum temperature requirements stated 8,1 the FSAR and Operations procedures were not woll supported in design documents. EC-M96-028 establishes and provides a basis for the requirements o' the EFW heat trace system.
SAFETY EVALUATION
- The function of the EFW heat trace is to prevent star. ding water in the EFW steam supply piping during idle periods and to reduce the condensate load during startup of the turbine. EC-M96-029 establishes a minimum overall temperature of 230 deg. F to prevent standing water and a minimum average temperature of 350 deg. F to prevent the formation of water slugs, The requirement established in EC-M96-028 provide additional margin and ensure that a heat trace failure does not affect operability of the Terry Turbine. There are no unreviewed safety questions associated with this change.
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- 22. LDCR-97-0134. Revises UFSAR Section 11.2.2.1. Boron Manaaement System DESCRIPTION This change revises the description of the operation of the Boron Management Systen; to agree with the system arrangement used by Operations.
REASON FOR CHANGE This change resolves a conflict between the UFSAR description of the BMS operation and how the system is actually operated.
SAFETY EVALUATION The safety evaluation states that there are no accidents or equipment important-to-safety that affect the BMS. The proposed change describes the normal operation of the BMS. It does not preclude operation of the system as originally described. All fluids in the BMS are retained within system boundaries. Without use of the flash tank, hydrogen will be removed from the water in the holdup tanks. The gas analyzer can be used to monitor hydrogen levels in the holdup tanks to preclude any buildup.
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- 23. LDCR-97-0138. Revises FSAR Section 9.5.1 DESCRIPTION Replace individual fire area equivalent fire severity values in FSAR Section S 3.1 with
- the stat 9 ment that the fire severity does not exceed the fire resistance rating of the area boundary barriers.
REASON FOR CHANGE -
- To eliminate the potential need for an FSAR change and 50.59 safety evaluation each time the combustible loading value changes in a fire area.
SAFETY EVALUATION
. Removal of the combustible loading information from the FSAR will not affect fire
. area / zone fire hazards analyses. Each fire arealzone discussed in the FSAR has a specific combustible load calculation that is a controlled design document. The data .'
contained in the area combustible loading calculations provides the technical input .-
necessary for controlling combustible loading in area boundary barriers, will limit fire damage to structures, systems and components important to safety so that the capability to safely shut down the plant in the vent of a fire continues to be ensured.
Fire 'arealzones will remain as analyzed. Neithcr will it alter the NRC evaluation of the
' fire protection program contained in the SER and its supplements. Removal of the specific combustible load information from the FSAR does not constitute a USQ.'
- ?3
r
- 24. kpCR-97-0140. Revises FSAR Sections 13.1 and 13.5. Table 13.1-2. and Fiaure 13.1-6 DESCRIPTION The proposed change revises FSAR Sections 13,1 and 13.5, Table 13.1-2, and Figure 13,1-6 to reflect the actual Plant Operations organization in place at this time, nuASON FOR CHANGE The change implements the Plant Engineering reorganization plan and the transfer of responsibilities from Radiation Protection to Chemistry. The Chemistry Department is now respunsible for radioactive effluents, radiological environmental monitoring, and gamma counting systems.
SAFETY EVALUATION The proposed change does not create an unreviewed safety question. The change is administrative in nature and does not eliminate any functions or responsibilities previously described. Only titles and organizational lines of accountability are affected.
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- 25. LDCR 97-0141. Revises FSAR Section 6.2.1. Containment Functional Desian DESCRIPTION The containment peak pressure analyses of FSAR Section 6.2.1 are revised to accommodate several revisions to the post-accident Containment Fan Coolers (CFC) and Shutdown Cooling Heat Exchanger (SDCHX) performance.
REASON FOR CHANGE The revised analyses are based on a different set of assumptions regarding CFCs and the SDCHX.
SAFETY EVAL.UATION According to the safety evaluation, the proposed change revised the minimum flow to the CFCs and SDCHX, but does not alter the operation of the CFCs (except to require both fans per train to be operable) or any other equipment. This change does not impact the severity of an event due to failure of a CFC(s) nor does it alter the worst single failure that has already been accounted for in the analysis.
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- 26. LDCR-97-0152. Revises FSAR Section 9.5.1 and FSAR Table 9.5.1-1 DESCRIPTION The propoeed change revises FSAR Section 9.5.1 and FSAR Table 9.5.1-1 to update certain aspects of the Fire Protection system configuration, programs, and site organizations to reflect the currently approved condition.
REASON FOR CHANGE FSAR Section 9.5.1 and FSAR Table 9.5.1-1 were identified as containing
' discrepancies in condition reports CR-96-1892, 96-1872, and SG-1706.
SAFETY EVALUATION The proposed changes are administrativa in nature and do not a ..r the functions of SSCs required by the approved Fire Protection Program. No accidents are affected, no equipment is changed, and no protective boundaries are affected.
a 126
271 LDCR-97-0170. Revises FSAR Section 9.5.1
- DESCRIPTION
' This change corrects the quantity of lubricating oil contained in the EDG crankcases and auxiliary components. It also replaces Fire Area RAB 15 and 16 severity / duration values in FSAR Section 9.5.1 with the statement that the fire severity of the areas does not exceed the fire resistance rating of the area boundary barriers.
REASON FOR CHANGE A corrective action document identified that the quantity of oil in the EDG crankcases and auxiliary components is considerably more than is shown in FSAR Section 9.5.1.
SAFETY EVALUATION
' A fire in either RAB 15 or RAB 16 will be contained in the area of origin, increased fire loading in these areas does not adversely impact either the Fire Hazards Analyses or the 10CFR50, Appendix R, Safe Shutdown Analyses for these areas. - The design -
features of the installed fire barriers exceed the revised combustible loading values; therefore, the change in combustiblo loading does not constitute an unreviewed safety
. question.
=_
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- 28. LDCR-97-0176. Revises FSAR Sections 8.1 and
8.2 DESCRIPTION
The proposed change revises the FSAR to revise the description of protective relaying provided on the W3 switching station to Waterford switchyard transmission line, changes the name of the operating company from Louisiana Power & Light to Entergy Louisiana, Inc., and corrects typographical errors.
REASON FOR CHANGE Protective relays will be replaced on the W3 switching station to Waterford switchyard transmission line as existing relays have become obsolete and spare parts are not available for relay maintenance. These changes require the description of transmission line protective relaying in the FSAR to be revised. Other cha f) s incorporate editorial or typographimi corrections.
. SAFETY EVALUATION The only event associated with the transmission line is Loss of Offsite Power (LOOP).
Installation of the new protective relays and improved relaying logic will increase the reliability of the transmission line and decrease the probability of a LOOP. No important-to-safety equipment is affected, no new system interactions or connections are created, and no margin of safety is reduced. Therefore, there are no unreviewed safety question associated with this change.
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- __-_-.-_m...__.._
- 29. LDCR-97-0201. Revises FSAR Section 9.2.5.3.2. Tables 9.2-1. 9.2-3. 9.2-9. 9.2-
- 10. and Fioures 9.2-4. 9.2-4a. and 9.2-Sa DESCRIPTION The proposed change removes the temperature and associated Wet (WCT) and Dry Cooling Tower (DCT) fan requirement restrictions placed on the Ultimate Heat Sink (UHS) for operability, it also revises the affected FSAR sections as a result of the potential additional heat input into the UHS.
REASON FOR CHANGE To support a change to the Containment Fan Cooler technical specification, an analysis was performed to determine the impact on the UHS assuming inputs that would provide the maximum heat input into the UHS The analysis determined that the maximum possible UHS heat load, including plant auxiliary heat load, exceeded the UHS peak accident load given in the FSAR. An operability determination concluded the UHS is operable provided all WCT and DCT fans remained operable, a peak dry bulb temperature did not exceed 95 deg. F, and a three day average dry bulb temperature did not exceed 84 deg. F.
SAFETY EVALUATION According to the safety evaluation, the UHS does not initiate any accident, therefore there is no increase in probability. Analysis and testing have demonstrated the accident mitigation function of UHS is not affected by this change so accident consequences are not increased. The proposed change does not alter the operation or function of the UHS and the failure modes associated with the UHS remain unchanged.
No new system interactions are created and the analyses provided by engineering calculations identify that the TS bases margin of safety is preserved.
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F. MISCELLANEOUS EVALUATIONS .
- 1. SPC-95-010-0. Wet Coolina Tower Basin Temperature Setpoint DESCRIPTION This SPC lowers the setpoint value for the Wet Cooling Tower (WCT) Basin temperature. The SPC is based upon engineering calculation EC-M95-008. The
)
setpoint for the WCT fans is lowered to ensure the basin temperature remains below 89 degrees F.
REASON FOR CHANGE As a result of testing (Refuel 6) on the Component Cooling Water Heat Exchanger (CCWHX), Design Engineering evaluated the Ultimate Heat Sink (UHS)in order to reduce the design heat duty on the CCWHX. UHS design meteorological conditions were determined to be overly conservativo and were reevaluated. The evaluation, calculation EC-M95-008, changed the WCT basin temperature for operability to 89 degrees F. This temperature provides an additional allowance for fouling in the CCWHX and is based on WCT design capability at the meteorological conditions given in EC-M95-008.
SAFETY EVALUATION According to the safety evaluation equipment important to safety which is affected by this SPC is the CCWHX and the WCT Basins. This equipment will continue to opaate and function within the accident basis limitations for the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) events. The WCT fans will continue to operate in a similar manner as previously designed except that the fan start and stcp signals will occur at lower values. These new setpoints will maintain the basin at a lower temperature value for heat removal at the CCWHX. The new high temperature alarm value will be 86 degrees F. which will allow the operator time to address system malfunction and to ensure the basin temperature is maintained below the 89 degree F.
value.
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- 2. SPC-94-005-0. Pressurizer Pressure Hi/Lo Alarm Annunciator H0501 RC IPAC0100-X.Y DESCRIPTION ,
Revises the safety evaluation to include the Pressurizer heater cutout setpoint that is part of the alarm annunciator.
REASON FOR CHANGE The original safety evaluation did not include the Pressurizer heater cutout setpoint, this revision corrects the safety evaluation. No field work is required this only applies to the safety evaluation.
NOTE: SPC-94-005 was reported in W3F2-96-0011, Report of Facility Changes, Test and Experiments, dated May 6,1996. See item F.7 of that report.
SAFETY EVALUATION According to the safety evaluation, SPC-94-005-0 is a conservative setp aint change that lowered the setting of the Pressurizer Pressure Alarm Annunciator' rom 2350 psia to 2270 psia. This will also result in the cutoff of the pressurizer proportional heaters at 2270 psia. FSAR Section 15.2 (Decrease in Heat Removal by the Secondary System) and Section 15.5 (Increase in Reactor Coolant System Inventory) credits the pressurizer pressure reactor trip on high pressure. This SPC will not affect this analysis in an adverse manner. The new setpoint will assist in limiting pressurizer pressure during a slow build up of pressure. The events in Sections 15.2 and 15.5 involve a rapid increase in pressure, this SPC has no adverse sffect on the accident e',alysis.
The SPC will not reduce the margin of se'ety as defined in the bases for any technical specification or safety analysis The change will lower the Hi Pressurizer Pressure Alarm Annunciation and pressurizer heater cutout from 2350 psia to 2270 psia based upon calculation EC-193-016. This conservative change will alert the operator when the Pressurizer pressure is close to initiating a reactor trip (limit of 2350 psia) and cutout the pressurizer heaters to assist in preventing the reactor trip on a slow pressure rise.
131
- 3. Condition Report 96-0543. CCW/ACCW Throttle Valve Positions DESCRIPTION The engineering evaluation (W4.101) performed for condition report 96-0543 determined that the Essential Chillers, Emergency Diesel Generators, and both trains of Auxiliary Component Cooling Water are operab!e with changed positions of valves ACC-127A(B), CC-406A(B), CC-407A(B), CC-5416, and CC-555 and apparent degraded a.ondition of flow in the ACCW Train 'B'.
REASON FOR CHANGE The following interim limitations are more restrictive than the UFSAR and Technical Specifications and will remain in place until corrective actions for the condition report are implemented: 1) Essential Chille 'B' remains operable as long as ambient dry bulb temperature is </= 98 degrees F and all'B' train DCT and WCT fans main operable; 2)in the event DCT is bypassed due to tornado missile damage, control ACC-126 to maintain Essential Chiller flow >/= 850 gpm; 3) in the event of a LOCA, maintain EDG engine manifold temperature < 120 degrees F.
SAFETY EVALUATION According to the safety eva!uation within the imposed operating limitations, the overall performance and reliability of the identified components and systems will not be affected. The imposed administrative limits ensure that the conditions under which the equipment is capable of performing its required safety function are adhered to.
132
W
- 4. CR-96-0543. Essential Chiller 'B' Operability without Restrictions DESCRIPTION Removes the temperature and associated WCT and DCT fan requirement restrictions placed on the Essential Chiller 'B' for operability.
REASON FOR CHANGE-A special test procedure determined that a maximum accident ACCW flow of less than -
4000 gpm is required for the CCW Heat Exchanger 'B' to dissipate its accident heat --
duty at design basis meteorological conditions. Therefore, the Essential Chiller 'B' can -
fulfill its safety function without rcstrictions since the flow balance testing on ACCW.
Train 'B' demonstrated that the design accident flow of 850 gpm to the Essential Chiller can be achieved at an ACCW flow of 4500 gpm to CCW Heat Exchanger 'B'.
- SAFETY EVALUATION According to the safety evaluation, this change will not alter the operation or function of the Essential Chiller or alter any Essential Chiller component, system, or structura.
- Calculation EC-M96-003 and the special test procecure show that UHS Train 'B' can fulfill its safety function, at design basis meteorological conditions, with an accident flow of 850 gpm to the Essential Chiller.
133
l S. SPC-96-003-0. Condensate Storace Pool Level DESCRIPTION The proposed change revises th6 level cetpoints for the Condensate Storage Pool to include vodoxing.
REASON FOR CHANGE New setpoints will include the TS limit of 170,000 gallons, loop uncertainty, transmitter offset, and the new vortex limit.-
SAFETY EVALUATION The EFW system is required for safe shutdown following a LOOP, LOCA, or = MSLB or
. other loss of feedwater event. The CSP is the main water reservoir for the EFW pump-suction. The proposed change does not affect the volume of water available for safe shutdown of the plant, but ensures that adequate alarms and indication are provided for operator actions, increasing the setpoint does not impact any of the previously mentioned accidents or their consequences. The setpoint change does not require any physical change to the plant and does not increase the probability or consequences of a malfunction of important to safety equipment. No new failure methods are created and the change does not affect the margin of safety.
134
I
- 6. SPC-96-004-0. Loss of instrument Air Accumulator CVR-101. CVR 201 Alarm Setooints DESCRIPTION Revises the Containment Vsnum Relief pressure switch setpoints and valve
- accumulativ pressure twitch setpcInti REASON FOR CHANGE Replacement of current pressure switch with pressure switches capable of tighter reset values.
4 GAPETY EVALUATION
_ There are no accidents evaluated in the FSAR that may be caused or affected by the proposed change. Therefore, there is no increase in accident probability or consequences The proposed change enhances indication in the control room but does not increase the likelihood or consequences of a malfunction of important-to-safety equipment. No new system interconnections are created, no new failure modes are created, and no margin of safety is reduced.
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- 7. W4.101. EFW Heat Tracina Evaluetion (Revision 0 and Revision 1)
DESCRIPTION l
The FSAR requires heat tracing maintain the temperature of piping at or above 250 degrees F to prevent thermal shock to the turbine and EFW steam piping. This engineering evaluation identifies that there is no adverse affect caused by not maintaining 50 feet of 616 feet of piping on Channel 1-8B in the EFW piping.
REACON FOR CHANr,F An electrical problem exists for the heat tracing on this segment of piping and replacement parts are not readily available.
SAFETY EVAL.UATION According to the safety evaluation, the reduced temperature (70 degrees F) of the 50 foot section of piping will produce more condensate during system startup. However, the steam traps and drain valve are capable of removing the additional condensate. In addition, the piping stresses are not adversely affected by the lower temperature.
Some condensate may ren.aln entrained in the steam flow and enter the Terry Turbine, but due to the piping configuration and the momentum of the condensate, most of the condensate will enter the drain lines and steam traps. Therefore, operation of the EFW system is not affected by this char.ge and there is no increase in probability of occurrence of a malfunction of equipment important to safety.
136
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- 8. SPC-93-017-0. CCW and Shutdown Coolina Room Temperature Setpoints DESCRIPTION This setpoint change lowers the setpoints for the CCW/SDC Heat Exchanger room temperature loops because the existing setpoints did not have adequate margin to allow for instrument uncertainty.
REASON FOR CHANGE The function of the safety related room coolers is to maintain a 104 degree F ambient temperature in the CCW/SDC hirt exchanger rooms during accident conditions.
During normal operations, coohv to tim rooms is provided by RAB normal ventilation.
The lower room 19mperature setpoints will ensure that the temperature in the CCW/SDC heat exchanger rooms does not exceed the design value . 'S4 degrees F.
SAFETY EVALUATION According to the safety evaluation, the setpoint change lowered the temperatt re control band for the CCW/SDC room coolers by 2 degrees F. Changing the setpoint does not affect chilled water flow or increase the heat load esn the Essential Chillers. With respect to the equipment in the rc oms, this is a change in the conservative direction.
Therefore, performance of the CCW/SDC heat exchangers as ',,3ll as other equipment contained in the rooms is not edversely affected.
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_ . _ _ _ _ _ _ _ _ _ _ _ _ - )
- 9. TRM-004. Table 3 3-11 and Section 3.7.
10.4 DESCRIPTION
Deletes portions of and revises portions of Technical Requirements Manual Table 3.3-11 and clarifies Section 3.7.10.4 regarding use of gated wye valves in compensatory action for RCB Fire Hose Stations.
REASON FOR CHANGE Design Change DC-3268 replaced the existing fire detection system with a new system that meets the requirements of NFPA 72D, Section 9.5-1 of the W3 UFSAR, and SSER
- 5. The TRM delineates the curveillance and operability requirements for portions of the Fire Protection System.
SAFETY EVALUATION According to the safety evaluation, this is an editorial / typographical correction that corrected fire area designations and number of instruments incorrect and transposed.
There is no unreviewed safety question and no adverse impact on nuclear safety, 138
- 10. Closure of isolation Valves CVR-401 A(B) (Revision 0 and Revision 1)
DESCRIPTION Thir. SE evaluates changes to operations procedures OP 903-001, OP-002-010, OP-903-031, OP 903-040, and OP 903-094 that are required as a result of maintaining !
containment isolation valves CVR-401 A(B) closed except during operation of !
Containment Atmosphere Purge (CAP) and surveillance procedure testing. ;
REASON FOR CHANGE Closing CVR-401 A will result in isolating pressure transmitters CAP-IDPT-5171 and CVR-IDPT-5017A. Closing CVR-401B will result in isolating pressure transmitters CVR-IDPT-50178 and C. With CVR-401 A closed and remaining closed, CAP-IDPI-5171 and Plant Monitoring Computer point A51000 will not be able to satisfy TS Surveillance 4.6.1.4 for measuring containment internal pressure and provide reliable )
indication. Changes to these procedures implement alternate instrumentation to satisfy l the TS Surveillance requirement mentioned previously.
l i
SAFETY EVALUATION l
According to the safety evaluation, maintaining these valves el ) sed does not increase the probability or consequences of an accident or affect the margin of safety. By keeping the valves closed during normal operation, the dose consequence of a tubing rupture downstream of the valves would be limited to leakage through the isclation valves which is included in bypass leakage. The valves would only be opened during containment purge. In the event containment isolation were required during
- containment purge, a CIAS signal would close both valves. Thus this change does not
-increase the probability or consequences of an equipment malfunction.
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o - . - - . _ _ . _ . _ _ _ _ _ _ _ _ , _ . _ _ _ _ _ . _ _ -
- 11. Control of Safety and Relief Valve Blowdown Rina Settinas, Root Cause Analysis 96-0463 DESCRIPTION The root cause analysis for CR 96-0463 addressed the generic issue of incorrect setting of safety relief valve blowdown rings, This evaluation addressed the effect of the ring setting on six valves that have not been reset to the correct ring setting.
REASON FOR CHANGE Address concerns surrounding the ability of six of the orig'nal twelve relief valves identified via corrective action for NRC IEN 92-64 to perfona as designed. Those valves are RC-603, SI-132A, SI-214, SI 2200, SI-404A, and SI-408A.
SAFETY EVALUATION According to the safety evaluatior., there are currently no credible mechanisms to cause pressure transients that would cause the valves addressed in this evaluation to open. The probability of a previously analyzed accident occurring with these six valves having mispositioned blowdown rings is not increased and the function of the valves does not serve as an initiator of any design basis accident. There are no new system interactions created and the mispositioned blowdown rings do not affect the function of important-to-safety equipment. Discharge by these valves does not affect a protective boundary, therefore the margin of safety is not affected.
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- 12. eost Accident Samplina System (PASS) Heat Exchanaer Operation and Chilled Water (CHW) Chemical Addition (PElR OM-117)
DESCRIPTION PElR OM-112 provides recommendations for operation of PASS heat exchangers including the basis for operation of PASS using only one heat exchanger aligned to CCW Tra:n 'A' and provides recommendations for operation of CHW chemical addition pots.
REASON FOR CHANGE PElR OM 112 addresses the acceptability for aligning non safety related, non-seismic components in the CCW and CHW systems to safety-related, seismic portions without automatic isolation valves. PElR OM-112 demonstrates that it is acceptable to opeiste PASS using only the CCW Train 'A' to provide sample cooling as long as the '
appropriate Limiting Condition for Operation is entered or an operator is stationed at the CCW inlet isolation valve.
SAFETY EVALUATION According to the safety evaluation, this change does not require abnormal operation of PASS and CHW nor does it degrade their performance. Therefore, there is no increase in probability or consequences of an accident. Precautions taken to enter the appropriate LCO and the separation of safety from non safety portions of the CCW and CHW systems ensure there is no increase in probability or consequences of a malfunction of equipment important to safety. Two new system alignments are created by the proposed change. The first uses only the CCW Train 'A' heat exchanger to cool he sample flow through PASS. However, an expansion reservoir in the CCW loop allows for thermal expansion of the cooling water. The second allows for use of the non safety 'AB' loop in CHW for chemical addition. However, this loop is isolated on SIAS, low surge tank level, or high non safety loop flow. Therefore a different malfunction of important-to-safety equipment is not created. No protective boundaries are affected, therefore the margin of safety is not reduced.
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- 13. SPC-96-011-0. CC Accumulators for CC-807A(B). CC 808A(B). CC-822A(B).
and CC-823A(B)
DESCRIPTION
- The proposed change increases regulator output pressure for the containment isolation valves for CCW to and from the Containment Fan Coolers to maximum output. It also increases the accumulator low pressure alarm setpoints for each of the valves to 92 psig with a reset of 108.5 psig.
REASON FOR CHANGE The accumulators for valves CC-807A(B), CC-808A(B), CC-822A(B), and CC-823A(B) require leakrate testing to ensure adequate motive air to allow valve stroke in the event of a Loss of Instrument Air without a SIAS present. By calculation, it was determined that the minimum required accumulator pressure is 90 psig to ensure adequate volume for post-accident operation on loss of instrument air provided no SIAS is present.
SAFETY EVALUATION The safety evaluation states that there is no unreviewed safety question for this change. The increase in the setpoint will not increase the likelihood of a LOCA or MSLB occurring and it will not affect the function of the valves to open to allow CCW flow to the CFC's to maintain containment pressure and temperature at acceptable levels. While increasing the regulator pressure willincrease the valves open stroke time, administrative controls will ensure that open stroke design limits are preserved.
The increase in setpoint does not create any new system interactiotis. The accident response of the CFC's and the containment isolation valves for CCW to and from the CFC's are not affected by this change; therefore, no margin of safety is reduced.
t 142 g . . .
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- 14. SPC 94-017-0. RAB Ambient Neoative Pressure. HVRIPAC5275 A and B DESCRIPTION The proposed change revises the setpoints for Alarm and Damper manipulation used to maintain the Controlled Ventilation Area System (CVAS) at a negative pressure of 0.25 INWG relative to surrounding areas following a LOCA.
REASON FOR CHANGE
- The old setpoint values did not include instrument uncertainty as determined by current methodologies.
.EA_FJTY EVALUATION According to the safety evaluation, the proposed change ensures that affected areas will remain below 0.25 INWG during normal, accident, and post-accident operations.
Thus the probability and consequences of an accident are not affected. The function of CVAS and its important to safety equipment is also not affected by this propose change and no new system interactions are created. Because the proposed change maintains radiological releases to 10CFR100 limits, the margin of safety is not impacted.
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- 15. W3 Pump and Valve Inservice Test Plan. Revision 8. Chance 2 i DESCRIPTION Added Dry Cooling Tower Bundle Isolation Valves to the IST Plan and to FSAR Table 3.9-9; added Section 5.2.1 to address use of digital instrumentation; deleted Relief Request 2.1.6; changed testing frequency of chiller cooling water supply valves to cold shutdown; deleted closed test requirements for CVR-302A(B) and added these valves to FSAR Table 3.9-9; added leak test requirements and removed stroke test requirements for CVR-401 A(B); corrected valve type from ' gate' to ' globe' for CVC-101 in FSAR Table 3.9 8; made other clarifications sind editorial changes REASON FOR CHANGE I l
Changes are being made to update the IST Plant to reflect additional / changed test l requirements and to clarifv information.
SAFETY EVAL.UATION Technical Specification 4.0.5 requires inservice testing of pumps and valves be I conducted in accordance with ASME Section XI All changes to the IST Plan are consistent with these requirements and other NRC guidance. The addition of the Dry Cooling tower bundle isolation valves and the CVR excess flow check valves to FSAR Table 3.9 9 is necessary since these are active components which apparently were never incorporated into the table. The valve type correction for CVC-101 in FSAR Table 3.8-8 is editorial only. No unreviewed safety question exists.
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- 16. Calculation EC-M95-012. Minimum Submeroence to Prevent Vortexino 3 Calculation (Revision il DESCRIPTION This calculation determines the critical height at which vortex formation begins in the Condensate Storage Pool (CSP), Boric Acid Makeup (BAM) tanks, EDG Oil Storage Tanks, EDG Oil Feed tanks, Refueling Water Storage Pool (RWSP), and Volume Control Tank (VCT).
REASON FOR CHANGE Corrective action document CR-96-0657 identified that the volume requirements for the CSP did not include margin for vortexing. The corrective action led to development of EC-M95-012, Revision 0, which calculated the level at which vortex formation would I
occur in the CSP. Other corrective action requirements included evaluation of the other tanks and storage pools identified in thle change.
SAFETY EVALUATION According to the safety evaluation, this calculation does not require any changes to the plant or procedures. It provides the basis for available volume determinations without entraining air into the flow stream for the evaluated tanks. Operation of the equipment evaluated in the calculation and the associated systems is not affected by vortex formation and entraining air into the pump suction piping, it also shows that the evaluated storage pools and tanks are not adversely affected by vortex formation.
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- 17. TRM Section 4.7.
11.2 DESCRIPTION
TRM surveillance 4.7.11.2.a requires a 31-day channel functional test of electrically supervised doors. Surveillance 4.7.11.2.b requires a weekly verification that doors are closed. These surveillances are being carried out on a daily basis in accordance with procedure PS-015-111 which is more conservative that the TRM.
REASON FOR CHANQg This change satisfies the corrective action for CR-96-1620 which identified that TRM surveillances 4.7.11.2.a and 4.7.11.2.b c m not being conducted.
SAFETY EVALUATION According to the safety evaluation, this change is not an unreviewed safety question.
- The changes involved are administrative in nature and delete information which is either not being used or is not applicable. The alternate testing being performed is .
more frequent and more conservative than the TRM surveillance requirements. The ability of SSCs to adequately respond to and mitigate the consequences of a fire, as described in FSAR section 9.5.1, are maintained.
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- 18. SPC-97-001-0. Emeroency Breathina Air low Pressure Alarm DESCRIPTION i
This setpoint change revises the low pressure alarm setpoint for the Emergency Breathing Air (EBA) system to 1840 psig.
REASON FOR CHANGE Technical Specifications require the EBA system to maintain a pressure of 1800 psig in the EBA system when the control room envelope is inoperable due to a known breach.
The current setpoint of 1800 psig does not allow for instrument uncertainty, tole.ance of the instrument.
GAFETY EVALUATION The proposed change does not represent an unreviewed safety question. There are no accidents in the SAR which are caused or affected by this change. The new setpoint will account for uncertainty in measuring EBA pressure and allow some margin between the alarm and the TS limit. The EBA system function and performanca are unaffected by the proposed change, There are no new system interactions or connections and no protective boundaries are affected.
147
- 19. Drawina G-587. Common Foundation Structure Masonry. Sheet i DESCRIPTION This change will update controlled drawing G-587 which also exists as FSAR Figure 3.8-43, to show four columns that are located at column line '6FH' and column line 'U' underneath the spent fuel pool, that were inadvertently left off this document but are shown on all other documents associated with this area.
REASON FOR CHANGE This change uodates controlled drawing G-587 and FSAR Figure 3.8-43 regarding four columns that were inadvertently left off.
SAFETY EVALUATION Updating the controlled documents to reflect this change does not create an unreviewed safety question. No accidents, equipment, or protective boundaries are affected by the change.
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- 20. Instrument Calibration Calculations and Instrument Calibration Data for Refuelina Water Storaae Pool Level Loops. SI ILO305 A. B. C. and D DESCRIPTION The proposed change implements revised calibration data for transmitters Sl ILT0305 A, B, C, and D, that account for ambient pressure variations at the transmitter reference legs. The revised calibrations will result in the actual level in the RWSP being greater than or equal to the indicated level in the RWSP for credible ventilation system lineups.
REASON FOR CHANGE Condition Report CR 97-05G2 identified that RWSP level indication is affected when the Controlled Ventilation Area System is started.
SAFETY EVALUATION The implementation of the revised Instrument Calibration Calculation and Instrument Calibration Data Records for loops SI lLO305A, B, C, and D does not result in an unreviewed safety question. The proposed change does not increase the probnbility or consequences of a LOCA or MSLB. It also does not increase the probability or consequences of a malfunction of equipment important to safety. The change wo result in the actual level in the RWSP being slightly greater than or equal to the indicated level in the RWSP. This change compensates for the effect of the operation of two CVAS trains, a single failure of a makeup damper, and other revised inputs for temperature and boron concentration.
149
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- 21. D_urnura Extension Proaram DESCRIPTION The proposed change involves an experiment (high burnup demonstration) not described in the UFSAR. The 220 fuel pins in a single Batch J assembly at the center of the core in this experiment are expected to accumulate one pin (axially integrated) burnups at 66,200 mwd /MTU at EOC-9.
REASON FOR CHANGE The purpose of this experiment is to demonstrate acceptable corrosion resistance of the current standard ABB CENO low-tin Zircatoy-4 fuel pin cladding at Waterford 3 to at least 63,000 mwd /MTU for fuel management considerations. At the end of Oucle 9, the information gained from this experiment on the Batch J assembly will be usc, .o support the development of an ABB CENO topical report demonstrating the acceptability of higher burnup fuel.
SAFET/ EVALUATION Operation of up to 220 fuel pins in a single assembly beyond 60,000 mwd /MTU exceeds a one pin burnup design limit of UFSAR Appendix 4.3A.3.1.3. The extension of burnup for this test assembly has been explicitly included in the core calculations performed in support of the reload safety analysis of Cycle 9 using methods approved by the NRC Additional explicit evaluations of CEA ejection, the fuel mishandling accident, and fuel performance for the lead Batch J assembly at the highest one-pin burnup anticipated indicates no licensing limits will be exceeded. No changes are required to the UFSAR or TS for a temporary experimental extension of the burnup design limit for a single test fuel assembly for Cycle 9. There are no unreviewed safety questions asscciated with this experiment.
150
- 22. Calculation EC-M92-015 DESCRIPTION The purpose of calculation EC-M92-015 was to structurally evaluate the new Type C-2 steam generator nozzle dams for operation under Waterford 3 loading conditions.
REASON FOR CHANGE Problems were encountered in RFO2 and RFO3 due to difficulty installing the nozzle dams and creating a good seal. The dams generally required several reinstallations to obtain a leak-tight seal, which impacted man-rem exposure and critical path time. The new nozzle dams allow shorter installation time, thereby saving man-rem exposure.
SAFETY EVALUATION Using a design loading of 20 psig, a normal operating pressure of 17.33 psig, and a normal test pressure of 25 psig, along with the single failure of a pin not engaged, and a design basis earthquake evaluation, the nczzle dams were evaluated to be structurally capable of performing their necessary functions. An overpressure condition of 50 psig was also analyzed and tested. The analysis and testing show the structural design of the dam system meets design criteria, will perform as designed, and no unreviewed safety question exists.
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- 23. Cycle 9 Core Re!oad (Mode 6 Ontv)
DESCRIPTION This safety evaluation determines whether or not the Cycle 9 fuel can be safely handled and placed in the reactor vessel. This safety evaluation is valid for Mode 6 operation only, except for boron dilution alarm reset times which are valid for all r,) odes of operation.
REASON FOR CHANGE Cycle 9 core reload.
SAFETY EVALUATION According to the safety evaluation, the accidents potentially affected by new fuel in Mode 6 are inadvertent boron dilution and a fuel handling accident (FHA). Boron dilution alarm reset times were analyzed and revised to ensure the alarms will provide the required advance warning to plant operators of an ongoing boron dilution at least 30 minutes prior 's the loss of Shutdown Margin. Thus there is no increase in probability of occurrence of a boron dilution event. New fuelis not the initiator of a FHA, therefore, there is no increase in probability. The consequences of a FHA have been analyzed and documented in the FSAR and the BOC 9 core burn-up is below the FSAR limit. Thus there is no increase in consequences. No important to safety equipment is affected by this change, thus there is no increase in probability or consequences of a malfunction. The spent fuel racks, containment temporary storage rack, and fuel carrier will continue to maintain a k.,less than 0,95, thus there is no reduction in margin of safety.
152
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- 24. Calculation MNf 0)-6-45 (Revision 1. Chance 1)
- QESCRIPTION The proposed change revises calculation MN(0)-6-45 to clarif;* the assumptions and methodology of the calculation. This calculation determines the duration of a water barrier in penetrations 27, 34, 35, 40, 41, 69, and 70 based on conservative assumptions for leakage.
REASON FOR CHANGE Condition report CR-96-1807 identified that the assumptions used in calculation MN(0)-6-45 were inadequate to ensure that a 30-day barrier would be maintained in penetration 27.
SAFETY EVALUATION The results of the calculation Indicate that a water barrier will be maintained on the affected penetrations in excess of 30 days at the assumed leak rates. Therefore, these penetrations do not represent a pot:qtlal containment atmosphere leak path during or following an accident considering a single failure of a system component and the basis of the Type 'C' test exemption for penetrations 27,34,35,40,41,69, and 70 is not affected. The proposed change has no affect on the operation or function of the containment isolation valves in these penetrations. No unreviewed safety question is created.
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- 25. Cycle 9 Core Reload DESCRIPTION OF CHANGE Fuel Manaoement and Core Neutronics The Cycle 9 core will utilize a " low leakage" fuel management scheme, with an estimated reactivity life of 510 effective full power days (EFPD) at 10 ppm boron concentration based on a Cycle 8 end point of 495 EFPD. The Cy:le 9 Reload Analysis supports an operation range of 501-531 EFPD for the long Cycle 8 end point (510 EFPD) and an operation range of 519-549 EFPD at the short Cycle 8 end point (480 EFPD).
The Cycle 9 core consists of 84 new Batch R assemblies and 133 previously irradiated assembliss (all 96 Batch P and 37 Batch J assemblies). Batch J and P assemblies contain Al:03-B C burnable poison shims, while the new Batch R assemblies contain Er:03 integral burnable absorber rods, called Erbia rods, consisting of UO2 and (U,Er)O pellets. Reload batch R will consist of 28 R1 assemblies containing 32 Erbia rods,8 R2 assemblies containing 48 Erbia rods, and 48 R3 assemblies containing 72 Erbia rods. The erbium shim loading is 2.1 wt. percent Er203 In order to minimize the batch size, the feed enrichments were raised from 4.32/3.92 wt. percent U-235 to 4.42/4.07 wt percent.
Power peaking in the Cycle 9 core design is similar to the previous cycle. The power distribution behavior versus burnup of the Erbium burnable poison is similar to B4 C shim rods. The maximum Cycle 9 Fry is 1.55 at BOC which is similar to the Cycle 8 value of 1.55 which also occurred at BOC. The Cycle 9 maximum assembly relative power density of 1.40 at BOC is similar to the Cycle 8 me c.num value of 1.43.
The Cycle 9 analyses are performed using approved NRC codes and methodologies.
The nuclear aspects are appropriately considered using methods as referenced in the Methodology section of the Waterford 3 Core Operating Limits Report (COLR) and Section 4.3A of the FSAR.
Due to the minimiza' ion of feed batch size by increasing the enrichment, the Cycle 9 core potentially has a more positive moderator temperature coefficient at beginning of cycle hot zero power conditions than does the previous cycle. However, the best
. estimate value (+0.34x10" Ap/ F) is less positive than the maximum used in the safety analysis input (+0.5x101Ap/ F). The MTC at full power is sufficiently negative, as required by the Technical Specifications and stated in the COLR.
As a demonstration that CEA reactivity worth is similar to the Reference Cy_cle, scram worth (assuming the most reactive control rod stuck out) was appropriately calculated for the end of cycle full power (most limiting) main steam line break transient.
154
Comparison calculations using Entergy methodology confirmed the results are acceptable.
The power dependent control rod insertion limits were established, in part, to ensure that adequate shutdown margin exists throughout the cycle. ABB-CE demonstrated that the limits established for Cycle 8 apply for Cycle 9. Note that the Physics Databook reactivity balance provides the means to satisfy the surveillance requirements on shutdown margin. The safety analysis appropriately assumes the minimum worth of the worst stuck rod is 1.2% Ap consistent with the assumption used in developing the Physics Databook. Data covering the full range of operating
! conditions is generated for this document. This data is then used to ensure the shutdown margin requirements are met throughout the cycle.
The fuel performance of the Guardian grid fuel designs at higher Cycle 9 burnups has been evaluated using NRC approved codes (FATES 38) and all design criteria were confirmed to be met. The projected Cycle 9 maximum integrated fuel rod burnup at the safety analysis upper burnup limit is 55,322 MWD /MTU. The burnup, except for the center assembly, is lower than that of Cycle 8 and remains below the 60,000 MWD /MTU limit imposed by the ABB-CE topical repon. The burnup of fuel rods in the center test assembly will exceed 60,000 MWD /MTU and is projected to be less than 67,000 MWD /MTU. The burnup of the center test assembly is addressed in a 10CFR50.59 safety evaluation independent of this reload evaluation. The fuel rod internal pressure remains below system pressure for the projected Cycle 9 maximum burnup. The Cycle 9 burnup will be within the industry experience base.
Amendment Number 108 was made to the Technical Specifications to increase the maximum enrichment from 4.1 to 4.9 weight percent U-235 for the spent fuel pool racks, the containment temporary storage rack, and the fuel carrier when the fuel assemblies contain sufficient fixed poisons to limit the assembly reactivity to that used in the criticality analysis. The maximum reactivity of the Batch R1, R2, and R3 was shown to be less than that of the base assembly design used in the analysis (fuel containing 4.5/4.1 wt. percent U-235 with eight shims at 0.016 g B-10/ inch). The as-built stack densities and enrichments were found to be within the tolerances assumed in the safety analyses used to support the enrichment increase. Therefore, the Cycle 9 fresh fut,I can be safely storea in the spent fuel pool racks, the containment temporary storage rack, and the fuel carrier. Irradiated fuel has considerably lower reactivity than the base bundle. Thus, the Batch R fuel assemblies at the end of Cycle 9 can also be safely stored in the spent fuel pool with no restrictions.
Note that although the New Fuel Storage Vault will not be used to store new Batch R assemblies, one new Batch R assembly was repaired in the new fuel storage vault.
Criticality analyses that bound this configuration had been performed to ensure all criteria are met.
155
l The design changes which are detailed below have been considered in the above
, criticality evaluation. Therefore, they do not affect the applicability of the criticality analyses. The as built variations in enrichment and stack height density are considered by the tolerances included in the study.
The neutronic fuel design changes (slightly shorter cycle length, higher enrichment, smaller batch size, higher burnup, Erbium poison) were properly incorporated in the
, neutronics models. Appropriate methodologies were used which have been approved l
by the NRC for application to W-3 reloads. The core loading pattern and Erbium integral poison rods were configured to maintain acceptable power peaking margin throughout the cycle while meeting the energy requirements. The resultant small differences in the neutronics parameters are consistent with expected variations.
Reload Fuel Assembly Desian Chances Cycle 9 core reload differs slightly from Cycle 8 core reload due to changes in fuel management scheme and modifications in Batch R fuel design. With the exceptions that are discussed below, the mechanical design of the Batch R and Batch P reload fuel bundle assemblies are identical. The mechanical design bases have not changed since the initial core loading fuel design.
- The design of the top spacer grid assembly has been revised to contain " backup arches" adjacent to all cantilevered springs in the interior cells of the grid to reduco the variation in the proload of the cantilevered springs during grid to-rod contact following the rod loading during fuel bundle assembly fabrication. This feature,
" backup arches" adjacent to all cantilevered springs, presently exists in the rows along the perimeter in all laser welded HID 1L and GUARDIAN spacer grid assemblies. Placing " backup arches" in the interior cells will limit compression of the Orid's springs during fuel rod loading in tho manufacturing process. By limiting the amount of compression a spring experiences there will be a greater margin of preload to operate within to preclude fretting sinca there is improved contact between the grid and rod. This change to the top spacer grid assembly design is expected to eliminate the types of third-burned fuel failures at the top zircaloy spacer grid locations observed in Waterford 3 Cycle 7. The evaluation of the new top spacer grid assembly design shows that there is no impact on the grid's form, fit, function, and performance requirements. These backup arches on the top spacer prid will have negligible impact on the bundle pressure drop. Since the design criteria have not changed the top grids are fub! compliant with all internal and external product requirements.
. The laser welded HID-il and GUARDIAN
- spacer grid assemblies that were used in Cycle 8 had the guide tube openings placed in the spacer grid assembly by a coring operation (i.e., holes were cut in the grid). The HID-1L and GUARDIAN
- 156
spacer grid assemblies that will be used in Cycle 9 are coreless spacer grid assemblies. A coreless grid is a spacer grid assembly that has the five guide tube openings made by using short grid strips rather than by a coring operation. The coreless design of the HID-1L and GUARDIAN spacer grid assemblies have no impact on the grids form, fit, function, and performance requirements. Since the design criteria have not changed these grids are fully compliant with all internal and external product requirements.
- The Zircaloy 4 clad used to manufacture the Batch R fuel (UO2 ) and integral burnable absorber (UO2 -(U,Er)O2) rod assemblies will have the final stages of its manufacturing process done at AB Sandvik Steel (ABSS), Sweden. The following key factors are identical for cladding tubes manufactured at Sandvik Special Metals fSSM) of Kennnewick, Washington, USA and AB Sandvik Steel (ABSS), Sweden:
- 1. Same starting material, Wah Chang supplied TREXes.
- 2. Same tube reduction schedule in pilgering steps.
- 3. Some annealing time / temperature steps resulting in same integrated annealing parameter.
- 4. Same cladding material specification with the same requirements.
Therefore, the PWR performance of the ABSS produced tubes is expected to be identical to that of the SSM produced tubes.
. Integral Burnable Absorber Rod Assemblies, Erbia rods, consisting of UO2 and (U,Er)O2 pellets replace the Al2 orb.C burnable poison rod assemblies that have been > J 'd since the first cycle of operation. The Erbia rod assemblies with the exceps, n of the pellets use the same components as the fuel rod assemblies. The genmetry of the (U,Er)O2 Pellets is identical to the UO2 pellets. The use of erbia affects certain fuel pellet properties, such as density, thermal conductivity and expansion, specific heat, etc., that are used in the fuel performance analyses. The evaluations of the Erbia rod assemblies show that all of the design criteria for the rods and fuel bundles are met.
The thermal performance of composite fuel rods that envelope the fuel rods of the fuel batches present in Waterford Unit 3 Cycle 9 have been evaluated using the FATES 3B fuel performance computer code of the C-E evaluation model, and the Erbia burnable absorber methodology described in CEN-382-P-A,
- Methodology for Core Designs Containing Erbium Burnable Absorbers', ABB-CE Topical Report, August 1993. The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak fuel rod for each batch at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of this cycle. Predicted maximum rod internal pressures are less than system pressure,2250 psia. The cold internal rod pressure will remain below the NRC Regulatory Guide 1.25 value of 1200 psig. Results of these burnup 157
! dependent fuel performance calculations were used in the transient analyses and in the ECCS Analysis performed for this cycle.
As discussed in Section 4.1 of *Waterford 3 Cycle 9 Reload Analysis Report" and Associated References, ABB-CE Letter L-97-013, File OR-026-27 dated April 1997, the metallurgical requirements of the fuel cladding, and per review of the Bill of Materials for the two contracts the fuel assembly components for the Batch R fuel are the same as those used in the Batch P fuel presently being used in Cycle 8. Therefore, the chemical and metallurgical performance of the Batch R feel will be similar to the Batch P fuel. To date the Batch P fuel bundle assemblies have performed satisfactorily. Calculated oxide thickness value for the Cycle 9 fuel will remain below 100 microns.
Adequate shoulder gap is predicted for all of the Batches of fuel in Cycle 9. This conclusion is based on the fuel rod growth used in *Waterford 3 Cycle 9 Reload Analysis Report" and Associated References, ADB-CE Letter L-97-013, File QR--026-27 dated April 1997, it should be noted that the overall envelopes of the fuel rod assembly, the poison rod assembly and the fuel bundle assembly all remain unchanged from the previous reloads. The HID 1L Zircaloy spacer grids are the same for Batch R as was used for Batches H, J, and P. The fuel bundle assembly widths and shoulder gaps are unchanged. The CEA guide tube assembly is also unchanged.
The dropped bundle LAR388 has beon disassembled, and all damagei pellets have been replaced. CE has evaluated the impact of the assembly repair on Waterford 3 Cycle 9 reload Analyses, and concluded that there is no impact on the Cycle 9 analyses or operation of Cycle 9 as a result of this event.
REASON FOR CHANGE Cycle 9 core reload SAFETY EVALUATION The Cycle 9 core has been analyzed using NRC approved computer codes and methodologies. The results of the Cycle 9 analyses are valid under the assumptions that the required Tech Sp .c and COLR changes due to the Cycle 9 core would be in place prior to startao. This ensures that the initial conditions used in the accident analyses remain valid.
The minimum RWSP/ SIT boron concentration Tech. Spec change package (NPF 189) resulting from Cycle 9 core has been analyzed, and submitted to NRC for their 158
l approval. None of the other Tech Spec change packages submitted for Cycle 9 are We to the Cycle 9 core.
The impact of Cycle 9 on COLR has also been evaluated and all necessary changes will be in place prior to startup. The current ASI LCO limits in the COLR include a symmetric ASI measurement uncertainty which is based on the axial offset of the fuel relative to the incore detectors. The axial offset caused by the long endcap debris resistant design, resulted in an asymmetric ASI uncertainty This was previously accounted for by appropriately adjusting the COLSSICPC conste > i by implementing an offset constant in COLSS (ASIOFF) which would automatically adjust the ASI measurement (for all power levels) to yleid a symmetric ASI uncertainty compatible with that assumed in the analysis. This cycle specific values are implemented each cycle in the COLSS database (<70% power level), and as an addressable constant (t 70%
power). For Cycle 9, the COLR is changed to specifically incorporate the asymmetric ASI limits. This approach is not considered a change, it simply applies the uncertainty to the ASI value stated in the COLR rather than to the COLSS constant (ASIOFF). The ASIOFF constant for Cycle 9 will be set to zero.
The Tech Spec and COLR changes ensure that Cycle 9 will be operated within its design basis. By operating the plant in this manner, the probability of occurrence of an accident does not increase.
All Cycle 9 accident analyses (LOCA & non-LOCA) have also yielded acceptable results meeting all required acceptance criterion. This is discussed in detail under Section B2 of this evaluation.
Cycle 9 core reload fuel design differs only slightly from Cycle 8 core reload due to changes in fuel management scheme and minor modifications in Batch R fuel design compared to Batch P and J fuel design.
The probability of a fuel handling accident [FSAR 15.7.3.4) will not be increased. The assemblies with Erbia integral burnable poisons and with the minor mechanical design changes, have the same structural cage as that previously esed at Waterford 3 and will be capable of withstanding the expected handling loads (FSAR 4.2.11, and 4.2.3.1.5). These assemblies will continue to be compatible with the fuel handling equipment. The manner of handling the new fuel assemblies will be unchanged. The envelope of the new fuel is no different than that of the past. The mass of these new assemblies remains unchanged compared to the previous batch. Hence, the probability of a fuel handling accident is not increased.
No changes to the plant equipment are required for due to Cycle 9 fuel design. Since the fuel itself does not initiate any accident, there is no impact on the probability of an accident due to the Cycle 9 fuel. Therefore, the probability of an accident previously 159
ovaluated in the FSAR will not be increased due to the Cycle 9 Fuel Management or Reload Fuel Assembly Design changes.
i As documented in the Cycle 9 Reload Analysis Report, ABB CE has reviewed all the accident analyses to determine whether these events are bounded by the Reference Analysis or need further evaluation and/or reanalysle. The non-LOCA Design Basis Events (DBEs) were evaluated with respect to four criterion:
- 1. Offsi's Dose,
- 2. Reactor Coolant System Pressure,
- 4. Loss of Shutdown Margin.
!n most cases, comparison of key input parameters between Cycle 9 and the previous cycle determined that Cycle 9 inputs were bounded by the previous cycle inputs, thus ne reanalysis was required (note that if a particular analysis is bounded by its previous cycle, it would also be bounded by the AOR/ Reference Cycle). However, ABB-CE performed specific analyses for all events for which comparison of key input parameters could not demonstrate that the Cycle 9 results would be bounded by the Referonce Analyses recults. These events include:
Increased Main Steam Flow with Loss of Offsite Power (Excess Load)
With respect to Offsite Dose, and Fuel Performance criterion
. Main Steam Line Break (Pre-Trip Power Excursion)
With respect to Offsite Dose, and Fuel Performance criterion
. Single Reactor Coolant Pump Shaft Seizure / Sheared Shaft With i spect to Offsite Dose, and fuel Performance criterion Uncontrolled CEA Withdrawal from Subcritical or Low Power With respect to RCS Pressure, and Fuel Performance criterion CVCS Malfunt. tion (Inadvertent Boron Dilution)
With respect to Shutdown Margin criterion
. Startup of an inactive Reactor Coolant Pump Event With respect to Shutdown Margin criterion
. Inadvertent Opening of a Steam Generator Atmospheric Dump Valve With respect to Offsite Dose, Fuel Performance, and Shutdown Margin critorion 160
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All other non LOCA e ents have less limiting core physics input for Cycle 9, and therefore, are baur.ded by the Reference Cycle (FSAR) analysis.
An ECCS performance analysis of the limiting large and small break size LOCA events was also performed. These events were analyzed to demonstrate compliance with 10 CFR 50.46, the NRC Acceptance Criteria for Light Water Nuclear Power Reactors, and will be discussed in the ECCS analysis scetion.
The analysis results for each of these events are discussea below:
Increased Main Steam Flow With loss of Offsite Power (Excess Load):
This event required analysis due to changes in pin census for Cycie 9. The event is analyzed to demonstrate that the amount of fuel failure remains less than the Reference Cycle analysis. The evaluation showed that the fuel failure due to the Increased Main Steam Flow in combination with a Loss of Offsite AC Power would result in 2.80% of the fuel pins experiencing DNB. This value is acceptable as it is smaller than that reported in the Reference Cycle (2.91%).
Conclusion:
This event meets all applicable offsite dose and fuel performance criterion as it remains bounded by the Reference Cycle.
Main Steam Line Break (Pre-Trio Power Excursion):
This event required analysis due to changes in the Cycle 9 pin census. This event, concurrent with loss of AC power, is analyzed to evaluate the maximum number of calculated pin failures for the site boundary dose calculation.
The Cycle 9 analysis used the criterion that all fuel rods for which CE-1 DNBR falls below the 1.26 SAFDL are assumed to experience cladding failure. The analyses are done with both an accident initiated lodine spike (doses must be a small fraction of 10CFR100 limits) and with a pre-existing iodine spike (doss must be within 10CFR100 limits). The results of the inside and outside containment cases for Cycle 9 indicate that 3.99 percent (Reference Cycle calculated 4.5 percent) of the fuel pins fail for the inside containment breaks and 1.19 percent (Reference Cycle calculated 3.0 percent) of the fuel pins fail for the outside containment steam line breaks. Since the predicted fuel failure for Cycle 9 is less than that of the Reference Cycle, the Reference Cycle offsite doses remain vLiid.
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The results of the pre trip analysis of SLB with LOCA for Cycle 9 demonstrates that a coolable geometry is maintained during this event as the number of fuel pins calculated to fail is small. Therefore, the conclusions for this event in the FSAR remain valid for Cycle 9.
Conclusion:
This event meets all applicable offsite dose and fuel performance criterion as it remains bounded by the Reference Cycle.
Sinole Reactor Coolant Pump Shaft Seizure / Sheared Shaft:
The single reactor coolant pump (RCP) shaft seizure / sheared shaft with Loss of Offsite AC power was rear.alyzed due to a change in Cycle 9 fuel failure pin census, and a more severe flow coastdown curve. The event was reanalyzed to ensure that a coolable geometry is maintained and that the site boundary doses do not exceed a small fraction of 10CFR100 guidelines. Both the RCF shaft seizure and sheared shaft were evaluated for Cycle 9. LOAC was evaluated for each, and the sheared shaft was determined to be more limiting with respect to fuel performance than the RCP seized rotor event.
The sheared shaft with LOAC results in a predicted fuel failure less than *he maximum allowable fuel failure of 7.6 %. This fuel failure limit ensures that the resultant offsite doses are less than 30 REM thyroid and less than 2.5 REM whole body. Additionally, the peak RCS pressure is less than 2750 osia and a
, coolable geometry is maintained. For the sheared shaft with the LOAC, the radiological doses are less than 10% of the 10CFR100 limits of 300 REM thyroid dose and 25 REM whole body dose.
Conclusion:
This event meets all applicable offsite dose and fuel performance criterit a as it remains bounded by the maximum allowable fuel failure of 7.6 %.
Uncontrolled CEA Withdrawal from Subcritical or Low Power The uncontrolled CEA withdrawal (CEAW) from subcritical conditions is analyzed to ensure that the departure from nucleate boiling ratio (DNBR) and the fuel centerline melt (CTM) specified acceptabin fuel design limits (SAFDLs) are not violated. Additionally, the CEAW from low powers is analyzed to verify that the peak RCS pressure is less tha,i the design limit of 2750 psia. The CEAW frum suberitical conditions is included due to an assumed increase in the suberitical reactivity addition rate and 3-D peaking factor. The CEA Withdrawal from Inw power conditions did not require reanalysis for the current cycle since its inpui wes bounded by the Reference cycle.
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The uncontrolled CEA withdrawal from suberitical conditions resulted in the minimum calculated CE-1 DNBR for this event to be greater than the design limit of 1.26. The peak linear heat generation rate (PLHGR) was calculated to be in excess of the steady state acceptable fuel centerline melt (CTM) limit of 21 kW/ft. The steady state acceptable PLHGR of 21 kw/ft is allowed to be exceeded for short durations in dynamic situations as long as the deposited energy in the fuel does not cause centerline melting. Because the power spike for this event is terminated in such a short time, the amount of heat in the fuel is L not enough to cause melting. Fuel centerline temperature did not exceed 4900
- F and fuel melt was not predicted to occur. Additionally, the peak RCS pressure remained less than the design limit of 2750 psia.
Therefore, an uncontrolled CEA withdrawal from suberitical power conditions will not exceed the DNBR or CTM limits.
Ccoclusion: This event meets all applicable RCS pressure, and fuel performance criterion as it remains bounded by the Reference Cycle.
CVCS Malfunction (Inadvertent Boron Dilutioni The CVCS Malfunction (Inadvertent Boron Dilution) was reanalyzed due to changes in the Critical Boron Concentration, inverse Boron Werth, and Stuck Rod Worth. The event was reanalyzed to ensure that the automatic Boron Dil!1 ion alarm will provide the required advance warning to the plant operators to an ongoing Boron Dilution at least 15 minutes prior to the loss of Shutdown Margin for operation in Modes 3-5 and 30 minutes for Mode 6. The results of the CVCS Malfunction analysis verify that utilizing Cycle 9 specific CBC and IBW values and the following revised minimum alarm reset times:
- 1) No sooner than one half hour after shutdown
- 2) At least once per hour if the reactor has been shut down < 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
- 3) At least once per 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> if the reactor has been shut down 210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> but <
25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />
- 4) At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the reactor has been shut down 2 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> but
< 21 days
- 5) At least once per week if the reactor has been shut down 2 21 days The automatic boron dilution alarm will provice the require'd advance warning to the plant operators. Tne Boron Dilution scenario need not be considared during 163
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! the initial half hour after reactor trip since the shutdown margin immediately after trip is much greater than the minimum amount assumed in the analysis.
The analysis also provides updated COLR Surveillance Tables to detect an on-going Boron Dilution and terminate the transient before the core reaches criticality, assuming failure of the Boron Dilution alarm (s)
The CVCS Malfunction (Inadvertent Boron Dilution) analysis concludes that for Modes 3-6, ensuring the above revised minimum alarm reset times and COLR Surveillance Table requirements are properly met, an inadvertent Boron Dilution to the RCS will not allow sufficient positive reactivity addition to the core to lose required shutdown margin.
Conclu'on: By updating COLR reset times and Tables 1 through 5, this event meets all applicable shutdown margin critorion.
Startup of an inactive Reacter Coolant Pump Event The Startup of an Inactive Reactor Coolant Pump Event was reanalyzed due to 9anges in ITC, and a mere reactive core. The event was reanalyzed to ensure 9 at the core remains subcriticc. ~nd fuel design limits are not exceeded during an inadvertent startup of a reactc, coolant pump during operation in Modes 3-5.
Two cases were evaluated to determine the reactivity addition. The first case is defined as the set of conditions where the saam generator temperature is less than the RCS temperature. The second case is defined as the set of conditions where the steam generator temperature is greater than the RCS temperature.
The more limiting case of the two, in terms of reactivity addition, is when the steam generator temperature is less than the RCS temperature with a negative isothermal temperature coefficient. Technical Specification 3.1.1.2 states that the shutdown margin must always be greater than the COLR Figure 1 value when all rods are inserted. This shutdown margin combined with the worth of the most reactive stuck rod is greater than the positive reactivity addition due to the inadvertent startup of a reactor coolant pump. The available shutdown margin is thus sufficient to preclude criticality due to an inadvertent reactor coolant pump startup.
An inadvertent Reactor Coolant Pump Startup event will not add sufficient positive reactivity to the core to exceed the required shutdown margin.
Therefore, the reactor will remain subcritical and the specified acceptable fuel design limits are not exceeded, thus maintaining clad integrity.
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Conclusion:
This event meets all applicable shutdown margin criterion.
Inadvertent Openina of a Steam Generator Atmospheric Dump Valve at Full Power with Loss of Offsite Power The IOADV in combination witl. a Loss of Offsite AC Power (LOAC) and the associated coast down of the reactor coolant pumps (RCPs) is an event for which the number of potential fuel failures is calculated for the performance of a radiological consequences calculation. Two analytical steps were taken to demonstrate the Full Powei RADV with LOAC Cycle 9 results were no more adverse than the Zero Power IOADV with LOAC results given in the FSAR.
These steps were:
- 1. Continued use of the 1-D HERMITE code to model the flow coast down -
reactor trip portion of the event. The NRC has been informed in past reloads that HERMITE is used when more detail of the core performance is necessary.
- 2. Crediting of the CPC Low RCP speed trip after the beginning of the coast down of the RCPs.
Sufficient thermal margin is preserved by COLSS and/or other Limiting Conditions for Operation to ensure that the increased Heat Removal event caused by the inadvertent Opening of an Atmospheric Dump Valve (IOADV) will not cause violation of the DNBR SAFDL. This requirement is met for Cycle 9.
The Full Power IOADV in combination with a Loss of Offsite AC Power will not result in DNBR SAFDL violation for Cycle 9. Therefore, the results are bounded by the Zero Power IOADV in combination with Loss of Offsite AC PoNer analysis.
No fuel failure due to the Full Power inadvertent Opening of an ADV in combination with a Loss of Offsite AC Power is predicted to occur for Cycle 9.
This implies that the radiological consequences of the Full Power IOADV with LOAC are no more adverse than the Zero Power IOADV with LOAC present in the UFSAR.
Conclusion:
This event meets all applicable criterion with respect to Offsite Dose, Fuel Performance, and Shutdown Margin.
ECCS Analvt An ECCS performance analysis of the limiting large break size LOCA was performed due to changes in fuel design / performance (i.e. higher enrichment, 165
erbia,...), and the minimum containment temperature reduction These events were analyzed to demonstrate compliance with 10 CFR 50 46, the NRC Acceptance Criteria for Light Water Nuclear Power Reactors. The input data for Cycle 9 small break LOCA is bounded by the 'rence Analysis and thus, the results continue to meet acceptance criteria.
Laroe Break Loss of Coolant Accident (LBLOCA)
An ECCS performance analysis of the limiting large break size was performed for Waterford 3 Cycle 9 to demonstrate compliance with 10CFR50A6, the NRC Acceptance Criteria for Light Water Nuclear Power Reactors. The analysis justifies a maximum allowable Peak Linear Heat Generation Rate (PLHGR) of 13.2 kw/ft. This PLHGR is a 0.2 kw/ft reduction from the Cycle 8 limit for Waterford 3. The Cycle 9 analysis allows the removal of the 0.2 kw/ft penalty to COLSS addressable constant T42 to account for lower containment temperature. This penalty may be removed since the new analysis is done at the reduced containment temperature of 90 'F.
The following table provides the NRC acceptance limits, Reference Analysis (performed in support of SIT level and pressure range expansion), and Cycle 9 results for the ECCS analysis for the limiting largs break LOCA:
NRC Acceptance Ref.
Limit Analysis Cycle 9 Peak Clad Temperature, 'F 2,200 2,177 2,170 Maximum Local Oxidation 17 % 8.55 % 8.39 %
Core Wide Oxidation 1% <0.805% <0.805%
The above table shows that the Cycle 9 results are bounded by the Reference Analysis, and meet the NRC acceptance criterion.
COLSS/CPC Marain Events Certain events, primarily AOOs, e.g., CEA drops, CEA deviations within CPCs deadband are analyzed to obtain the required overpower margin that needs to be set aside in COLSS/CPC to prevent fuel failure for AOOs. Margin and setpoint requirements resulting from the transient analyses are incorporated into COLSS/CPC setpoints. Database and/or addressable constants will be modified as needed prior to cycle startup to ensure acceptable results.
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I Fuel Manaaement Chanaes The characteristics of the Cycle 9 fuel assemblies are bounded by the assumptions used in the spent fuel pool rack, temporary fuel storage rack, and fuel carrier criticaiity analyses. As the criteria are not exceeded, the consequences of any accident previously evaluated in the FSAR will not be increased for the Cycle 9 fuel assemblies.
Criteria have been established for moderator temperature coefficient (MTC) at various power levels to ensure that the consequences of various accidents are acceptable. The Cy'.le 9 MTC values have been evaluated and have been determined to meet these criteria. As these criteria are not exceeded, the consequences of the accident will not be increased for the Cycle 9 fuel.
The evaluation of the consequences of these accidents appropriately accounts for the changes due to the higher enrichment and the Erbia integral burnable absorber rods of the Batch R fuel assemblies.
Reload Fuel Assembly Desian Chanaes The thermal performan' ' composite fuel rods that envelope the fuel rods and Erbia rods of the fuel L as present in Waterford Unit 3 Cycle 9 have been evaluated. The analysis was performed using a power history that enveloped the power and burnup levels representative of the peak fuel rod for each batch at each burnup interval, from beginning of cycle to end of cycle burnups. The burnup range analyzed is in excess of that expected at the end of this cycle.
Predicted maximum rod internal pressures are less than system pressure, meeting FSAR section 4.2.1.2.1, " Fuel Cladding Design Limits", requirement for the peak rod internal pressure.
The cold internal rod pressure (640 psic) will remain below the NRC Regulatory Guide 1.25 value of 1200 psig. The mass of the Batch R assemblies is the same compared to the previous batch. Hence, the number of fuel pins that will fail during a fuel handling accident will not be more than the present analyzed pin failures. Therefore, consequences of a dropped bundle accident (FSAR 15.7.3.4.2.2 and 15.7.3 4.5] are also not increased.
These Batch R fuel assemblies, with Erbia integral burnable poisons and with the minor mechanical design changes, have the same envelop, materials, dimensions and the same structural cage as that previously used at Waterford 3. Adequate shoulder gap is pr idicted for all of the Batches of fuel in Cycle 9. The chemic'l and metallurgical performance of the Batch R fuel will be similar to the Batch P fuel. As such, no change will occur in the radiological release rate / duration, no new release 167
mechanisms can be postulated, and no impact will occur to any radiation release barriers. Therefore, the consequences of any accident previously evaluated in the FSAR will not be increased because of the use of Batch R fuel assemblies.
The use of these design change features in Batch R assemblies has not significantly altered the enrichment and burnable poison loading scheme of pellets and rods in a fuel assembly, nor the low-leakage loading scheme of fuel assemblies in the Cycle 9 core. ABB/CE has successfully provided full batch application of these design features to five other Combustion Engineering plar'ts. There are no required changes in any vendor's quality control procedures, quality surveillance programs, or fabrication processes to ensure correct loadings of fuel and burnable poison in assemblies and in the core. Therefore, the probability of erroneous loading of fuel pellets or fuel pins of different enrichment in a fuel assembly or erroneous placement or orientation of fuel assemblies in the core (FSAR 15.4.3.1) duo to these aesign change features in Batch R assemblies is not increased.
The probability of fuel failure due to mechanical or flow induced vibration and fretting with the spacer grids [FSAR 4.2.1.2.1.g,4.2.3.1.1,4.2.3.1.3,4.2.3.2.1 and 4.2.3.2.4) will not be increased. The Batch R fuel assemblies with Erbia integral bumable poisons and with the minor mechanica; design changes have the same structural cage as the previous reioad. Its fuel rods and poison rods have the same external dimensions, materials, clad thickness, and approximate mass as the Cycle 8 rods.
The probability of CEA misoperation [FSAR 15.4.1.4) is not increased. The dimensions and positions of the CEA guide tube assemblies are unchanged compared to the assemblies used in the previous cycles. Also, any dimensional changes due to irradiation, such as assembly bow, will not be altered since no changes in the guide tubes material have occurred.
All equipment important to safety will function in the same manner with the Cycle 9 reload core as with the previous core. Thero is no characteristic of the Cycle 9 core, with the Batch R reload assemblies containing Erbia integral burnable poisons and with the minor mechanical design changes, different from the cores from previous cycles that would increase the probability of a malfunction of equipment important to safety.
Therefore, the probability of occurrence of a malfunction of equipment important to safety is not increased due to Cycle 9 core reload, or the utilization of Erbia as a burnable absorber.
No equipment or operadonal change is caused by the Cycle 9 fuel due to the increased enrichment or the utilization of Erbia integral absorber. All equipment important to 4
safety will function in the same manner with the reload core as with the previous core.
The function and duty of the equipment important to safety is not altered. No changes in the assumptions concerning equipment availability or failure modes are made. Thus, the consequences of a malfunction of equipment important to safety are not increased.
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The minor fuel assembly design mechanical changes and the Erbia integral burnable absorber rods do not increase the consequences of a malfunction of any equipment important to safety.
The FATES 3B fuel performance analysis for the Cycle 9 reload assemblies containing Erbia integral burnable poisons has demonstrated that no change will occur in the '
radiological release rate / duration, no new release mechanisms can be postulated, and no impact will occur to any radiation release barriers.
There are no new system interactions or connections associated with W-3 Cycle 9 core reload with the Batch R reload assemblies containing Erbia integra! burnable poisons and with the minor mechanical design changes. The changes associated with the higher fuel assembly enrichment and utilization of Erbis burnable absorber will not require new equipment or alter the way in which the plant operates. No changes in the failure modes of the equipment impor! ant to safety were assumed in these analyses.
No initiators to any of the accidents are impacted. Therefore, operation of Waterford 3 with the Cycle 9 reload core will not cause an accident of a different type than any previously evaluated in the FSAR.
Installation of a reload core cannot cause the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the F'3AR.
Equipment important to safety will function in the same manner with a reload core as with the previous core. The impact of changes in core characteristics on any parameter that would affect the function of equipment important to safety, has been accounted for in the analysis. The Waterford 3 testing and verification program ensures that all required calibrations and setpoint changes resulting from Cycle 9 core are performed. There are no new methods of failure associated with any of the changes identified previously for Cycle 9.
There is no new equipment associated with the use of Batch R fuel assemblies containing Erbia integral burnable poisons and with the minor mechanical design changes. No new systems or substructures are involved. The changes will not alter the way in which the plant operates. No changes in the failure modes of the equipment important to safety, including the fuel were identified in the Cycle 9 analyses.
Therefore, the possibility of a malfunction of equipment important to safety of a different >
type than any previously evaluated will not be created due to the fuel management or reload fuel assembly design changes.
All accidents have been shown to have results within the appropriate NRC acceptance limits. There is no reduction in any margin of safety as defined in the bases of any Tech Spec.
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COLSS/CPC setpoints will be modified to ensure acceptable results for relevant transients.
The fuel performance of both the Guardian (Batch P assemblies and Batch R reload assemblies which contain Erbia integral burnable poisons) and non-Guardian (Batch J) grid fuel designs at the Cycle 9 burnups has been evaluated using NRC approved codes (FATES 38) and all design criteria were confirmed to be met. The maximum cladding plastic strain will remain below 1.0% within the anticipated fuel assembly burnup, and fuel melt will continue not to occur, thus the margins of safety are not reduced.
The Cycle 9 core was evaluated using NRC approved computer codes and methodologies. All design criteria were confirmed to be met. The core loading pattern, and use of Erbium integral poison rods were shown to maintain acceptable power peaking throughout the cycle with no degradation in margin of safety. Therefore, the margin to safety will not be reduced for the Cycle 9 core due to fuel management or reload fuel assembly design changes.
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- 26. Enaineerina Reauest ER-W3-97-0064-00. Evaluation of Plant Fire Doors DESCRIPTION Provide an evaluation to allow generic application of UL Field Report 84NK2996 to all plant fire doors listed in FSAR Table 9.5.1-1.
REASON FOR CHANGE Corrective action program identified the need for_the UL report to bound all plant fire doors.
SAFETY EVALUATION This evaluation demonstrates the acceptability of the generic application of the UL ,
Report to all plant fire doors and further provides no adverse impact to the previously approved fire protection program and features. There is no increase in the probability or consequences of an accident or equipment failure, and no new failure mechanisms are created. The margin of safety provided by the fire protection program has been maintained.
}~
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- 27. Calculations for DC Battery Systems A. B. AB. and TGB DESCRIPTION The proposed change consolidates all outstanding items associated with the A, B, AB, and TGB DC battery systems. Open items include the modifications made as a result of DC-3402 and revisions to tables and figures as a result of incorporating design changes, plant changes, SPEERs, and NCis into calculations.
REASON FOR CHANGE Battery calculations have been revised to incorporats design changes, plant changes, SPEERs, and NCis. As a result of these incorporations, information associated with the battery systems in the UFSAR requires updating.
SAFETY EVALUATION The function, rating, and capacity c'the A, B, AB, and TGB battery systems have been previously evaluated and reviewed. The changes to the battery systems enhance their performance capabilities.- All battery systems mentioned, both non safety and safety related, will continue to perform their functions. Therefore, it has been determined that no unreviewed safety question exists.
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- 28. SPEER 9501423. Replacement of EGA-136A(B) and EGA-137A(B)
DESCRIPTION The SPEER replaces the existing 1" piston check valves EGA-136A(B) and EGA-137A(B), inlet to Emergency Diesel Generator (EDG) Air Receivers, with 1/2" in-line (nozzle) check valves.
REASON FOR CHANGE implementation of the SPEER willincrease the reliability of the check valves to hold tight. This will ensure that the EDG Air rieceivers maintail the required air supply during and after an accident.
SAFETY EVALUATION According to the safety evaluation the replacement of the check valves will not inuease the probability of an accident or affect the consequences of an accident. The change willincrease the reliability of the EDG starting air system. The new valves have been correctly sized and supplied with soft seats to ensure the proper function of preventing backflow from the EDG air receivers.
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- 29. SPEER 9501467 Replacement of EDG Cool Down Trin Circuit Check Valves DESCRIPTION Existing EDG Cool Down Trip Circuit Check Valves (NUPRO model B-4 CPS-1) will be replaced with NUPRO model SSF90 CHM 4F48U-1 check valves.
REASON FOR CHANGE The existing NUPRO model B-4 CPS-1 check valves have been the cause of several inadvertent diesel engine cool down trips. The replacement check valves have an 1 integral filter to prevent debris from becoming lodged in the seats and preventing a j positive shutoff.
SAFETY EVALUATION According to the safety evaluation, replacement of these check valves will not have an
[ adverse affect on operation of the EDG's since the cool down trip circuits are bypassed in the emergency mode. This chmge should decrease the probability of an inadvertent EDG trip by providing filters which should prevent debris from lodging in the valve seat and preventing a positive shut off.
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- 30. SPEER 9701667. NUKON Blanket Insulation - Reactor Vessel Closure Head DESCRIPTION This is an evaluation of the use of NUKON thermal blanket insulation to replace the original Transco insulation on the reactor vessel closure head. The scope of the evaluation is limited to include only the peripheral region on top of the head (from the 58" radius outward inside the cooling shroud) where the original insulation was designed and identified for removability.
REASON FOR CHANGE This change supports ASME Section XI code inspections of the reactor vessel head required during Refueling Outage 8. It allows replacement of any or all of the insulation within the evaluated scope that may be damaged during the inspections.
SAFETY EVALUATION NUKON insulation has been previouW , valuated and installed on other components at Waterford 3, including the pressurizer, steam generators, and reactor coolant pumps.
Its use on the reactor vessel closure head will not impact any safety features described in the FSAR, will not reduce any margin of safety as defined in the basis for any Technical Specification, and will not create an unreviewed safety question. Insulation does n_ot perform any safety-related function, has no impact on any accident or
- important-to safety equipment described in the FSAR.
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- 31. SPEER 9701684. Wet Coolina Tower Nozzle Replacement DESCRIPTION Replacement of the PVC spray nozzles in the Auxiliary Component Cooling Water (ACCW) system with stainless steel.
I REASON FOR CHANGE Many of the PVC nozzles have cracked and failed. The change to stainless steel is a better overall design that will add strength and life extension to the nozzles.
SAFETY EVALUATION The nozzles do not initiate any accidents; therefore, their replacement will not increase the probability of an accident. Changing nozzles will result in a smaller droplet.
Analysis has shown that this will result in better heat transfer capability but will not issu(t in an increase in drift losses. Thus, no accident consequences are affected by
/} the change. In addition, the new nozzles have been evaluated and shown to have no g adverse affect on the ability of the related systems to perform their intended safety E functions. No protective boundary or margin of safety is affected by the change and there are no unreviewed safety questions.
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I G. COMMITMENJ_QHANGES 1; ppgjpjdefety Evaluations included in PORC Meetina Minutes During conduct of NRC Inspection 89-07, the NRC discovered that the Safety Review
' Committee (SRC) had not reviewed four 10CFR50.59 safety evaluations. Notice of Violation 89-07-01 was issued. In response to that NOV, Waterford 3 committed to '
include 10CFR50.59 safety evaluations as attachments in the Plant Operations Review Committee (PORC) meeting minutes.
In 1989, UNT-001-004 was revised to require that 10CFR50.59 safety evaluations be sent to the SRC as attachments to the PORC meeting minutes. This eventually caused SRC to receive 10CFR50.59 safety evaluations via two methods: (1) the PORC Secretary submitted 10CFR50.59 Safety evaluations to SRC in accordance with Attachment 6.3, item ' cf UNT-001 -004 and (2) Safety Review submitted 10CFR50.59 safety evaluations ' R after preparation of PORC meeting minutes. Thus, the SRC received two copies " ...a same evaluation. In an effort to eliminate duplication, Safety Review pe,1ormed a 10CFR50.59 safety screening that inconectly justified not including 10CFR50.59 safety evaluations as attachments to the PORC meeting minutes. Condition Report 96-0886 was written to document that discrepancy. Since $
Attachment 6.3, item 1, of UNT-001-004 (copy attached) requires the PORC Secretary to submit all 10CFR50.59 safety evaluations to the SRC, there is no need to include the evaluations as attachments to the PORC meeting minutes. Furthermore, this CCEF demonstrates that the change has no safety significance.
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- 2. QA Manaaer Review of Procedures That Implement the QA Proaram in October 1984,2._e NRC issued Notice of Violation (NOV) 8431-01 because the QA Manager (contrary to the requirements of the QAPM) did not select, review and concur with those procedures that implement the OA Program, As corrective act;on, Waterford 3 established a list of procedures that require QA review and reiterated current program requirements: (1) quality-related nuclear services procedures, quality-related management procedures and plant quality procedures are reviewed by QA, and (2) the Corporate Quality Assurance Group is required to sign the cover page of quality-related NSP's and PMP's and the Plant Operations Review Committee review sheet for plant Quality Procedures. Section 4.2 of QAPM Chapter 5, " Instructions, Procedures and Drawings," states that the Quality Assurance manager is responsible for assuring that a quality review of safety related Waterford 3 procedures, instructions, drawings, and specifications is conducted to insure the inclusion of applicable Quality Assurance Program requirements. The QAPM does not require the review to be conducted by QA.
Prior to the development of the Commitments Management System (CMS), the best way to ensure the inclusion of applicable QA program requirements was to have QA review all safety-related procecures. Since the development of CMS, a separate review by QA is redundant, inefficient and unnecessary. Section 4.1 of QAPM Chapter 5, ' Instructions', Procedures, and Drawings", states that directors and managers are responsible for ensuring that Waterford 3 commitments and obligations are addressed in safety-related procedures for which they are responsible.
The QAPM is not affected by this change. Reviews are being conducted to ensure the inclusion of QA program requirements and the QA organization periodically reviews the adequacy of those reviews. Therefore, this change is justified.
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3.- Testina the Post-Accident Samplina System The commitment text states that the Post Accident System (PASS) at Waterford 3 will be tested every six months by obtaining a Reactor Coolant System (RCS) sample through the PASS and comparing the results with a concurrent RCS sample obtained by normal means. At the same time, on-line instrumentation will be calibrated and tested. This commitment as stated in letter W3P83-2141 was generated to address the piovisions of Criterion 9 of NUREG-0737, item II.B.3. This commitment is reiterated in SSER 06 ll.B.3.
-Proposed change is to increase the testing frequency of the PASS system as stated above from 6 months to 7 months. This is only a one time change and all subsequent testing will resume to its regular interval of 6 months. The present testing method requires that non nuclear safety portions of the CCW be aligned to both loops of the Safety portion of CCW. Under current plant operating philosophy, when a non safety portion of a safety class portion of a system with no automatic isolation function is aligned to the safety po-tion of the system, the system must be declared inoperable and the appropriate LCO must be entered. Current testing method would require both loops of CCW to be declared inoperable which places the plant in T.S. 3.0.3 An attemate testing method is being evaluated which will precluda the plant from entering T.S. 3.0.3.
This evaluation will be completed approximately 2 weeks after the current six month testing period expires. Therefore a one time change in testing frequency from 6 months to 7 months is proposed so that the evaluation of the alternate testing method can be completed and the new testing method put into effect.
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- 4. Evaluation of Emeraency Classifica' ion D.acedure in response to an NRC observation as documented in IR 96-06 involving an activity in en Emergency Plan drill, the Emergency Planning department committed to evaluate the emergency classification procedure to determine if additional enhancements can be made in identification of fuel cladding failure criter:. . This activity is scheduled to be completed by 12/31/96.
The proposed change is to change the scheduled completion date of this activity from 12/31/96 to 4/1/97. This change is requested because this activity, although started, will not be completed per the original scheduled date of 12/31/96. This commitment was made to address an NRC observation which does not require a docketed response. This commitment is an enhancement which by delaying it does not reduce the effectiveness of the emergency plan organization or i+4 activities.
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- 5. Incorporate Firewatch into the Manaaement Observation Prooram As a result of insocetion IR 93-10, a commitment was made to incorporate Firewatch into the Manegement Observation Program. This commitment is being deleted based on implementation of the original commitment. Firewatch patrols were incorporated into the Management Observation Program effective June 1,1993. In March,1996, a management decision was made to eliminate the formal process of management observations and management encouraged daily observations per letter from the General Manager dated 3/20/96. The firewatch program has been enhanced and increased emphasis ensures adequate implementation of firewatch requirements. In addition, a Quality Assurance Surveillance Commitment adequately observes the Firewatch Program, i.
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- 6. Makeup and Function of the Condition Review Board The Condition Review Board will function to review all Condition Reports (CR's) and Condition Identification (CI's) on the front end to ensure proper priority and dedication of resources. Membership will be General Manager Plant Operations, Director of Nuclear Safety, Manager of Licensing, Manager of Operational Experience Engineering, Director of Design Engineering.
The review of Condition identifications (Cis) is being transferred to the Condition Review Committee. These reviews will be conducted by responsible department (mostly management) personnel during this Operations chaired meeting. It is believed that this level of review is sufficient since these are the individuals most likely to know if and when a Condition Report needs to be generated concurrently with a Condition identification. Additionally, recent experience has shown that the review of Condition Identifications has produced very few additional Condition Reports and those were of low or no safety significance.
The membership of the CRB has also changed. This portion of the commitment is being revised to indicate that the General Manager Plant Operations will appoint members to the CRB. This will facilitate organization and functional changes while still onsuring a high management presence.
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- 7. Retrieve Missina Steam Generator Pluas This commitment was made to identify and retrieve steam generator plugs which were identified missing in December,1986. The commitment text states, 'No further action is planned other than removal of the remaining four loose plugs when they can be retrieved, most likely during the 10 year inservice inspection of the Reactor Vessel
. intervals Of the original five plugs missing, only one and a partial plug have not been located and retrieved. The 10-year ISI of the reactor internals did not reveal the plug and piece. A question was raised as to whether additional searches for the plugs are warranted in the future. An evaluation was performed by Design Engineering stating that the remaining plug and partial piece do not creata a saiety. This evaluation is supported by the original evaluation of the missing plugs performed in 1986. Design Engineering determined that the remaining plug and partial piece could remain in the RCS until the end of plant life with no safety concein. The intent of this commitment is met in that the plugs which were identified and retrievable have been removed. No further searches for the remaining items is necessary.
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- 8. Completion of EQ Data Record Form The commitment specifies that "these procedures require that the EQ data record form from procedure MD-001-020 be completed when any maintenance is performed on EQ equipment. The EQ data record forms identify the EQ requirements for each piece of equipment."
The two main reasons to clarify the intent of the commitment are: 1) Some EQ equipment has no maintenance actions that are required to maintain qualification. This leads to a form being filled out that contains little more than the component number. In this case, the form verifies nothing and serves no purpose (no value added). 2) There are times when work is being performed which involves EQ equipment, but the work neither fulfills an EQ maintenance function nor will be affected by EQ requirements. In cases such as these, every single line on the form will be marked "N/A". Again, this does not add value nor does it document that EQ status has been preserved.
The commitment is revised to read: "...these procedures require that EQ data record form from MD-001-020 be completed when any maintenance which is necessary for maintainina EQ status is performed on EQ equipment. The EQ data record forms identify the EQ requirements for each piece of equipment, should reouiremer,ts exist."
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- 9. Use of " Auto Seouential" Control Rod Insertion At Waterford 3, procedure OP-010-001, ' General Plant Operations' specifies the use of Boration or ' Manual Sequential" control rod insertion to affect a unit shutdown. After ihe March 1996 INPO visit, Operations committed to establish a Rapid Plant Power Reduction section to our General Plant Operating Procedure, OP-010-001. One method of performing the Rapid Plant Power Reduction will be by using the ' Auto Sequential' rod control method. However, Information Notice 83-075 was concerned with operators following approved rod sequence procedures. ' Auto Sequential' rod control meets this commitment due to the fact that the Reactor Regulating system will be controlling rod movement in accordance with plant design. The control rods will move in the exact same sequence as " Manual Sequential' except the Reactor Regulating System will be moving control rods instead of an operator moving rods in ' Manual Sequential'.
OP-010-001, ' General Plant Operations' will be revised to allow "Boration, ' Auto Sequential' control rod insertion, or ' Manual Sequential' control rod insertion can be used to affect a unit shutdown."
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- 10. Trainina and Qualification Requirements for Health Physics Contract Personnel LER 92-011 identified that a Technical Specification required sample was obtained late.
- as a result of personnel error. Training and qualification requirements for contract personnel were evaluated and a commitment was made to ensure that contract personnel, working in the health physics count room, were aware of the importance of obtaining tech spec required samples within the frequency required by the applicable action statement. This was incorporated into procedure NTC-230.
Due to a recent organizational change involving the health physics and chemistry departments, there is no longer a need to bring in contract staff to work in the count room. The affected sections of NTC-230 are no longer applicable since the health physics contractor count room qualification cards have been deleted from the health physics training course descriptions.
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11, Departmental Self Assessments to include Desian Enoineerina At the ACCW Enforcement conference on March 5,1996, a slide was shown that called out departmental self assessments. These later became defined as assessments performed to a specific questionnaire developed by Prism and issued to the design department manager's for their use on June 20,1996. This became item 1.d of the Focus Plan. Since that time, significant system and program self assessments have ,
been perurmed by W3. These were coupled with major NRC inspections and the most recent corporate design and licensing basis assessments and 10CFR50.54(f) response. As a result of the many recommendations and commitments made in the engineering programs area, the departmental self assessments, as originally planned, will be modified to mean compliance in accordance with the applicable program enhancements detailed in the corporate design basis recommendations issued in the final report on Feb. 7,1997. Also, confirmation of this will be handled administratively prior to the end of the year by the manager when he reports to the Director of Design Engineering during the performance planning and review process.
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- 12. Shelf Life of Okonite Electrical Tape and Nuclear Splice Cement During construction, a commitment was made that the shelf life of Okonite #35, T95 electrical tape and nuclear splice cement would be posted at each storage location.
Limited shelf life material is controlled by procedure SSP-827. The procedure describes the control of limited life material while in storage and ensures that material with expired shelf life is not issued without an engineering evaluation. The architect / engineer did not have a comprehensive shelf life program comparable to the program currently in place for Waterford 3. Since this was a construction issue, and since there is a shelf life program in place at Waterford 3 it is recommended that this commitment be deleted.
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- 13. LER 91-006 Commitments LER 91-006 commitments required,1) that a PM-7 portal monitor be installed at the -4 MSL RAB control point; 2) the source check periodicity be increased from daily to every shift during outages; and, 3) the event be reviewed during an upcoming ALARA committee meeting. This commitment has been deleted since items 1 and 2 are redundant with another commitment and item 3 was completed on July 16,1991, d
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l 14. Review of Permanent Control Room Modifications
- This is a change in the responsibility for review of all permanent Control Room modifications or other equipment for human factors aspects from " Engineering &
Nuclear Safety (NOE)" to 'Waterford 3." Due to organization changes, Engineering and Nuclear Safety is now call Design Engineering ElC. To preclude such furthre changes, the commitment is ravised as identified.
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- 15. Component Coolina Water Makeuo Commitments in response to IR 96-24, Waterford 3 committed to develop a design basis for CCW Makeup following design basis accident, revise DBD-003, Emergency Feedwater, and update calculation EC-191-003 with results by March 31,1997. A separate commitment required Waterford 3 to determine if Component Cooling Water Makeup Pumps are required to be tested in accordance with ASME Section XI. This change extends the completion date for these commitments from M6rch 31,1997, to July 31, 1997.
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- 16. Plant Operations Review Committee Review of Licensee Event Reperts Several Waterford 3 commitments required Plant Operations Review Committee (PORC) approval of Licensee Event Reports (LERs) prior to submittal to the NRC.
Commitment P-4656 discue.9s a permar;ent onsite Event Evaluation Committee which is to coordinate reportable events. This committee provides PORC with a report of its completed investigation After PORC approval, the Plant Manager approves disposition of PRES as LERs for transmittal to the NRC. Commitment P20787 discusses a steering function to the LER re-- .s. EAR &R personnel along with plant management develop a plan for dispositis i the LER.
These commitments are being deleted as PORC approval of LERs will no longer be required prior to robmittal to the NRC. The Quality Assurance Program Manual, Chaptar 1, only requires PORC to review reportable events. Technical Specifications require that each reportabte event shall be reviewed by the PORC and the results of this review shall be submitted to the Safety Review Committee and the Vice President Operations. Neither the QAPM nor the TS specify that the review must be done prior to NRC submittal. Procedure W2.501, ' Corrective Action', requires that all Category 1, significant adverse conditions, will have a root cause analysis conducted unless the RCA is waived by the Condition Review Board (CRB), CRs which are identified as reportable aie classified as Category I and M RCA is performed in the majority of instances. On occasion, the CRB may waive the RCA for simple, reportable events.
The RCA must be reviewed by the CRB which is made up of the GMPO and various director and management level personnel. LERs are routed to the appropriate level of management for review and approval thus ensuring they are correct prior to NRC submittal. Any substantial changes to the LER as a result of PORC review would be provided to the NRC in a follow-up submittal.
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- 17. Breakdown in the Plant Overtime Policy As a result of a programmatic breakdown in the plant's overtime polley, LER 97-012 was issued which committed to the following procedure changes:
- 1. The administration of the working hour policy shall be validated monthly by comparison to time reports.
- 2. An enhancement to Attachment 6.1, " Authorization of Working Hour Policy Deviatic,ns," to indicate, in addition to the reasons for the deviation, the number of hours required to be worked.
- 3. A clarification on the requirement for approval of working hours deviation prior to the commencement of the work activity.
- 4. A clarification on the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception for the " late watch relief'.
- 5. A clarification on who the working hour policy applies to, in addition, management will ensure that periodic QA audits are performed and, if necessary, onhanced to evaluate compliance with the requirements of the working hour policy.
The proposed change is to delete line item 1 regarding mon'hly validation of the working hour policy and to reword the statement regarding management ensuring that periodic QA audits will be performed and, if necessary, enhanced to evaluate compliance with 'ha raquirements of the working hour policy.
The change is made to match the corrective actions documented in the Root Cause Analysis breakdown in the plant's overtime policy which was approved by plant inanagement subsequent to the issuance of revision 0 of LER 97-012 The corrective actions are now being incorporated into revision 1 of LER 97-012, 193
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I 18. Use of Portable Radios and Telephones Based on IEC 80-09, use of portable radios and telephones in (1) the control room, (2) the relay room (+35 RAB), and (3) the computer room is prohibited. These areas are posted with signs prohibiting use of such devices.
This change allows the installation of an on site cellular phone system for use at W3.
This cellular phone system will help minimize traffic on the paging system. The existing communications systems will continue to function as designed and will not be affected by this new cellular phone system. The main components for the cellular phone system include a control unit, base stations, and handsets. The control unit will be mounted in the telecommunication room at RAB +7 and will receive power from PDP 387 2A. The base stations will be mounted as necessary throughout Waterford 3 using existing telephone lines. Several locations will require a change of cable to allow a greater number of conductors to accommodate the multi-channel base stations. The cellular phone system can be used with a restricticn requiring the handset or base station to remain 1 foot away from the surface of any electrical and electronic
- <olpment control panels. Special tags shall be placea on the handset to inform the user that the handset must remain at least 1 foot from EMI sensitive equipment.
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- 19. Use of Portable Radios and Telephones i Based on IEC 80-09, use of portable radios and telephones in (1) the control room, (2) the relay room (+35 RAB' and (3) the computer room is prohibited. These areas are posted with signs prohlbb.ig use of such devices.
. This change allows the installation of GN NETCOM'S MPA Satellite cordless telephone headset in the Control Room and Technical Support Center without causing an EMI related incident. The headsets and cellular telephone system can be used with a restriction requirirg the devices to remain i foot away from the surface of any electrical and electronic equipment control panels. Specla' tags shall be placed on the headset's basestation and handset to inform the user that the devices must remain at least i foot from EMI sensitive equipment.
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ll. PROCEDURES A. PLANT PROCEDURES
- 1. CE-002-001. Maintainino Steam Generator Chemistry (Revision 12)
DESCRIPTION Revision 12 allows the waiver of steam generator cation conductivity limits when ethanolamine (ETA) is used in the Secondary System for pH control. The revision also removes Primary to Secondary Leak Rate determinations and refers the determinations to a new procedure, CE-003-705. The revision also includes revising some steps to improve the Hideout Study that is conducted on the steam generators.
REASON FOR CHANGE ETA is designed to provide improved protection against Flow Assisted and Erosion Corrosion of Secondary Cycle piping. Changes in the determination of Primary to Secondary Leak Rates resulted in this important function being addressed by a specific procedure.
SAFETY EVALUATION According to the safety evaluation the FSAR Chapter 15 Accident Analyses identifies Steam Generator Tube Rupture, Steam Line Break, and Feedwater System Pipe Break as events that could be affected. The introduction of ethanolamine into the secondary cycle has been evaluated for use at Waterford 3, with regards to material compatibility, steam generator effects, balance of plant effects, turbine effects and industry experience and testing. No adverse effects have been identified as discussed in the
" Evaluation of the Application of Alternate Chemical Control for the Waterford 3 Secondary Cycle," W3C5-93-117, which was prepared for DC-3389, ' Alternate Chemical Addition for the Secondary System." (Reported in W3F2-96-0011, dated May 6,1996, Report of Facility Changes, Tests and Experiments.) The breakdown, into organic compounds, of ethanolamine in the secondary cycle is expected and does not pose any significant problems. The breakdown products therraally decompose into acetic, formic and glycolic acids. Industry experience indicates the increase in these acids have a negligible effect on the secondary cycle and do not contribute to corrosion mechanisms or compromise steam generator integrity. The organic related increase in cation conductivity will mask any increase in cation conductivity due to an intrusion of anodic impurities. The improved monitoring capabilities of other on-!ine instrumentation and periodic evaluation and correlation of analysis data in recent years, has reduced the need to rely on cation conductivity as a primary indicator of low level impurity intrusion into the secondary cycle. Adverse secondary cycle chemistry is more accurately identified and mitigated using other more reliable parameters such as sodium, chloride and conducting correlations and mass balances rather than relying on cation conductivity.
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The reactor coolant pressure boundary performance will not be affected or compromised and with the reduction of iron deposition in the steam generator, the sludge pile vdume would decrease thus reducing the risk of intergranular attack and stress cracking of steam generator tubes.
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- 2. CE-002-002, Maintainino Condensate and Feedwater Chemistry (Revision 8)
DESCRIPTION 3
Revision 8 allows the waiver of the feedwater cation conductivity limits when ethanolamine (ETA) is used in the Secondary cycle and increase the limit for pH in the feedwater system. The revision also changed the method of introducing ammonium chloride to the secondary cycle for steam generator molar ratio control.
REASON FOR CHANGE ETA is designed to provide improved protection against Flow Assisted and Erosion Corrosion of Secondary Cycle piping.
SAFETY EVALUATION According to the safety evaluation the FSAR Chapter 15 Accident Analyses identifies Steam Generator Tube Rupture, Steam Line Break, and Feedwater System Pipe Break as events that could be affected. The introduction of ethanolamine into the secondary cycle has been evaluated for use at Waterford 3, with regards to material compatibility, steam generator effects, balance of plant effects, turbine effccts and industry experience and testing. No adverse effects have been identified as discussed in the
" Evaluation of the Application of Alternate Chemical Control for the Waterford 3 Secondary Cycle," W3C5-93-117, which was prepared for DC-3389, " Alternate Chemical Addition for the Secondary System." (Reported in W3F2-96-0011, dated May 6,1996, Report of Facility Changes, Tests and Experiments.) The breakdown, into organic compounds, of ethanolamine in the secondary cycle is expected and does not pose any significant problems. The breakdown products thermally decompose into acetic, formic and glycolic acids. Industry experience indicates the increase in these acids have a negligible effect on the se ~ Ondary cycle and do not contribute to corrosion mechanisms or compromise steam generator integrity. The organic related increase in cation conductivity wi!l mask any increase in cation conductivity due to an intrusion of anodic impurities. The improved monitoring capabilities of other on-line instrumentation and periodic evaluation and correlation of analysis data in recent years, has reduced the need to rely on cation conductivity as a primary indicator of low level impurity intrusion into the secondary cycle. Adverse secondary cycle chemistry is more accurately identified and mitigated using other more reliable parameters such as sodium, chloride and conducting correlations and mass balances rather than ralying on cation conductivity.
The reactor coolant pressure boundary performance will not be affected or compromised and with the reduction of iron deposition in the steam generator, the sludge pile volume would decrease thus reducing the risk of intergranular attack and stress cracking of steam generator tubes.
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- 3. CE-002-013. Maintainina Essential Services Chill Water Chemistry (Revision 10)
DESCRIPTION The revision to the procedure rewords and renumbers the body of the procedure, adds chemical addition to the non-safety loop, adds non-a0gressive system flush, and adds bleed and feed.
REASON FOR CHANGE Provide clarity and consistency with other cooling water procedures. Also, addition of chemicals using plant installed non-safety chemical addition pots causes Operations to enter cascading technical specifications.
SAFETY EVALUATION According to the safety evaluation, implementing these changes will not reduce chill water flow to safety-related air handling units during accident conditions. Bleed and feed will be performed one loop at a time and that loop is required to be declared out of service prlor to bleed and feed. The only new connection created is when Essential Chill Water system is interconnected to the Liquid Waste Management system during bleed and feed. However, the LWM has the capacity to receive water drained from the Essential Chill Water system. No margins of safety are affected by this revision.
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- 4. CE-002-036. Chemical Control of Zebra Mussels in Circulatina Water System (Revision 01 DESCRIPTION The proposed procedure provides guidance for injecting bentonite clay, a blocide, into -
I the Circulating Water Systene to eradicate Zebra Mussel infestation and for deactivating the chemical prior to discharge into the Mississippi River.
REASON FOR CHANGE Injection of the blocide will eradicate Zebra Musselinfestation of the Circulating Water System.
SAFETY EVALUATION According to the safety evaluation, the probability nd consequences of an accident will not be affected by this procedure since it will not alter operation of the CWS, in the worse case scenario, there is a potential for loss of condenser vacuum if enough condenser tut'as were plugged by mussels as they dio off. However, loss of condenser is already an analyzed accident in the FSAR. The possibility of an accident being a caused by the interaction of the blocide skid and other components was evaluated by engineering, which determined that CWS is not compromised by use of this procedure.
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- 5. HP-001-220. Bioassay Prooram (Revision 8)
DESCRIPTION The proposed change deletes whole body counting as mandatory for in-precessing, termination, and annual evaluations.
REASON FOR CHANGE These changes are required in order to initiate the passive whole body monitoring program as pedormed by the personnel contamination monitors (PCMs) and the l personnel monitors (PM-7s) l l
SAFETY EVALUATION ;
1 The proposed ' change allows implementatbn of a passive whole body monitoring program. It does not affect any accidents or equipment important to safety, does not <
create any new system interactions or connections, and does not affect any margin of safety. i 201
- 6. Ml-003-102 (Revision 6) and OP-903-102 (Evaluation is also for Cl-301153/WA-01144191. NI Loa Power Channel Calibration Safety Channel A. B. C. or D - see item 1.B.2 )
DESCRIPTION The procedure revisions and Cl/WA reflect a conservative calibration of excore log power to 100% to match 100% reactor thermal power. Performance of the revised calibration will allow the exiting of the Excore Nuclear Instrumentation (ENI) Log Power Indicator LCO.
REASON FOR CHANGE The revision enhances performability of the procedure. The revised tolerances and values for the " Log Calibrate" switch ensure that the ENI log power indicators used aro calibrated to reactor thermal power.
Change in tolerance and values for the " Log Calibrate" switch positions are due to a change in the general bias required by the changes in neutron flux al the log power neutron detector.
SAFETY EVALUATION According to the safety evaluation the only accident which has radiological release consequences for the log power trip is the Uncontrolled CEA Withdrawal from Suberitical Conditions.
The log power trip remains Out of Service untilit can be returned to service. Since the equipment is Out of Service no credit la taken for its safety function and all Technical Specification LCOs will be met. it is not Applicable in Mode 1. The change in the general bias does not create any new system interactions or connections that did not previously exist. Indicated log power is altered to better and more conservatively match actual power. The adjustment will improve the accuracy of all log power indicators and the indications will remain within the loop accuracies.
The change in the general bias will maintain the margin of safety and assumptions used for the safety analysis by ensuring that the log power signal used for the log power trip correctly represents reactor power.
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- 7. Ml-005-415. Main Turbine Eccentricitv/ Kev-Phasor/Zero Speed Tachometer /
Acceleration Monitorina System (Revision 4. Chance 2)
DESCRIPTION The proposed procedure change it being made as a result of design change PC-8006 which added an additional interlock to the Main Turbine Zero Speed circuit. Steps are being added to the procedure to check the calibration of the tachometer setpoint.
REASON FOR CHANGE The proposed change is required to support the calibration of the probe added into the turning gear circuit under PC-8'X)6. The tachometer probe exists in the field but performs no control function so only the probe gap is adjusted under procedure Ml-005-415. A contact from the tachometer module is being added into the turning gear permissive circuit. This requires a functional check and calibration to ensure proper operation of the circuit.
SAFETY EVALUATION This evaluation is for the procedure change only. 'According to the safety evaluation, there are no unreviewed safety questions associated with this change.
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- 8. MI 005-46'LPlant Protection System Bistable Calibration (Revision 1. Chance B). OP-004-004 Control Element Drive (Revision 7. Chance A). OP-009-007.
Plant Protection System (Revision 4. ChanaqA). OP-010-001. General Plant Operations (Revision 17. Chance D). OP 903-107. Plant Protection System Channel A.B.C.D Function Test (Revision 12. Chance B)
DESCRIPTION The procedure deviations provide corrective actions associated with Waterford 3 Corrective Action Program document CR-96-0116. The deviations add an additional factor of 1000% in order to bound uncertainties associated with log power trip setting.
Also added is the restnction that Control Element Assemblies (CEAs) cannot be withdrawn until Reactor Coolant System (RCS) temperature is above 520 degrees F.
REASON FOR CHANGE All uncertaintesa in the log power trip setting may not have been accounted for when the setting was developed. The deviations conservatively bounds the magnitude of the uncertainty in the change in neutron flux reaching the excore detectors duo to core configuration, core temperature, core power, e.nd RCS boron concentration.
SAFETY EVALUATION According to the safety evaluation, the log power trip is credited for only the Uncontrolled CEA Withdrawal from Suberitical Conditions. The log power trip protects the fuel as a protective barrier. The change in the trip setting will ensure that the barrier functions as assumed in the safety Analysis by initiating the mitigating actions at a power equal to or lower than that assumed in the analysis. The change in the log power setting will not increase the probability of this accident.
The requirement that the CEAs be incapable of movement below 520 degrees F. will actually reduce the probability for the credited accident. This action reduces the length of time that the rod drivo mechanism has the chance to initiate the accident. The temperature limitation ensures that the change in the power indication due to temperature shadowing effects is bounded by the log power trip setting.
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- 9. MM-008-002. Conts* ment Penetration Modification for Refuelina (Revision 1)
DESCRIPTION Revision 1 of MM-008-002 adds steps to ensure ccMainment penetrations are returned to their original configuration after completion of the temporary configuration for refueling outages - Penetration 13 is used for temporary cables and penetration 63 is used for temporary service air.
REASON FOR CHA!Mg The procedure revision adds steps to ensure that penetrations 13 and 63 are returned to their proper configuration when the 'amporary modifications are removed. The revision also removes the reqc.rement to monitor pressure on the service air connection, this is accomplished by the Containment Closure Log, OP-001-003.
Sf FETY EVALUATION Acwrding to the safety evaluation, this procedure is only implemented during Modes 5 or 6. The Containment Closure Log, OP-001-003, provides administrative controls to ensure that penetration 63 is isolated if required. Penetration 13 has a temporary seal plate that forms a seal between the containment building and containment annulus, in the event of an accident this seal would prevent the transfer of air from the containment building into the annulus area. Seal integrity will be maintained during a seismic event.
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- 10. MM TEM-043. Installation and Removal of the Temporary Reactor Coolant Pump Seal (Revision 0)
DESCRIPTION A temporary cover designed to 50 psig and 250 deg. F will be installed in place of replacing the RCP seal to allow flood up for refueling. This cover is also designed to be used as a jacking point for use in determining the axial movement of the RCP motor's rotating element.
REASO_N FOR CHANGE To allow flood up for refueling activities without reinstalling RCP seats.
, SAFETY EVALUATION According to the safety evaluation, there are no accidents affected by this procedure s'id thus no change in consequences. The sealing mechanism is the same as used P during normai operation, therefore leakage is highly unlikely, in addition, the temporary ,
seal has been analyzed and found capable of withstanding the maximum anticipated pressuro and temperature. There are no unreviewed safety questions associated with this temporary procedure.
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- 11. J-001. Development. Revision. and Deletion of Procedures. Standard and -
Guides (Revision 4)
DESCRIPTION Revision 4 Iraorporates the requirements of Facility Operating License, NPF-38,
- Amendment 100 into the procedure.
REASON FOR CHANGE Amendment 100 establishes the Qualified / Technical Reviewer process for procedure reviews. This revision incorporates that program into the Design Engineering administrative procedure.
SAFETY EVALUATION According to the safety ovaluation this procedure does not control the content of any operating procedure, A change to the facility resulting from an application of this procedure will require a separate safety evaluation for the specific change. Every procedure is prepared in a way that commitments are maintained (e.g., maximum containment design pressure) or a separate analysis exists to justify the new parameter in the procedure (e.g., setpoints) such that the margin of safety is not reduced. A search for conflict between this procedure and the dasign basis of the plant has been completed and no conflicts were uncovered.
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- 12. NOECP-402. NPIS Common Foundation Basemat inteority Check (Revision 1)
DESCRIPTION The proposed revision deletes the requirement that the basemat settlement, the crack width, and the groundwater level measurements must be taken the same week. The procedure states they must be taken concurrently, i.e., as close together as practical to prevent any groundwater level changes from affecting the settlement and crack width measurements.
BEASON FOR CHANGE Correspondence W387-1123, from W3 to the NRC, indicated that the measurements would be taken concurrently not tha' ' hey must be taken in the same week.
- SAFETY EVALUATION The safety evaluation states that this revision will not create an unreviewed safety question. No system is affected and no equipment will be operated in an abnormal manner. The EPP and radiological waste systems are not affec'ed by this change.
The procedure continues to meet the requirements of TS 6.8.4.e.
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- 13. HQ1CP-405. Installation of Pre Enaineered Access Platforms and Ladders.
(Revision 0)
DESCRIPTION .
This procedure provides pre-engineered details for construction of platforms and ladders anywhere in the plant.-
REASON FOR CHANGE Using these pre approved guidelines and details, a design change is not required to add platforms or ladders.
SAFETY EVALUATION The safety evaluation states that the addition of an access ladder or platform does not impact the structural integrity of any adjacent equipment. Any changes on the nuclear island will be seismically supported. Any changes in areas of High Energy Line Breaks will be reviewed for jet impingement prior to installation. The proposed change does not challenge any safety system or cause a system to be operated outside its design limit, not does it affect any important-to-safety equipment. Any changes inside the containment building will be reviewed to ensure TS performance boundaries are not affected and the margin of safety is not reduced.
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- 14. OP-002-006. Fuel Pool Coolina and Purification DESCRIPTION The proposed .nge will allow the fuel pooling pump (s) to be secured as long as one pump remains op able. An upper limit of 1?O dagrees F will be placed on the pool indicated temporai le while the pump (s) are secured. Before the pool temperature reaches that limit, one pump should be p' aced back in operation. In addition, there is a maximum time limit of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> that the pumps may be secured.
REASON FOR CHANGE Temporary securing of the fuel pool pump (s) will be required to allow work to be performed in the spent fuel pool prior to, during, and following re-racking of the fuel pool. It may also be required to allow maintenance on the Fuel Pool Cooling System.
SAFETY EVALUATION According to the safety evaluation, the probability or consequences of an accident will not be affected by this change. No fuel movement will be allowed during the time the pump (s) are secured. The 120 degree F temperature and 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> time limits will ensure the regulatory limit of 140 degrees F is not exceeded.- No equipment is modified or used in a different manner due to this change. While both pumps may be secured, one pump must remain operable at all times. There are no new system interactions cr connections required by the change and no new accident modes are created. The regulatory limit of 140 degrees F is maintained, thus the margin of safety is not reduced.
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- 15. OP-003-014. Control Room Heatino.and Ventilatino (Revision 5. Chance A)
DESCRIPTION The proposed change ac js limitations that should the control room ventilation system be in isolation for greate that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during normal operations, normal intake air will be restored for at c:-' '.wo hours, if the control room envelope is to be isolated for
, greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the ability to restore normal ventilation would not be possible, access will be restricted to 16 people at all times.
REASON FOR CHANGE Limit carbon dioxide concentration in the control room and eliminate the need to account for personnel.
SAFETY EVALUATION No accidents in the FSAR are affected by this change; therefore, there is no increase in accident probability or consequences. No plant equipment is modified by this change; therefore, no important-to-safety equipment is affected, no new system connections are created, and no new methods of failure are created. The changes do not adversely affect the basis for any TS. Controls are added to protect the margin of safety by adding barriers to limit carbon dioxide concentration in the control room. No unreviewed safety question is created.
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- 16. OP-010-001. General Plant Operations (Revision 18)
DESCRIPTION This revision adds a caution stating that CEA breakers should not be .:losed until
- Reactor Coolant temperature is greater than or equal to 520 degrees F. It also changes the Log Power Hi trip setpoint to 0.0257%.
REASON FOR CHANGE All of the uncertainties in the Log Power Trip setting may not have been accounted for when the setting was developed. The only restriction is the minimum temperature required for criticality. .
SAFETY EVALUATION According to the safety evaluation, the Log Power Trip is credited in the FSAR only for the Uncontrolled CEA Withdrawal from Suberitical Conditions, The change in the Log Power Trip setting will not increase the probability of this accident. The requirement that CEA's be incapable of movement below 520 degrees F will actually reduce the probability for the credited accident by reducing the length of time that the rod drive mechanism has the chance to initiate the accident.
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- 17. OP-100-009. Control of Valves and Breakers (Revision 13. Char,r O DESCRIPTION The proposed revision adds or removes valves from the " Locked Valve List" and the
" Inaccessible Locked Valve List" o' moves valves betwe3n the two lists.
REASON FOR CHANGE By their nature, manual valves are not needed in the immediate aftermath of an accident. Thus operators are able to manipulate manual valves as needed regardless of whether they are locked in position or not because time permits the actions needed to unlock a valve anu reposition it. Valves are locked as an administrative measure to guard against inadvertent cperation.
SAF1TY EVALUATION The safety evaluation states tnat the procedure ma!ntains FSAR commitments regarding containment isolation valves, SlT loolation valves, demineralized water system valves, ECCS valves needed for SDC, condensate polisher bypass valves, and ADVs. Adding a manual valve to either locked valve list can neither affect current accident analyses, create new accidents, nor reduce the margin of safety, Any valves removed fro:.1 the list must meet several rigorous conditions: they are non-safety related, they are not said to be locked by any statement in the licensing basis, they are irrelevant to actions needed to either add water to or control level in a steam generator, they do not prevent inadvertent off site releases, and they cannot initiate a reactor trip.
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- 18. OP-901-520. Toxic Chemical Release (Revision 2. Chance A)
DESCRIPTION The proposed change adds a caution to limit access to the control room envelope to 16 people when the control room is isolated during a toxic gas event.
REASON FOR CHANGE Limit carbon dioxide concentration in the control room and eliminate the need to account for personnel.
SAFETY EVALUATION No accidents in the FSAR are affected by this change; therefore, there is no increase in accident probability or consequences. No plant equipment is modified by this change; therefore, no important to safety equipment is affected, no new system connections are created, and no new methods of failure are created. The changes do not adversely affect the basis for any TS. Controls are added to protect the margin of safety by adding barriers to limit carbon dioxide concentration in the control room. No unreviewed safety question is created.
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- 19. PE-005-040. Diaonostic Different!al Pressure T'astina of Motor Ooersted Valves (Revision 1)
DESCRIPTION The proposed change adds sections for differential testing of valves SI-225, -226, -227,
-228, -219, and -506.
REASON FOR CHANGE Consolidate required differential testing procedures for the Safety injection (SI) system.
SAFETY EVAL.UATION Though the equipment added to this procedure is required for accident mitigation, it cannot cause any accidents; therefore, there is no increase in probability or consequences. System function and operat,lity are not affected and there are no new system connections; therefore, no equipment important-to-safety is affected. No margin of safety is affected and no unreviewed safety question is created.
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- 20. RF-003-003. Steam Generator Studae Removal (Revisionf)
DESCRIPTION The proposed change adds a section to perfre an upper bundle flush of the Steam Generators (SG) prior to sludge lancing. It inu des instructions for the preparation, operation, and disassembly required to perform the flush. It also includes human factors improvements and reference updates.
REASON FOR CHANGE Performing an upper bundle flush will provide a means for removing sludge buildu ) in the upper regions of the SGs down to the tubesheet where it can be removed by sludge lancing. This will minimize tube degradation in the upper region due to sludge buildv.
SAFETY EVALUAllON No accidents are affected or created by the proposed change and the consequences of a radioactive release are not increased. The pressure of the flush water is approximately 50 psi which is well below the design pressure of 1100 psi of the secondary side of the SGs. No new interactions are created and no margin of safety is reduced.
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- 21. RF-TEM-001. Dropped Fuel Assembly Stabilization DESCRIPTIQN, This is a temporary procedure for securing fuel assembly LAR338 preventing it from sliding further from its current position in the Spent Fuel Pool (SFP).
REASON FOR CHANGE New fuel assembly LAR338 was dropped in the Spent Fuel Pool and this is the method to temporarily secure it in its dropped position in the SFP.
SAFETY EVALUATION The fuel handling accident analyzed in the FSAR assumes an irradiated assembly with a decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, implementation of this procedure, which is for a dropped new fuel assembly which is resting in an area of the SFP which contains no irradiated fun, will not increase the probability or consequences of a fuel handling accident. No important to-safety equipment is affected, no new system interconnections are created by this temporary procedure, and no margin of safety is reduced.
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- 22. RF-TEM-002. Recovery of Fuel Assembly LAR338 >
DESCRIPTIQN This is a single use procedure used for recovering fuel assembly LAR338 from its dropped position in the Spent Fuel Pool (SFP).
REASON FOR CHANGE New fuel assembly LAR338 was dropped in the SFP.
SAFET'r EVALUATION The probability and consequences of a fuel handling accident (FHA) will not be increased by use of this procedure. The FHA in the FSAR assumes irradiated fuel with a decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This is a new fuel assembly which will be carried only over empty locations, locations with equipment, or locations with new fuel. Fuel handling equipment will not be operated in an abncrmal manner and no new system interactions will be created. No protective boundary will be changed and no margin of safety reduced.
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- 23. STA-001-004 Local Leak Rate Test (LLRT) (Revision 1. Chanoe 3)
DESCRIPTION The proposed revision changes Penetration 63 from non-Bypass to Bypass leakage.
E REASON FOR CHANGE I
Condition Report 96-0361 identified that Penetration 63 was incorrectly ider.'.ified as non-Bypass leakage.
SAFETY EVALUATION According to the safety evaluation, this is an administrative change only, Penetration 63 is being added to the Bypass leakage total of 63,069 sccm. It does not increase this l- total leakage thus the TS limit is maintained and the consequences of a LOCA are not increased by the change.
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- 24. l
_)NT-005-014. Offsite Dose Calculation Manual (Revision 5. Chanae 1 and Chance 3)
DESCRIPTION The proposed change revises tables to indicate the source and type of release, adds a note to clarify channel cl.cck and chsnnel source check pre-discharging action requirements for BWM, LWM, and GWMS menitors, changes sample frequencies for the REMP, and revises the annual land use census.
Change 3 adds a one hour time period in which effluent releases may continue when the particulate sampler and iodine sampler channels are inoperable for radioactive gaseous effluent monitoring instrumentation. It increases decay time prior to analysis of gaseous effluent samples for gross alpha and reassigns resporsib;;N to the Chemistry Superintendent.
REASON FOR CHANGE Implement corrective actions to ensure there is no potential for different interpretation of original action statements and to clarify the current channel check and source channel check. Historical data indicates changing REMP cample frequencies is warranted.
Change 3 allows performance of gaseous effluent sampling surveillances without
- having to actually install auxiliary sampling equipment as specified in the action requirement for these channels. This will be achieved by implementation of a one hour
" grace period" before either: 1) the release is secured; or 2) auxiliary campling equipment is installed. The action statement is intended to provide that these radioactive release pathways are monitored continuously. There is no allowance for routine effluent surveillances in which filter media is changed or a grab sample is obtained for these release pathways. There is also no time requirement allowed for installing auxiliary sampling equinment. Decay time for gaseous effluent gross alpha needs to be increased.
SAFETY EVALUATION According to the safety evaluation, the change in sample frequencies and RMS instrument operation and surveillance tables do not have any impact on accident probability or consequences, on function of equipment important to safety, or on the margin of safety. Sampling frequencies are being changed as allowed by the Environmental Report based on the historical data which indicates the ability to predict the rad;ological impact associated with any effluent pathway or with plant operation in General. Instrument changes are for clarification purposes and do affect use of the equipment.
Allowing a one hour time period in which the particulate z.nd iodine samplers are out of service on gaseous effluent radiation monitors without actually instelling auxiliary 220
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sampling equipment will not affect the consequences of an accident. Monitoring of continucus effluent release streams will still be performed in such a manner that the health and safety of the general public are not compromised. No unreviewed safety question exists for this change.
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25,- UNT-006-019$ Control of Local Leak Rate Testina (Revision 1. Chance 3)
DESCRIPTION The proposed change provides additional guidance for determining penetration leakrates in penetrations with danger-tagged containment isolation valves, includes a description of the Reference Volume Method of performing LLRT, corrects certain
- administrative limits, and changes Penetration 63 to Bypass Leakage.
REASON FOR CHANGE The proposed changes will ensure Total and Bypass leakrates are within TS limits when CVR-401 A(B) are opened, add the Reference Volume Method, correct typographical errors, and implement corrective action for CR-96-0361.
SAFETY EVALUATION According to the saf6;y evaluation, this is an administrative change only which became necessary when a removable spoolpiece was replaced with hard pipe resulting in the Penetration 63 leakage communicating directly with an area outside the RAB and Turbine Building. This change does not increase the probability or consequences of an-accident. No physical change will be made to either of the barriers in Penetration 63,-
thus the probability or consequences of an equipment malfunction will not_be affected '
by this change, in addition, no test method will be changed so no new accident is L created and the TS limit for bypass leakage will be maintained. No margin of safety is reduced.
L B. SPECIAL TEST PROCEDURES (STP)
- 1. STP-01135315. On-line Leak Test of SI-120A(B) and SI-121 A(B) pESCRIPTION This procedure measures leakage though SI-120A(B) and SI-121 A(B) by applying 1500
, psig upstream of the valves using a hydrostatic pump. The procedure renders one ECCS train out-of-service for the duration of the test.
REASON FOR CHANGE SI-120A(B) and SI-121 A(B) together seal off recirculating ECCS water so that HPSI maintains minmum NPSH from the safety injection sump. These valves also limit the
- off-site dose and control room dose by limiting the amount of contaminated recirculated water in the RWSP. The procedure checks the leak tightness of each recirculation isolation valve by having them closed (one at a time) before the ECCS train pressurizes.
SAFETY EVALUATION According to the safety evaluation, this test procedure requires one ECCS subsystem
- to be rendered inoperable and non-functional, thus invoking the action of TS 3.5.2 and 3.6.2.1. However, this is done pursuant to TS 4.0.5. The benefit to sefety derived from meeting surveillance requirements for SI-120 and SI-121 ;s considered to more than compensate for the risk to safety from operating the facility in an LCO action statement for a small fraction of the allowed out time (AOT). The AOT is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and each ECCS subsystem should be out-of-service approximately one hour. Because the, _
procedure implements surveillance requirement 4.0.5, it is not necessary to postulate a single active failure and accident initiator during the test. In addition, the procedure poses no unusual conditions or stresses on the tested pipe and boundary valves.
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- 2. - STP-01145052. Nitrocen Iniection to Condenser DESCRIPTION The proposed test will allow injection of nitrogen into each condenser section (shell side). The nitrogen will be injected at test connections beneath the low pressure turbine skirts.
REASON FOR CHANGE To evaluate its effectiveness in reducing condensate disselved oxygen.
, , SAFETY EVALUATION To preclude an increase in the likelihood of a loss of condenser vacuum accident, the procedure limits the amount of nitrogen to be injected at a combined rate of 18 scfm.
The condenser air evacuation system is capable of handling this amount of nitrogen.
. Because nitrogen is inert, it will have no impact on the ability of the condenser air evacuation pump radiological monitor to perform its intended function. No new system interconnections or failure modes will be created, no protective boundaries are affected, and no margin of safety is reduced.
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- 3. STP-01147312. CCW Heat Exchancer Performance Test (Modes 1-4)
DESCRIPTION This procedure provides for collecting temperature and flow data for the 'B' CCW Heat Exchanger to be used to calculate the Overall Heat Transfer Coefficient.
REASON FOR CHANGE This procedure will be used to determine CCW Heat Exchanger 'B' heat transfer capability with ACCW Train 'B' low flow conditions identified by WA 01146595.
SAFETY EVALUATION According to the safety evaluation, the CCW System is required to mitigate the consequences of a LOCA or a MSLB by removing heat from the containment and rejecting the heat via the cooling towers to the atmosphere. The CCW System is also required to supply cooling water to the Essential Services Loop. If required to achieve adequate test conditions, the CCW Trains 'A' and 'B' will be separated with Train 'B' supplying the nonessential loops. Essential Chiller B' and Containment Fan Cooler 'B' and 'D' will be secured. Train 'B' Dry Cooling Tower Fans will be under manual control and Dry Cooling tower Fan flow will be bypassed, in this case, the applicable Technical Specification Action Statements will be entered and the 100% capacity CCW Train 'A' will be available to provide accident mitigation.
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- 4. STP-01149859. Control Room Carbon Dioxide Collection DESCRIPTION This special test will place the Control Room Ventilation System (HVC) in isolate modc and record the carbon dioxide concentrations independent of occupancy levels.
REASON FOR CHANGE This test will provide data to allow justification of changing the Control Room 16 person occupancy limit to unlimited while in isolate mode for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and not in a Toxic Gas Event.
SAFETY EVALUATION According to the safety evaluation, the probability or consequences of an accident will not be increased by this special test. There is a test limit of 1/3 of the toxicity limit of carbon dioxide and a reduction of staff levels from 16 to 10. No plant equipment is being modified by this test and there is no increased reliance on any equipment or system important to safety. There are no new system interactions or connections and no new methods of failure are created To maintain the margin of safety, the test will be exited if carbon dioxide levels reach less than 1/3 the toxicity limit of 1.0% or if a Toxic Gas Event occurs.
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- 5. _STP-01150154 CCW System Flow Balance Test (Revision 0 and Revision 1)
DESCRIPTION This special test will verify that each safety-related component cooled by the Component Cooling Water (CCW) system receives the proper flow during accident conditions.
REASON FOR CHANGE The CCW system, as part of the Ultimate Heat Sink (UHS), is required to mitigate the consequences of a Loss of Coolant Accident (LOCA), a Main Steam Line Break (MSLB), or a Main Feedwater Line Break (MFLB) in containment by rejecting heat from containment. CCW is also required to supply cooling water to the Essential Services Loop (Containment Fan Coolers, Emergency Diesel Generators, Shutdown Heat Exchangers, Essential Chillers, HPSI Pumps, LPSI Pumps, and Containment Spray Pumps).
This special test will verify that each safety related component cooled by the CCW system receives the proper flow during accident conditions. This will be accomplished by putting the CCW system in its accident lineup and documenting indicated flow through each component.
SAFETY EVALUATION The function of the CCW system to remove heat from mechanical components and heat exchangers during normal and accident conditions will not be changed for this test.
The test will be accomplished during normal power operation by isolating the nonessential seismic and nonessential nonseismic loops from the train being tested.
Cooling water for the RCPs and CEDM coolers will be supplied from the train not being tested. According to the safety evaluation, this test will align the CCW system to its accident (SIAS or CSAS) line up with administrative controls to prevent opersting the system outside its design limits. There will be no new interconnections to other systems and appropriate valves will be placed in their accident-required positions so that the probability or consequences of an accident will not be increased by this change. During the test, each CCW train will be tested separately, so the opposite train will be available should any equipment be tested fail. Thus there will be no affect on equipment important to safety. The margin of safety will not be affected since the .
CCW system will be operated within its design limits. No unreviewed safety question exists as a result of this test.
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- 6. ' STP-01150615. CS-118A Flow Test
- DESCRIPTION ..
. This special test performs a low flow test of CS-118A to gather data to be used in '
determining past operability of the valve.
RF_ASON FOR CHANGE LThis specid test procedure was needed to determine if backleakage through CS-118A to the RWSP had potentially exceeded TS requirements SAFETY EVALUATION According to the safety evaluation, the proposed test will have a minimal impact on the CS system because the system will be briefly configured in an abnormal alignment -
-i.e., CS-118A will be throttled 1 and 1/2 turns open and SI-346 will be opened as the .
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- collection point. There will be no affect on the probability or consequences of an accident since the only possible radiological consequence would be a Recirculation
- Actuation Signai_ (RAS) during the time CS-118A is throttled and SI-346 is open.
However, they will only be in this configuration for no more than one minute and arr RAS would not occur until 20 minutes post-LOCA. Operators will be stationed at both valves with specific direction to close them in the event of a SIAS or CSAS. In addition, - .
the proposed test does not create the possibility of a fai!ure or accident nor reduce the margin of safety.~
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7, STP-01151879. CC-181 A(B) and CC-135A(B) Leak Test DESCRIPTION The purpose of this procedure is to determine the amount of leakage through the Dry Cooling Tower (DCT) check valve, CC-181 A(B), and the DCT olet isolation valve, CC-135A(B)
REASON FOR CHANGE
. Condition Report CR-96-1652 iden'.ified the need to account for leakage through CC-181 A(B) and CC-135A(B) while the DCT is_ bypass mode, during a design basis tornado event. This laakage should then be reflected in Condensate Storage Pool and Wet Cooling Tower basin inventory margin.
SAFETY EVALUATION According to the safety evaluation, the DCT will not be operated in an abnormal manner and one train of CCW will be available for accident mitigation. Therefore, the probability and consequences of an accident are not increased by this test, The -
proposed evolution is the normal evolution when DCT maintenance is required and has already been evaluated in current licensing documents; therefore, no new adverse ;
affect on important-to-safety equipment is created. There are no now system
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interconnections created by this test. No protective boundaries are affected; therefore,
. the margin of safety is not reduced.
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- 8. STP-01156079. Shutdown Coolina Heat Exchanaer 'A' Performance Test <
DESCRit i.ON The proposed test will evaluate the current condition of the Shutdown Cooling Heat l.
Exchanger (SDCHX) 'A' by measuring temperatures and flows after the plant has entered Mode 4 during plant cooldown. A zero cooldown rate will be established. -The data collected will allow extrapolation of the SDCHX overall heat transfer coefficient to accident conditions.
REASON FOR CHANGE The SDCHX has recently been included in the GL 89-13 program which mandates this -
type of performance testing. The information and results from this test will be trended over a period of time to predict the rate at which fouling is occurring in the SDCHX, allowing a fouled condition to be corrected prior to impairing the SDCHX safety function -
SAFETY EVALUATION The proposed test installs temporary instrumentation to allow adequate data collection.
An array of RTDs will be non-intrusively strapped to the external surface of the SI and CCW piping and covered with insulation to allow very accurate temperature profiles of the fluids. High accuracy transmitters will be connected '.o the test connections of the
.SDC flow transr.iitter and also the CCW f;ow transmitte.; Where required, a seismic evaluation has been performed of the installation 'o ensure no radiological consequences would result from a seismic event. The SDC and CCW systems will bei operated in accordance with approved procedures, the FSAR, and TS. : Data will be -.
collected when the proper conditions are established. During the time the tests are being performed, TS 3.4.1.3 will be in effect and the applicable LCO will be entered if required. There is no unreviewed safety questionL associated with this special test.
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- 9. STP-01156080. Shutdown Coolina Heat Exchanaer 'B' Performance Test DESCRIPTION The proposed test will evaluate the current condition of the Shutdown Cooling Heat
. Exchanger (SDCHX) 'B' by measuring temperatures and flows after the plant has entered Mode 4 during plant cooldown. A zero cooldown rate will be established. The data collected will allow extrapolation of the SDCHX overall heat transfer coefficient to accident conditions.
REASON FOR CHANGE The SDCHX has recently been included in the GL 89-13 program which mandates this 1 type of performance testing. The information and results from this test will be trended over a period of time to predict the rate at which fouling is occurring in the SDCHX, ailowing a fouled condition to be corrected prior to impairing the SDCHX safety function SAFETY EVALUATION The proposed test installs temporary instrumentation to allow adequate data collection.
An array of RTDs will be non-intrusively strapped to the external surface of the SI and CCW piping and covered with insulation to allow very accurate temperature profiles of me fluids. High accuracy transmitters will be connected to the test connectbns of the SDC flow transmitter and also the CCW flow transmitter. Where required, a seismic evaluation has been performed of the installation to ensure no radiological consequences would result from a seismic event. The SDC and CCW systems will be operated in accordance with approved procedures, the FSAR, and TS. Data will be collected when the proper conditions are established. During the time the tests are being performed, TS 3.4.1.3 will be in effect and the applicable LCO will be entered if rcquired. There is no unreviewed safety question associated with this special test.
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- 10. STP-01156545; Vacuum Fill of LPSI Penetration Noi 39 DESCRIPTION This special test procedure will use a vacuum pump to attempt to remove gas from the Penetration 39 piping between the LPSI Flow Control Valve SI-138A and LPSI header to RC Loop 2B inside Containment Check Valve SI-142A. Once the gas is removed,-
the piping will be_ refilled with water by opening SI-138A with the LPSI pump secured.
REASON FOR CHANGE' The amount of gas presently in the piping is more than has been previously analyzed to be acceptable.- This makes the 'A' train of LPSI inoperable. The amount of gas must be reduced to less than the previously analyzed acceptable amount.
SAFETY EVALUATION There are no unreviswed safety questions related to the performance of this special
- test. - During the performance of this special test procedure, the _'A' train of LPSI will be
- out of service and the applicable LCO entered in accordance with TS 3/4.5.2.-
Therefore, this special test procedure will be conducted with the system within its analyzed cases. The test will be conducted while maintaining double isolation from the RCS. The system will be refilled in a controlled manner to ensure that an adverse hydraulic transient is not introduced.
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- 11. STP-01156593. Vacuum Fill of LPSI Penetrations No. 36 and 37 DESCRIPTION This special test procedure will use a vacuum pump to attempt to remove gas from the
- Penetration 36 and 37 piping between the LPSI Flow Control Valve SI-138BISI-139B '
fand Si header to RC Loop 2B inside Containment Check Valves SI-142BISI-143B, Once the gas is removed, the piping will be refilled with water by opening SI-1388 or .
SI-139B with the LPSI pump secured; REASON FOR CHANGE The amount of gas presently in the piping will be vented via a vacuum pump to comply with TS 3/4.5.2.J.'
'. SAFETY EVALUATION There are no unreviewed safety questions related to the performance of this special __
test 7uring the performanc,s of this special test procedure, the 'B' train of LPSI will be -
out c; service and the applicable LCO entered in accordance with TS 3/4.5.2.
Therefore, this special test procedure will be conducted with the system within its analyzed bases. The test will be conducted while maintaining double isolation from the - '
RCS. The system will be refilled in a controlled manner to ensure that an adverse hydraulic transient is nct introduced,-
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- 12. ETP-01157063. Pressure Test of Class 1 RCS Vent and Drain Lines
< DESCRIPTION L This special test involves pressurization of 23 separate and normally isolated vent, !
= drain, and seal lines of the Reactor Coolant System (RCS) to RCS nominal operating
- pressure (NOP) of 2250 psi during Modes 3 or 4. After appropriate hold times, visual examinations are performed for through wall leakage.
- REASON FOR CHANGE '
This test satisfies the pressure testing requirement of ASME Section XI for Class 1 lines by applying ASME Code Case N-498-1 in lieu of hydrostatic tests. Pressure testing supports and preserves system design and accident analysis.-
' SAFETY EVALUATION This test was developed to meet ASME_ test requirements by opening the inner valve of a double isolation boundary. Each test pressurizes lines, manual valves, or bolted flanges that are designed to 2485 psi or greater, to the RCS NOP of 2250 psi. This does not impose any unanticipated loads to those components. - Expanding the boundaries in this limited manner to existing passive safety components, with identical characteristics, in the same locations, allows the existing small break LOCA analysis to -
remain the same; System performance is enhanced as a result of this test because >
normally isolated lines are proved to comply with acceptance criteria, and boundaries are ensured to have maintained and continue to retain their safety functions, There is no unreviewed safety question involved with this testing.
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- 13. STP-01157743. Chance in RWSP Level with One CVAS Makeup Damper Failed Closed DESCRIPTION The purpose of this special test is to determine the change in RWSP level indication with one CVAS makeup damper failed closed. This will be accomplished by failing closed damper HVRMVAAA303B with both CVAS trains in operation.
REASON FOR CHANGE This test is being performed to verify that a single failure of a CVAS makeup damper will not cause incorrect RWSP level indication that could adversely affect the operability of the Safety injection System.
SAFETY EVALUATION This test will not result in any unreviewed safety questions since the system will operate within its original design during the test. Failing a damper closed will place it in -
its fail-safe position and will ensure that negative pressure is maintained for the CVAS areas. The ability of the system to respond to an accident will not be compromised.
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- 14. - STP-01158622.- CVAS Boundary Test : A DESCRIPTION The proposed test will verify that the RAB Control led Ventilation Area System can perform its safety related function with one CVAS isolation valve failed open. The test -
will fail open one isolation valve with one CVAS filter train running and verify that the system maintains the required -0.25 inwg negative pressure in the CVAS areas. The test will be performed for each of the isolation valves.
REASON FOR CHANGE This test is being performed to verify that a single active fai!ure of a CVAS isolation valve will not prevent the system from maintaining the required negative pressure in the L CVAS areas. .
SAFETY EVALUATION According to the safety evaluation, the test will not initiate or affect any accidents or.
affect any impoitant-to-safety equipment. The test will make one train of CVAS
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inoperable and the appropriate LCO will be entered. The operable train will remain available to perform its safety function as will its associated isolation valves. No protective boundaries are affected, no safety margins are reduced, and there are no unreviewed safety questions. .
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15l STP-01158897.- Shutdown Coolina Flow Control Valves DESCRIPTION The proposed test will determine the throttled positions for the Low Pressure Safety -
Injection (LPSI) neader flow control valves in the event SI-129A(B) were to fail open while on shutdown cocling with Reactor Coolant System (RCS) level less.than 18 feet
- MSL. 1 REASON FOR CHANGE Evaluation of the RCS drain down procedure recommended a reduction to the maximum allowed shutdown cooling flow through a single train. This test will allow determination of the correct LPSI flow control valves' throttle position to mitigate the consequences in the event valve Sl-129A(B) should fail open.
SAFETY EVALUATION According to the safety evaluation, this test will be conducted within the existing SDC system operations requirements land limitations, and in compliance with applicable Operating proceduras and Technical Specifications. At all times, the required SDC rate will be met or exceeded. No accident or important-to-safety equipment will be affected.
No new system connections or interactions are required during this test so no new
- accident or equipment malfunction will be created No margin of safety _will be reduced -
. and there is no unreviewed safety question associated with this test.
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- 16. STP-301702. Throttlina Condenser' Outlet Isolation Valves DESCRIPTION -
The STP provides for the reduction oi Circulating Water (CW) through the condenser
.by throttling the ( andenser discharge is7lation valves, CW-127A, B, & C and CW- ;
128A, B, & C. The goal of the test is to meintsia dissolved oxygen levels to below 10 ppb without adversely affecting plant or equipment operation. '
REASON FOR CHANGE The flow reduction will decrease the amount of condensate subcooling which allows -
higher levels of dissolved oxygen to be released from the condensate. Dissolved oxygen may increase corrosion of the steam generators and is generally a problem when the river water is below 45 degrees F.
SAFETY EVALUATION The safety evaluation identifies two accidents that may be caused by the STP: Loss of -
Condenser and Loss of Condenser with a Single Active Failure. The STP will reduce CW flow to the condenser which will increase condenser temperature resulting in an -
increase in condenser pressure. Condenser pressure is normally maintained at approximately 1 inHg absolute. : At 5 inHg absolute there is a high pressure alarm and at.10 inHg absolute the turbine will trip.
The STP contains a 1,5 in Hg average increase in condenser pressure as a limitation to prevent condenser pressure from reaching the high pressure alarm. ,
CW pump motors are limited to a current requirement of 350 amps.- This will prevent-motor damage which could kad to a pump trip. In the event that a pump would trip, there would be two CW pumps operating.' The flow prior to the pump trip is limited to 450,000 GPM minimum per the STP. Assuming that the total flow from one pump is discharged through thu closing CW pump discharge valve the remaining flow being
- supplied to the condenser should be approximately at pump runout or 300,000 GPM. _
- PEPSE model analysis shows that this flowrate is sufficient to maintain condenser pressure below the trip set point.
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- 17. STP-307622. DC-3435 Acceptance Test DESCRIPTION The proposed test will perform 125% load tests of Refueling Machine (RFM) hoists, establish current limits on hoist drives, validate hoist operation, demonstrate hoist brake operation, and demonstrate the 1,1stallation and operation of the communication cable.
REASON FOR CHANGE Validase acceptability of DC-3435, RFM hoist drive improvements, establish current limiting settings on hoist drives, and collect reference information on the hoist brakes.
SAFETY EVALUATION The proposed test will operate the equipment in a manner not performed during actual fuel handling to test the various features of the hoists. However, rt o of these tests will be performed with new or irradiated fuel and w.11 not be performed over any fuel.
No new system interactions or connections are required for these tests and use of the equipment during actual fuel movement is not affected and will not be changed. There is no unreviewed safety question associated with this test.
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- 18. STP-99100019A. Ultimate Heat Sink Cable Re-route Train 'A' Retest DESCRIPTION This special test will verify proper operation of UHS components affected by the cable repull effort associated with exposed conduits that could be affected by a tornado i missile. The affected components are CC-134A, CC-135A, ACCILT7079A, and DCT Fans 7A-15A.
REASON FOR CHANGE Condition report CR-96-1591 identified several conduits supplying control power to UHS components which are not protected from a tornado missile.
SAFETY EVALUATION According to the safety evaluation, the proposed test will not initiate or cause a LOCA, MSLB, or design basis tornado. The train 'A' CCW Makeup pump will be isolated to prevent overflowing the CCW Surge Tank; however, the 'B' train pump will remain in service. While none of the equipment being tested will be operated in a manner that would increase the probability or consequences of its malfunction, the 'B' train will also be available to mitigate the consequences of any event that might occur. In addition, the test does not change any protective boundary or affect a margin of safety.
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- 19. STP-99003468A. Essential Chilled Water Flow Balance. Minimum Room Temperatura Test, and DC-3468 Acceptance Test for Train A DESCRIPTION
' This procedure will record and adjust, as necessary, Essential Chilled Water Train 'A' flow rates with Essential Chilled Water Pump 'A' or 'AB' operating and appropriate air handling units' flow control valves failed open, it will measure minimum room temperatures at full flow conditions as required by DC-3468. - It will serve as the acceptance test for the 'A' Chilled Water train by establishing that acceptance criteria -
are met for sub-loop and total flow and for maximum pump motor amps as required by DC-3468.
REASON FOR CHANGE l The temparature control valves at each air handling unit supplied by the Essential Chilled Water Loop, with the exception of the control room air handling unit and the switchgear area air handling units, will be permanently failed open as part of DC-3468.
In order to maintain design water flow to all air handling units, this will require a re-balance of water flows in the loop to account for both 'A' train component air handling .
units and 'AB' train component air handling units having full water flow simultaneously, -
since the original system flow balance and subsequent Train Maximum Flow Tests :
were performed with either the 'A' or the 'AB' component air handling units having full-flow at any one time. Also, the minimum room temperatures in various spaces will be recorded with air handling units operated manually and low internal heat loads in order to verify that room temperatures remain within design values.
SAFETY EVALUATION Conduct of the proposed flow balance on Essential Chilled Water Loop 'A' and minimum room tempere*are test as described do not result in an unreviewed safety question. The flow balance is an adjustment of system throttle valves to ensure that water flow through each air handling unit remains within design values after DC-346E is implemented. - The minimum room temperature test only involves recording room equilibrium temperatures with air handling units in manual operation and internal room heat loads kept to a minimum. The procedure requires the test to be terminated before any room temperature reaches the lower design temperature value.
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- 20. STP-990034688. Essential Chilled Water Flow Balance, Minimum Room Temperature Test. and DC-3468 Acceptance Test for Train B DESCRIPTION This procedure will record and adjust, as necessary, Essential Chilled Water Train 'B' flow rates with Essential Chilled Water Pump 'B' or 'AB' operating and appropriate air handling units' flow control valves failed open. It will measure minimum room temperatures at full flow conditions as required by DC-3468. It will serve as the acceptance test for the 'B' Chilled Water train by establishing that acceptance criteria are met for sub-loop and total flow and for maximum pump motor amps as required by DC-3468.
REASON FOR CHANGE The temperature control valves at each air handling unit supplied by the Essential Chilled Water Loop, with the exception of the control room air handling unit and the switchgear area air handling units, will be permanently failed open as part of DC-3468.
In order to maintain design water flow to all air handling units, this will require a re-balance of water flows in the loop to ac,Dunt for both 'B' train component air handling units and 'AB' train component air handling units having full water flow simultaneously, since the original system flow balance and subsequent Train Maximum Flow Tests were performed with either the 'B' or the 'AB' component air handling units having full flow at any one time. Also, the minimum room temperatures in various spaces will be recorded with air handling units operated manually and low internal heat loads in order to verify that room temperatures remain within design values.
SAFETY EVALUATION Conduct of the proposed flow balance on Essential Chilled Water Loop 'B' and minimum room temperature test as described do not result in an unreviewed safety question. The flow balance is an adjustment of system throttle valves to ensure that water flow through each air handling unit remains within design values after DC-3468 is implemented. The minimum room temperature test only involves recording room equilibrium temperatures with air handling units in manual operation and internal room neat loads kept to a minimum. The procedure requires the test to be terminated before any room temperature reaches the lower design temperature value.
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- 21. STP-99003470. Acceptance Test for DC-3470 (Revision 0 and Revision 1)
D_ESCRIPTION This special test procedure verifies the equipment installed or modified by DC-3470 will function as required.
REASON FOR CHANGE DC-3470 installed equipment in the plant to prevent waterhammer in the Auxiliary Component Cooling Water (ACCW) System.
SAFETY EVALUATION The CCW, ACCW, and CHW systems are required to mitigate the consequences of a LOCA or MSLB. The EDGs are required to provide an emergency source of AC power to the safety buses 3A(B) during loss of offsite power and standby AC power supplies.
All testing on these systems will be performed during Modes 5 or 6 with the train being tested inoperable and not required to support plant operations. Any temporary equipment or instrumentation will be installed on the inoperable train only and will be removed prior to returning the affected train to operable status. No accidents or equipment important-to-safety will be affected by this test and no margin of safety will be reduced. Thus there is no unreviewed safety question associated with this special test.
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- 22. STP-99003478. SIT Nitroaen Supply PCV Leak Tes_t DESCRIPTION This STP provides instructions for performing seat leakage testing of the Safety injection Tanks (SIT) nitrogen supply isolation valves, NG-161 A(B) and NG-162A(B).
REASON FOR CHANGE Design Change "^s-3478 will replace the existing hard seat WKM globe valves with soft seat BNL vaives to reduce or eliminate recurring seat leakage problems. This STP is the acceptance test for these new valves.
SAFETY EVALUATION The proposed special test does not create an unreviewed safety question. The test requires the SIT boundaries to remain as normally required by procedure OP-010-001 for Mode 5 operations. The non-safety containment nit gen gas supply header is !
isolated and depressurized when checking for lenkage irom the SITS to the nitrogen header. The non-safety reactor drain tank, non-safety quench tank, and non-safety post-accident sampling system will be isolated during this test. The differential test' pressure is well below the maximum design pressure for the SIT nitrogen supply isolation valves. '
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- 23. STP-99003492. DC-3492 Acceptance Test DESCRIPTION The Component Cooling Water (CCW) Filtration Skid, installed by DC-3492, will be placed into. service to ensure that the components normally cooled by CCW are not adversely affected. CC-413A will be failed open to simulate operation of an Emergency Diesel Generator. Procedure OP-903-118 for CC-200A/CC-727 and CC-200B/CC-563 will be performed to verify the amount of water diverted to the CCW filters does not adversely affect system operation when the non-safety header isolation valves are closed during quarterly IST valve stroke testing.
REASON FOR CHANGE DC-3492 instatied a CCW system filtration skid on the CCW non-safety header. The filtration skid uses bag type filters to remove suspended solids in the CCW system bulk water. The filtration skid will assist in maintaining CCW system chemistry. The filtration skid requires acceptance testing to ensure the flow diverted through the filters does not adversely affect the components normally cooled by CCW.
SAFETY EVALUATION CCW is required to mitigate the consequences of a LOCA or MSLB by rejecting heat from containment to the atmosphere via the cooling towers. The proposed test places the CCW filtration skid into service to ensure the flow diverted through the filters does not adversely affect the components normally cooled by CCW In the event of an accident, the CCW filtration skid is automatically isolated with the non-safety header, if a single active failure on one of the headers were to occur, the opposite train equipment is available to mitigate the accident consequences. No new accidents are created by the proposed test and no margins of safety are affected.
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- 24. STP-99003523. Acceptance Test for DC-3523 l
DESCRIPTION
, This test functionally tests the Containment Spray Actuation Slanal (CSAS) manual override switches and annunciator installed in the circuitry for CS-125A(B). It also tests the Safety injection Actuation Signal (SIAS) manual overrido switches and annunciator installed for CC-807A(3), CC-808A(B), CC-822A(B), and CC-823A(B).
REASON FOR CWANGE Key switches allow operator to override the CSAS and SIAS signals and close the affected valves if it is de'. ermined t ' be in the best interest of protecting the health and safoty of the public.
SAFETY EVALUATION The CS, CC, and CCW systems are required to mitigate the consequences of a LOCA or MSLB but their failure will not initiate any accident. During thic test, the flow path from CS into containment will be isolated to prevent Inadvertent spray down of containment. No new system interactions will be created by this test. No protectiva boundary will be affected and no margin of safety will be reduced.
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- 25. - STP 99100019.B. Ultimate Heat Sink Cable Reroute Train B Rete.st DESCRIPTION This test will verify proper operation of UHS components s.ffected by the cable repull effort associated with exposed conduits that could be affected by a tornado mitcile.
REASON FOR CHANGE Condition Report CR-CS 1591 identifiej several conduits supplying control power to UHS components which are not protected from a tornado missile.
SAFETY EVALUATION The evaluation has concluded that no unreviewed safety questions exist for this test.
The test will functionally test the components affected by the cable rerouting but will not initiate or cause a LOCA, MSLB, or tornado, in the unlikely event of a LOCA or MSLB during the tost, redundant CCW Train A will be available. To prevent overflowing the CCW Surge Tank while the simulated low level is present, the B Train CCW Makeup Pump will be isolated from the CCW header and allowed to operate on recirculate.
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