ML20086N187
| ML20086N187 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/18/1991 |
| From: | Burski R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F291-0039, W3F291-39, NUDOCS 9112190133 | |
| Download: ML20086N187 (138) | |
Text
. , 10CFR50.59 O_Q .g. g, Entergy Operations,Inc.
I P r) B o T r ,,, 'ur
,,. y ' n y R. F. D urski e :
W3F291-0039 A4.05 QA December 12, 1991 U. S. Nuclear Regulatory Commission ATTN Docament Control Desk washington, D. C. 20555 ._
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 1991 Report of Facility Changes, Tests and Experiments Enclosed is the 1991 Report of Facility Changes, Tests and Experiments for Waterford 3 which is submitted pursuant to 10CFR50.59. This annual report covers the period from June 19, 1990 through June 18, 1991.
If you have any questions regarding this report, please contact G. E. Wuller, at (504) 739-6424.
Very truly yours,
- . sw RFB/GEW/pi Enclosure cca R. D. Martin, NRC Region IV D. L. Wigginton, NRC-NRR -
R. B. McGehee N. S. Reynolds ,
NRC Resident Inspectors Office i
/[i l a u '
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91121YO135 Y10618 \
4' Entergy operations, Inc.
Waterford 3 SES Docket No. 50-382 License No. NPF-38 k
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REPORT OF FACILITY CHANCES, TESTS AND EXPERIMENTS FOR 1991 PER 10CFR50.59
3 4
Entergy Operations, Inc.
Waterford 3 10CFR50.59 Annual Report for 1991 TABLE OF CONTENTS I. FACILITY CHANGES Report _ Page Number Document / Title No.
Summary................................................. 1 A. STATION MODIFICATIONS (SMs)
- 1. S.1-69, Installation of Isolation Valves for Charging Pump Relief Valves (Revision 0)....... 2
- 2. _SM-138, _ Installation of Refueling Water Level Indicating System (Revision 4)................. 3
- 3. SM-188, Install Strainers on Inlet Lines of Process Radiation Monitors (Revision 2)....... 4
- 4. SM-624, Concentrator Retrofits (Revision 3)........... 5
- 5. SM-818, Waste Gas Analyzer Modifications (Revision 7).................................. 6
- 6. SH-1297, Seismic Monitor SM-IYR-6020 Relocation (Revision 2)................................. 7-
- 7. SM-1539, Addition of Vent valves to Extraction Steam Piping.from Moisture Separator Reheater "B" (Revision 0).................... 8
- 8. SM-1844, Repair Leak of Post Accident Sampling System Undiluted Grab sample cylinder Connectors (Revision 0)...................... 9
-B. DESIGN CHANGES (DCs)
- 9. DC-3013, Snubber Reduction (Revision 4)............... 10
- 10. DC-3036,-Refuel Cavity Deep End to Containment Sump Hot Particle Filter Pipe Tee Installation (Revision 4)................... 11
- 11. DC-3081, Installation of Dionex On-line Ion Chromatograph (Revision 1).................. 12
- 12. DC-3093, Ventillation of Noble Gas in Pipe Chase Below Charging Pump Rooms (Revision 0)....... 13
- 13. DC-3104, Dissolved Hydrogen Analyzer and Recorder (Revision 1)....................... 14 i
Report Page Number Document / Title No.
- 14. DC-3105, Feedwater Heater Drain Modifications (Revision 1)................................. 15
- 15. DC-3120, Fuel Transfer System Surgo Suppressors (Revision 0)................................. 16
- 16. DC-3135, Containment Spray Pump Seal Leakage 17 (Revision 0).................................
- 17. DC-3141, Control Room Annunciator Reduction (Revision 0)................................. 18
- 18. DC-3147, Emergency Dienel Generator Emergency Start and Governor Control Circuitry (Revision 0)................................. 19
- 19. DC-3164, Additional Toilets for -4 RAB Women's Locker Room (Revision 0)..................... 20
- 20. DC-3170, Master Benchmark for Basemat Survey (Revision 0)................................. 21
- 21. DC-3177, Drain Lines Rerouted (Revision 0)............ 22
- 22. DC-3186, Steam Generator Manway Tensioner Rail Assembly Deletion (Revision 0)............... 23
- 23. DC-3202, Chemical and Volume Control System Purification Filter Changeout (Rerision 0)................................. 24
- 24. DC-3204, Replacement of Non-Safety Piping Components Located Outside Containment (Revision 1)................................. 25
- 25. DC-3209, Emergency Diesel Generator Local Annunciator Logic (Revision 0).............. 26
- 26. DC-3217, Main Steam Isolation Valve Partial Stroke Test and Soft Closure (Revision 0).......... 27
- 27. DC-3222, Pressurizer Safety Relief Valves Monorails (Revision 0)............. ................... 28
- 28. DC-322f. Permanent Test Cables to Support Procedure OP-903-069 (Revision 2)...................... 29
- 29. DC-3233, Elgar Bypasa Supply Transformer Input Breaker (Revision 0)......................... 30
- 30. DC-3254, Improved Charging ramp Blocks (Revision 1)................................. 31
- 31. DC-3258, Pressurizer Insulation Replacement (Revision 1)................................. 32 LL
Report Page Number Document / Title No.
- 32. DC-3266, Cutting and Capping Valve Leak-Off Lines 33 (Revision 2)................................
- 33. DC-3270, Main condenser Waterbox Sight class Addition (Revision 1)........................ 34
- 34. DC-3272, Auxiliary Component Cooling Water Pump Bearing 011 Cooler Modification 35 (Revision 1)..................................
- 35. DC-3286, Valve Actuator Replacement for Valves SI-405 A&B (Revision 1)...................... 36
- 36. DC-3294, Pressurizer Spray Valves Actuator Replacement (Revision 1)..................... 37
- 37. DC-3302, High Pressure Turbine Horizontal Joint Pipair (Revision 0).......................... 38
- 38. DC-3303, Movable and Fixed Incore Instruments (Revision 0)................................. 39
- 39. DC-3308, Reactor Coolant Pump Seal Replacement (Revision 1)................................ 40
- 40. DC-3316, Feedwater Heater 2B Impingement Plate (Revision 1)................................. 41
- 41. DC-3326, Reactor Coolant Pumps Insulation Replacement (Revision 1)..................... 42
- 42. DC-3327, Emergency Diesel Generator Air Dryer Reinstallation (Revision 0)................. 4?
- 43. DC-3328, Addition of Hoists Inside Reactor Containment Building (Revision 0)........... 44
- 44. DC-3331, Reactor Coolant Pump Gasket Leakoff Drain Lines (Revision 0)........................... 45
- 45. DC-3340, Reactor Coolant Pump Stud Elongation Test Hole Depth Increase (Revision 0)............ 46
- 46. DC-3344, Control Room Envelope Enhancements 47 (Revision 0).................................
C. WORK AUTHORIZATIONS / CONDITION IDENTIFICATIONS (WA/CI)
- 47. WA-01007647/CI-252577 - Boric Acid Makeup Tank Level Instrument Tubing..................... 48
- 48. WA-01068959/CI-272242 - Qualification of Base Plate for Main Steam Pipe Support MSRR-286........ 49
- 49. WA-01073263/CI-273778 - Reactor Fuel Reconstitution (Refuel Outage #4).......................... 50 LiL 1
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Report Page Number Document / Title No.
- 50. WA-01076246/CI-274976 - Installation of Dummy Instruments in Reactor Core.................. 51
- 51. CI-274719 (Revision 1) - Steam Generator Blowdown Containment Isolation Valves Instrument Air Regulators............................... 52 D. TEMPORARY ALTERATION REQUESTS (TARS)
- 52. TAR-89-052, Removal of Relief Valve on Feedwater Heater 6C and Installation of Blank Flange.................................... 53
- 53. TAR-90-013 (Rev. 1), Provide Acid Supply to the Demineralized Water System................ 54
- 54. TAR-90-014, Gag Condensate Valve CD 204A.............. 55
- 55. TAR-90-015, Install Mechanical Jumper from condensate Makeup to Primary Makeup System.................................... 56
- 56. TAR-90-016, Temporary Chillero for Containment Air conditioning.............................. 57
- 57. TAR-90-018, Reactor Coolant Pumpo 1B & 2A Gasket Leakoff to Reactor Drain Tank............. 58
- 58. TAR-90-019, Reactcr Coolant Pump 2A Replacement Inau1ation............................... 59
- 59. TAR-90-021, Installation of Water Softener for Supplementary Chiller condensing System.................................... 60
- 60. TAR-91-003, Disconnection of Two Defective Heated Junction Thermoce .e (HJTC) Sensor Heaters........ ........................ 61
- 61. TAR"91-004, Spent Resin Tank Liquid Level Indication. 62
- 62. TAR-91-005, QSPDS Channel 1 HJTC #5 Heater Disconnection............................ 63
- 63. TAR-91-006, Reduction of Containment Noise Levels..... 64
- 64. TAR-91-009, Temporary Chillers for Refueling 4 Outage 65
- 65. TAR-91-010, Temporary Access Point for Security to Maintain Accountability............... 66
- 66. TAR-91-012, Install a Plug in RCP 2A Case to Allow Repair............................. 67
- 67. TAR-91-020 (Rev. 1), Blind Flange for Valve SI-108B... 68 iv l
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.-.-_-,.._______m _ _ _ _ _ _ _
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Report Page Document / Title No.
Number
- 68. TAR-91-022, Provide Station Air to containment During LLRT of Penetration #8............. 69
- 69. TAR-91-024, Installation of Blind Flange Upstream Side of Valve SI-108B..................... 70
- 70. TAR-91-025, Temporary Instrument Air Service to Containment............................... 71
- 71. TAR-91-030, QSPDS Channel 1 HJTC 41 Heater D3sconnection............................. 72
- 72. TAR-91-031, QSPDS Cnannel 2 HJTC #4 Heater Disconnection.............................. 73 E. SPECIAL EVALUATIONS (1) DOCUMENT REVISION NOTICES (DRNs)
- 73. DRN #E-8902088, -2089, Emergency Diesel Generator 4.16 kV Safety Bus - Manual Synchronization...................... 74
- 75. DRN #M-8800406, Instrument Air System - riow Diagram.. 76
- 76. DRN #M-8800851, component Cooling Water - Flow Diagram.............................. 77
- 77. DRN #M-8800920, component Cooling Water System -
Flow Diagram......................... 78
- 79. DRN 8M-8801799, HVAC Air Flow Diagram................. 80
- 60. DRN #M-8900525, Component closed Cooling Water System............................... 81
- 83. DRNs #M-9100144, et al, Inconsistencies in Piping Designations........................ 84 E. SPECIAL EVALUATIONS (2) LICENSF DOCUMENT CHANGE REQUEST (LDCRs)
- 84. LDCR-89-147, Diesel Generator Loading Sequence....... 85
- 85. LDCR-90-0085, Net Positive Suction Head for Component / Auxiliary Component Cooling Water Pumps..................... 86 v
s Report Page l No.
Number Do :ument/ Title
- 86. LDCR-91-0157, Steam Generator Tube Rupture Analysis FSAR Update............................. 87
- 87. LDCR-91-0190, Component cooling Water Flr Under Accident Conditions (previt 'ly LDCR-90-0084).......................... 88
- 88. LDCR-91-0191, control Room Outside Air Intake Radiation Mcnitors..................... 89
- 89. LDCR-91-0207. Containment Penetrations and Isolation Valves................................. 90 E. SPECIAL EVALUATIONS (3) MISCELLANEOUS EVALUATIONS
- 90. Pump and Valve Inservice Test Plan (Change 1 - Revision 5)............................... 91
- 91. Pump and Valve Inservice Test Plan (Revision 7)...... 92
- 92. Discrepancies in Pump and Valve Inservice Test Plan.. 93
- 93. Pressurizer Safety Relief Valves..................... 94
- 94. SPEIR #91-E-001, Reactor Coolant Pump Seal Upgrado... 95
- 95. SPEER 191-608. Oil Level Probes in Reactor Coolant Pump Motor Reservoirs................. 96
- 96. SPEER #91-Gil, Emergency Diesel Genurator Electronic Governor Controle..................... 97
- 97. Cycle 5 Reload....................................... 98
- 98. Materials Management Organization Changes............ 100
- 99. Steam Generator Tube Plugging....................... 101 100. Firmware Change Package FCP No. 91-01............... 102 II. P?OCEDURES A. PLANT PROCEDURE CHANGES 101. OP-001-002 (Change A - Revision 8), System Operating Procedure - Reactor Coolant Pamp Operation........... 103 102. OP-003-003 (Change 1 - Revision 9), System Operating Procedure - Condensate-Feedwater..................... 104 ,
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Report Page Number Document / Title No.
103. OP-003-016 (Change 2 - Revision 5), System Operating Procedure - Instrument Air.......................... 105 104. OP-004-009 (Revision 4), System Operating Procedure -
Incore Nuclear Instrumentation....................... 106 105. OP-004-017 (Revision 4), Operating Procedure -
Vibration and Loose Parts Monitoring................. 107 106. OP-007-003 (Change 7 - Revision 8), Operating Ph ;edure - Gaseous Waste Management................. 108 107. OP-901-046 (Revision 7), Off-Normal Operating Procedure - Shutdown Cooling Palfunction............. 109 108. OP-903-030 (Change 2 - Revision 7), Surveillance Procedure - Safety Injection Pump Operability Verification........................................ 110 109. OP-903-032 (Change 5 - Re ision 7), surveillanco Procedure - Quarterly IST Valve Tests............... 1;.1 IV. OP-903-032 (Revision 8), Surveillance Procedure -
Quarterly IST Valve Tests........................... 112 111. OP-903-033 (Change 6 - Fevision 8), Surveillance Procedure - Cold shutdown IST valve T9sts............ 113 112. OP-903-033 (Change B - Revision 8), Surveillance Procedure - Cold Shutdown IST Valve Tests (Deviation). ....................................... 114 113. OP-903-117 (Change 3 - Revision 2), Surveillance Procedure - Containment Purge Valve Leak Test........ 115 114. OP-903-114 (Revision 3), Surveillance Procedure -
Local Leak Rate Test (LLRT).......................... 116 115. OP-904-005 (Revision 7), Surveillance Procedure -
Sprinkler and Spray Systems Alarm Test............... 117 116. UNT-005-013 (Revision 2), Administrative Procedure -
Fire Protection Program.............................. 118 117. HP-001-210 (Revision 7), Administrative Procedure -
Health Physics Instrument Control.................... 119 118. RF-002-001 (Revision 3), Refueling Procedure -
Fuel Receipt......................................... 120 119. Specifying Local Manual Action to Close Valves CS-117 A&B to Initiate Shutdown Cooling............. 121 vii l
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fleport Page Number Document / Title No.
B. SPECIAL TEST PROCEDURES (STPs)
A20. STP-027069-1, controlled Ventilation Area System (CVAS) Flow and Differential Pressure T0nt ................................................ 123 121. STP-255644-C (Change 1), Testing of Ambient and Servico Temperatures Inside Cantainment ............. 124 122. STP-01062461, Chemical and Volume Control (CVC)
Lutdown Radiation Monitor Flow Teet.................. 126 123. STP-01074040, Test to Establish Shutdown Cooling Flow Limitation........................: .......... 127 124. STP-99000404, Insta11ftion Testing of L A ' tor Coolant Shutdown Level Measurement s,etem (RCSLMS)... 128 125. STP-99003147, Emergency Diesel Generators A and B Start and Governor Control........................... 129 viii
Wattrford 3 SES 1991 Report of Facility Changes, Tests and Experiments
SUMMARY
This report provides the Waterford 3 Facility Changen made rursuant to 10CFR50.59(a)(1). The report covers the period from June 19, 1990 through June 18, 1991. None c f the items in this report represent an unreviewed safety quattion. No experiments or teste not described in the FSAR wate conducted at Waterford 3 during the report period.
The report identifies 100 Facility Changes (8 station Modifications, 38 Design Changes, 5 Work Authorizations, 21 Temporary Alterations, 11 Document Revision Notices, 6 License Document Change Requests and 11 Hiscellaneous Evaluations) and 25 Procedure Changes (19 Plant Procedures and 6 Special Tests).
1
- l. QC11,1TY CHANCES A. STATION MODIrlCATIONS (6Me)
- 1. SM-69, Installation of Isolation Valves for Charging Pump Relief Valves (Revision 0)
Poveri,pt_ ion of Change The modification installed three isolation valves, one for each of the three Charging Pump Discharge Relief Valves in the chemical and Volume control System (CVCS). The isolation valves are 2-inch, manually operated globe valves.
Reason for Change The addition of the isolation valves ellows the Charging Pump relief valves to be isolated for maintenance or testing purp ses during plant operations.
The lack of valve isolation became an identified problem in 1981.
T:le architect-engineer later recommended a plant change to add isolation valves via a Desigr. Change Notification, DCN-MP-992, initiated on 1/11/04. The station modification was initiated to implement the design change, subsequently, it was discovered that safety evaluation documentation per 10CFR50.59 could not be located with the old modification package and a Quality Notice was issued to rectify the document omission.
Safety Evaluation The facility change by SM-69, with the addition of isolation valves for maintenance and testing, preserves the safety functions of CVCS. The isolation valveu at' locked open and are only utilized to isolate the Charging Pump discharge relief valves when necessary. The isolation valves do not impact the function of the reliet valves and will not affect accidents associated with CVC3.
FSAR Figure 9.3-6 (Sheet 2), Flow Diagram Chemical and Volume control System, reflects the isolation valves installed by SM-69 for the three charging pumps.
2
- 2. SH-138, Installation of Refueling Water Level Indicating System (Revision 4)
Description of Change The station modification installed a Refueling Water Level Indicating System (RWLIS) onto the Reactor Cooling System (RCS).
RWLIS provides level indication in the RCS when conducting maintenance activities requiring partially drained RCS conwitions.
Both local and remote indication in the Control Room is provided, together with alarms for high and low water levels.
The modification installed narrow range and wide range level transmitters connected to an RCS hot leg drain and to a RCS high point attached to a pressuriser level sensing line. Also included in the modification package was a personnel platform to allow technicians to have easy access to the RWLIS transmitters for calibration and maintenance.
Reason for Chango The modification was made to provide improved operator awareness of the water level in the RCS during various maintenance and refueling evolutions. Continued practice was deemed unacceptable for reliance on the unreliable temporary standpipe for local indication of RCS water level during refueling. The current level indicating system could not be consistently relied upon to provide 6.1 curate indication during RCS draining or refilling operations.
SM 138 was implemented to provide a reliable, durable monitoring system during all phases of a refueling and/or enintenance outage.
SM-138 fulfilled the commitment made in response to the NRC Generic Letter 87-12 to install a permacent water level monitoring system.
Safety Evaluation RWLIS does provide increased operator awareness for RCS level.
The station modification does not affect any safety related equipment or any existing analyzed accident. Plant hardware required for safe shutdovn of the plant is not impacted by the facility change. RWLIS ;e described in FSAR Section 5.4.16.
3
- 3. SH-188, Install strainers on__ Inlet Lines of Process Radiation i
Monitors (Revision 2)
Description of Change Strainers were installed in the inlet lines for the Process Radiation Monitors located in the Dry Cooling Tower Sump, Industrial Sump, and the Circulating Water Sump.
Reason for Change The strainers were added to prevent clogging of the process radiation monitors. Implementation of the station modification was completed in January 1988, however, the evaluation at the time failed to revise the affected FSAR figures. Subsequently, the safety evaluation performed as part of the corrective acticq for a Quality Notice revised the drawings for FSAR figures to reflect the installed strainers.
Safety Evaluation The precess equipment is not altered or operated in an abnormal manner and the radiation n iitorn continue to perform the intended design functions. The installation of the strainers had no affect on any procedures describad in the FSAR, and no Technical Specification change was needed. The radiation monitors serve no accident mitigating action and do not affect plant equipment important to safety.
The operation of the Process Radiation Monitors (PRM-IRE-6775, 6776, 6777, 6778 and 1900) is enhanced by the straincts. The addition of the strainers on the process lines by SH-188 necessitated the updating of FSAR Figure 9.3-3 and 10.4-5.
4
- 4. SM-624, Concentrator Retrofits (Revision 3)
Description of Chance The station modification was designed to improve the operation of the three plant concentrators, namely, the Doric Acid, Boric Acid /Wasta and Water Concentrators. The proposed modification involved (a) Installation of a vent condenser to condense and remove water vapor to eliminate corrosion in the vont header, (b) Helocation of the flow element in distillate discharge system, and (c) Adding a protective trip to the concentr? tor heaters to prevent burning out low concentrator
- c level.
Revision 3 of SM-624 alter ;he modificatir.n package to deleto specific items of the original plan. Changes by Revision 3 wares (1) the installed vent condenser was not tied into the Vent cas Collection llender, (2) the flow element for the Waste concentrator was not relocated, and (3) the requirement to modify the concentrator heater circuit to cut-out on low level was deleted.
Reason for Change The facility changes were intended to improve the operation of the three plant concentrators.
Safety Evaluation The facility change made by SM-624 are deemed to be design and operational enhancements. The plant concentrators are not required for sufe shutdown of the plant. The modifications completed have no offeet on pinnt safety systems, an6 do not impact the FSAR accidents previously evaluated. Implementation of the facility changes did not require a change to the Technical Specifications.
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S. SM-818, Waste Oas Analyrer Modifications (Revision 7)
Description of Change The station tuodification, over a period of time and modification revisions, replaced the Waste Gas Analyzer System. The new system including panel, analyzers and sampling system meets or exceeds the requirements of the previous system. The purpose of the Wasto Oas Analyzer system is to sample the waste games from various sources atid provido alarm should high levels of oxygen or hydrogen occur. Annunciation is provided in the control room and locally for any abnormal condition.
Reason for change The original Automatic Gas Analyzer Panel was not reijable and proved to be susceptible to numerous problems. Difficulties included moisture affec*.ing the hydrogen analyter, crud build-up in the sample solenoid valves, poor pressure regulttion, and maintenance and operational problems.
Safety Evaluation The new system provides design and operational enhancements and improves safety and sample conditioning. The Waste Gas Analyser System installed via SM-838 is non-safety, non sciamic, quality related. The analyzer room and devices inside the room are installed explosion-proof.
The new Wasto Gas Analyzer does not create a possibility for an accident or equipment malfunction of a different type than ,
previously evaluated. Margin of safaty is not reduced in any a Technical Specification bases by the installed new system.
The implementation of SM-818 did require a change to the plant technical upacifications because monitor calibration was increased from quarterly to monthly. Technical Specification Table 4.3-9 was changed and issued by the NRC under License No. NPF-38, Amendment No. 56.
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- 6. SH-1297, Soismic Monitor S't-IYR-6020 Relocation (Revision 2) l Description of change Seismic Monitol SM-IYR-6020, was relocated from the top of the preocurizar to .he lower lifting lug of Safety Injection Tank 18.
Peason for change The seismic monitor was found to be damaged by heat from its location on top of the pressurizer.
Safety Evaluation The seismic monitor has no part in mitigating the consequences of an accident. The monitor performa no active function and its f ailure does not af fect other equipment. The relocation and installation of the seismic monttor by SM-1297 continues to meet the seismic catagory I requirements. Revision 2 of SM-1297 provided drawing change to reflect in FSAR Figure 1.2-17 the correct location of Seismic Monitor SM-IYR-6020.
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- 7. SM-1539, Addition of Vent Valves to Entraction Steam Piping from ~
Moisturn Separator Reheater "hd (Revision 0)
Description of Chance Three 1-inch vont valves were added to the Extraction Steam Piping coming from Moisture Separator Reheater "B". The architecta engineer /constructos made the facility change and udded the vent valves in July 1978 via Design Change Notification, DCN-MP-34. At that time the safety evaluation was not prepared for the change and inadvertently the engineering drawings (flow diagram) and applicable FSAR figure were not reviced :n show the vent valves.
Reason for Change The three vent valves serve as high point vents for the extraction steam lines. The valves are used only for venting during off-normal operations (i.e., testing, maintenance, startup, shutdown). During normal plant operations the vent valves are closed and capped.
The 10CFR50.59 safety evaluation was prepared much later in May 1990 for a document correction etation modifictition (SM-1539) which was issued to add the vont valves to the flow diagram drawing and revise FShR Figure 10.2-4.
Safety Evaluation SM-1539 af f acted only the Extraction Steam Systern which is non-safety, non-esiomic. No protective safety boundary or margin of sefety is imparced. The addition of the vent valves does not altee the ability of the Extraction Steam. System to function as described in ths FSAR and the valves provide no interface with safety systems, structuren or components.
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B. SH-1844, Repair Leak of Post Accider.t Sampling System Undiluted Grab sample cylinder Connectors (Revision 0) pescription of Change The station mol.ification provided the temporary addition of a shott section of braided stainless steel hose for use in quick-connection of the undiluted grab sample cask. SH-1844 was only partially implemented and further redesign of the Post Accidont Sampling System (PASS) was nouded and subsequently handled via Design Change DC-3102 to allow cloneout of SM-1844.
Rog,oon for change Due to restricted visibility around the PASS undiluted grab sample cart, because of shielding, 2.t was difficult to determine if cask connectors were engaged and frequent sample spillage resulted.
,S,afety Evaluation Post Accident Bampling System (PASS) is not a safety related eyetem but is considered important to safety. The grab sample provision is not an essential part of PASS but is a part of the syste:n to be used during poet accident recovery activities. The facility change by SH-1844 provided a temporary jumper until an enhanced system redesigr. wt. implemented. The alteration did not increase the possibility or consequences of any FSAR analytod accident. No T<achnical Specification change was needed for SH-1844.
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I. B. DESIGN CHANGCS (DCs)
- 9. DC-3013, Snubber Reduction (Revision 4)
Description of Change The design change removed or replaced existing snubbers througbcut the plant by rigid supports and support modifications.
Re, aeon for Change impoll Corporation perforr.ied a snubber reduction evaluation of piping systems for Waterford 3 to avoid high testing failure ratos of snubbors, and to eliminato the over-conservatism in seiomic analyses of the piping systems. The evaluation was done for 49 piping system calculations and resulted in the elimination of 153 anubbers. DC-3013 was initiated for the removal or replacament of the unneeded existir:g snubbers, and te perform modifications to compensato for revised loadings or pipe movemente.
Safety Evaluation Per the re-designs, the piping and supports altered by DC-3013 are qualified to meet the design requirements. The snubber reduction was accomplished without impacting the current licensAng requirements. The design chango achieved both ecoromic and rollability gains by reducing the use of snubbera.
Elimination of snubbers improves plant operability, outage management, ALARA, and maintenance and repair costN associated with snubbers. DC-3013 did not change or modify the pipe routing or configuration; changen it.volved only the removal or modification of piping supports.
The reanalyrod piping and supports are fully qualit19d to meet the FSAR design connitments. Increases in pipe thermal, solemic, Loos of Coelant Accident movement were ovaluated and are acceptablo.
The facility change only involved the removal and/or modification if existing pipe supports. No changes or modifications were mado to pipe routing or configuration, postulated pipe break locationo, nor pipe whip restraint locations. No change occurred in the function, capability. or service life of any of the piping systems or any associated equipment. The analycos which support the design change address and meet all curront I.icensing TSAR, and Dissign Crit. aria requirementa.
The design change package provided document update of piping isometri.e drawings, piping pupport shotches, and supporting calculations.
Implenentation of 0c-3013 did not require a change to the Pisnt Technical Specifications, or any cesign bases.
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__ _ = _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ ____ _ _ _____ _ _ __
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- 10. 4efuel cavity Daeg End to Contalrut. ant Hump Hot Particia ~
DC-3036, $ ter Pipe tee lntTtallation (Re vTsion 31 ,
Tioge,Lytion l of Change, A f.n',ty roanaal ball valva (SP-1014) in the drain lino from the doop end of the roactor cavity to the containment aump was raptaced by another tranual ball valve of dif ferent manut'acture.
A sacond manan1 bali valve, iduntical to the replacement valve above, is inrnallod su an isolation valvo in the drain line frorn Fan Cooler AH 1(3D-SB) the.t connects to the above sump line (SP-10131).
f<eaaon for change Phile perfortning its function of containtnant Sump isointion during Hafueling Outage 2, a manual ball v/nive (flP-1014) allowed contaminated water to enter the loop seal of Contairenant Fan Cooler AH-1(,la-4a). To prevent recutrence, a ball valve identical to the SP-1014 replacement was installed as .sn isolation valve (SP-10131) for the cooler.
Saf1 ty Evaluation The only changes in this goodificat!.on are enhancaments to dreinage incido contaiturient for ALARA purposen and not to any material or the functioning of safety-related equipment. 110 changee hAve been made to thn functioning of nyetcme that may releare radioactive material to the environment. The modifi':ations are enhancements to tha plumbing and nump syatem and have no arfect on any enfety-related equipment or nycton. This change has no equipment that could be considered important to safety and so cannot increase the probability of rt.altunction of equipment which 10 Arnportant to anf oty. The change has not af fected any equitxtent or system that may releano radioactive rnatorial to the envirotunant.
The material in this modification has no safety-related function-and the changea completed to the plumbing and nump system have no impact on Technical specifications or their basen.
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- 11. DC-3081, Instal 11tdon of_ Diognx on-line Ioa, chromatograph (Revieton 1)__
legel l ption i of Change, The design change insta11od a Dionex on-line ion chromatograph which provides connections to nineteen sample lines (eleven permanent and eight potential), 115V power, air at 120 pai, llelium at 50 poi, and domineralir.ed water. This desigt. change only affected the secondary sample lines.
Roanon for Change The Dionex lon chromatc<"'aph etthancou abomii,try ability to evaluate the chamistry of the secondary systura condensate.
Sa f ety_ _ Evaluation, The f acility change af fected the nocondary narnpling system only which is a non ostety, non~ssismic system with no special code requiremante. The pabability or consequences of an accident will not be increased because the analysis used for the Secondary campling is still valid and all isolation remains intact. The possibility of an acaddent which is different wil.'. iot be increased because
- o 3ps are in the secondary gartpling system which hoc been evai. t + :. . 'the malfunctio.s conalderations of equiteent iraportant to saf ety are unaf fectesd bncause samplir.g only is af f ected and all isolations ca.tne.in intact, No othat' equipment irnportant to safety is af fected. Tha margin of safety will not be reduced. Secondary su pling is not addressed in the Technical Specifications.
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- 12. DC-3093, Ventillation of Noble Gas in Pipe Chant Delou Charging I'unp Rooms (hevision O)
Description of Change The design change added a f our inch diatacter ventilation duct from the reactor auxiliary building (RAD), normal ventilation system oxhaust duct (overhead in charging } vm room A/P) down, through the floor and into the pipe chase (ns adiation or fire barriers are penetrated). This duct provides an air exchange approximately every 51.19 minutes and prevent excessive build-up of Noble gas and hydrogen. A volume control darper is provided for flow regulation and isolation when not in use.
Reason for Change This change will prevent Noble Gas from migrating into the Doric Acid Make-up Tank (DAMT) rooms because of accumulations present an the pipe chase below the charging pump rooms.
Safety Evaluation There were no safety-related or important to safet systems or components modified. This modification providas s stilation to a pipe chaos which has been a stagnant area. The hangers for installation are categorized as non-seismic but qu7lity related, and are designed to retain structural integrity during and after a meiomic event. No safety or other quality considerations are applicable and no fire or radiation barriers are being penetrated.
Therefore, the accident and equipment considerations are bounded by existing analysis. The margin of safety is not reduced.
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- 13. DC-3104, Dissolved Hydrogen Analyrer , i Eevordor (Revision 1)
Description of Change The design change replace. a ten.g srcry hydrogos. analyzer 1 ith a permanent installation, and added a rocorder and a calibrktion gas line.
Reampn for change The analyzur installation was unsatamfactory in the present configuration. Present routing of the datector cable and the gas calitration line breacles the Primary Sampling Panel sealed cabinet which is required for ALARA purposes. Also, the gas calibration line presents a tripping hazard as presently used.
The sample line from the Primary Sampling Panel to the hydrogen detector la suspended with temporary brackets which could f ail and cause an ALARA problem.
Safety Evaluation There are no accident evaluations la the FSAR for the primary sampling system (PSS). Therefore, the probability of an accident previously evaluated cannot be increased, or can the consequences be increased. Tha PSS is a passive systen and does not af fect other systems or equipment. Therefore, an accident different than any already evaluated jt the FSAR canrot be created. Since there is no safety-related egolpment in tai- installation, the pLobability of malfunction of equipnent important to safety previously evaluated in the FSAR will not be increased. Since thio does not contain anc toes not interact with any safety-related ogtipment, a malfunction will not increase the consequence to any sarety-related equipment. Since the system is not safety-relatnd and isolated ! rom other safety-related dystems, the possibility of a malfunction of equipment important to safety different thte) day already evaltated in the FSAR will not be created and margin of stinty cannot be reduced.
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- 14. DC-3)05, FePdwater Heater Drain Modificatiend (Revision 1)
Deseriptton of Chage The design modifleation changed the actuators on the emergency Feedwater Heater Drain Line control Valves $6A, B and C to fail open instead of fail closed. The design change initially installed a perforated plate over the existing slotted openings of both the normal and emergency discharge pipes. This change was made to reduce the discharge area and increase line backp essure to ensur! single phase flow during plant operations.
Subsequently, via Revision 1, new drain valves of higher capacity were installed on the f 6A,15 and C feedwater heater normal low pressure drain lines at a lower elevation and closer to the condenser to ensurf rdequate head and required design flow. The sparger discharge area was also increased for eain normal drain.
Reason for Change The normal drain lines for she 86 Paedwater Heaters did not provide sufficient capacity to pass the required flow at normal 100% flow conditions. The lack of flow capacity resulted in the emergency drain lines opening in an attempt to maintain a proper heater level.
Safety Evaluation The normal and emergency feedwater heater drain lines are non-safety related, non-seismic and ic -quality related. The feedwater heater drains help to protect the non-safety related main turbine from water induction.
The feedwater heater drain system does not connect or interact with any system or equipment important to safety. No accidents analyzed in the FSAR are initiated or mitigated by the heater drain system.
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- 15. DC-3120, Fuel Transfer System Surge Suppressors (Revicion 0)
Description of Change The design change permanently installed ten surge suppressors in Oc3 trol console #2 cf the Fuel Transfer Machine. Surge suppressors had previously been temporarily installed, and used with satibtactory results during Refueling Outage $2, to overcome control circuit interference difficulties. Following the outage, the suppressors were removed and the temporary alteration was clos.4. The permanent installation is equivalent to the previously proven installation.
Reason for Change During Refueling Outage 42, a problem was expurienced with the Fuel Transfer Machine. The machine would trip and the overload alarm would ring when a carriage transfer was initiated from either control console. The machine vendor identified the problem as control circuit electromagnetic interference caused by contact arcing in motor starters. DC-3120 was initj ated to permanently install surge suppressors to avert future control circuit problems. Surge suppressors were installed across the Ac input to the electronic weighing panels and across each contact of the transfer machine carriage motor forward, reverse and brake contactors.
Safety Evaluation No credit is taken for components or subsystems of the fuel handling equipment to prevent or to mitigate the consequences of the postulated fuel handling accident. The Fuel Transfer Equipment Let is designed as non-nuclear safety and non-safety seismic. The facility change does got alter the operation or possible failure mode nf the Fuel Transfer Machine. No Technical Specification r.hange was needed for DC-3120. The facility change in a maintenance design change to prevent falso overload trips and enhance the operation of the Fuel Transfer System.
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- 16. DC-3135, Containment Spray Pump seal Leakage (Revision 0)
Description of change The design change replaced the seals in the Containment Spray Pumps. The pumps had previously been equipped with mechanical anals backed up by packing which required constant cooling by compnent cooling water. The installed newly designed mechanical seal employs a carbon steel bushing that does not require cooling water. A drain linu was also added to remove mechanical seal leakage.
Reason for Change The packing for the old pump seal dcsign required constant water cooling. The pressure of the Component Cooling Water system was too high for the packing type seal and caused excessive leakage part the packi.sg onto tho floor. the new pump seal design does not require cooling water and leakage is minimized.
Saf6ty EvaluatAon The Containment Spray Pumps provide water to be sprayed into the Reactor Containment Building for cooling during an accident. The pumps are safety related equipment designed to mitigate the conocquences of accidence. The new seals are essentially the same as the old seals except for use of a carbon steel bubbing for backup rather than packing which required cooling water.
The new seal design is a design and oporational enhancement as seal leakage is minimized. The equipment change does not materially alter the function of the safety related containment Spray Pumps for the mitigation of accident consequences. No Technical Specification changes were needed for the facility change.
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- 17. DC-3141, Control Room Annunciator Reduction (Re, vision 0)
Description of Change The design change covered the elimination or modification of 17 control room annunciators. These annunciators are not safety-related. The BLOWDOWN MAGNETIC FILTER TROUBLE, RADIATION MONITOR ACTIVITY HI AND SYSTEM TROUBLE, STORAGE TANKS AREA INST CAB TEMP HI/LO, INTAKE AREA INST CAB TEMP HI/LO AND TURBINE BUILDING TEMP HI/LO are being eliminated. REACTOR AUXILIARY BUILDING (RAB) UPPER LEVELS INST CAB TEMP HI/LO, CNTMT PEN AREA INST CAB TEMP HI/LO and eight TRAVELING SCREEN TROUBLE annunciators were modified to reduce the number of instruments that cause each annunciation.
Reason for Change Operations had identified 17 control room annunciators that are unnecessary and cause distractions to control room personnel.
The design objective was to eliminate or modify the circuitry controlling the 17 annunciators to reduce control room distractions, eliminate the annunciator if there is redundant indication or the annunciator is absolutely not necessary, and modify the circuitry to eliminate unnecessary actuation and retain the valid actuations.
Safety Evaluation The blowdown filter trouble annunciator is duplicated at the local control panel, only local operation is used. The rad monitor high activity annunciator is duplicated at the RM-ll panel in the control room. The accident probauilities or consequences are not affected by the annunciator operation. There are no new failure modes or plant responses introduced by this design change. Plant equipment does not change because of this change. The probability or consequences of malfunction of equipment important to saftty previousil evaluated in the FSAR will not be increased because the annunciators are non-safety and the functions are duplicated. No new f ailure modes are introduced and the equipment af fected is not important to safety. By eliminating duplication and distraction, operations personnel will be able to concentrate on possible malfunctions of equipment important to safety. The annunciators are not required by Technical Specifications and no margins of safety are reduced by this design change.
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- 10. DC-3147, Emergency Diesei cenerator Emergoney Start and covernor control circuitry (Revision _01 Descriptic n of Change The design change replaced the control relays f or the Emergency Diesel Generators (EDGs) with mercury wetted rolsys as reconmonded by the EDG vendor, Cooper Beanomer; and modifA.ed the control circuitry.
Five design objectivas for DC-3147 werwt
- a. change out the control relays to a typo less susceptible to oxidation,
- b. add a new relay for transfer of the EDG governor to the emergency mode to improve reliability and eliminate falso alerms and provide the redundancy feature to the omorgency start circuit,
- c. provide circuit changes to allow paralleling across the bus tie or the generator breaker for the EDG omergency modo paralleling to facilitate testing and reduce the number of tout starts,
- d. automatically terminate generator field flash when the EDG fails to start in the emergency mode, and
- e. dosi n a method to provido start air to recover from Station Blac..out.
Reason for Change
- 1. The original control relays associated with the EDG engino governor were subject to contact oxidation which caused changua in resistance and affected EDG frequency control.
- 2. The initial transfer of the EDG to the emergency mode was not sufficiently reliables gave falso alarms and did not have a redundancy feature.
- 3. Emergency modo paralleling was desired to facilitate testing.
- 4. Improvements were needed to handle emergency modo start failures.
- 5. The facility change provided a backup starting air supply for the EDGs via a portable skid mounted diesel driven compressor unit.
Safety Evaluation The Emergency Diesel Generators are the standby power supplies for the onalto power distribution system. The design chango was initiated to enhance control performance and provide redundancy features.
The EDG system is classified safety class 3 and Seismic Category I. The design changes made by DC-3147 did not change the safety function tf the EDGs.
Facility changes made by DC-3147 do not fundamentally change EDG control or function and do not compromise plant safety er licensing basis. No Technical Specification change was needed for the design change.
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- 19. DC-3164, Addist mal Toilets for -4 RAB Women's Locker Room (RevieTon 0)
Description of Change This is an interim modification that installed two new commodes and repositionei :wo existing sinks. Waste, vent and supply lines in the Sanitati... and Potable Water Systems are affected. There are no safety or seismic considerations.
Reason for Change The increase in the number of female workers has shown the need for additional facilities in the ~4 el RAD women's locker room.
safety Evaluation This modification adds two toilets and repositions two sinks in the -4 el RAB women's locker room which is outside the radiation control area. The nature and location of this work, therefore, provides no basis for accident condition changes, malfunctions of equipment important to safety, or a change in the margins of safety. This evaluation documents a review for floor plan changes appear'.ng in FSAR Figures 1.2-10, 9.5.1-4 and 12.3-1b.
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- 20. DC-3170, Master nonchmark for Basemat survey (Ravision 0)
Description of Change The design change installed two benchmarks on the plant site with protective bumper posts. The benchmarks are a steel bar approximately one inch in diameter and are driven approximataly sixty-five feet into the ground.
Reason for Change The new benchmarks will reduce the amount of time required to complete the basemat settlement survey, as the existing benchmark is approximately one-balf to one mile from the plant site.
Safety Evaluatiorj .
The basemat provides the foundation for the reactor building, fuel handling building, and reactor auxiliary building. Basemat integrity is fundamental to all facets of plant operation and all transient responses. The change affects only measuring basemat settlement. Basemat settlement is not a direct factor in accident analysis assumptions. Thus, the me.suring program change cannot affect the probability of accidents.
The change affects only the master benchmark location from which the basemat is surveyed. The physics of how or why the basemat settles is unaffected by this change.
The surveillance program for the nuclear plant island structure common foundation basemat provides continuing basemat integrity assurance. The points selected for monitoring will show any unusual behavior of the basemat and will be useful in addressing the significance, if any, of such behavior. The program has action limits such that if measurements reach a limiting value, the cause and its effect upon the integrity of the basemat can be assessed. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously synluated.
The basemat measurements provide data to compare to action levels for basemat settling. A new benchmark provides a new reference point from which to measure basemat settlement. No operating related failure modes are created or changed. More accurate measurements will be obtained by bringing the benchmark closer to the site. Accuracy implies measurement repeatability. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated. Basemat settlement is not included in Technical specification bases and no safety margins are altered by moving the 5enchmark. FSAR Figure 1.2-1, Plot Plan, shows the locations of t.le benchmarks.
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- 21. DC-3177, Drain Linen Herouted (Revision 0)
Description of Change The design change disconnected the existing drain connection on fire protection sprinkler test valve FPM-21 from the radioactive floor drain system and reconnected it to the non-radioactive sanitary Waste system. Additionally, the existing sink drain in the Health Physics Count Room was discot.nected and reconnected to the radioactive Detergent waste drain system.
Reason for Change The fire protection sprinkler system must be tested by procedure periodically which unnecessarily burdens the Waste Management system with non-radioactive waste water.
Health Physics had previously discontinued use of the sink located in the Health Physics Count Room for surveying radioactive liquid samples because the sink drain had been connected to the non-radioactive manitary waste and vent systems.
Safety Evaluation . . ,
The terouted service sink drain was provided both a "P-trap" and manual isolation valve to reduce the possibility of radioactive gases in the Laundry Drain System from entering the Health Physics Laboratory.
Neither function of the Sanitary Waste nor the Detergent Waste Systems are affected by the facility changes which in essence are minor piping changes. No testing is needed for the rerouted drains that requires systems to be operated in an abnormal manner.
Rerouting of the Health Physics sink drain minimites the possibility of an unintentional, unmonitored radioactivity release from this source and it provides a design /cperational enhancement.
Technical specifications were not impacted by the DC-3177 facility changes.
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- 22. DC-3186, Steam Generator Manway Tensioner Rail Assembly ~ Deletion (Revision 0)
Description of Change The design change permanently removed four steam generator manway rail assemblies and their structural steel supports, and installed grating in the vacated areas.
Reason for Change Haintenance has not used the manway cover handling devices since the Refueling fl Outage because of the awkwardness and inefficiency when using the system. Maintenance currently uses Westinghouse special tools for handling the steam generator manway covers.
Safety Evaluation Removal of the rail casemblies provided a valuable space gain in the congested confines of the steam generator platforms. Use of the speciality toolr for handling the manway covava results in a significant reduction in maintenance manhours and increases the manway handling efficiency. the design change is considered an enhancement for conducting a maintenance / refueling activity.
The facility changen by DC-3186 do nec involve changes to any equipment, componente or systems triportant to safety. No system testing is required by the changes and there is no Technical Specification involvement. No r..u system interactions or connections result from the design hange.
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- 23. DC-3202, Chemical and Volume Control System Purification Filter Changeout (Revision 0)
Description of Change The design change permits the Chemical and Volume Control System (CVCS) purification multiple filter cartridge sizing to be less than or equal to 20 microns absolute. The Chemistry Department is responsible for evaluating, determining and selecting the appropriate filter sizes, msdels and brands to be used in various stages at different plant conditions.
Reasons for Change A significant portion of radioactive crud in the CVCS was not being removed by the current use of 20 micron absolute filter cartridges. Significantly more efficient removal of cr-51, co-58 and co-60 is expected by the use of smaller size filtering cartridges.
Safety Evaluation The use of various size filter cartridges in the CVCS purification train, dependent upon the plant conditions, will taximize the radioactive crud reduction from the Reactor Coolant System, and will minimize dose rate.
The CVCS purification filter removes particulates from the letdown flow upstream of the Purification Ion Exchanger. The replacement of the cartridge filters to achieve removal of smaller particles does not affect the safe operation of the plant. The design change only permits changeout of the filter cartridge. Other aspects of the filter vessel design, operation and location are unchanged.
The facility change by DC-3202 has no affect on the initiation or mitigation of the FSAR analyzed accidents. The filter cartridges are not safety related components and the function or failure has no impact on equipment or systems important to safety. No new systems or system interfaces are created by the DC-3202 design changes.
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- 24. DC-3204, Replacement of Non-Safety Piping Components Located Outaide containment (Rovision 1)
Description of Changa The design change provided the information for nocessary replacement due to erosion / corrosion of the piping componente examined by the Inservice Inspection program during the Refuel 4 Outage. A large number of oncondary piping components located outside the containment wore selected for examination. The degraded carbon steel componento rejected by the inspection were replaced with alloy steel of low chromium / molybdenum content (i.e., 2-1/4% Cr; 1% Mo). Revision 1 of DC-3204 provided the Document Revision Noticos on changes for the affected documents, e.g., piping isometric drawings, line lists, piping stress calculations on the replaced piping components.
Roanon for Chorge Wall thinning occuro in carbon steel piping due to oronion/ corrosion. The ISI program examinations identified the unacceptable, degraded piping components that needed to be replaced during the Refuel 4 outage.
Safety Evaluation The replacement piping components meet the original design code requirements. Replacement of the carbon steel components with the low chromium / molybdenum alloy steel is considered a design enhancement as the new piping components ars more corrosion ronistant. The replacement piping continues to maintain the proosure boundary for the fluid systems involved. Only non-safoty piping components in secondary systema outside containment were involved in DC-3204. Replacement of piping componente with similar components of different material does not decrease but rather increases the expected lifetime of the components.
The facility changes do not result in a challenge to the safety of systems involved and do not cause plant systems to be opurated outside design or test limits. The likelihood of an accident is not increased by the design change. No Technical Specification change was needed for DC-3204, 25
- 25. DC-3209, Emergency Dieral Generator Local Annunciator Logic (Revision O_)
Drncription,of Change DC-3209 changed the Emergency Diesel Generator (EDG) local annunciator window 8-3 r.omenclature and logic.
DC-3209 added three jumpers and changed one annunciator windov at the EDG - A and B control panels. These changen allow the local sync selector switch, 43 MAS, to be in manual, maintain an annunciator black board and eliminate nuisance alarms. The new off normal condition of " Local Manual / Auto syne Switch Aligned for Auto Sync" is alarmed.
Reason for Change The EDG local sync eclector switch must be left in auto to prevent getting the local annunciator alarm window 8-3, " Control Switchen Not Proper for Remote Auto Operation". With the switch in auto, a local start or transfer of the EDG to local control with the EDO running will result in the immediate paralleling of the EDG with off-site power by the auto-synchronizer. Operatione desiren to leave the local sync selector switch in manual and still maintain the black board concept. This is consistent with the policy of manual paralleling only.
Safety Evaluation The design change prevents inadvertent auto sync from occurring when EDG control la transferred from the reactor turbine generator board to the local station. The change has no effect on the EDG emergency mode and accident response.
DC-3209 doen not effect the emergency mode response of the EDGs.
The EDGo will continue to operate in the same manner as designed to limit the consequences of LOOP (lose cf offsite power) and LOOP with eingle active failure (SAF). DC-3209 does not change the emergency mode controls of the EDGs and cannot increase the probability of an EDG malfunction. The EDG annunciator logic for auto sync is the only circuit involved.
Failure of one diesel generator will not change the consequences of a LOOP, or LOOP with SAF. The system is designed and evaluated to consider a single active failure.
A change to the auto syne annunciator window will not create a new method of failure for the EDGE.
The proposed change does not change the emergency mode response or loading of the EDGs and will not change the margin of safety as previously defined.
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- 26. DC-3217, Main Steam Isolation Valve Partial Stroke Test and Soft closure (Revision 0)
Description of Change The design change modified the control circuit logic for the Hein Steam Isolation Valves (MSIVs) partial stroke test from totally automatic (upon initiation by the operator) to a manual test by the eperator. Additionally, the design change provided local manual control of the inlet valve to give soft closure of the MSIVs during shutdown.
Reason for Change Improvements were desired to increase the reliability of the partial stroke testing of MSIVs. Enhancements of the plant configuration were also planned to extend the life of plant safety equipment.
DC-3217 improved the reliability of the partial stroke test of the MSIVe with the modified control circuit logic. By providing for soft closure of the isolation valves the valve stems are protected from unnecessary excessive closure forces aa created by the design basis requirement of 3-second closure.
The two features provided by the design change are expected to extend equipment life.
Safety Evaluation The Main Steam Isolation Valves (MSIVs) are provided to isolate the steam generators from the remaining portions of the secondary system in the event cf a loss of coelant accident or a main steam line break. The safety function of the MSIVs is not affected by DC-3217.
The partial stroke and full stroke tests of MSIVs are described in the FSAR. The test requirements for the MSIVs are maintained by the design change and no other tests are required by the f acility changes. The consequences of an accident are not increased by the changes to the control circuit logic. The plant modifications likely reduced the probability of an inadvertent closing on an MSIV. The design change did not introduce any new system interactions, connections or system perturbations. No Technical Specification change was required for DC-3217.
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r 1222, Pressurizer Safety Relief Valves Honorails (Revision 0)
[ ! 7,tcaiption of Change The design change instat.ed two monorails ano two trolley S4 ses;.*Llies in the Pressurizer Cubicle to aid in 9moval and jf5>
e di
_r. allation of pressurizer safety relief valver :SRVs). The mu .. rails a>w armanently attached to the ceiling . af the Presourizer i C above each SRV, The trolleyL are equipped 1
aith side plr assure structural integr1*y while experiencing vibratioc, or ) ' r , loads.
' Re 7su .. for Change.
It cone.inutd to be difficult and time consuming to remove and instal, the pressurizBr safety relief valves. Four padeyes were welded to embeddou plates located in the ceiling of tne q, Pressurizer Cubt.le under a temporary alteration in 1988. The C padeyes wera provided to assist in the ramevnl and installation of safety relief valves RC-317A and RC~31?B. Although they provided
< a cafte method for removing and installing the valves it still required an excessive amount of task time resulting in unnncessary 4rsonnel radiation doses. DC-3222 was implemented to alleviate b% maintenance activity.
Sufuty Evaluation The new equipment neet 3eismic Category I design reqairements.
The monoral; cdditions re engineered to ensure operability of the structures and any ocner miscellaneous components.
The design changes did not alter or adversely impact the function of equipment or operating proceduren. The plant configuratita changes by DC-3222 do not increase the probability of occurrence or the consequences of accidents previously evn.uated. No impacts to cause malfunction of systems, e--:' ment or components important to safety are created by the facil. changes. No Technical Specification changas were required tor DC-3222.
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- 28. DC-3223, Permanent Test cables to Support Procedure OP-903-069 k1 ] Revision 2)
Description of Change
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Surveillance Procedure OP-903-069, Integrated Emergency Diesul Generator / Engineered Safety Features Test uses chart recorders in the control room to monitor test parameters. Temporary cables in the past have provided the relay room signals to the recorderb.
The design change waw initiated to permanently add a Bendix connector in each emergenc, diesel generator sequencer, frequency indicator, and safety but c rcuit to supply the chart recorder inputs. In summary, DC-3223 installed permanent test connectora for the chart recorders used in implementing surveillances per procodure OP-903-069.
Reason for Change The current uL' if temporary cables to provide the relay room signals to the .ecorders was deemed to be unsatisfactory for long-term continuance of procedural surveillances.
Safety Evaluation The Bendix connectors that were added by the design change are used only during the performance of Surveillance Procedure OP-903-069 to connect chart recordere to record test data. The connectora do not change the function of the af fected equipment.
The facility changes by LC-3223 do not compromise plant safety or licensing basia. The Bendix connectore are passive componente used to monitor test parameters and the likolihood of single active failure accidents is not increased by the permanent addition.
Ratings of the new cast connectors are compatible with existing system ratings and will not increase the occurrence of malfunction of equipment important to safety. No new system interactions are created by the facility changen, and there le no possibility of causing an accident of a type differant than previously evaluated.
No Technical Specification change as. roquired for DC-3223.
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- 29. DC-3233, Elgar Bypass Supply Transforn.er Input Breaker (Revision 0)
Description of Change The design change deleted the redundant main circuit breakers which provide alternate (bypass) AC power to Static Uninterruptible Power Supplies (SUPS) 3A-3 and 3B-S and replaced the 60 amp Motor control Center (MCC) feeJe* Sreaker to SUPS 3MA-8 bypass AC input with a 70 amp circuit Dreaker.
Reason for Change The design changes are the result of the Waterford 3 response to NRC Bulletin 88-10 which determined that the b.eaksrs had inadequate traceability.
Safety Evaluttion The probability or connequences of an accident previously evaluated in the FSAR will not be increased. The 70 amp circuit breakers to SUPS 3A-S and 3B-S are not required. Each of SUPS bypass transformer feedern are protected by a class lE, 30 amp circuit breaker in the MCO. This breaker also provides protection for the transformer. The owletion removes a potential cause for SUPS 3A-S and 3B-S failure. The replacement of an unqualified 60 amp breaker with a qualified 70 amp brer.ker removes a potential cause for failure of SUPS 3MA-S.
Upon loss of bypass AC power, SUPS 3A-S, 3B-S and 3MA-S continue to provide uninterrupted output power via the Class lE batteries, therefore the possibility of an accident other than that already evaluated is not created.
The probability or consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will be reduced by the removal of three unqualified circuit breakers.
Also, the removal of the unqualified breakers will enhance the availability of SUPS 3A-S, 3B-S and 3MA-S. Therefore, this change will not create any malfunctions other than already evaluated for the SUPS.
Both SUPS 3A-S and 3B-S bypass circuitry are provided overcurrent protection by qualified 30 amp circuit breakers located at the motor control centers (MCCs). The safety margin for SUPS 3A-S a~-
3B-S is not af fected because the deletion of the 70 amp breakers will not change the SUPS design function or operation.
Additionally, the increase of the feeder breaker from 60 amp to 70 amp will not affect the safety margin for SUPS 3MA-S. The 70 amp breaker will provide adequate overcurrent protection for the cabling rated at 93 amps. The 60 amp circuit breaker located at SUPS 3HA-S will continue to provide protection for the bypass tranoformer and its load.
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- 30. DC-3254,' Improved Charging Pump Blocks (Revision 1) -;
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Description of Change The design-change'provided the engineering and documentation to
. procure and install replacement charging pumps _ blocks made of improved design und material.- The facility changes _ replaced the series 300. stainless steel blocks on Charging Pumps A, B and A/B with blocks _made from forgings of 17-4PH material, an improved material of high fatigue resistance. The block suction and discharge flanges are bolted to the block in place of the-
.previously welded configuration.
Reason for Change s
Cracking of the series 300 stainless steel charging pump blocks-is a generic problem on Combustion Engineering plants. At Waterford13i t he blocks cracked on Charging Pumps A and B.-
_ safety Evaluation g The material change is a design enhancement which increases the reliability and life of the charging pumps.- The new blocks-are forgings of an improved material-for the application; i.e., more p fatigue-failure resistant than the previous stainless steel type.
=The-replacement blocks.are dimensionally identical to the previous-blocks and no piping!cha.nges were required.
-The design change does not alter _the function of the charging >
J. pumps or any other. equipment or plant procedures.- The changes made to the-facility by DC-3254 are within the system design parameters and preserve the safoty function 1of tho Chemical'and-
' volume control system. The facility changes do notfincrease the onsequences of previously evaluated accidents. RNo Technical Specification change'was required for DC-3254.
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- 31. DC-3258, Pressurizer Insulation Replacement (Revision 1)
Description of Change The function of the insulation on the pressurizer is to reduce the heat lose to the centainment atmosphere. DC-3258 replaced the current insulation with a removable type insulation of a thermal wrap blanket system for the pressurizer bottom head. The new insulation is made of fiberglass wrapped in woven fiberglass blankets. The design changs was considered an enhancement as an improvement in the heat dissipation rate of the pressurizer results in improved thermal efficiency for the roector coolant system. The new insulation design is also an improvement for removal and reinstallation.
Reason _for Change The design change is a design enhancement to improve thermal efficiency of the reactor coolant system. The n.odification results in insulation materiale which are easier to remove and reinstall.
The design change considered the evaluation of the NRC Generic Letter ?'o. 85-22, ' entitled, Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage and the related Regulatory Guide 1.82 (Rev. 1), Sumps for Emergency Core Cooling and Containment Spray Systems.
Safety Evaluation There are no accidents postulated which could be caused by the plant modifications made by DC-3258. The insulation installed meets the seismic design requirements and does not affect seismic a .11yses. The new insulation meets or exceeds the heat loss specifications of the insulation replaced. The heat dissipation rate of the pressurizer is likely decreased by the new insulation and should Lmprove operational efficiency and natural circulation capability.
The design change does not adversely affect hydraulic performance and the probability of malfunction of a pressurizer is not increased. There are no new system interactions created by DC-3258. The new plant configuration does not create the possibility of an accident of a different type than any previously evaluated.
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- 32. DC-3266, cutting and cepping Valve Leak-Off Lines (Revision 2)
Description of change The design change cut and capped the leak-off lines to valves (CVC-101, -103, -132, ~510 and PMU-144) in the Chemical and Volume Control and Primary Make-up Water Systems.
Reason for Change The facility changes were made to eliminate leakage by the valves with the potential for violation of exceeding Technical Specification limitatiorm for the Reactor Coolant System. The valve packing was replaced with new Chesterton-style packing, which utilizes one seu of packing. The leak-off lines only have a funct: ion when conventional style packing is present, which includes two sets of packing neparated by a lantern ring. The use of Chesteron valve packing renders the leak-off lines useless and a source of unnecessary high lesk zate.
Safety Evaluation The design change will not cause the valves to fail or to stick open or otherwise r*1 function in a manner inconsistent with previous analysis. None of the accident consequences or probacilities which have been evaluated in the FSAR are adversely affected by the change. The mode of " fall open" by mechanical binding is not affected by leak-off line removal nor does the removal create new causes to fall the valves open.
The consequences associated with any accidents in which thtae valves are required to perform a safety function will remain unchanged because cutting and capping the leak-off lines does not affect the function of the valves. The leak-off lines do not perform a safety function. No new system interactions or connection are created. The failure of the nap is bounded by the small break LOCA analysis and thus does not create the poselbility of an accident or malfunction of equipment of a different type than previously evaluated.
Leak-off lines connected to a valve that is part of the Reactor Coolant System (RCS) pressure boundary are located downstream of f the valve seat and outside of the RCS boundary.
DC-3266 caps that leak-off line which was previously open to tho' reactor containment building. The line and cap have been designed to withstand RCS prcosure and temperature even though ASME does not classify the capped leak-off line as pressure retaining.
Thus, the margin of safety is not reduced.
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- 33. DC-3270, Main Condenser Waterbox Sight Glass Addition (Revis!on 1)
Description of Change The design change affected only the circulating water portion of the plant main cor. denser. DC-3270 added a magnetically couple 9 sight glass to each condenser outlet waterbox at the existing valve penetrations to give level indication.
Reason for Change
?ne condense.' circulating water outlet waterboxes can contain trapped air which reduces condenser efficiency. No means were currently available to check for air entrapment in the condenser outlet waterboxes. -The sight glasses installed by the design change enable Operations personnel to monitor levels to ensure the waterboxes are not partially air bound during operation and to gauge level during draindown.
Safety Evaluation The circulating Water System serves no safety function and the design change by DC-3270 affected only the circulating water portion of tha main condenser.
The new sight glasses are connected to existing isolation valves to provide level indication only and they have nc affect on equipment function.or plant operating proct: dure. There are no postulated accidents related to the main condenser. No new eystem interactions or connectinr.s are created by the facility changes.
No Technical Specification change was required for DC-3270.
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- 34. DC-3212, AuxillarLeoreponern. Cxling Water Pump Bearing 011 Cooler ~ ~ ~ ~
Ed*ficatiK(Revidicn Q l_ gescrjption of change The denign change connected the existit.g Auxiliary Component cooling Water (ACCW) Jmtp Bearing Oil Ccolor to sparo connectionti j on tha uuction and discharge flanges of the ACCW pumpa.
i 99ason fbr Changg The design change wao implementred to onoure lower bearing temperatur<ae 'or the ACCW pump bearing oil cooler. DC-3272 was inite.nted as a result of a Root ceuve Investigation (RCI) conductud itter bearings tailod, The betting temperatures reached 240 degrees Fahrenheit. The betrings are monitored by thormocouples which alarm in the Control Anom upon reaching 175 degreen Tahrenheit. Impicroe oil 2evel was determined to be th.
priuary cause of ocaring fla's. lure, but the MCI identified chat a contributing factor was the sbMence of cooling water to the oil I cocler e . Connection of the existing oil cooler to a cool'.ng water i ocurca reduces tha operating teropuratutes of the lubricating oil thereby lostring cearing temperatures and increasing equipment reliability.
$apty Evaluatity The operation of the ACCW nystem is'not adversely affected by the design change, the facility changes made by DC-3272 ensures the integrity of the ACCW oystem, which is required for safe shut 6 awn of tho plant following a loss-of-coolant or occam line breah accidente The Lantalled oil cooler connection meeta quality and eclamic design requirements. No special testing or abnormal equipment operation was required for th'a facility modifications.
The probability or consequences of an accident previously svaluated are not increased by the design change.
The probability of occurrence of malfunction of equipment impertant to safety ic reduced as the change lowers the operating temperature of the lubricating oil and providen plant operators more time to respond to temperature related anomalien. No possibility of a new type accident la created and a Technical specification change was not required for DC-3272.
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- 35. DC-3286, valve Actuator Replacement for Valves SI-405 AnB jnevision 7)
Description of Change The design change replaced the existing actuatora for Shthdown Cooling System isolation valves with actuators of an improved design from a different manufacturer.
Reason for Change The design change was initiated to improve the reliability of actuatora for valves SI-405 A&B to aneure the valves function as designed during shutdown cooling operations. Problems were experienced with the actuators on valves SI-435 A&B resulting in failures of the valves to open and in inadvertent clonure of the valves with subsequent loss of chutdown cooling.
Safety Evaluatlon The design change has no impact on equipment functions. The replacement valve actuatora fulfill the functione and meet the design criteria of the previous actuatote. The replacement actuators operate in the same basic way as the original actuators.
No accidents as previously evaluated are impacted by the design change. The changeout of valta actuators does -,t increase the probability of occurrence of an accident. "c. design change was initiated to improve actuator reliability and the prob 2bility of a malfunction of the valves is not increased. There are no new system interactione or connections created by the facility changes. No Technical Specification change was required for DC-3286.
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- 36. DC-3294, Pressurizer Spray, Valves Actuator Replacement (Revision 1)
Description of Change The design change replaced the actuators on the pressurizer spray valves with new larger type Size 60 actuatora.
Reason for Change The pressurizer spray valves had been experiencing cloaure and valve position indication problems. DC-3294 was initiated to solve the closure /indicstion probleme during the Refueling 4 Outage.
Safety Evaluation Engineering analysis determined that the minimum soray flow requirement of 375 gym was ottisfied by the actuator eepisonnent which reduced the valve ottoke length. The new replacement valve actuators were seismically qualified. The acfuty of the plant La not affected as credit for t!.e presourla$r sps ay su not taken in the TShR accident analynus. Thw rcictor coalknt t'yntem pressurts boundary is not affweted by the deNign changa under DC-3294, Boplacement of the prevsurizar spra.*r v alvo ectuatore is cen siderod a design enhancement aa valve perfoamanen to expensed to taprove.
PronourAzar spray valves aid in maintaining prnogure controJ of :
the reactor coolant system during normal operatton, heatup and cooldown, and transient conditions, by admitting wAt.ee from the cold log tc the steam space of the prwasurizor.
The probability of occurrenen of consequences of previously ovaluated accidents is not lacreased by the facility chaugus made by DC-3294.
There are no new system connecticht or interactions by the udulga chango. Th9 functional and pasatvo doelgn features of the ep;uy valve.e are retained and no new equipewmt failurAt tnothodo are introduced by the facility changes, No Tfachnical Specification change was required for DO-3294.
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.O . DC-3302, Itigh Pre 9uura Turbino llorizontal Joint Rep ~ air
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,JbiTITEA]'dT Deneription_of Chang Dt:-3102 was initiated to stop steam leaks f rom the horizontal joint of the high yressure turbine. The dauign chango wn.
Imp'.cmentod during tno Refueling 4 Outago. The Westinghouse llorizontc.1 Joint Program installed new stainloan stcol gland diaphrsgme with thermal shields to resist crosion and ruduco thermat otronsoa.
.tenann for chango Stuam W4u lea'ri.n.J from the harizontal joint of the high pressure turbine in titea area near the governor and gonorator end gland diaphragma. Injectlonu of noalant to stop th9 leakago were unoucceentul.
pafoty Evaluttion The dortign change invoA -ed enly maturial changen and machining at parts to titop etoam leaks and did not af f ect equipment.
laportant to safaty. The plant modifications made by DC-3302 were non-unfoty, non-notsmic ar.d noc~tp2411ty design changes. No component vao altered in th9 facility changes which could incrouaa the probability of accidentn previouuly analyca:d in the FSAR.
The deaign basis of the high pressure turbine wan not t.ffected by tha knp',owntation of DC-3302. The 17tprovements to the turbine hortirontal joint do not affret thre coticsrxtuoncos of an accident due to turbino minulles. No forecapable necidunt oennarios are created by the DC-2102 tacility changes.
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- 38. DC-3303, Movable and Fixed Incore Instruments (Revision OL Description of Change DC-3303. involved design changen for two plant ey tems, nnmely, the Movable Incore Detection System (MICDS) and the Fixed Incore Detection System (i.e., Incore Nuclear Instrumentation - INI).
The design change wse implomanted dur.Ing the Refueling 4 Outage.
MIt'DS was permanently isolated from all interf acing systems. In January 1990, the calibration tubes were isolated from the reactor via a Temporary Alteration Request. DC-3303 converted the temporary alteration into a permanent plant configuration (i.e.,
guide tubec wero sealed). T1e HICDs components were abandoned in place and the electrical cablen were removed.
In the INI system, 50 fixed incore nucl9ar instruments were replaced with re-designed inutruments that do not have calit, ration tubes as part of the assembly.
Ruanon for Change A temporary alteration made in January 1990 needed to be converted to a permanent plant configuration. The temporary alteration installed pressure caps on all the guide tubes of the movable incore instruments and removed the lower section of the guide tuben. The movable incore instruments were never operable and the caps were required because of Reactor Coolant System leakage and boric acid buildup on the missile shield, vessel head lift rig, reactor head and cavity.
Safety Evaluation The replacement fixed incore nuclear instrumente meet design requirements.
The operability of the non-safety system MICDS is not required for the mitigation of any accidents. Non-operability of MICDS does not increase the probability cf occurrence of a malfunction of equioment important to safetp No new system interactiona or i connections are created by t a design change.
Technical Specificaticn changes (TS 3.3.3.2 and Table 3.8-1) were requested and justified by Technical Specification Chr. age Requrst No. TSCR NPF-38-111 via Entergy Operations, Inc. letters to tia NRC in W3P90-1182 dated 11/9/90 and W3F1-91-0003 dated 3/5/91.
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- 39. DC-3308, Reactor coolant Pump Seal Raplacement (Revision 1)
Descri,ption of change =
The design chance replaced the existing Reactor Coolant Pump
-(RCP 2B) eeal with an Atomic Energy of Canada. Limited (AECL) type CAU4 seal. An AECL CAN4 seal was installed during Refuel 4 and its performance is being monitored.
Reason for Change The design change was initiated to minimize the amount of time the RCPs are shut down due to seal failure or leakage. The new type CAN4 seal for Reactor Coolant Pumps (RCPs) is dasigned to be more reliable. The AECL seal is compatible and interchangeable on the existing RCPs. Extensive testing has shown the AECL CAN4 seal to be a significant improvement over the current seal designo. The replacement seal is expected to provide greater reliability, improved staging, reduced pressure oscillationn and reduced interstage leakage due to improved denign and materials of construction. Design features of the new seal should improve RCP maintenance and reduce radiation exposures.
Safety Evaluation The modification only affected the three main ntages of the RCP 28 seal cartridge and did not require any change to the vapor seal of the fourth stage. Both the controlled bleed off and compenent cooling water requirements for the teal rencain unchanged. No piping changes resulted from t.he implementation of DC-5308.
Seist.ic qualification of the RCP.was evaluatad and it was determined that weight changes associated with the new seal are inolgnificant. Thermal loads on the heat uxchanger remain unchanged.
No new or special' tests were required as a recult of the facility change. The AECL raal provides the same fit and function of the existing RCP seals and the probability of an accident previounly evaluated is not increased by the replacement seal. The consequences of a RCP failure are not increased. The new seal design provides g; eater reliability and decreases the possibility of; malfunction'of equipment important to saf.ety. No Technical Specification change was required for DC-3308.
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- 40. DC-3316, Feodwater Heater 2B Impingement Plate (Revielon 1)
Description of Change DC-3316 installed a new impingement plate and thus restored Feedwater Heater 2B to its original design configuration. The impingement plate serves to deflect water entering the heater through nozzle N-7 to provent tube erosion.
Heanor for Char.ge Ccnditi.on Identification CI #269367 documented tnat the impingement plate on nozzle N-7 of Feedwater Heater 2D was missing (had broken loose or was never installod). Nozzle N-7 in a 4-inch diameter racirculation line from the Heater Drain Pumps which serves to rtgulate the level inside the heater shell. Work Author.1 zr .on WA #01057988 made temoor>try repair to plug leaking tubee and several adjacent tubes which were sacrificed as a proccutionary measure until the missing impingement plato could be installed.
Safety Evaluation Feedwater Heater 2B is non-safety, non-selemic and not considered important to safety to achieve safo shutdown of the plant. The new impingement plate design lo stronger than the original and in expected to be more reliable, i.e., longer lasting. No new system interactiona or connections are created by the installation of the impingement plate under DC-3316.
The facility change testored equipment to its original design configuration and does not affect the function or operation of safoty equipment. No new or special tests were required for the design change. Previously analyzed accidents cannot b9 caused or affected by restoring the impingement plate to original design conditions. No Technical Specification change was needed for
-DC-3316.
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- 41. DC-3326, Reactor Coolant Purrps Insulation Hoplacomont (Revision 1)
Doucription of Changa The design change replaced the insulation on the Reactor Coolart Pumps (RCPr; with an improved removable type. The replacement insulation is a NUKON fiberglaan, blanket t)pe, reinforced with stainlano steel mech and stool flash chields to protect the RCPs from oil spray and 'fater.
Roanon for Chango The design chango was initiat;d to provide an improved insulation for the Seacter Coolant Pumps with more effective thermal efficiency and easier to install, remove and reinstall.
Safety, Evaluation The design change does not affect naismic analyses of the equipment. The new insulation complies with the requiremonta of NRC Regulatory Guido 1.36. The cuntom manufactured and installed blanket type ir.sulat ion thorn.a3.
- y insulates more ef fectively than
+.he prevjously used metallic reflective insulatica.
The facility change ham no impact on the function of the reactor coolant pumpe. No special testa or experimente are involved in the insulation replacement. The plant modifications do not increase the consequences of previously evaluated accidents or malfunctions of equipment important to eafecy. There are no new system interactions created by the design change. No Technical Specification change was needed for DC-3326.
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- 42. -DC-3327. Emergency Diesel Generator Air Dryer Reinataljation (Revision 0)
-Description of Change DC-3327 was a maintenance design change as it installed the manufacturer repaired air dryer for the Emergency Diesel Vanerator (EDG) air start system. The installed repaired air dryer had aqalpment differences from the original equipment and flexible home connections were used for the dryer system.
Reason for Change The leaking obnolete precooler for the air dryer assemblr could not be repaired and the vendor replaced it with a new precooler.
Receipt inspection identified differences between the repaired air dryer and the original air dryer assembly and the design change was initiated to install the altered equipment and provide flexible hose connections for the dryer system. To facilitate future repaire and rejnstallations for the other three EDG Air l Dryers the desien change-installed flexible hose connections for their inlet and outlet connections.
1 Safety Evaluation The EDG air dryers are clkssified as non-nuclear safety components. The dryets are qualified to seismic category I requirementa and ars supported to prevent interaction. with safety related equipment during a safety shutdown earthquake (SSE).
The repair and reinstallation of the air dryer does not increase the probability of occurrence of accidents previously evaluated.
Nc new systen interactions are introduced and the changes to the facility do nct increase the cor. sequences of previously analyzed accidents.
Because the design change provided maintenance benefits with the installed flexible aose connections, the probability of malfunction of equipment important to safety is likely reduced.
No new type accident-in created by the design change and no new equipment failure methods are introduced.
No Technical Specification change was needed for DC-3327.
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- 43. DC-332P, Addition of Hoists Inalde Reactor Containment Building (Revision 0)
Doncription of Change The design change provided additional hoisting capabilities inside the Reactor Containment Building (RCB). One 2-ton motorized hoist was installed with column mounted jib on top of the steam Generator No. 1 framing. Brackets were also installed for use of a temporary jib crane on the Steam Generator No. 2 framing over the Raaetor Coolant Pump 2A Heason for Chango The addition of the hoist and jibs was to reduce the usage of the polar crane for small lifts. The additions allow better schedal'.ng for Iolar crane usage to improve ef ficiency and permit more timely completion of tasks inside the RCB which require use of the polar crane. The additional hoist capability is designed for uso as mechanical tools to support outage related activities and material transfers during outages.
Safety Evaluation The hoist and jib over the Steam Generator No. 1 framing is seismically supported and permanently stays in the RCB. Jib cranes are used only during refueling or maintenance outages. The holet and jib crane over Reactor Coolant Pump 2A is removed at the completion of each outcge and stored outside tF, containment for futurn uso.
The design change did not alter any present configuration of syctems, supports, components or equipment. The changas enhance the working-environment inside the RCD during outages.
The addition of hoists did not require any new test or experiment and had no affect on plant operations. The facility changes do-not interfere with any safety related equipment. The probability of occurrence or consequences of a malfunction of equipment important to cafety is not increased by the facility changes. New accident potential is not created by the dasign change. No Technical Specification change was needed for DC-3328.
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l 44. DC-3331, Reactor Coolant Pump Caskat Leakoff Drain Lines I (Revision 0)
Description of Change The design change avaluated the temporary tubing installed under Temporary Altiration Request TAR-90-018 and made the installation a permanent part of the plant configuration. The temporary alteration had connected drain lines from Reactor Coolant Pumpe (RCPs) 15 and 2A flange gasket leakage instrument tubing to the Paactor T rain Tank (RDT) to provide a controlled leak path.
Reason for Change Condition Identification No. 271478 identified that Reactor Coolant Pump 2A inner and outer gaokuts between the pump case and the driver mount / pump cover were leaking in October of 1990. As a temporary repair, Temporary Alteration Request TAR-90-018 connected 1/2" tubing to the test valves on the pressure switch for RCP 18 and 2A and routed it to the Reactor Drain Tank to provide a controlled leak path.
Prior to DC-3331 there were ao provisions in place to determine if reactor coolant la leaking past the outer gasket. DC-3331 permanently connected the RCP case and pump cover / driver mount gasket lekkeff to the Reactor Drain Tank.
Safety Evaluat!.;', f The dealgn change is a da'inite plant enhancement as the leakoff flowpath to the reactor orcin tank was not previously available.
The drain lines, isolativi 'alves, and existing pressure switches allow gasket leakof f to be nonitored, counted as " identified" leakage, and identified with a specific RCP. The information provides for determination of corrective action for gasket leakage.
No new release paths are created by the permanent connection of RCP leakoffs to the Reactor Drain Tank. The design change has no impact on equipment function or operation. No special or unusual testing was required for the facility changes.
' The probabiltty of occurrence or consequences of previously evaluated accidents is not increased by the design changes. Plant modificttions made by DC-33?1 do not increase the probability of occurrence or consequences of a malfunction of equipment important to safety. No Technical Specification change was required for DC-3331.
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- 45. DC-3340, Reactor Coolant Pump Stud Elongation Test Hole Depth Tncrease (Revision 0)
Description of Change The design change increased the depth of the Reactor Coolant Pump (RCP) driver mount to pump case stud elongation test hole from 24 inches to 32 inches.
There are 16 driver mount to pump case stude in each RCP. Both the stud with the 24 inch deep elongation test hole and the 32 inch deep elongation test hole are acceptable for use in the RCPs with the provision that each RCP must have 16 stude of the same type.
Reason for Change The design change was made to facilitate the non-destructive in place ultrasonic (UT) test of the RCP driver mount to pump case stude. The change allows the entire length of the load bearing and threaded sections of the 33 inch long stude to be non-destructively tested. Extending the stud hole depth allows stud integrity to be verified without the arduous task of removing stude.
Safety Evaluation The stud elongation testing is enhanced by the change and it likely detreases the probability of occurrence of failure or malfunction of the component.
The design change does not alter the function or the ability to function of the RCP driver mount to pump case studs or the function o' the RCPs. The test hole depth increase was included in the original vendor design basis of the stud.
The increase in depth of the elongation test hole lLn tha RCP stude has no impact on the occurrence or consequences of previously evaluated accidents. There are no new system interactions or connections creat ed by the f acility changes. The possibility of a new type accident is not created by the changes to the=stude. No Technical Specification change was needed for ' 3340.
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- 46. DC-3344, control Room Envelope Enhancements (Revision 0)
Description _of Change The design change installed a loop seal in the +35.00' elevation Cable Vault area on the Radioactive Waste System drain header; and added removable mechanical pipe pluge en seldom used floor draine on the Sanitary Waste System.
The design change also added a repetitive task in the Station Information Msnagement System database to ensure that water le added on a regular basis to loop esals which are not routinely used in the Control Room envelope.
Reason for Change The design enange was initiated to enhance the Control Room envelope configuration and ensure regulatory compliance relative to loaatightness of the preanure boundary. Piping penetrations of the plumbing and drainage eyetems were determined to be a contributor to Control Room envelope leakage.
The Sanitary Drainage system presently han loop seale or P-trape on virtually every connection to the Control Room envelope. Tha loop neale rely upon routine use to sneure that the trap remains fu21 of water. Some of the loop seals are not used frequently enough to remain full. The plumbing vent system, in addition to the drain lina itself, allows air from the preneurized control Room envelope to escape.
Safety Evaluation Tne Control Room envelope is designed to maintain a positive preenure of 9.125 inch water gauge or greater with respect to outside air, with a makeup rate of 200 cfm or leen, during normal operations or high radiation conditions.
DC-33A4 is a system enhancement to reduce the leakage of air in the Control Room envel;pe through the drainage systems. The added loop seal and floor drain pluga reduend the leakage rate of the Control Room envelope and provides a greater margin at cafety egainst air leakage. These changen atc also beneficial during temporary removal of individual seals for routine maintenance.
The drainage systems affected by the design change are not safety relat96 and are not degraded. The DC-3344 facility changea do tot increase the probability of occurrence or consequences or previously evaluated accidents. No new system interactions are created by the plant modifications. The probability of occurrence or consequences of a malfunction of equipment important to safety are not increased by the destgn change. No Technical Specification change was needed for DC-3344.
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I. C. WORK AUTHORIZATIONS / CONDITION IDENTIFIyATIONS (WA/CI)
- 47. WA-01007647/CI-252577 - Boric Acid Hakeur Tank Level Instrument Tubing Description of Change The facility change replaced the Boric Acid Makeup Tan) (BAMTs) level indicators with Piping Class 3 components to retu.a the plant corfiguration to the original equipment design basis.
No physical change was made to the plaat configuration by WA-01007647. The BAMT level instrumento tubing and isolation valves were replaced with Code Clana 3 components by the temporary alteration TAR-88-041. CI-252577/WA-01007647 documented the nonconformance condition and corrective action; and updated the applicable drawings to close out the TAR-88-041.
Reason for Change The nonconformance condition was reported in detail in License Event Report, LER-88-026-01, entitled, " Tubing and Supports Not Seismically Qualified Due to Personnel Error."
The following is an excerpt from the test of LER-88-026-01:
Waterford 3 was in cold shutdows. in 1988 when it was discovered that the wrong piping class tubing was installed for the Boric Acid Makeup Tanks (BAMTa) Level Indicators.
The tubing and supports were to have been installed as polemically qualified; however, when some supports were discovered not inspected, a non-seismic qualification was justified. When the BAMT Technical Specification (TS) requiremcnto were changed, the supporta were inspected te seismically qualify the installation. A later review diocovered that the installed tubing welda could not be determined to be seismically qualified. Therefore, the plant operated in a condition outside the design basis since January 8, 1987.
The cause of thic event was personnel error in that inadequata reviews of the design at goveral points in this event allowed the condition to exist since initial t construction.
Safety Evaluation The BAMT level instrument tubing and isolation valves had been changed from Piping Clase 7 to Class 3 to conform to the FSRR design bauin.
Engineering evaluation indicated the potential for unacceptable loan of borated makeup water if the tuLing or volves should rupture. Replacement of the instrument tubing and valven with class 3 components ensured integrity in a celsmic event and remcved the unacceptable nonconformance condition.
WA-01007i47 in a noncoaformance report which restorsd the proper design bAsio documentation for the BAMT level instrument tubing.
Accident t.robabilities of occurrence and consequences are not impacted by the crawing correctiona and restoration to design baute requiremento. No Technical Specification change was needed for WA-01007647/CI-252577.
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- 48. WA-01068959/CI-272242 - Qualification of Base Platn for Main Steam
! Pipe Support MSRR-286 Description of Cnange The identified plant condition of a nonconformance design deficiency was dispositioned on use-as-is basis supported by an engineering calculation. Engineering Calculation EC-M90-005 determined that the main steam pipe support function is not significantly degraded by the use of three effective functional anchor bolts. Engineering concluded the structural integrity of the pipe support is assured and deemed it acceptable on the use- as-is basis.
Reason for Change During an Inservice Inspection examination of the Main Steam Pipe Support MSRR-286 it was found that only 3 of the required 4 bete plate anchor bolts were functional as one bolt could not be torquad.
Safety Evaluation The function of pipe support MSRR-286 is not significantly impairec with three anchor bolts. The design deficiency is acceptably dispositiened on the "use-au-is" basis as the engineering celculations demonstrated the support is capable of withstanding the ori;inal design loads. Acceptance of the facility change has no impact on the Main Steam System since the pipe support function is not impaired.
The change was characterized as an acceptable discrepancy to design baels and did not encall a procedure change or special test of abnormal system operation. The change does not affect overall system performance or reliability in a mannar which could increase the probability of occurrence or consequences of previonaly evaluated accidents. The pipe support maintains its aiginal design basis function and does not increaan the probanility of occurrencu or consequences of a malfunction of equipment important to safety.
Support MSRR-286 is requalified and will perform its design activity and does not create the possibility of an accident or equipnent malfunction of a different type than previously analyzed. No Technical Specification cnange was needed for the Work Authorization WA-0;068959, 49 1
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- 49. WA-01073?63/cI-273778 - Reactor Fuel Reconstitution (Refuel outage #4)
Dea [ription of change The work authorization covere the reconstitution of fuel assemblien which were found to be leaking by the fuel inspection (ultrasonic testing performed via CI-273777/WA-01073264) during the Refueling 4 Ou*ago.
Reason for change The fuel reconstitution was implemented to eliminate leaking fuel rods found by the ultranonic testing of reactor fuel during the Refueling 4 Outage.
Safety Evaluation During fuel reconstitution, a potential impact exists by the change due to abnormal operations, in that a fuel assembly with its upper end fitting removed la extending above the top of the apent fuel storage rack.
The Fuel Handling Accident might possibly be affected by the fuel reconstitution activity. The safety evaluation for tuel reconstitution determined the physical configuration of the fuel assembly being repajred does net increase the likelihood of the Fuel Handling Accident, and fuel pin damage would be within the bounds of the analyzed accident even in the unlikely event of an accident during fuel reconstitution.
No equipment important to safety lo affected by the fuel reconstitution activity. Fuel assembly repair could not create a different type accident than previously evaluated. The fuel reconstitution does not change protective boundaries or margins of safety in the design bases.
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- 50. WA-01076246/CI-274976 - Installation of Dummy Instrumento in Beactor core Description of change The facility change installed a dummy incore instrumentation (ICI) assem'a ly and hydro plug and lef t remaining about a ten foot section of the old failed ICI in the reactor core locations 0-13 and T-16. The change eliminated the normal neasuremonta of core exit temperature and local core power at core locatione 0-13 and T-16.
Reason for Change The work authorization was initiated to replace the damaged incore instrumentation (ICI) during the Refueling 4 Outage. Possibly Fuel Alignment Plate wear of the thimble tubes cauned the ICIe to wear and then break on removal. .Use of the hydro plugs and dummy
- ICIe as replacement for the miosing ICI was the proper action to resume plant operations.
Safety Evaluation The Reactor Vessel Internale Manual specifies and requiron the une of dummy ICI when no instrumente are used at a given core location.
Additional or special testing was not required for WA-01076246 since the hydro plug to functionally and physically equivalent to actual Incore Instrumentation (ICI).
No safety equipment in affected by the work authorization and the hydro plug ensures that the Reactor Coolant System boundary la maintained.
Combustion Engineering (CE) certified that all coder. standards and requiremonto that apply to the quality /pafety related portion of ICI also applied to the dumr.y ICI and hydro plug used as replacemento. CE also certified that the dummy ICI &nd hydro plug are designed to withstand reactor pressures and temperature, thermal / dynamic loads and environmental conditions.
There are no accidents associated with ICI. and margin of safety is not impacted by use of dumey ICI and hydro plug replacement.
No new system interactions arc created. A Technical Specification change was not needed for WA-01076246.
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- 51. CI-274719 (Revision 1) - Steam Generator Blowdown Containment i Isolation Valves Instrument Air Regulatcra Description of Change The nonconformance CI #274719 (Revision 0) was initiated to replace the plastic instrument air regulators for the Steam 9 Generator Blowdown Inside and Outside Containment Isolation Valves with Fisher controle type 67ATSR stainless steel regulators with fluoroelantomer plugs and diaphragme.
Reviolon 1 to CI-274719 changed the planned replacement to Fisher controla type 67AFR aluminum regulators due uo the unavailability of the stainleno steel 67AFSR regulators.
Reason for Change The current plastic air regulators were experiencing numerous failures due to plastic parta breaking apcrt. Revision 1 to CI-274719 was needed because of unavailability of the initially planned use of stainleen steel air regulators.
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Safety Evaluttion The nonconformance repair CI-274719 alightly increased the amount and surface area of aluminum as given in the FSAR. The use of the aluminum regulators will not significantly interfere with the ability of the Combustible Gas Control System to mairtain hydrogen concentration below acceptable limite following the unlikely event of a Lons-of-Coolant Accident.
The change in type of air regulatora does not alter the function of the operators for the Steam Generator Blowdown Isolation valves. Replacement of the air regulatora does not increase the probability of occurrence or the consequences of an accident previcusly evaluated.
The regulator material la considered a replacement improvement and should provide more reliable service than the original plastic regulators. The aluminum regulatora do not increase the probability of occurrence or the consequencea of a malfunction for equipment important to safety as previously evaluated. The replacement of the plastic air regulators with metal regulators does not create any new accident scenarios or any different type of safety equipment malfunction. No Technical Specification change was needed for C1-274719 (Revision 1).
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I. D. TEM _PORARY ALTERATION REQUESTS (TARS)
- 52. TAR-89-052, Removal of Relief Valve on Feedwater Heater 6C and Installation of Blank Flange Description of Change The temporary change to the facility removed the relief valve CD 181C on Feedwater Heater 6C and replaced it with a blank flange. Valve CD 181C is the liquid relief valve on the feedwater
- heater channel head and its function is to protect the heater R tubes and channel head from overpressure. Duration of the temporary alteration was to tne next plant outage to permit the relief valve to be repaired, bench tested and reinstalled.
Reason for Change The liquid relia! valve CD 181C on Feedwater Heater 6C failed open and it had to be removed and temporarily replaced with a blank flange to al?.ow the plant to remain at power.
Pafety Evaluation Relief valves on the other feedwater heaters provide redundancy for protection against overpressurization. The plant equipment affected by TAR-89-052 is not required for safe shutdown of the plant or to nitigare the consequences of an accident.
The feedwater heater relief valve serves no other purpose than to relieve pressura from the aeater and its failure or removal has no impact on the orobability of occurrence or consequences of previously ana'.yved accidents. Th9 temporary alteration does not 1 affect any soforg ratsted systems or equipment.
Tr; temporary change to the plant configuration did not create the possibility of a cifferent type accident or m&lfunction of equipmert biportant to safety than previcualy analyzed. No margin of safety waJ impacted; and a Technical Specification change was not neaded for TAR-8f-032.
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- 53. TAR-90-013 (Rev. 1), Provide Acid Supply to the Domineralized Water system Description of Change The temporary alteration installed tubing from valve CF-420 (Cation / Mixed Bed-l Concentre;ed Acid Inlet Strainer Drain Valve) to valve DW-110A (cation A outlet Strainer Upstream Drain). The change provided acid supply for the chemical treatment of water to aid in the removal of carbon d'. oxide. The duration of tii temporary alteration was untic permanent installation was implemented via a Design Cha';e.
Reason for Change The Domineralized Water Syston (DWS) purifies dif ferent wat er supplies by removal of ions, t*pability was needed to inject acid into the DWS prior to water purification to aid in carbon dioxide removal. The facility change was made upstream of the Forced Draft Degasifier in the Water Treatmant Building. Revision 1 of TAR-90-013 was initiated solely to provide additional descriptive information in response to a Quality Notice which cited an inadequacy in the description of th f lant configuration ir. the TAR document.
Safety Evaluation TAR-90-013 was a temporary field enhancement to later be made permanent to the plant configuration. The temporary alteration did not alter the function or operation of the Deminernlized Water System (DWS) which serves no safety function.
The DWS is not required fcr safe shutdown of the plant or to mitigate the consequences of an accident. The temporary facility changs did not require special testing or abnormal operation of a plant system.
TAR-90-013 did not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. The facility change did not create the possiblity of an accident of a different type than previously analyzed. No Technical Specification change was needed for TAR-90-013 (Revision 1).
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- 54. TAR-90-014, Gag Condensate valve CD 204A Description of Change The temporary alteration installed a gag to disable the relief valve Condensate Valve CD 204A. The valvs was gagged by installing a mechanical device to prevent the valve from lifting.
Duration of the gag was for temporary use through Refueling 4 Outage.
Reason for Change Condensate Valve CD 204A is a relief valve which had failed open.
The intent of the temporary alteration was to stop flow through relief valve CD 204A.
Safety Evaluation CD 204A is the non-safety liquid relief valve for the tube side of Feedwater Heater #4A. The function of valve CD 204A is to protect the heater tubes and channel heads from overpressure. Use of tags on the stop valves to effectively lock open CD 204B and CD 234C providea administrative control to ensure overpressurization protection for Feedwater Heater 4A. This action ensured conformance to the requirements of ASME Boiler & Pressure Vessel Code Section VIII.
There was no impact on plant operation or equipment created by the temporary alteration as the function of the gagged valve CD 204A was provided by other relief valves in the Feedwater Heater System.
No procedure changas were required for the temporary alteration as the condensate system functioned without change to the normal operating procedures. The temporary alteration did not require any special test or abnormal operation of safety systems or equipment.
The temporary facility change did not alter the probability of occurrence or consequences of any previously evaluated accident.
Feedwater Heater 44 is not important to safety and its malfunction has no Ompact on any equipment important to safety.
The temporary alteration provided no new system connections or interfaces to create an accident of a different type than previously evaluated. TAR-90-014 did not require a Technical Specification change.
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- 55. TAR-90-015, Install Mechanical Jumper from condensate Makeup to Primary Makeup System D_escription of Change The temporary alteration provided a connection between the primary water system and other demineralized water sources (Condensate and Domineralized Water Storage Tanke) while the Primary Water Storage Tank was being purified. A temporary techanical jumper was used for this connection. The duration for the temporary deviation of the plant configuration was set for seve al weeks to assure that the Primary Water Storage Tank (PWST) water met specification grade.
Reason for Change Condition Identification CI-271118 identified the problem of primary makeup uater being chemically out of specification. The temporary facility change established an alternate path for refill of the PWST.
Safety Evaluation The temporary facility change deviated from FSAR drawings. The temporary alteration did not involve safety or important to safety systems or equipment. All plant equipment remained functionally unchanged by the TAR-90-015. No operational or chemistry procedures were affected by TAR-90-015.
Accident and equipment malfunction probabilities are not affected by the temporary change and margins of safety for plant operation are not impacted. No Technical Specification change was needed for TAR-90-015.
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- 56. TAR-90-016, Temporary Chillers for Containment Air Conditioning
-Description of Change The temporary alteration provided chilled water to the containment Cooling HVAC System (CCS) tn support a forced maintenanca outage.
The temporary chillers were located on the east side of the B train dr Mode 5, ..e.,
y cooling Coldtowers and were used only during operational Shutdown.
An identical temporary alteration was previously implomonted via TAR-89-17 to sapport the Refueling 3 Outage. The duration of TAR-90-016 set for approximately one month was to ensure adequate containment cooling for the entire maintenance outage.
Reason for Change Temporary chillers were needed to provide adequate containment cooling during mode 5 of a forced maintenance outage. It is planned to later make the temporary connections a permanent plant configuration via design change DC-3073.
Safety Evaluation The temporary addition is to the non-nuclear safety supply and return portions of Compunent Cooling Water (CCW). The CCW system is nc: required to be operable in Mode 5 per Technical Specifications.
The Containment Fan Coolere (CFCs), along with Containment Spray, are designed to operate and maintain the containment within required pressure and tamperature limits during normal plant operation and under LOCA and Main Steam Line Break accident conditions. The CFCs reject heat to the outside environment via the CCW system. TAR-90-016 isolated CCW to the CFCs and an alternato path is established from the CFCs to the temporary chillers. Heat is thus rejected to the outside environment via the temporary chillers instead of the CCW system.
The temporary facility change does not require special testing or system operation in an abnormal manner. TAR-90-016 was implemented only in mode 5 and therefore had no impact on the FSAR analyzed accidents or operational equipment malfunctions.
TAR-90-016 did not create new accident possibili*d es or new methods of equipment failure. Margins of safety were not affected by the temporary facility change and no Technical Specification change was needed for TAR-90-016.
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- 57. TAR-90-Ol8, Reactor Coolant Pumps 1B & 2A Gasket Leakoff to Reactor Drain Tank
[..
Desciption of Change The temporary alteration provided drain lines from Reactor Coolant Pumps (RCPs) IB and 2A flange gasket leakage instrument tubing to the Reactor Drain Tank (RDT) .
The change created a RCP gasket leakage path that could be monitored by routing leakage to the RDT. The tubir.g is rated to withstand RCS pressure and the RTD is provided with a safety relief valve which is routed to the containment Sump.
Reason for Change A pressure switch monitors the space between the double gaskete on the RCP to detect an increase in pressure, which indicates a leak past the inner gasket. The pressure switch only provides annunciation in the control room. The RDT collects effluents within the Reactor Containment Building.
The local pressure indicators and pressure switches will not provide RCP leak detection monitoring while the drain lines are installed. Since flange gasket leakage was detected ou RCPs 18 and 2A, the leak detection instrumentation serves no function until the pump leaks are repaired. The temporary change rerouted leakage past the RCP inner gasket to the RDT.
Safety Evaluation The temporary change actually improved the Reactor Coolant System (RCS) operation by providing a flow path for known RCC pump gasket leakage which helped to prevent damage to the RCS pre?ouro retaining components (i.e., RCP stude and RCS piping) from boric acid corrosion.
The temporary facility change did not adversely affect the overall RCS or Boron Management system performance or any other cafety systems.
The RCP flange gasket leakage instruments and RDT are non-safety, non-quality and safety class 4.
The temporary alteration did not increase the consequences of accidente previously evaluated. No safety equipment or equipment important to safety was adversely impacted by the fccility change.
TAR-90-018 did not create the possibility of an accident of a different type than previously analyzed and a Technical Specification change was not needed.
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- 58. TAR-90-019, Reactor Coolant Pump 2A Replacement Insulation Description of Changa 3
The temporary alteration replaced the damaged insulation on Reactor Coolant Pump RCP 2A with a temporary thermal wrap blanket to provida heat retention. The replacement insulation was not identical to the original inanla' ion but it complied with the requirements of Regulatory Guide . 36, Nonmetallic Thermal Insulation for Austenitic Stainless Steel.
A permanent change out of the insulation for all the RCPs was planned for a later outage and was implemented via Design Change DC-3326 (see this Report Item #41).
Reason for Change The original rigid thermal insu.:sion on RCP 2A was damaged.
Safety Evalaation The function of the insulation is to reduca the heat lose from thL RCP to the containment atmosphere. The temporary alteration has no impact on the function of the pump or the ability to properly operate.
No special tests or experiments were involved in the temporary insulation replacement. The temporary modification to the plant configuration did not increase the consequences of previously evaluated accidents or malfunctions of equipment importan*. to safety. There were no new system intertctions created by the temporary change. No Technical Specification chango was needed for TAR-90-Ol9.
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- 59. TAR-90-021 Installation of Water Softener for Supplementary Chiller Condensing System Description of Change The temporary alteration installed a temporary water softener into the makeup flow path of the Supplementary Chiller Condensing (SCC)
Cooling Tower System.
Reason for Change The water softener was added to reduce the hardness of the potable water which is fed to the wet cooling towers. The objective of the softener is to prevent scaling (calcium carbonate) in t t.u conder.ser tubing of the supplementary chiller units.
Safety Evaluation The temporary alteration af fected the Potable Water and the Supplementary Qhiller Condensing Systems both of which are non-safety atalliary systems. The temporary facility change deviates from the normal status of some components on drawinga in FSAR figures.
No special testing of abnormal system operation was required for the temporary alteration. TAR-90-021 did not affect the probability of occurrence or consequences of any previously evaluated accident.
Connecting a water softener in the potable water makeup flow path can not cause malfunction of any safety related equipment. There are no new system interactions involved in the facility change to create the possibility of an accident of a dif ferent type than previously evaluated. Margins of safety are not impacted und no Technical Specification change was needed for TAR-90-021, 60
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- 60. TAR-91-003, Disconnection of Two Defective Heated Junction Thermocouple (HJTC) Sensor Heaters Description of Change The temporary alteration replaced the defective HJTC #5 and HJTC
- 7 source heatern with 25 ohm, 75 watt resistors; and jumpered out the unheated and heated thermocouples for HJTC #7 to Loturn the Qualified Safety Parameter Display System (QSPDS) Channel 4' probe to service.
Reason for Change The temporary alteration was performed to allow QSPDS Channel 2 to be operable per Technical Specifications 3.3.3.6 (Table 3.3-10).
Similar temporary alterations, TAR-91-005, TAR-91-030, and TAR-91-031, were performed for other defective heaters (see this Report Items Nos. 62, 71 and 72). (NOTE: TAR-91-003 was closed 5/4/91.)
Safety Evaluation The function of the HJTC system is to allow determination of the water inventory in the reactor vessel above the fuel alignment plate. There are two channele of HJTC instrument monitoring.
With the channel 2 sensors #5 and #7 out of service, the information is still available from chann31 1 sensors 95 and #7.
Also, the remaining six sensors in channel 2 will remain operable.
The HJTC provides information to the operators and will not affect the ability to shut down the plant. No other equipment will be affected. The reactor vessel level monitoring will remain operable. '.'he change is confined to wiring changes in a monitoring system outside of containment and cannot cause a different accident than previously evaluated. The change will lower the number of operable sensors in the HJTC channel 2 probe
-from eight to six (three in the head and three in the plenum).
The Technical Specification allows the channel to remain operable with a minimum of 1 sensor in the head and 3 in the plenum. Thus, the basis and the margin of safety is not reduced.
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- 61. TAR-91-004, spent Resin Tank Liqu'd Level Indication Description of Change The temporary alteration installed a mechanical jumper (i.e., a temporary sight glass) from an available test tap at the bottom of the Spent Resin Tank to the test tap for the reference leg tap for the level tranamitter. The temporary sight glass for liquid level indication consisted of clear tygon tubing secured to the wall, external to the Spent Resin Tank Room.
Duration of the temporary alteration was set to the return of the permanent level transmitter to service.
%eason for Change The local indication (transmitter RWMILT0644) for Spent Resin Tank liquid level was out of service, safety Evaluation The temporary facility change involved only the Resin Waste Management System. TAR-91-004 had no effect on radiological release consequences of the analyzed accidents.
The Spent Resin Tank is isolated from safety related equipment and can not increase the probability of occurrence or consequences of a malfunction of equipment important to safety.
The temporary sight glass does not create the possibility of an accident or malfunction of equipment Omportant to safety of a different type than previously evaluated.
The temporary facility changes did not involve a protective boundary and margins of safety are not impacted. No Technical Specification change was needed for TAR-91-004.
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- 62. TAR-91-005, QSPDS Channel 1 HJTC #5 Heater Disconnection Description of Change The temporary alteration disconnected the heated and unheated thermocouples for the Qualified Safety Parameters Display System (QSPDS) Channel 1 Heated Junction Thermocouple (HJTC) Sensor #5 and installed electrical jumpers in their place to return the channel 1 to service.
Reason for Change The temporary alteration was performed to make the remainder of the Reactor vessel Level System (RVLS) operational per Technical Specifications 3.3.3.6 (Table 3.3-10).
Similar temporary alterations, TAR-91-003, T AR-91-030, and TAR-91-031, were performed for other defective heaters (see this Report Items Nos. 60, 71 and 72). (NOTE: TAR-91-005 was closed 5/16/91.)
Safety Evaluation The function of the HJTC system in to measure the water inventory in the reactor vessel above the fuel alignment plate. The QSPDS Channel 1 probe will continue to function as designed with HJTC is sensor inoperable by using the remaining 7 sensors to measure vessel water level.
The HJTC provides information to the operators and will not affect the ability to-shut down the plant. No other equipment will be affected. The reactor vessel level monitoring will remain operable. The change is confined to wiring changes in a monitoring system outside of containment and cannot cause a different accident than previously evaluated. The change will lower the number of operable sensors in the HJTC Channel 1 probe from 8 to 7 (3 in the head and 4 in the plenum). The Technical Specifichtions allows the channel to remain operable with a minimum of 1 sensor in the head and 3 in the plenum. Thus, the Technical Specifications basis and the margin of safety is not reduced.
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- 63. TAR-91-006, Reduction of Containment Noise Levels Description of Change The temporary alteration made electrical wiring changes so allow the Control Room operators to run the Containment Fan Coolers (CFCs) in either fast or slow speed during operational modes 5 or 6 (i.e., Cold Shutdown or Refueling) of the Refueling 4 Outage.
Reason for Change The temporary alteration was implemented to reduce noise levels in containment (during modes 5 or 6 only) by allowing the CFCs motors to be operated in slow speed.
Safety Evaluation TAR-91-006 altered the operation of the CFCs by allowing the fans to operate in slow speed without an Engineered Safety Features Actuation Signal (ESTAS).
The CFCs, along with Containment spray, are designed to operate and maintain the containment within required pressure and temperature limits during normal plant operation and under Lous-of-Coolant Accident (LOCA) or Main Steam Line Break (MSLB) accident conditions. The LOCA and MSLB accidents are analyzed for operational moden 1 to 4. The operational change by TAR-91-006 is implemented only during operational modes 5 or 6 and it has no affect on previously analyzed accidents.
The temporary f acility change has no impact on the probability of occurrence or consequences of a malfunction of equipment important to safety as previously evaluated. Under TAR-91-006 the CFCs are not connected to and are not operated by an ESFAS. An accident or equipment malfunction of a different type than previously evaluated is not created by the temporary facility change.
Margins of safety are not affected by the wiring changes for special operation of plant equipment. No Technical Specification change was naeded for TAR-91-006.
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- 64. TAR-91,009, Temporary Chillers for Refueling 4 Outage Description of Change The temporary alteration provided chilled water to the Containment Fan Coolers (CFCs) via the Component Cooling Water System (CCWS) .
TAR-91-009 also provided operations personnel (by procedure OP-002-003) with control of the CFCs emergency discharge dampers and the CCWS inlet / outlet valves.
Similar temporary alterations were previously implemented via TAR-89-17 and TAR-90-016 to support containment activities respectively during the Refueling 3 Outage and a forced mafe.?,enance outage.
Reason for Change The temporary alteration was implemented to enhance the containment environment in support of work activities in containment during the Refueling 4 Outage.
The TAR-91-009 facility changes later are to be made a permanent plant configuration by Design Change (DC-3073).
Safety Evaluaticn The temporary addition is to the non-nuclear afety supply and return portions of CCWS. TAR-91-009 is applicable only to Operational Modes 5 and 6 (i.e., Cold shutdown and Refueling) during the Refueling 4 Outage. The Technical Specifications do not require CCWS or CFCs to be operable in modes 5 or 6.
The Containment Fan Coolers (CFCs), along with Containment Spray, are designed to operate and maintain the centainment within required pressure and temperature limite during normal plant operation and under LOCA and Main Steam Line Break accident conditions. The CFCs reject heat to the outside environtbnt via the CCW system. TAR-91-009 isolated CCW to the CFCs and En alternate path is established from the CFCs to the temporary chillers. Heat is thus rejected to the outside environment via the temporary chillers instead of the CCW system.
The temporary facility change does not require special testing or system operation in an abnormal manner. TAR-91-009 was implemented only in modes 5 or 6 and therefore had no impact on the FSAR analyzed accidents or operational equipment malfunctions.
TAR-91-009 did not create new accident possibilities or new methods of e;uipment failure. Margins of safety were not affected by the temporary facility change, and no Technical Specification change was needed for TAR-91-009.
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- 65. TAR-91-010, Temporary Access Point for Security to Maintain Accountability Description of Change The temporary alteration established a temporary personnel access point from the Health Physics Trailer to Containment and Q-Deck areas during the Refueling 4 Outage. TAR-91-010 installed 2 temporary cardrea.ders at the 2 temporary access doors.
Duration of TAR-91-010 was to the completion of the Refueling 4 Oatage.
Reabon for Change The temporary alteration was implemented to activate the Reactor containment Building west side access point to facilitate outage activities, and permit the Security Department to maintain personnel accountability during the Refueling 4 Outage.
Safety Evaluation During refueling outages, a Temporary Health Physics Trailer is used to allow additional personnel accet3 to the containment area.
A security officer is posted to ensure proper utilization of the temporary access point. The disabling of a cardreader and installation of 2 doors temporarily altered the access control system of the approved Physical Security Plan.
The temporary alteration had no impact on the probability of occurrence or consequences of previously evaluated accidents or malfunctionsaof equipment Lmportant to safety. The TAR-91-010 facility changes did not affect safety related or Ltportant to safety equipment or systems. Margins of plant safety were not affected and a Technical Specification change was not needed for 7.TR-91-OlO.
The security measures are the same special security measures used during Refueling 2 and 3 Outages and agreed to by the NRC Region IV personnel. Chapter 10 of the Waterford 3 physical security plan allows for special security measures during outages. The special security measures ensure that accident, equipment malfunction and margin of safety considerations are not adversely affected.
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- 66. TAR-91-012, Install a Plug in RCP 2A Case to Allow Repair Description of Change Tne temporary alteration installed a volute plug in the case of i Reactor Coolant Pump RCP 2A to sllow pump repair during the )
Refueling 4 Outage. The plug allowed the Reactor Coolant System J (RCS) to be filled above the level of the RCP. 1 Duration or the TAR-91-012 was set for about 25 days.
Reason for Change The volute plug was installed by TAR-91-012 to allow the concurrent support of refueling activities while the pump RCP 2A was maintenance serviced.
Safety Evaluation RCP 2A only functions as a primary boundary during refueling operations. Installation of the volute plug did not change the function of RCP 2A as a primary coolant boundary. The volute plug does not obstruct reactor coolant (shutdown cooling) flow.
RCS leakage around the installed plug was prevented by a nitrogen inflated seal and a setv..d nitrogen inflated seal and an 0-ring functioned as backups against leakage. The volute plug is not an ASME B&PV Code component, however, it functioned as a pressure boundary against static head during refueling when the RCS was filled above the pump level. The plug has three redundant sealing devices (two inflatable seals and one 0-ring seal). The two inflatable seals have independent nitrogen supplies to prevent a single failure ffom disabling both scals. In the unlikely event that all three sealing devices were to fail, the expected leakage would be approximately 332 gpm which is within the capacity of one HPSI pump, and within the FSAR analysis. While in reduced inventory conditions two HPSI pumps a.re required to be availabla.
Additionally, one LPSI or containment spray pump would be capable of supplying required RCS makeup.
The volute plug was seismically qualified and capable of withstanding a maximum pressure of 35 psig deemed to be the worst case condition during a loss of shutdown cooling event. The plug
-is designed to prevent falling inward or have upward movement.
The probability of occurrence or consequences of an accident or malfunction of equipment important to safety as previously evaluated le not increased by the temporary alteration. No new Reactor Coolant System interactions or connections are introduced by the temporary facility change. The margins of safety are not l
affected and a Technical Specification change was not needed for I TAR-91-012.
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- 67. TAR-91-020 (Rev. 1), Blind Flange for Valve SI-1088 Description of Change The temporary alteration installed a blind flange on the downstream flange face in place of the wafer check valve SI-108B.
TAR-91-020 was removed prior to entry into operational Mode 4, Hot Shutdown.
A similar temporary alteration, TAR-91-024, installed a blind flange on the upstream flange face in place of EI-108B (cee this Report Item No. 69).
Reason-for Change Valve SI-108B needed repair and the temporary alteration provided a blind flange so the valve could be removed and repaired for reinstallation.
Safety Evaluation TAR-91-020 blocked the flow path from the Refueling Watcr Storage Pool (RWSP) to Low Pressure Safety Injection (LPSI) Pump B and prevented operation of LPSI B on safety injection initiation. The temporary alteration did not affect shutdown cooling operations except the LPSI B side was unavailable and the LPSI A nide had to be available when the pool level was less than 23 feet above the reactor vessel flange. With the reduced decay heat load following refueling, the use of one HPSI pump and one LPSI pump assured sufficient flow for shutdown cooling.
The temporary alteration did not alter the probability of occurrence or the consequences of accidents or malfunctions of equipment important to safety, as previously evaluated.
The temporary facility change made no new system interactions or connections and did not create the possibility of an accident or equipment malfunction of a different type than previously evaluated. Plant margins of safety were not reduced and a Technical 3pecification change was not needed for TAR-91-020.
68 I.
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- 68. TAR-91-022, Provide-Station Air to Containment During LLRT of Penetration 48 Description of Change TAR-91-022 provided statior r.r to containment while Penetration
- 8 Station Air was local larx rate tested (LLRT). The temporary alteration was approved.or.ly-for Plant Modes 5 and 6 (i.e., Cold Shutdown and Refueling) with an expected duration of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The temporary alteration was implemented by installing a 150 lb.
rated rubber hose (mechanical jumper) betweer valve SA-912 (station air inside containment) and the temporary station air manifold inside containment.
Reason for Change The temporary alteration was initiated to conduct the Local Leak Rate Test (LLRT) of Containment Penetration #8.
Safety Evaluation TAR-91-022 did not af fect containment closure. The Station Air l System was not impacted by TAR-91-022 which temporarily created an l interface between the permanent station air header in containment and the temporary station air and deviated from the plant as described in the FSAR.
The temporary alteration was applicable only during operational modes ! and 6 and did affect the probability of occurrence or consequences, of accidents or malfunctions of equipment important to safety, as previously evaluated.
The temporary facility change provided a new interconnection between similar systems and the potential for a differnt type accident or equipment malfunction than previously_ evaluated was not created.
TAR-91-022 did not affect a protective boundary and margin of safety was not reduced. A Technical Specification change was not needed for TAR-91-022.
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- 69. TAR-91-024, Installation of Blind Flange Upstream Side of Valve SI-lOBB Description of Change The temporary alteration installed a blir.d flange on the upstream flange face in place of the wafer check valve SI-1088.
TAR-91-024 was removed prior to entry into Operational Mode 4, Hot Shutdown.
A similar temporary alteration, TAR-91-020, installed a blind flange on the downstream flange faco in place of SI-108B (see this Report Item No. 67).
Reason for Change Valve SI-1083 needed repair and the temporary alteration provided a blind flange so the valve could be removed and repaired for reinstallation.
Safety Evaluation TAR-91-024 blocked the flow path from the Refueling Water Storag_
Pool (RWSP) to Low Pressure Safety Injection (LPSI) Pump D and prevented operation of LPSI-B on safety injection initiation. The temporary alteration did not affect shutdown cooling operations except the LPSI B side was unavailable and the LPSI A side had to be available when the reactor coolant was less than 23 feet above the reactor vescel flange. With the reduced decay heat load following refueling, the une of one HPSI pump and one LPSI pump assured sufficient flow for shutdown cooling.
The temporary alteration did not alter the probability of occurrence or the consequences, 6! accidents or malfunctions of equipment important to safety, as previously evaluatea.
The temporary facility change made no new system interactions or connections and did not create the possiblity or an accident or equipment malfunction of a different type than previously evaluated. Plant margins of safety were not reduced and a Technical Specification change was not needed for TAR-91-024.
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!~ 70. TAR-91-025, Temporary Instrument Air Service to containmont Description of Change The temporary alteration installed a mechanical ;umper (15n lb.
rated rubber hose) to maintain instrument air (IA) service to containment during the Local Leak Rate Testing (LLRT) of the IA containment penetration isolation valves.
TAR-91-025 was valid only in Operational Moden 5 and 6, i.e., Cold Shutdown and Refueling (and valid in Mode 6 only when core alteration or fuel movement was not in progreen). The temporary IA hose was routed through the containment personnel airlock.
Reacon for Change The temporary alteration was implemented to maintoin Instrument Air earvice to the Reactor Containment Building concurrent with performing the LLRT of the containment isolation valves for the IA penetration.
Safety Evaluation The temporary f acility char.ge deviated frca the normal plant configuration as given in tr.e FSAR.
TAR-91-025 was approved only for Operational Modus 5 and 6. The
' temporary change had no impact upon the IA system which serves no safety function and is not required for safe shutdown of the plant or to mitigate the consequences of an accident.
The temporary hose was provided with quick disconnects at the personnel airlock to ensure the requirement for 4-hour containment closure could be met ir, the event of loss of shutdown cooling.
TAR-91-025 did not increase the probcbility of occurrence or consequences, of accidents or malfunctions of equipment important to safety, as previously evaluated. The temporary alteration did not change a protective boundary and maegin of safety was not reduced. A Technical Specificatich changt wse not needed for TAR-91-025.
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- 71. TAR-91-030, QSPDS Channel 1 HJTC #1 Heater Disconnection Description of Change The temporary alteration replaced the HJTC source heater #1 with a 25 ohm, 75 watt resistor. The temporary alteration disconnected the heated and unheatod thermocouples for the Qualified Safety Parameter Display System (QSPDS) Channel 1 Heated Junction Thermocouple (HJTC) Sensor #1 and installed electrical jumpers in their place to return the Channel 1 to service.
Reason for Change The temporary alteration was performed to make the remainder of the Reactor Vessel Level System (RVLS) operational per Technical Specifications 3.3.3.6 (Table 3.3-10).
Similar teaporary alterations, TAR-91-003, TAR-91-005 and TAR-91-031, were performed for other defective heaters (see Report Items Nos. 60, 62 and 72). -(NOTE: TAR-91-003 was closed 5/4/91 and TAR-91-005 was closed 5/16/91.)
Safety Evaluation The function of the HJTC system is to measure the water inventory in the reactor vessel above the fuel alignment plate. The QSPDS Channel 1 probe will continue to function as designed with HJTC #1 sensor inoperable by using the remaining 7 uensors to measure vessel water level.
The HJTC prcvides information to the operators and will not affect the ability to shut down the plant. No other equipment will be affected. The reactor vessel level monitoring will remain operable. The change is confined to wiring changes in a monitoring system outside of containmene and cannot cause a different accident than previously evaluated. The change will lower the number of operable eensors in the HJTC channel 1 probe from 8 to 7 (2 in the head and 5 in the plenum). The Technical Specifications allows the channel to remain operable with a minimum of 1 sensor in the head and 3 in the plenum. Thus, the Technical Specifications basis and the margin of safety is not reduced.
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- 72. TAR-91-031, QSPDS Channel 2 HJTC #4 Heater Disconnection Description of Change The temporary alteration replaced the hJTC soutco heater #4 with s 25 ohm, 75 watt resistor. The temporary alteration disconnected the heated and unheated thermocouples for the Qualified Safety Parameter Display System (QSPDS) Channel 1 Heated Junction Thermocouplo (HJTC) Sensor #4 and installed electrical jumpers in their place to return Channel 1 to service.
Reason for Change The temporary alteration was performed to make the remainder of the Reactor Vessel Level System (RVLS) operational per Technical Specifications 3.3.3.6 (Table 3.3-10).
Similar temporary alterations, TAR-91-003, tat-91-005 and TAR-91-030, were performed for other defective teaters (ese Report Itema Nos. 60, 62 and 71). (NOTE: TAR-91 -n03 w;s closed 5/4/91 and TAR-91-005 was closed 5/16/91.)
Safety Evaluation The function of the HJTC system is to measure the water inventory in the reactor vessel above the fuel alignment plate. The QSPDS Channel 1 probe will continue to function as designed with HJTC #4 sensor inoperable by using the remaining 7 sensors to measure vessel water level.
The HJTC provides information to the operators and will not affect the ability to shut down the plant. No other equipment will be affected. 2ne reactor vessel level monitoring will remain operable. The change is confined to wiring chnages in a monitoring system outside of containment and cannot cause a different &ccident than previously evaluated. The change will lower the number of operable sensors in the HJTC channel 1 probe from 8 to 7 (3 in the head and 4 in the plenum). The Technical Specifications allows the channel to remain operable with a minimum of 1 sensor in the head and 3 in the plenum. Thus, the Technical Specifications basis and the margin of safety is not reduced.
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I. E. SPECIAL EVALUATIONS (1) DOCUMENT REVISION NOTICES (DRNs) 9 -
- 73. DRN #E-8902088, -2009, Emergency Diesel Generator 4.16 kV Safety Bus - Manual Synchronization Description of Change The DRNs implemented changes for FSAR corrections to Figure 8.3-1, Onsite Power System Diesel Generator Logic Diagram; and the applicable Control Wiring Diagrams (CWDs).
The circuitry allows either Emergency Diesel Generator (EDG) to be manually synchronized to the respective 4.16 kV safety bus.
Previously, a permissive had existed on drawings with a bus tie breaker contact for synchronizing the EDGs to the 4.16 kV safety bus.
Reason for Change The FSAR and CWDs were updated by the DRNs to reflect the as-built plant configuration. The circuitry feature involved is requ! red for testing; and securing the EDGs follow..ig restoration of of f site power subsequent to a Loss of Of f eite Power (LOOP). The applicable CWDs and FSAd drawing were inadvertently overlooked during previous document updating at the time design change notices were initiated.
Safety Evaluation The original hardware change was implemented prior to plant commercial operation and t.he current plant configuration was not impacted by the DRNs. The system descriptib' was correctly given in the FSAR Section 8.3, Onsite Power Systems.
The DRNs corrected documen7ation only of electrical interlocks in the EDGs circuitry. No changes were made to the plant configuration or the operat lon of the EDGs or any subsystem.
Loss of-Offsite Power (LOOP) in the only applicable FSAR evaluated accident due to the interface with the EDGs.- The physical arrangement of the generator controls and bus tie breakers were-not modified by the DRNs. The EDG circuitry and the procedural administrative controls ensure that the probability of occurrence or-consequences of a LOOP accident are not increased.
l The probability of occurrence or consequences of a malfunction of i -the EDGs or the 4.16 kV safety buses was not increased by the DRNs
! -since the operating methods were not altered. The DP.Ns did not i
create any new system interactions, connections or modes of l operation. A Technical Specification change was not needed for DRN d2-8902088 and #E-8902089.
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74 DRN #I-9001835, Acceptable Pickup Voltage for HFA AC Relaya Description of change The DRN altered the Setpoint Document to reflect the correct j Design Engineering resolved pickup voltage for .4FA I.C relays.
setpoint differences associated with the undervoltage relays for the AB bee tr ansf er scheme and provided recommendations in response to thw Proble? Evaluation /Inf ormation Request, PEIR-61288.
Reason for Change Plant Engineering identified disparaties on pickup voltage for HFA AC relays between vendor carvice advice letters and the Setpoint Document. Design Engineering responded to PEIn-61288 and recommended that the Setpoint Document be revised to the correct values of 73 to 81 percent of rated (i.e., 87.6 to 97.2 VAC) on acceptable pickup voltages for HFA AC relays.
Safety Evaluation The affected relays are used to assure a dead bue transfer of the AB buses and to monitor for a blown fuse of their potential transformers.
The setpoint clarification had no affect on the function of equipment or procedures and did not compromise plant safety or licensing basis. Correcting the setpoint 'f the affected relays in the Setpoint Document created no new fai 'e modes and hed no nuclear safety significance. Ine AB bus tra.afer scheme undarvoltage relays continued to operate as designed and described in FSAR Section 8.1.3.
The bus transfer scheme affected by the change is used only when the reactor is not at power and thus does not contribute to the probability of occurrence of accidente previously evaluated. No new system interactions or connections were created by the DRN; and margin of plant safety is not affected. A Technical Specification change was not needed for DRN #I-9001835.
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- 75. DRN #M-8800406, Instrument Air System - Flow Diagram Description of Change The DRN updated-and corrected drawing discrepancies for FSAR Figure 9.3-1, Sheet 4, Flow Diagram Service and Instrument Air Systems. The change corrected instrument identification numbering and added missing Inste ment Air Isolation Valves on the drawing.
Reason for Change Problem Evaluation /Information Request No. PEIR-10468 identified drawing discrepancies relative to the Instruments List. FSAR Figure 9.3-1 contained incorrect instrument identification numbers and missing instrument air isolation valves.
Safety Evaluation The drawing changes, and additions, implemented by DRN fM-3800406, only involved number designations for non-safety valved and instruments.
The probability of occurrence and consequences, of accidents and malfunction of equipment important to safety, as previously evalutted are not increased by the DRN which provided only for the drawing update.
The DRN did not alter the physical plant configuration. Margin of safety was not affected by the drawing changes. A Technical SpeciC mtion was not needed for ORN #M-8800406.
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- 76. DRN #M-8800851, Component Cooling Water - Flow Diagram i Description of Change The DRM was initiated to revise drawings, and the related FSAR Figure 9.2-1, Flow Dingram Component closed Cooling Water System.
The drawing changa corrected the identification numbers on Blowdown Sample Cooler valves.
Reason for Change Drawings and the related FSAR figure needed a revision to give correct valve identification numbers. Problem Evaluation /
Information Request No. PEIR-10287 identified drawing discrepancies regarding incorrect identification ntunbers for Blowdown Sample Cooler valves.
Safety Evaluation The DRN was a documentation change only to revise the drawing discrepancies un valve identification nurtering. No phy91 cal alterations to the facility occurred. The changes had no affect on accidents or malfunctions of equipment. A Technical Specification change was not needed for the FSAR revision made by DRN #M-8800851.
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- 77. DRN (M-8800920,- component cooling Water System - Flow Diagram Description of Change The DRN' was initiated to revise drawings, and the related- PSAR
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Figure 9.2-1, Flow Diagram component closed Cooling Water System.
The. drawing change reflected the correct and as-built conditions for the Component Cooling Water system for drains and pressure test connections for the HPSI and LPSI pumps sealing and bearing cooling water.
Reason for Change Drawings and the related FSAR figure needed a revision to correct drafting draving errors and show the correct as-built
- onfiguration for portions of the Component closed cooling Water System.
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Safety Evaluation The DRN was 4 documentation change only to revise the drawing discrepanicas on as-built configuration. No physical alterations to the facility occurred. The changes had no affect on accidents or malfunctions of equipment. A Technical Specification change was not needed for the FSAR revision made by DRN fM-8800920. ,
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- 78. DRN #M-8801798, HVAC Air Flow Diagram Description of Change The DRN identified the drawing discrepancies of existing instrumentation and root valves not shown on the HVAC drawings.
Similar DRNs are reported as follows:
DRN #H-8801799 - Report Item No. 79 DRN fM-9001008 - Report Item No. 81 DRN #M-9001009 - Report Item No. 82 The DRNs identified the drawing changes needed for later incorporation by the planned conversion of all UVAC Flow Diagrams into Piping and Instrumentation Diagrams (P&lDs).
Reason for Change Problem Evaluation /Information Request No. PEIR-71032 identified the discrepancies on drawings and related FSAR figures of missing instrumentation and root valves on various HVAC systems. HVAC root valves were assigned unique identification numbers.
Safety Evaluation A generic safety evaluation was prepared and applied to the related DRNs M-8801798, M-8801799, M-9001008 and M-9001009.
The addition of identification numbers to the HVAC root valves had no impact on the function of any equipment or procedures. The coot valves serve to isolate the differential pressure transmitters during testing and maintenance for the air handling units and filtration units.
The DRNs provided additional information for the drawings to assist in the identification of root valves and add tags to the root valves, but did not involve any actual changes to the plant systems, equipment or procedures and had no impact on any plant accident scenarios.
The DRNs did not create any new systaa interactions or connections to provide the possibility of any new types of accidents or malfunctions of equipment important to safety. The DRNs did not change any margin of safety and did not affect any protective boundary. A Technical Specification was not needed for the DRNs.
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- 79. DRN iM-8801799,15fAC Air Flow Diagram Description of Change l The DRN identified the drawing diserspanicies of existing !
instrumentation and root valves not chown on the HVAC drawings. l Similar DRNs are reported as follows:
DRN #M-8801798 - Report Item No. 78 DRN #M-9001008 - Report Item No. 81 DRN #M-9001009 - Report Item No. 82 The DRNs identified the drawing changes needed for later incorporation by the planned conversion of all HVAC Flow Diagrams into Piping and Instrumentation Diagrams (P& ids).
Reason for Change Problem Evaluation /Information Request No. PEIR-71032 identified the discrepancies on drawinge and related FSAR figures of missing
- instrumentation and root valves on various HVAC systems. HVAC root valves were assigned unique identification numbers.
Safety Evaluation A generic safety evaluation was prepared and applied to the related DRNs M-8801798, M-8801799, M-9001008 and M-9001009.
The-addition of identification numbers to the HVAC root valves had no impact on the function of any equipment or proceduren. The root valves aerve to isolate the differential pressure transmitters during testing and maintenance for the air handling units and filtration units.
The DRNs provided additional information for the drawings to assist in the identification of root valves and add tage to the root valves, but did not involve any actual changes to the plant' systems, equipment or procedures and had no impact on any plant accident scenarios.
The DRNs did not create any new system interactions or conne1tions to provide the possibility of any new types of accidents or malfunctions of equipment important to safety. The DRNr did not change any margin of safety and did not affect any protective-boundary. A Technical Specification was not needed for the DRNs.
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- 80. DRN #H-8900b25, c,omponent closed Cooling ' dater system Description of Cha m The DRN restored valves cc-80310B and CC-8062 in the component closed cooling Water System to their original design basis position as normally closed.
Reason for change Previously, DRH #H-8800868 had erroneously defeated the original design basis and changed the position of valves cc-8,310B and cc-8062 from normally closed to normally open.
(The reportable event of operating in a condition outside t:
plant design basis was reported to the NRC on June 14, 1990 bicensee Event Report Number LER-89-006-01.)
Saf ety Evaluation DRM fM-8900525 restored the valves to the correct design basis rasition of normally closed.
valves cc-80310B and cc-8062 are manual isolation valves which supply corpunent cooling water (CCW) to the post accident sampling system (PASS) chillet and heat exchanger #2. Both components are
- ASME code break valves and are located at the boundary between the Safety class 3 and noa-nuclear astoty (NNS) portions of the CCW eystem. Because theon valves do not have remote operators, they are requirad to be normally closed f or leak isolation in the event of a NNS piping break.
The restoration of the valves to the design basis position did not increase the probabilj.ty of occurrence or consequences, af accidents or malfunct!.ons of equipment important to safety as pteviouely evaluated. Tt.2 DRN did not create any new system interactions or connections. Margins of plant safety wure not aff(cted; and a Techt.ical Specification change was not needed for DRN fM-8900525.
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- 81. DhN fM-9001008, HVAC Air Flow Diagram Description or Change The DRN identified the drawing discrepanicies of existing '
instrumentation and root valves not shown on thw HVAC drawings.
similar DRNs are reported as follows:
DRN #H-8801798 - Report item No. '8 DRN f M-8801799 - Report Item No. 79 DRN $H-9001009 - Report Item No 82 The DRNs identified the drawing changes needed for later
- incorporation by the pinnned conversion of all HVAC Flow Diagrams into Piping and Instrumentation Diagrams (PGIDs).
Reason for Change Problem Evaluation /Information Request No. PEIR-71032 identified '
the discrepancios on drawings and related FSAR figures of missing instrumentation and root valves on various HVAC systems. HVAC root valves were assigned unique identification numbers.
l Safety Evaluation A generic safety evaluation was prepared and artlied to the related DRNs M-8801798, H-8801799, M*0001008 and M 9001009.
The addition of identification numbers to the HVAC root valver had-no impact on the function of any equipment or procedures. The root valves serve to isolate the differential pressure transmitters during testing and maintenance for the air handling units ar.d filtration units.
The DRNs provided additional information for the drawings to assist in the identification of root valves a.J add tags to the root-valves, but did not involve any actual changes to the plant systems, equipment or proceduros and had no imps-t on any plant ,
accident scenarios.
The DRNs did not create any new system interactions or connections to provide the possibility of any new types of accidents or malfunctions of equipment important to etafety. The DRNs did not change any margin of safety and did not affect cny protective boundary. A Technical Specification was not needed for the DFNa.
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i The DRN identified the drawing discrepanicies of existing instrumentation and root valves not shown on the HVAC drawings.
Similar DRNs are reported as follows:
DRN 4M-8801798 - Report Item No. 78 DRN fM-8001799 - Report Item No. 79 DRN #H-9001008 - Report Item No. 81 ,
The DRNs identified the drawing changes needed for later incorporation by the planned conversion of all MVAC Flow Diagrams into Piping and Instrumentation Diagrams (P& ids).
Raason for Change Problem Evaluation /Information Request No. PEIR-71032 identified the discrepancies on drawings and related TSAR figures of missing instrumentation and root valves on various HVAC systems. HVAC root valves were assigned unique identification numbers.
( Safety Evaluation A generic safety evaluation was prepared and applied to the related DRNs M-8801798, M-8801799, H-9001008 and H-9001009.
The addition of identification numbers to the HVAC root valves had no impact on the function of any equipment or procedures. The root valves serve to isolate the differential pressure transmitters during testing and maintenance for the air handling ,
units and filtration units.
The DRNs provided additicnal information for the drawings to assist in the identificatioa of root valves and add tags to the root valves, but did not involve any actual changes to the plant systems, equipment or proceduren and had no impact or any plant-accident scenarios.
The DRNs did not create any new sy3 tem interactions or connections to provide the possibi'ity of any new types cf accidents or malfunctions of equipment important to safety. The DRNs did not change any margin of safety an.d did not affect any protective
- boundary. A Technical Specification was not needed for the DRNs.
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- 83. DRNs 4H-9100144, et al, Inconsistancies in Piping Designations Description of change The DRNs reestablished some piping design designations to overcome inconsistencies that existed due to connections of system piping to other systems. Component cooling Water System linsa to condensate lines; and Chenical Feed lines to the Emergency Feedwater suction piping were involved in the piping designation changes.
The DRNs resulted in required changes for tag numbers in FSAR Figure 9.2-1, Flow Diagram Component C1nced Cooling Water System, and Figure 10.4-2, Flow Diagram Feedwater Condensato and Air Evacuation System.
The DRNs relocated the change in pressure and temperature design requirements in piping runs such that the boundary between pressure and temperature qualificatior. is placed at isolation valves.
Reason for change Design inconsistencies were i entified via Problem Evaluation /
Information Request No. PEIR-> 930. Resolution of the piping designation inconsistencies w,o needed to satisfy ASME hydrostatic test requirements.
Safety Evaluation some piping affected by the DRNs is used to provide cooldown of the Reactor Coolant System following a main steam or feedwater line break or loss of normal foodwater in the event the condensate storage pool is et a low level. Other piping af fected by the DRNs serves as chemical feed lines f rom the hydrazine and ammonia pumps to the Emergency Feedwater System suction piping.
The DRNs altered FSAR inf ormation but not the oporation, function, or ability to perform the funcsion of any system, structure or component. The FSAR paper changes involved modifying the design pressure and tsmporature of piping sections to be consistent with surrounding systems.
The probability of occurrence or consequences of an accident previously evaluated was not increased by the DRN changes.
The DRNs did not increase the probability of occurrence or consequences of a malfunction of equipment important to safety as previously evaluated. No new system interactions or connections resulted from the DRNs. Margin of plant safety was unchanged. A Technical Spocification change was not needed for the DRNs 4M-9100744 et al.
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- 1. E. SPECTAL EVALUATIONS (2) LICENSE DOCUMENT CilANGE RFQUEST (LDCRs)
B4. LDCR-89-147, Diesel Generator Loading Sequence Description of Change LDCR-09-147 corrected an illogical energizing sequence shown in the FSAR ft.c Emergency Diesel Generator loading. The LDCR revised a) FSAR Subso;*.lon 8.3.1.1.2.13, Diesel Generator Load Shedding Circuits, b) FSAR Table 8.3-1, u 3 eel Generator Loading Sequence, c) FSAR Figure 8.3-27a, (Diesel Ge.srat or Loading Sequence)
Loading Prior to Load Block 1 During LOCA ca :tELB with LOOP, and d) FSAR Figure 8.3-27b, Loading Prior to Load Block 1 During Loop.
Reason for Change The FSAR changes were made by LDCR-89-147 to provide agreement with the correct Control Wiring Diagrama (CWDs) and to conform to the an-installed plant equipment.
Safety Evaluation The CWDs represent the current and correct system operation for the emergency diesel generator loadings and are the design basis doeurae nt s . LDCR-89-147 did not require design changee or alterations of the plant equipment, components or the physical configuration. The load sequencer and associated relays continue to function as designed and installed. No system changes to connections or settings were performed.
The FSAR changes did not increase the probability of occurrence or consequences, of an accident or a malfunction of equipment important to safety, as previously evaluated.
The PSAR changes did not create the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated.
Margin of safety was not reduced; and a Technical Specification change was not needed for LDCR-89-147.
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,, , da-M - dt . Sig ;'oeitive Suct',on Head for Component / Auxiliary i
Jenqhnent Cooling Nater Fumps
( 09?s s;g . of Change uDCH-90-0085 changed the available Not Positive Suction llead (t:PSH) values in FSAR Table 9.2-1, Design Data for the Component cooling Water (CCW) System and Auxiliary Component Cooling Water (ACCW) System Components. The correct design values are 85 feet for the CCW pumps and 26 feet for the ACCW pumps as given in plant design documentation.
Reason for change Problem Evaluation /Information Request No. PEIR-10783 identified open items (errors) in the design basis document for CCW and ACCW, i.e., W3-DDD-004. Document Revision Notice, DRN 4H-9001066 was initiated to correct the design basis document. LDCR-90-008$
corrected the FSAR.
Safety Evaluation There were no physical changes made to the plant configuration by correcting the available NPSH values in the PSAR and Design Bauia Document. The CCW/ACCW pumps had been sized for the correct NPSH values.
The FSAR changes did not increase the probability of occurrence or consequences, of an accident or a malf unction of equipment important to safety, as previously evaluated.
The FSAR changes did not create the possibility of an accident or a malf unction of equipment important to safety of a different type than previously evaluated.
Margin of nafety was not reduced; and a Technical Specification change was not needed for LDCR-90-0085.
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- o. LDCR-91-0157, 3 team Generator Tube Rupture Analysis FSjiR Update Description of change LDCH-91-0157 updated the FSAR accident analysis for a steam Generator Tube Rupture (SGPR) with concurrent Loss of Offsite l Power (LOOP). The FS AR update gives the newly calculated design basis radiological consequetces for the SGTR event.
Reason for Change j The SGTR accident analysis currently in the FSAR did not account for potential U-tube uncovery during the event. The update analysis was prompted by a review done pursuant to the NRC Information Notice No. 88-31 'intitled, " Steam Generator Tube Rupture Analysis Deficiency." The updated analysis is based upon more realittic assumptions concerning operator actions in respone, to the SGTR. t Safety Evaluation NO changes to the Emergency Operating Procedures were required as a result of the new analysis of the SGTR event. The on4y change by LDCR 491-0157 is the revision of the FSAR, as there was no physical change to the plant equipment, structures or components; and rse procedural changes were needed. Implementation of the FSAR revision via the change requebt LDCR 491-0157 ensures that the FSAR contains a SGTR analysis which conservatively bounds the potential radiological consequences which could result during an actual SGTR event. The reanalysis demonstrated that the NRC acceptance limits (in the safety Evaluation Report - NUREG-0787) are met in consideration of the potential for U-tube uncovery.
The FSAR change did not result in any system interactionD or connections which did not previously exist. The change merely documents an updated analysis of a SGTR and has no effect on the physical structure of the plant or in hov the plant is operated.
The FSAR changes did not increase the probability of occurrence or consequences, of an accident or a malfunction of equipment important to safety, as previously evaluated.
The FSAR changes did not create the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated.
Margin of safety was not reduced; and a Technical Specification change was not needed for LDCR-91-0157.
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- 07. LDCR-91-0190, component cooling Water Flow Under Accident conditionn (previously LD6fAF-10iT4)
Description of Change LDCR-91-0190 revised the accident response description for chillers in FSAR Table 9.2-2, CCWS and ACCWS Temperature and Flow Control Description.
Reason for Change The FSAR changes were made by LDCR-91-0190 to more accurately reflect the automatic and manual operator actions related to component Cooling Water (CCW) supply flow to the chillers under accident conditions. The change deleted the operator action, to reduce CCW flow to each Shutdown ilsat Exchanger to 2000 gpm to avoid runout of the CCW pump, as an engineering calculation demonstrated that pump runout would not occur.
The FSAR change action was taken to avert any gotential confusion for operator action in response to Observation H-5 identified in the Safety System Punctional Inspection of the component cooling Vater System.
Safety Evaluation The PSAR change describes the actual system operation and no physical alterations were made to equipment, systems or procedures. The system operates automatically per design and the capability exists for the CCW system to be aligned manually by the operators if it should be neceesary.
The PSAR changes did not increase the probability of occurrence or consequences, of an accident or a malfunction of equipment important to safety, as previously evaluated.
The PSAR changes did net create the possiblity of an accident or a malfunction of equipment important to safety of a different type than previously evaluated.
Margin of safety was not reduced; and a Technical Specification change was not rueded for LDCR-91-0190.
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- 88. LDCR-91-0191, Control Room Outside Air Intake Radiation Monitors Description of Change LDCR-91-0191 corrected and clarified infcrmation in FSAR Section 12.3.4, Area Radiation and Airborne Radioactivity Monitoring Instrumentation, concerning the operation and use of the control Room Outside Air Intake (CROAI) Radiation Monitors.
Reason for change, The F3AR revision af fected the description of the operation of the Control Room Outside Air Intake (CROAI) radiation monitore. The revision clarified the description of the function of the CROAI radiation monitors with the emergency outside air intakes, revised tho description of the gamma only detectors by listing their function as installed spares, and corrected the FSAR by changing the number of alarms required to cause a control room ventilation system actuation from 2 to 1.
Safety Evaluation The CROAI ra< ation monitors monitor the airborne radiation activity in :.e control room outside air intakes and alarm when high activity is detected. The alarm causes the control room normal air intake to isolate and the control room ventilation emergency system ta start.
The FSAR change affected the function of the CROAI radiation monitor's gamma only detector. The gamma only detector is an installed spare ui*.h no operation function. LDCR-91-0191 had no effect on the bota/ gamma detector or the CROAI radiation monitor's ability to perform its alarm and actuation functions.
The FSAR changes did not increase the probability of occurrence or consequenese, of an accident or a malfunction of equipment important to safety, ao previously evaluated.
The FSAR changes did not create the possibility of an accident or a malf unction of equipment important to safety of a dirierent type than previously evaluated.
Margin of safety was not reduced; and a Technical Specification change was not needed for LDCR-91-0191.
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- 89. LDCR-91-0207, containment Penetrations and Isolation valves Description of change LDCR-91-0207 revised the FSAR Table 6.2-32, Containment Penetrations and Isolation Valves. The designation for secondary r.. ode of actuation was changed to NONE for twelve valves listed in FSAR Table 6.2-32. The FSAR change more accurately reflects the actual design of the valves which have no uncondary mode of operation.
Reason for change Document Revision Notice. DRN fH-9001728 updated the deaign basis documer*. W3-DBD-013 for 1 Containment Spray System and in the process identified the 1. I for clarification on modes of valve actuations in FSAR Table o.2-32.
Smfety Evaluation The FSAR changes via LDCR-91-0207 more accurately reflect the actual design of the valves which have ne recondary mode of operation. The affected valves are not physically altered by this activity and the valvis are still required to perform a safety function in the event of a loss of coolant accident. The F3AR accident analysis have not taken credit for operation of these valves by any method other than the primary method of actuatien as given jn FSAR Table 6.2-32.
The FSAR changes did not increase the probability of occurrence or consequences, of an accident or a malfunction of aquipment inpurtant to safety, as previously evaluated.
The FSAR changen did not create the possibility of an accident or a malfunction of equipment important to safety of a different type than previously evaluated.
Margin of safety was not reduced; and a Technical Specification change was not needed for LDCR-91-0207.
90
I. E. SPECIAL EVALUATIONS (3) HISCELLANEOUS EVALUATIONS
- 90. Pump and Valve Inservlee Test Plan (Change 1 - Revision 5)
Description of change Change 1 to Revision 5 of the Pump and Valve Inservice Test (P&V IST) Plan addressed and it..lemented the six concerns expressed by the NRC in the Safety Evaluation Report dated 2/7/89. The Plan change deleted 3 unnecessary previous relief requests; and clarified 2 relief requests. Relief Request 2.1.3 was revised and added flow to the parameters list for comparing to limits when determining operability of pumps tested using mini-recirculation flowpaths.
Reason for Change Changes and clarifications were needed to resolve the NRC conectne expressed in the Fafety Evaluation Report for the P&" IST Plan.
The amended P&V IST Plan (Change 1 - Revision 5) was transmitted to the NRC on 6/19/89 via letter W3P89-3063.
Safety Evaluation The P&V IST Plan change did not affect the design or operation of safety related equipment and the probability of occurrence or consequences of accidente previously evaluated were not increased.
The design and operation of safety related equipment is not affected by the Plan changes and the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased.
The possibility of an accident or equipment malfunction of a different type than previously evaluated ware not created by the Plan changen. Margin of plant safety is not decreaped; and a Technical Specification change was not needed for the P&V IST Plan (Change 1 - Revision 5).
91
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- 91. Pump and Valve Innervice Tes: Plan (Revision 7)
Description of Change The Pump and Valve Inservice Test (P&V IST) Plan (Revision 7) '
incorporated changes to address the generic potential plan deficiencies identified by the NRC in the Generic Letter No.
89-04, Guidance on Developing Acceptable Inservice Testing Programe.
Reason for Change Revision 7 of the P&V IST Plan was initiated to ensure compliance ;
with the requirements of the ASKE B&PV Code section XI and to document conformance to the NRC guidance in Generic Letter No. i 89-04. [
Safety Evaluation The majority of the Plan changes in Revision 7 constituted program enhancements and adherence to the detailed guidance of the NRC Ceneric Letter No. 89-04. A new approach had recently been
, initiated by the NRC regarding P&V IST Plans. NRC no longer
! reviews.and approves the plans and changes thereto; rather, the l NRC relies upon inspection teams to determinsi the adequacy of licensee plans towards meeting the regulatory requirements for the inservice testing of pumps and valves. .
The changes in P&V IST Plan (Revision 7) are consistent with the 1e 'el of testing assumed in the Fafety Analysis Report. The plan changes did not affect.the probability of occurrence or ,
consequences, of accidents or equipment malfunctions Lmportant to safety, as previously evaluated.
The Plan changes did not alter the plant design, configuration, or operation. No new system interactions or connections were created by the changes. The margin of safety is maintained by the Plan changes. A Technical Specification change was not needed for P&V IST Plan (Revision 7).
92
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- 92. Discrepancies in Pump and valve Inservice Test Plan Description of change Three testing discrepancies were identified in the approved Pump and Valve Inservice Test (P&V IST) Plane Engineering evaluation determined that the alternate testing provided was acceptable for interim "use-as-is" of the affected components.
The Plan discrepancies consisted oft a) omission of testing of r.mergency Teodwater (ErW) System ;
valves which prevent diversion of ErW flow upon a loss of ;
main feedwater, b) omission of testing of the Emergency Diesel Generator fuel oil transfer pump, and c) f ailure to include code relief request for exemption of the test supervisor qualifications for safety valve testing.
Reason for Change Condition Identification CI-270317 documented three deficiencies
-in the Pump and Valve Inservice Test-(IST) Program. The Engineering Input for CI-270317-(Revision 1) concluded there was no impact on nuclear safety ce the alternate testing is reasonable alternative to the ASME Boiler & Pressure vessel Code requirements ,
for establishing operability of the involved components. t safety Evaluation In each discrepancy case, alternate testing provided a reasonable 1
-alternative to the ASKE code-requirements to establish, on an interim "use-as-is" basis, operability of the components involved.
Engineering evaluation determined there was a high level of confidence that the components would perform their intended safety
. function if required.
Details on the inconsistencies _r le P&V IST Plan were reported to the NRC in Licensee Event Report Number LER-90-010-01,
" Inconsistencies in the Pump and Valve In-Service Test Program" ,
which was transmitted to the.NRC via W3BS-91-0099 on T/17/91. 1 l
l' 93- ,
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- 93. Pressuriser safety Relief Valves Description of Change Pressurizer safety relief valves (purchased on P.O. #WP035057) deviated from the original valves relative to ASME BGPV applicable Code year, time of hydrostatic test, and performance of ASTM A-262 Practice E Test for susceptibility to stress corrosion.
Reason for Chan2" Discrepancy Notice, DN-5569-90, identified all the deviations of the purchased valves. Engineering evaluation showed that the quality of the valves, with the Dressor modifications, is equal to or better than the original valves; the probability of a valve malfunction is not increased. The engineering evaluation showed that the valves meet all the ASME requirements and FSAR connitments.
Safety Evaluation The function of the pressuriser safety relief valves (SRys) is to protect the reactor coolant system from overpressurization. The changes in the purchased pressuriser SRVs did not affect the ability of the equipment to perform the design function. The purchased valves were reconciled to the code of construction for the existing safety valves.
The method of operating the pressurizer SRVs was not altered and no special test requiring system abnormal operation was involved in the changed SRVs. The evaluation of the valves differences determined there were no chatges to FSAR commitments or licensing basis.
The new valves met the same design commitments of the original valves and the probability of occurrence or consequences, of accidents or malfunction of the valves, as previously evaluated is not increased. The valve differences did not affect the possibility of an accident or equipment malfunction of a different type than previously analyzed. No margin of safety is affected.
A Technical specification change was not needed for the pressurizer SRVs.
94
- 94. SPEER $91-E-001, Reactor Coolant Pump seal Upgrade Dameription of Change j spare Parte Equivalency Evaluation Report, SPEER 91-E-001 is a '
design enhancement which upgraded the current N-9000 seal for the !
Byron Jackson manufactured Reactor Coolant Pumps (RCPs) by replacing semis with a new designed N-9000 ;eal cartridge. l Reason for Change ,
The N-9000 seal cartridges installed on the RCPs in 1987 and 1988 ,
gave unsatisfactory performance. The seals displayed symptoms of ,
high outleakage, high temperature differentials across the seal ,
cartridge and destaging. ;
Safety Evaluation l Reactor Coolant Pump seals control reactor coolant leakage around the pump shaft. No pressure boundary materials are changed by ,
SPEER 91-E-001; changes are dimensional only to the RCP internal e i parts on the three main stages. The 4th stage of the pump seal (vapor stage) is not altured and can withstand full system i pressure upon failure of the other 3 stages. The failure probability of the RCPs im expected to decrease because of the seal irprovements. The new seal parts are compatible and .
Interchangeable with the existing N-9000 seal. '
No additional accident scenaries or equipment malfunctions are created by the.RCP replacement seals.-
i The probability of occurrence or consequences, of accidents or equipment malfunctions, as previously evaluated are not increased by the new RCP seal design. The .argin of safety has not been reduced by the change because the pressure boundary role of the upgraded N-9000 RCP seal is the same as the existing seal. A Technical Specification change was not needed for SPEER 891-Es001.
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- 95. SPEER #91-608, oil Level Probes in Reacter Coolant Pump Hotor !
Reservoirs Description of Change Spara parts Equivalency Evaluation heport, SPEER #91-608 replaced the 1/4-in-h diameter oil level probes with 1/2-inch diameter ,
probes in the upper and lower reservoirs for the Reactor coolant Pumps (RCPs) motors. ,
Reason for Change The existing capacitance proben which are subcomponents of the RCP .
motor oil level transmitters were unreliable.
Safety Evaluation ,
The RCP motors are non-safety related. The probe change by SPEER
- 91-608 improved the reliability of the RCP motor oil level instrumentation. The probe change did not impact any proceduress and no special tests were required involving abnormal system operation.
The replacement of the new probes did not increase the probability of occurrence or consequences of accidents or malfunctions of equipment important to safety, as previously evaluated, because the reactor coolant system was not degraded.
Increased reliability of oil level indications is provided by the change thus decreasing the pro'oability of occurrence of motor bearing failure from oil starvation. No new system interactions or connections are introduced by '.he replacement oil probes.
Margin of plant safety is not affected by the oil probes. A Technical specification change was not needed for SPEER #91-608.
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I 96
- 96. SPEER $91-611, Emergency Diesel 'enerator riectronic Governor control _s Description offhange Spare Parts Equivalency Evaluation Report, SPEER #91-611, replaced the dual parallel voltage dropping resistor assembly for the Emergency Diesel Generator (EDG) electronic governor controls with a single path resistor assembly.
Reason for Change The SPEER 891-611 replacement was a result of the problem identified in the NRC Information Notice IN 90-51, " Failures of Voltage-Dropping Resistors in the Power Supply circuitry of d Electric Governor Systems". The new resistor assembly provides increased reliability. over the old assembly, as it allows the mechanical gcVernor tu take over speed control whereas failure of the old assembly rendered the EDG inoperable.
Safety Evaluation The electronic governor provides the primary means of speed control for the EDGs.
The resiotor assembly replacement by SPEER 991-611 likely will decrease the probability of a station blackout because of improved reliability. Similarly, the likelihood of EDO malfunction is decreased because of more reliability and failure of the resistor assembly does not make the EDG inoperable.
SPEER #91-611 did not increase the consequences of an EDG failure.
Replacement of the EDG resistor assembly did not introduce any new system interactions or connections as the new resistor assembly required no wiring changes. No new types of equipment failures are created by the now resistor assembly. Margins of plant safety are not reduced. A Technical Specification change was not needed for SPEER 891-611.
97
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- 97. Cycle 5 Reload Description of Change cycle 5 Reload changed the plant configuration by the replacement of 84 fuel assemblies with fresh (Batch c) fuel having a nov ,
debris-resistant fuel design.
Reload Batch 0 consists of 16 type 00 assemblies (no poison rods),
PC type 01 assemblies (8 burnable poison rods per assembly), and 4 4d type 02 assemblies (16 burnable poison rods per assembly). The '
core was loaded with quarter core rotational symmetry.
The debris-Jesistant fuel design was incorporated into the Batch 0 assemblies. This design elevates the fuel and poison columns so that debris which is caught by the inconel grid interacts with a lengthened solid end cap instead of fuel (or poison) rod cladding.
The Cycle 5 core will consist of 84 fresh Batch C fuel assemblies, 84 Batch F assemblies initially inserted in Cycle 4, 40 Batch E assemblies initially inserted in Cycle 3, and one twice burned Batch C assembly which was dircharged at the end of Cycle 2 and
- will be reinserted for Cycle 5. The cne Batch C assembly and the l 40 Batch D assemblies in the cycle 4 core will be discharged to I the spent fuel pool, as will 36 of the 84 Cycle 4 Batch E assemblies.
Reason for Change The facility change provided the reactor fuel core for operational cycle 5.
Safety Evaluation The reactor core is the heat source for the power plant. The fuel assemblies which constitute the reactor core provide a primary barrier to the release of radioactive material that must be maintained within acceptable limits.
The Reload Analysis Report (RAR)-describes the analyses performed '
for Cycle 5. Cycle 2 is the Reference cycle used for the RAR.
The specific changes from the Reference Cycle that had potential impact --4 were specifically evaluated are:
- Burnup for some fuel rods are calculated to exceed the i
52,000 MWD /T peak rod burnup value discussed in the CE High Burnup Topical Report approved by the NRC.
- Cycle 5 will have a flatter power distribution than the
. Reference cycle.
- Datch 0 fuel incorporates a debris resistant fuel design, which is designed to reduce debris-induced failures of fuel pins and thus reduce the number and/or potential for leaking fuel pins.
- Cycle 5 CEA shutdown worths are decreased comparad with previous cycles but still remain within Technical Specification limits.
(continued) 98 ,
(continued) 97. Cycle B Peload The safety analyses for the Excess Leto and Pre-trip Hain Steam Line Break events are based on an assumption that the CPC addressable constant for radial peaking is not reduced in value late in the Cycle, when the predicted peak value is past.
The safety anal"ses documenced that all potential accidents using Cycle 5 fuel and design basis events have consequences bounded by the Reference Cycle 2 and are below the HRC acceptance limits.
The IOCFR50.59 safety evaluation was performed at the Cycle $ Core Meload has t!.o potential to alter the core t.tutronics and/or thermal-hydraulic characteristics beetase of the adoption of a debris rosistant fuel design. The active fuel lengch is slightly shorter and results in an upward shif t in the fuul. The safety analysis clearly addresses the f uel asson61y chaages to show that the Referonce Cycle 2 analyses are bounding. Thu adoption of the debris-resistance fuel design in the Cycle 5 new fuel is expected to decrease the probability of debris-related fuel failures.
There are no accidents evr'unted in the TSAR that are initiated by the reactor core. Aed.
malfunctions which potenti[ ally resultents areininitiated by equipment The cora transienth.
fuel itself has no effect on the likelihood of occurrence of an accident.
As documented in the Cycle S Reload Analysis Report, the consequences for all previously eva'uated accidents romain bounded by the Referencu Cycle analyses anj within NRC acceptance limits.
This change does nat inccesse the cu. sequence of any accident previously evalunted in the Fahr.
All equi e.snt t importent to s af ety will iunction in the manie manner with the reload core as witn the previous core. There is no characteristic of the Cycle 5 core (including the debris resistant design of Baten 0) different from the cores from previous cyc1 9e which would tend to increase the probability of a malfunction of equipment laportant to safety.
There are no new nystem interact, tons er cennections associated with core reload. Therefore, operation with the Cycle 5 reload core will not cause an accidttnt of a different type than any previously traluated in the FSAR.
Installation of the reload core cannot cauue the possibility of a malfunction of equipmont important to saf ety of a dif ferent type tnan any previously eve.luato6 in the FSAR. Equipment important to safety will function in the orme manner with the reload core as with the previous core.
All accidents have been shown to have consequences bounded by the 1tuf erence Cycle and below the appropriata NRC acceptance limits.
- Therefore, there is no reduction in any margin of safety.
f h Technical Specification change was not needed for Cycle 5 Reload.
99
- 98. Hat erials Management Organization Changes v.acription of change The organizational change constitutes a newly established, onsite integrated Haterials, Purcharing & Contracts Departmunt which repotta to the Director Site Support. The function of the new organisation is to manage the purchase and control of materials and contracts at Waterford 3. The purchasing and materials control process currently in place will remain in place and equipment procurement controls are not affected by the organisation change.
Peason for Change Entergy Operations, Inc. conducted a study of the materials organizations at its nuclear stations and Corporate headquarters.
The study revealed the existing site organizational structurez sites in terms of scope and organizational were inconsistent approach to procurement among/ materials management. Material functions were fragmented at the sites with various functions performed in various organizations.
as a result of the study, a now corporate materials, purchasing and contract organization was established to provide plant support and take advantage of synergies that exist between the nulcear sites. Additionally, a new integrated materiaio, purchasing and contracts organization was established at each nuclear station for Entergy Operations. The now approach of a site consolidated / integrated material organization was adopted to provide improved material control.
Safety Evaluation The organitasion change had no impact on equipment function. None of the processes or requirements to purchase and control equipment were modified as a result of the organization change.
The change is organizational only and had no impact on accidents or malfunctions of equipment important to safety. Margin of plant safety was not affected. A Technical specification change was not needed for the Haterials Management Organization changes.
E 100 l
l 99, steam cenerator Tube plugging Description of Change The facility change left plug top remnants in six hot leg tu'oes of the Steam Generators (SG); three in SG #1 and three in SG #2.
Reason for Change Westinghouse (Nuclear Services Integration Division) had issued Nonconformance Report CWTR-91-001 to cover the acceptable method for SG Tube plugging with plug top remnants left in the tubes. Westinghouse also preparad a report, entitled, " Safety Evaluation to Support operation with steam Generator Hochanical plug Remnants in Steam Generator Tubes with Replacement plugs."
The Westinghouse report included a Safety Evaluation Check List SECL-90-612 which was prepared and approved on 11/30/90.
Safety r'raluation The function of the steam generators is to remove heat from the Reactor Coolant system. Leaving plug remnants in tubes with replacement plugs does not impact the ability of the SG to fanction as designed with tube plugr.
The plug top remnant does not form a pressure boundary since a good plug is insta11sd in a leaking steam generator tube. Since the remnant is behir d a good plug, it cannot increase the potential for opera'.lons with a loose part in the SG or thn Reactor Coolant System. The plug remnant be:tng lef t in a plugged 50 tubo does not lucrease the probability of occurrence or consequences of a steam Generator Tube Rupture accident previously evaluated.
The probability of occurrence or consequences of previously reevaluated malfunctions of equipment important to safety are not bar easod by SG tube plugging with plug remnants left in the tubes.
No new system itteractions or connections are created by this approach to SG cube plugging. plug remnants in plugged 50 tubes does not change a protective boundary and margins of plant safety are not impacted. A Technical Specification change was not needed for the Westif.ghouse Nonconformance Report, NR-CWTR-91-001.
101 I l
100. Firmware change Package TCP No. 91-01 Description of Chay e, Firmware Change Package TCP-91-01 revised firmware used in the Particulate, Iodine, and Gas (PIG) radiation monitors to allow the RH-80 to independently control the eartple flow on two flow paths.
Reason for Change The revised firmware was required to support the implementation of design changen, DC-3010 and DC-3033 whic.4 replaced the Barkdale flow switches with Kurs mass flow meters for PIG radiation monitors.
Safety Evaluation The PIG radiation monitors are used to monitor plant effluents and areas for radioactive particulates, iodines and noble gases. The firmware change enhances the ability of the PIG radiation monitors to perform their design functions. Aevising the firmware does not cause the plant to be operated in an abnormal manner. The PIG monitors have no effec; on the ability to safely shutdown the plant or to cause or mitigate the consequences of accidents.
No radiation alarm functions are affected by the firmware change.
The firmware change will decrease the prnbability of malfunction of radiation monitors by providing a note reliablo means of sample flow control.
No new system interactions or connections are cre6ted by the firmware change. No accident or equipment malfunction of a different type than previously evaluated is created by the firmware change. Margin of safety or protective boundary are not offected by the sample flow control changes. A Technical specification change was not needed for FCP-91-01.
102
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I. PROCEDURE
S A. PLANT PROCEDURE CHANGES 101. OP-001-002 (Change A - Revision 8), System Operating Procedure -
Reactor foolant Pump operation Description of change Change A to Revision 8 of OP-001-002 was a procedure dovietion to the system Operating Procedure. The change deleted the requirement of clearing high vibration alarms prior to starting a Reactor Coolant Pump (RCP). The procedure deviation was applicable in Operational Mode 5, Cold shutdown, and expired before entry into Mode 4, Hot Shutdown.
Reason for Change The vibration alarm setpoints on NCPn need to be set while the RCPs are running. The procedure deviation deleted the alarm clearing verification for starting a RCP.
Safety Evaluation The temporary procedure deviation did not affect the function of the RCPs but allowed the RCPs to be operated _without vibration detection for switch setting and reactor coolant system (RCS) loop flow while RCS temperature is less than 200 degrees F.
FSAR Subsection 5.4.1.5.5 merely mer.tions the RCP vibration detection system. The temporary change allowed the startup and operation of the RCPs so the setting and adjusting of the vibration switch setpoints can be done while the RCP is running.
The accident analyses involving the use of RCPs, in FSAR Chapter 15, have 100 percent full power conditions and are not applicable to the procedure deviation which was in affect in Operational Mode 5.
The RCP vibration detectors have no function other than to altrm when vibration threshold levels are reached. The procedure change did not increase the probability of occurrence or consequences of accidents or equipment malfunctions previously evaluated.
The possibility of an accident or equipment malfunction of a different type than previously evaluated was not created by the procedure deviation. Margin of plant safety was not impacted. A Technical Specification change was not needed for OP-001-002 (Change A - Revision B).
103 t
102. 0P-003-003 (Change 1 - Revision 9), System Operating Procedure -
Fondensate-FoeT ater Description of Change OP-003-003 provides instructions for the operation, startup and shutdown of Condensate, Feedwater, Feedwater Heaters, and Auxiliary Feedwater Systems. Change 1 to Revision 9 of OP-003-003 corrected the StanGay Valve Lineup to reflect the changes made by design change DC-3105. The procedural change also corrected nomenclature to agree with current valve nomenclature, and corrected condensate discharge valve throttle position required to satisfy pump start interlocks.
Reason for Change DC-3105 was a system modification that changed Feedwater Heater Drain Line control valves to f ail closed. The procedure change corrected standby valve lineups for the design changes and additionally corrected valve nomenclature.
Safety Evaluation The procedure change altered plant optration cue to the modifications made by a design change. ine change did not involve a test that required a system to be opirated in an abnormal manner.
The feedwater heater drains, normal and alternate, help ptotect the non-safety related main turbine from water induction and resulting failures. No new system internations or connections result from the procedural change. No FSAR analyaed accidents are initiated or mitigated by the heater drain system. A Technical Specification change was not needed for OP-103-003 (Change 1 -
Revision 9).
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103. OP-003-016 (Change ~ 2 - Revision 5), System Operating Procedure -
Instrument Air Description of Change OP-003-016 provides instructions for startup, normal operation and shutdown of the Instrument Air (IA) System. Change 2 to Revision 5 of OP-003-016 changed the standby valve lineup position for valves IA-5631 and IA-5632 from open to closed; and added three new instrument air valves to the standby valve lineup in Attachment 11.1.
R9ason for change The existing approved safety evaluations performed for the design changes DC-3195 and DC-3250 addressed the changes made by Change 2 to Revision 5 of OP-003-016. DC-3195 changed SI-602 A&D from air operated valves to motor operated valves and thus instrument air valves IA-5631 and IA-5632 no longer supply air to the air operators and the valves should be closed as is done by the procedural change. Also the three new instrument air valves added by DC-3250 were added to the Standby System Valve Lineup list by OP-003 016 (Change 2 - Revision 5).
Safety Evaluation Changes made to the procedure did not affect the function of the procedure as the components supplied by the valves involved do not require instrument air.
The probability of occurrence or consequences, of an accident at malfunction of equipment important to safety previously evaluated, are not increased by the procedural changes.
No new system interactions or connections were created and the procedure change did not cause the possibility of an accident or equipment malfunction of a different type than previously evaluated.
Margin of plant safety was not affected by the changeu. A Technical specification change was not needed for OP-003-016 (Change 2 - Revision 5).
105
104. OP-004-009__(Revision 4), System Operating Procedure - Incore Nuclear Instrumentatter.
Description of Change OP-004-009 provides instructions for normal operation of the Incore Nuclear Instrumentation System. Revision 4 of OP-004-009 removed the use of the Movable Incore Nuclear Instrumentation which was previously rendered not usable by the temporary alteration TAR-90-003. The TAR installed incore instrument pressure caps on all the guide tubes of the movable incore instrumentation.
Reason for Change The procedure revision was initiated per the biennial procedure review purauant to the Procedure Upgrade Program. The procedure was reformatted in Revision 4 in accordance with the Writer's Guide for Operating Procedures. The procedure revision deleted unused and unnecessary relegences to the inoperable Movable Incore Detector System.
Safety Evaluation A design change (DC-3237) was planned to remove the Movable Incore Detector System during the Refueling 4 Outage.
The procedural revision in OP-004-009 (Revision 4) changed the use of previously evaluated plant ?quipment but did not change the plant configuration. The use of movable incore detectors is not required for accidents as previously evaluated and no new accidents are postulated by elimination of the detectors.
A Technical Specification Change Request (NPT-38-111 - NRC TAC No.
79116) was submitted on 11/9/90 requesting the removal of movable incere detectors and deletion of requirements for overcurrent protection devices for containment penetrations associated with the movable incore detectors. NRC approved the Technical Specification change by issuance of Amendment No. 70 to Facility Operating License No. NPT-28.
106 1
105. OP-004-017 (Revision 4), . operating Procedure - Vibration and Loose Parts Monitoring Description of Change The purpose of OP-004-017 is to provide instructions for startup, shutdown, and actions for alarm conditions of the Vibration and Loose Parts Monitoring (VLPH) System to ensure compliance with Technical Specifications. The VLPH System is designed for continuous monitoring of anomalous conditions due to a vibration or a presence of loose parts in the Reactor Coolant System.
Revision 4 of OP-004-017 resulted from the biennial procedure review program. The revision made minor editorial improvements in accordance with procedure OP-100-013, Writer's Guide for operating Procedures. The major change of Revision 4 provided that the completed Noise Investigation Forms (Attachment 11.1 of the Procedure) be sent to Plant Engineering f or review and disposition in lieu of the Reactor Engineering and Performance (RE&P) organization as specified in FSAR Subsection 4.4.6.1. Plant Engineering to deemed to be more competently staf f ed than RE&P to analyze and interpret the vibration noise data. (NOTE: Changes to FSAR Subsection 4.4.6.1 are being made via LDCR-91-02?6 due to f acili',y enanges by Design Change, DC-3155. )
Reason for Change Revision 4 of OP-004-017 provided an enhanced pro.edure in that it was more complete and easier to use.
Safety Evaluation No plant equipment was affected by the procedure revision.
Revision 4 of OP-004-017 did not affect the probability of occurrence of previously evaluated accidents or malfunctions of equipment important to safety.
Since the procedure revision on'y added information and redirected monitoring data to the proper department, there were no new system interactions or connections created. The possibility or an accident or equipment malfunction of a different type than previously evaluated was not created by the procedure revision.
Margin of plant safety is not impacted by the chanjes made to the operating procedure. A Technical Specification change was not needed for OP-304-017 (Revision 4).
107
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106. OP-007-003 (change 7 - Revisiot4 8), operating Procedure - caseous Waste Management Description of Chance operating procedure OP-007-003 provides instructions for the operation of the caseous Waste Management System (CWHS). Change 7 to Revision 8 of OP-007-003 changed the method of adding nitrogen to a Gas Decay Tack and clarified steps for purging the Gas Surge Tank.
Reason for Change Change 7 to Revision 8 of OP-001-On3 was initiated in respones to an undesirable incident when the contents of a Gas Decay Tank had back-flowed through the Gas Decay Tank nitrogen regulator, which was failed open at the time due to a faulty in-line check valve, and contaminated the Domineralized Water Storaga Tank. The procedural change prevents the recurrence of the described incic*nt.
Safety Evaluation The procedure change did affect. a radioactive waste system, i.e., the caseous Waste Management System which is clasuified as non-seismic and non-nuclear safety related.
The procedural change involves the manipulation of components, and causation of parameter changes, directly related to the GWHS.
The procedure change is not a system modification as the changen only provided a different method of accomplishing an existing, and normal system function. The procedure change did not represent an alteration of plant equipment or a facility change. The procedure changeo continue to comply with the regulatory guidance f or design and operational criteria for radioactive waste systems.
The procedure change had no impact on the probability of occurrence or consequences of previously evaluated accidents or malfunctions of equipaent important to safety.
The procedure change did not create the pcasibility of an accident or equipment malfunction of a different type than previously evaluated. Margin of plant nafety is not affected by the procedure change. A Technical Specification change was not needed for OP-007-003 (Change 7 - Revision 8).
108
107. OP-901-046 (Revision 7), Off-Normal Operating Procedure - Shutdown Cooling Halfunction Description of Change OP-901-046 provides Operator actions to mitigace and restore the plant from the postulated Shutdown Cooling (SDC) Malfunction.
Reviston 7 to OP-901-046 made numerous changes for procedure enhancsments to more closely follow tht esquence of events following a SDC malfunction.
Reason for Ch&nge The procedure revision considered the commite nts made in response to the NRC Information Notice IN-88-36. The major change in Revision 7 for safety evaluation is the use of LPF.~ pumps and shutdo n Coolinu Heat Exchangers (SDCHX) for long term cooling of the Reactor Coolant System (RCS) following a SDC malfunction.
Safety Evaluation All connections for LPSI pumps to desw from CIS were already installed per the plant design. The procedure revision does not alter the plant equipment. It doss change the operational method of the 11. stalled equipment because it allows the use of LPSI pumps and SDCHX as a cooling medium for the RCS following a SDC raa'iunction
. when no other method exist.s. Also the procedural steps do not deviate from equipment design for test purpsnes.
The procedere revision did not alter any squipment or facility components. The procedurul steps iv. OP-901-046 are for the mitigation of 4 SDC malfunction and do not deviate from the equipment design at dascrioed in the FSAR.
OP-901-046 (Revision 7) does not increase the probability of oceverence or consequences of a loss of heat capability during
' shutdown cooling as previously evaluated.
The procedur0 revision does 9t create the possibility of an acc.ident or equipment malfu - ion of a different type than 3 previously evaluated. No ne - stem interactions or connections resulted from the p'rocedure t
- non.
Margin of plant safety La not n . god by the procedure revision.
A Technical Specification change was not needed for OP-901-046 (Revision 7).
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108. OP-903-030 (Change 2 - Revisitn 71, Surveillance Procedure -
Safety InIection Pump Operability Verification, DescriF.aon of Chanqe surveillance Procedure OP-903-030 provides the procedural steps to verify operability of the Safety Injection Pumps. Change 2 to Revision 7 of OP-903-030 added the requirement for recording High Pressure Safety Injection (HPSI) Pump recirculation flow fullowing the throttling of the Recirculation Stop Check Valve in the procedure sections covering all three HPSI pumps. Change 2 also added recirculation flow acceptance criteria for HPSI pumps to the procedure attachments.
Reason for Change The procedure change was initiated to ensure pump data is obtained as required by the ASME Boiler & Pressure Vessel Code,Section XI, IWP Table 3100-1.
Safety Evaluation The procedura change did not cause any alteration to the facility configuration or plant operation. Acceptance criteria was changed for inservice testing of tne HPSI pumps to improve ability to detect pump degradation and likely should increase pump reliability.
Change 2 to Revision 7 of OF-903-030 did not increase the probability of occurrence or conseq2ences, of accidents or malfunctions of equipment impurtant to safety previously evaluated. The added acceptance criteria for pump testing will likely reduce the probs.bility of a HPSI pump malfunction but has no impact on the consequences of a pump malfunction.
No new sistem interactions or connee' ions are created by the procedure change and the-possibility of an accident or equipment malfunction of a different type than previously evaluated is not created.
Margin of safety is not reduced by the procedure change. A Technical Specification change was not needed for OP-903-030 (Change 2 - Revision 7).
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109. OP-903-032_,1 change 5 - Revision 7), Surveillance Procedure -
Quarterly IST Valve Testa Description of Change, Change 5 to Revision 7 of OP-903-022 added in procedure Section 7.4 reference to Shutdown Cooling in service. The change allows Containment Spray B Train to remain operable during its valve testing.
Rpason for Change The reference to shutdown cooling being in service was inadvertently omitted in previous procedure changen. The procctare change provided consistency in valve testing of the A and a trains of the Containment Spray System. The change removed th6 requirement to place Containment Spray B Train in an innperable condition during the test.
Safety L' valuation Section 7.4 of OP-903-032 tests the containment Sprhy valves.
The function of the Containment Epray System is to remove energy and radioactivity from the containment atmosphere during a Loss-of-Coolant Accident (LOCA) or a Main Steam Line Break (MSLB).
The change lessens the consequences of a failure of Containment Spray Train A should an accident occur during a test of the Centainment Spray B Train. ,
The procedure change did not alter the plant design or configuration and did not require system operation in an abnormal manner. The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased by the procedure change.
No new system interactions or connections are created by the procedure change. The possibility of an accident or equipment malfunction of a different type than previously evaluated is not created. Margin of safety is not impacted, and a Technical Specification change was not needed for OP-903-032 (Change 5 -
Revision 7).
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h 110. -OP-903-032 (Revision 8), survoillance Procedure - Quarterly IST Valve Tests-Description of Change Revision 8 of OP-903-032 made the following procedure changes:
- 1. Deleted test of Instrument Air Check Valves from procedure.
- 2. In Att. 10.1 for section 7.3 changed stroke times for SI-602A&B from 28 to 34 wec.
- 3. Added step in section 7.7 to reset Fuel Pool Temperature Control Valve (TCV), CC-620.
- 4. In Att. 10.1 for section 7.18 changed Acceptance Criteria for Nitrogen Accumulators.
- 5. In Att. 10.1 for eection 7.20 changed stroke time acceptance criteria for Emergency Feedwater (EFW) valves from 25 to 24 sec.
- 6. Added steps to section 7.5 to lock Emerogency Diesel Fuel (EGF) valves in their proper positions.
Reason for Change
- 1. DCP 3195 changes SI-602ALB from air-operated to motor-operated valves. This section is no longer required to be performed.
- 2. SI-6d2A&B are changed per DCP 3195 from air-operated to motor-operated valves and require different stroke times, j
- 3. Steps were needed in procedure to reset Fuel Pool TCV. ,
- 4. Acceptance Criteria for Nitrogen Accumulators changed to agree with Design Basis Document W3-DBD-014.
- 5. l'FW valves stroke time limit changed to ensure total time rcaponse limit is within Tech Spec limits of T.S. 4.3.2.3.
- 6. St;ps were needed to lock EGF valves in proper positions sicco they are locked valves per OP-100-009.
Safety Evaluation OP-903-032 provides instructions for the performance of quarterly inservice testing of specific valves listed in the Pump and Valve Inservice Test (P&V IST) Plan.
The pertinent procedure changes in OP-903-032 (Revision 8) were evaluated on the basis of the existing approved safety evaluation prepared for Design Change DC-3195 which changed valves SI-602A&B from air-operated to motor-operated valves.
Valves SI-602A&B are used only to mitigate the consequence of a LOCA. SI-602A&B cannot initiate or cause any accident to occur.
Changing the actuators does not increase the probability of occurrence of an accident.
Revision 8 to OP-903-032 did not increase the probability of occurrence or consequences, of accidents or eq,sipment malfunctions previously evaluated.
No new system interactions or connections were resultant from OP-903-032 (Revision 8) and the possibility of an accident or equipment malfunction of a different type than previously evaluated is not increased. Margin of safety is not reduced.
A Technical Specification change was not needed for OP-903-032 (Revision B).
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111.. OP-903-033 (Change 6 _ Revision 8), Surveillance Procedure - Cold Shutdown IST Valve Testt Description of Change change-6 to Revision 8 of OP-903-033 changed the Reactor coolant System Loop Shutdown Cooling Isolation Valves, SI-405A and SI-405B, stroke time limit from 10 seconds to 34 seconds.
Reason for Change change 6 to Revision 8 of OP-903-033 was initiated for procedure compliance to the design cha"ge implemented by DC-3286 which replaced the existing Borg Warner actuators to Paul Munroe actuators for isolation valves SI-405 AGB. The new actuators stroke time in the closed direction is approximately 25 seconds which required a stroke time limit change to an acceptable 34 seconds from a previous maximum of 10 seconds.
Safety Evaluation The procedure change was addressed by the exioting approved 10CFR50.59 safety evaluation prepared for Design Change DC-3286 (see this Report Item No. 35). The actuators for valves SI-405A and SI-405B were replaced by DC-3286 and had increased stroke time.
The procedure change had no impact on equipment functions. The replacement valve actuators fulfill the functions and meet the design criteria of the previous actuators. The replacement actuators operate in the same basic way as the original actuators and testing of the valves does not affect the valve functions.
Testing of valves per the procedure change does not increase the probability of occurrence or consequences of accidents or equipment malfunctions as previously evaluated. There are no new system interactions or connections created by the procedure change. A Technical Specification change was not needed for OP-903-033 (Change 6 - Revision 8),
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112. OP-903-033 [ Change B - Revision 8), Surveillance Proceduro - Cold Shutdown IST Valve Testt (Deviation)
Description of Change Change B to Revision 8 for OP-903-033 is a procedural deviation which added procedure steps to allow testing of tce Steam Generator il and #2 Feedwater Upstream Check Valves FW-181 A & B to meet the ASME Section XI requirements. The expiration date for the deviation was specified as the completion of Refuel 4. The procedural change provides a new test method for verifying that check valves FW-181 A & B can be exercised to their closed positions.
Reason for Change OP-903-033 provides instructions for performing cold shutdown testing of specific valves listed in the Pump and Valve Inservice Test (P&V IST) Plan.
The procedure change was made to satisfy the requirements of ASME Boiler & Pressure Vessel Code Section XI for testing check valves.
Safety Evaluation The procedure deviation did not alter system operation or function. The procedure testing is performed at cold shutdown intervals and has no affect on the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated.
The temporary connections used to test the Feedwater Check Valvos did not create the possibility of an accident or equipmn..t malfunction of a different type than any previously evaluated.
Margin of safety is unchanged by the procedure deviation. A Technical Specification change was not needed for OP-903-033 (Change B - Revision 8).
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l 113. OP-903-112 (Change 3 - Revision 2), Surveillance Procedure -
Containment Purge Valva Leak Test Description of Change Change 3 to Revision 2 of OP-903-ll2 revised procedure Section 7.4 to preclude overpressurization of test volumes during the initial pressurization.
Reason for Change Surveillance Procedure providou instructions to perform the Local Leak Rate Test (LLRT) of the Containment Atmosphere Purgo Penetrations. Change 3 to Revision 2 of OP-903-ll2 is a procedure enhancement to preclude an overpressurization condition.
Safety Evaluation The procedure change did not alter the intent of the procedure for leak rate testing. No iscility equipment chan es or plant configurations ara invoiced in the procedure change.
The procedure change did not increase the probability of occurrence or consequencoe of an accident or malfunction of equipment important to safety previoucly evaluated.
No new system interactions or connections are made by the procedure change and the possibility of an accident or equipment malfunction of a different type than any previoucly evaluated is not created.
Margin of safety is i ,c reduced by the procedure change. A Technical Specif. cation was not needed for OP-903-112 (Change 3 -
Revision 2).
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114. OP-903-114 (Revision 3), Surveillance Procedure - Lncal
~
teak Rate Test (LLRT)
Deecription of Change Revision 3 of OP-903-114 completely reformatted the procedure and added previously approved changes which stem from the LLRT performed upon completion of Refuel 3 activities.
Reason for Change The procedure revision resulted from the biennial procedure review pursuant to the Proceduree Upgrade Program. The procedure was reformatted in accordance with procedure OP-100-013, Writer's Guide for Operating Procedures and was enhanced by the incorporation of previously approved changes.
Safety Evaluation Procodure OP-903-114 provides instructions for a) performing and documenting the LLRT of containment isolation boundaries and reworked isolation valves, b) satisfying stroke requirements of valves per the Pump and Valve Insetvice Test Plan, c) visually inspecting containment isolation valven for signa of degradation, and d) meeting the pertinent Technical Specifications requirements.
The procedure revision did not physically alter the facility configuration or operation of the plant. The changes made in the procedure revision affected details but did not alter the LLRT requirements ao described in FSAR Section 6.2.6.3.
The probability of occurrence or consequences, of an accident or malfunction of equipment important to safety, previously evaluated are not increased by the procedure revision. Revision 3 of OP-903-114 did not create the possibility of an accident or equipment malfunction of a different type than previously evaluated.
There were no new system interactions or connections created by the procedure revision. Margin of safety is not impacted. A Technical Specification change was not needed for OP-903-114 (Revision 3).
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115. OP-904-005 (Revision 7), Surveillance Procedure - Sprinkler and Spray Systems Alarm Test Description of Change Revision 7 of OP-904-005 provided clarification for the conduct of Drain Tests on the system designators FPMs (Fire Protection Mechanicals) in the Radiation Controlled Areas (RCAs). The revision also exempted systems in the RCAs from the quarterly tests, provided the systems were flow tested following each valve repositioning (i.e. , reset, drain down, valve cycling, etc).
Reason for Chance Difficulties were experienced in conducting quarterly rise flow tests of sprinkler systems within the RCAs. Specific difficulties included waste water generation and floor drain overflove. The testing exemptions were concurred with by the Nuclear Insurance Underwriter.
Safety Evaluation OP-904-005 provides instructions for performing alarm and drain tests on spray and/or sprinkler systems every 92 days and following system restoration after maintenance or valve operation in a system flow path ao required by Insurance Underwriter.
The procedure revision did not modify the installed fire protection system and did not affect the function of the plant equipment. The revision claritied the testing requirements and did not include tests that require systems to be operated in an abnormal manner. Previously evaluated accidents and equipment malfunctions are not impacted by the changes made in the surveillance procedure tents.
OP-904-005 (Revision 7) providas that the water drained from the system by the conduct of Drain Tests is an influent to the Liquid Waste Management System (LWMS). The LWMS is not modified in configuration or altered in operations by the procedure revision.
The design provisions and commitments to the applicablo regulatory guides for radioactive waste syotems are not affected by the procedure revision.
Margin of plant safety is impacted by the procedure revision. A l Technical Specification change was not required for UP~904-005 (Revision 7).
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l 116. UNT-005-013 (Revision 2), Administrative Procedure - Fire Protection Program Dencription of Change Revision 2 of UNT-005-013 provided corrections and additional changes for human factor considerations and organizational structure updates. The revision also provided editorial changes to reflect current definitions in the Technical Specifications.
Reason for Change The procedure revision was initiated to correct an erroneous entry p:eviously made regarding the Fuel Handling Building Fire Zone FHB-4. New position titles were given as a result of current corporate reorganizations at Entergy Operations, Inc.
Safety Evaluation UNT-005-013 describes, delineates responsibilities, control. and implementing requirements for the Fire Protection Program. It gives administrative and equipment operability guidance for the Fire Protection System components.
The level of fire protection is maintained in the procedure revision consistent to that previously approved and identified in the FSAR Section 9.5.1.
No new or additional tests are introduced by the procedure revision. The probability of occurrence or consequences of a fire accident or fire equipment malfunction previously evaluated are not increased by UNT-005-013 (Revision 2).
The procedure revision did not cause any new system interactions or connections. The possibility of an accident or equipment malfunction important to safety is not created by the procedure revision.
No safety margin or protective boundary is affected by the procedure revision. A Technical Specification change was not needed for UNT-005-013 (Revision 2).
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117. HP-001-210 (Revision 7), Administrative Procedure - Health Physics l Instrument control Description of Change HP-001-210 provides instructions for the control of Health Physics instruments and calibration c'evices. Revision 7 was initiated primarily for procedural updating due to the newly procured PM-7 portal monitors to replace the IRT portal monitors currently used at the Primary Access Point (PAP).
The procedure revision also corructed a previous error and made editorial changes.
Reason,for Change The replacement of monitors was prompted by the unavailability of parts for the existing portal monitors. Additionally, the procedure revision corrected a previous error relative to air samplers which cannot be source checked; and to provide editorial changes for procedure enhancement and clearer instructions.
Safety Evaluation The procedure revision does not involve any plant system for operating the plant. The procedure does not affect any FSAR design basis accident sa the PAP portal mo?!. tors are not eafety related and ttuy do not interface or connsc: to a safety related system or component. TP) PM-7 portal monit ars have better sensitivity than the existing mcnitors and their use will provide an improvement in the detection of personnel contamin fon at the PAP.
Margin of plant safety is not redeced by f.he procedure remision.
A Technical Specification change was not needed for HP-005-210 (Revision 7).
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118. RF-002-001 (Revision 3), Refuelina Procedure - Fuel Receipt Description of Change The procedure revision changed the storage location of now fuel, and reformatted the procedure.
RF-002-001 provides instructions for a) unloading, dismantling and reassembling New Fuel Assembly Shipping containers, b) inspection and removal of new fuel assemblies from shipping containers, c) for handling and storage of new fuel, and d) pre-operational checks o' equipment.
Reason for Chance RF-002-001 (Revision 3) directed the plar9 ment of new fuel into the Spent Fuel Pool, not the new fuel sturage racks, to minimize fuel handling. The procedure revision, pursuant to the Procedures Upgrade Program, reformatted the procedure in accordance with OP-100-013, Writer's Guide for Operating Procedures.
Safety Evaluation The procedure revision did not change the function of the procedure or plant equipment. There were no facility changes made by the procedure revision which provides instructions for the operation of equipment already installed. The storage of now fuel (up to 4.1 weight percent U-235) in the spent fuel storage racks is acceptable por the FSAR Section 9.1.2.3. Fuel receipt does not require special testing or system operation in any abnormal manner.
RF-002-001 (Revision 3) did not increase the probability of occurrence or consequences, of the design basis fuel handling accident or malfunction at fuel handling equipment, previously evaluated. The possibility of an accident or equipment malfunction of a different type than previously evaluated was not created by the procedure revision.
No protective boundary is impacted and margin of safety is not reduced by the procedure revision. A Technical Specification change was not needed for RF-002-001 (Revision 3).
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119. Specifying Local Manual Action to Close valves CS-117 A&B to Initiate Shutdown Cooling Description of Change Plant Operational action was taken that deviated from the license basis in the FSAR. Local manual action outsido of the main control room was required to close containment Spray System valves CS-117A and/or CS-117B to isolate the containment spray headers and initiate shutdown cooling.
Reason for Change A system configuration existed different from that described in the plant license basis. Specifically, alignment of the Shutdown Cooling System to the shutdown cooling mode could not be accomplished solely by means of remote alignment of valves from the main control room as indicated in the FSAR in *ubsection 9.3.6.3.2.
Although it was intended to comply with the regulatory guidance, that residual heat removal systems bc capable of being operated from the control room, valves CS-117 A&B appear to have been overlooked it. the design / construction process.
To Anttiate shutdown cooling, the contairtent spray flow path through the shutdown cooling heat exchangers (SDCHXs) must be isolated by closing valves CS-ll7A and CS-il7B. Otherwise, the cooling flow from the SDCHXe would be diverted from the RCS to the cantainment spray headerb.
Safety Evaluation Upon discovery of the nonconformance plant configuration, Potential Reportable Event, PRE-91-035 was initiated and the condition was determined to be not reportable. Nevertheless, Entergy Operations, Inc. informed the NRC, documented the finding and provided a summary of the safety evaluation to the NRC via letter W3F1-91-0424 dated 6/18/91.
The 10CFR50.59 safety evaluation determined the impact of specifying local (vice remote) manual operation for CS-ll7A and CS-1178. The evsluation concluded that local manual action to close the two valves in order to initiate shutdown cooling is technically acceptable. Branch Technical Position RSB 5-1 permits limited manual actions outside of the control room to initiate SDC provided the actions can be suitably justified; this is the case for CS-ll7A and CS-117B.
The technical concern for local manual operation of CS-Il7A and CS-117B is the post-Loch radiation levels at the valve operators.
A detat ed evaluation of the post-LOCA radiation levels at the valve operatots was performed, using the standard post-LOCA source terms from the FSAR and a datailed modeling of piping geometries.
A 10 minute stay time was postulated at six hours into the event, consistent with previous analyses. The calculated accumulated doses were 3.84R and 2.68R for operation of CS-ll7A and CS-ll1B, respectively. These doses are acceptably less than the SR limit of NUREG-0737 and GDC-19.
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(continued) 119. Specifying Local Manual Action to close Valves CS-ll7 A&h to~
Initiate Shutdown cooling Actual dose levels which would exist if these valves are operated after a LOCA are expected to be lower than the calculated values.
Operation of these valves is only required after a small-break LOCA, for which core damage and fission product releases are expected to be smaller than for the design basis large break LOCA, on which the source terms are based.
Since manual post 50c% operation of valves CS-ll7A and CS-ll7B has been shown to De acceptable from the standpoint of accumulated radiation dose, ne physical changes to the plant are currently envisioned.
The FSAR will be updated to reflect the requirement that valves CS-ll7A and CS-117B be operated manually to initiate shutdown +
cooling.
The operational change for alignment of valves for shutdown cooling entry did not increase the probability of occurrence or consequences of an accident or equipment malfunction previously evaluated.
The operational change did not result in any system interactions or connections which did not previously exist. The possibility of an accident or equipment malfunction of a different type than previously evaluated is not created by the change in plant operations from that described in the FSAR. A Technical Specification change was not needed for local manual closure of valves CS-ll7 A&B to initiate shutdown cooling.
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II. B. SPECIAL TEST PROCEDURES (STPs)
.120. STP-027069-1, Controlled Ventilation Area System (CVAS) Flow and Differential Pressure Test Description of Change The purpose of Special-Test Procedure STP-027069-1 was to ensure Control Ventilation Area System (CVAS) filtration unit differential pressure setpoints sustain CVAS air flow rate at 3000 cfm +10% when operating with the normal Reactor Auxiliary Building Supply / Exhaust System secured. Major Change 1 to STP-027069 added instructions for adjusting CVAS makeup damper open/ closed positions.
Reason for Change Doors to areas served by CVAS were difficult to open when the system was in operation.
The plant configuration problem was initially identified, under Condition Identification / Work Authorization CIWA-027069, as CVAS negative pressures being too great to permit door opening with reasonable effort. Special Test Procedure STP-027069, CVAS Negative Pressure Control, was conducted and proved the orifice plates on makeup dampers were not needed to control pressures.
Temporsry Alteration Request, TAR-86-091 temporarily removed the orifica plate on the makeup dampers to allow proper airflow to CVAS filter train when' negative pressures in CVAS are achieved.
Subsequently, Station Modification, SM-1706 closed out the.
temporary alteration and made permanent the deletion of the makeup dampers arifice plates. Major Change 1 to STP-027069-1, CVAS Flow Differential Pressure Test, was initiated to eliminate variations in differential pressure which had been experienced and were outside the routine operating bands for system operation.
Safety Evoluation The special test procedure change permitted the appropriateness of the plant alterations to be verified. The STP change did not increase the probability of occurrence or consequences, of an accident or malfunction of equipment important to safety previcusly avaluated.
The procedure changes did not create the possibility of an accident or equipment malfunction of a different type than previously evaluated. No new system interactions or connections were made by the procedure change. Margin of plant safety was not reduced, and a Technical Specification change was not needed for STP-027069-1.
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i 121. STP-255644-C (Change 1), Testing of Ambient and Service Temperatures Inside Contai.nment Description of Change The facility change to the configuration in containment involved the placement of resistance temperature detectors (RTDs) on and around selected equipment to monitor ambient and service temperatures during power operation from Refuel 4 to Refuel 5.
The intent of this RTD placement is to address the concerns of NRC Information Notices 87-65 and 89-30 pertaining to potential containment temperature deviations in excess of the 120 degree F Design Basis Temperature identified in Table 3.11-1 of the FSAR.
The re-calibration as required and the relocation of the RTDs to allow tmplement ation of the special test procedure and to measure the ambient and service enmperatures inside containment during power operation from Refuel 4 to Refuel 5 was done under Work Authorization WA #01070880.
Reason for Change The purpose of STP-255644-C is to address the concerns of NRC Information Notices 87-65 (Plant Operations Beyond Analyzed conditions), and 89-30 (Figh Temperature Environraents of Nuclear Power Plants). The intent of the special testing in to develop baseline data to support qualification analysis for various plant equipment within the Environmental Qualification Program.
Parameters being measured by STP-255644-C are ambient and contact temperatures at various locations inside containment.
Change 1 to STP-255644-C was initiated to delete the temperature monitoring computer point #A42117 which was intended to measure Steam Generator #1 cold leg ambient temperature. The Resistance Temperature Detector (RTD) extension cable for this monitoring location was inadvertently removed by Plant Personnel during the Refueling 4 activAties after the preliminary installation and the missing extension cable could not be located. Deletion of this monitoring location did not violate any specific regulatory commitments but it was deemed a major change to the STP.
Safety Evaluation The relocation of RTDs in containment had potential to altcr system operation and required in-depth safety evaluation. The safety evaluation considered the RTD placement for Refuel 4 to Refuel 5 to perform the containment Temperature Monitoring Program.
The RTD locations provide for temperature monitoring only and do not impact the intended functions of the monitored equipment. The changes via the work authorization relocations of RTDs and the special test procedure do not result in any functional alteration of plant equipment or electrical systems. The special test procedure did not require equipment or plant system tc be operated in an abnormal menner.
(continued) 124 l
(continued) 121. STP-255644-C (Chango 1), Testing of Ambient and Service Temperatures Inside Conthinment The primary coolant system pressure boundary or any other boundary is not breached in any fashion to perform this monitoring function. The RTDr and the circuits are 1.ot electrically or physically interlocked with any other existing plant electrical circuits. The change does not result in any 5.retional alteration of any plant equipment or electrical system.
The probability of occurrence or consequences of an accident or equipment malfunction previously evaluated are not increased by the opoeial test procedure changes. The possibility of an accident or rquipment s malfunction of a different type than previously evaluated is not created by STP-255644-C (Change 1).
The change activity does not reduce margin of safety. A Technical Specification change was not needed for tha changes in the Containment Temperature Monitoring Program.
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l 122. STP-01062461, Chemical and Volun'.c
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control (CVC) Letdown Radiation Monitor Flow Test Description of Change i STP-0106:461 was a pre-implementation special test for Design Change, DC-3299, to determine if the CVC letdown process radiation monitor can obtain sufficient flow with the process sample auction pump deleted. The STP was also to determine, with the radiation monitor in service, if the Cyc letdown system can be operated at maximum flow rate without backpressure control valve problems or relief valves Itfting.
Reason for Change The auction pump on the CVC Letdown Radiation Monitor failed and allowed water to leak through the motor to damage part of the electrorlew associated with the radiation monitor skid. An engineering' evaluation was conducted and it recommended the euction pump be. removed and replaced with stainless steel tubing via the design change process. The special test procedure was implemented to verify for che design change if sufficient flow is achieved with the suction pump deleted.
Safety Evaluation L The CVC Letdown Radiation Monitor serves a monitoring function only and it has no safety function except to alert operators of a possible fuel rod failure. The process radiation monitor (PRM) in letdown system monitors the reactor coolant system for gross gamma activity and alarms at a pre-set level. The special test procedure did not require the CVCS or Letdown PRM to be operated in any abnormal manner.
STP-01062461 did not increase the probability of occurrence or consequences of an accident or equipment malfunction previously evaluated. The procedure change only affected the CVC Letdown System PRM which is not important to safety.
No new system interactions or connections were created by the special test. There were no changes made to a protective boundary and margin of safety was not reduced. A Technical Specification Mhange was not-needed for STP-OlC62461.
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123. STP-01074040, Test to Establish Shutdown cooling Flow Limi;tation Description of Change STP-01074040 was implemented to identify the positions of LPSI Header to Reactor Coolant Loop Flow Control Valves SI-138 AEB and SI-139 A&B required to limit flow in each shutdown cooling train to 4000 to 4100 GPM with the LPSI pump discharge header control valves SI-129 A&B and shutdown Heat Exchanger outlet Header Flow control Valves SI-415 A&B fully opened.
Reason for change The special test procedure was to obtain information to assist operations in responding to reduced inventory conditions and possible off-normal plant conditions ir. shutdown cooling operational modes. Based on the special test results, operations developed a new procedure to cover Shutdown Cooling operations for reduced inventory in the reactor coolant system.
Safety Evaluation STP-01074040 is performed only in operational Modes 4, 5 and 6 and requires Shift Superrisor Approval to perform the test. The special test procedure operates within Technical Specification limits and system design criteria. The special test procedure affected only the Shutdown Cooling System and reactor decay heat removal.
STP-01074040 made no changes to the function of plant equipment.
The information gained from the special test was used to reduce the possibility of a loss of shutdown cooling.
The probability of occurrence or conrequences of an accident or equipment malfunction previously evaluated was not lucreased by the special testing. STP-01074040 did not create the possibility of an accident or equipment malfunction of a different type than previously evaluated. The special test operated equipment within design limits and did not impact margin of safety. A Technical Specification change was not needed for STP-01074040.
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124. STP-99000404, Installation Testing of Reactor Coolant Shutdown ~
Level Measurement System (hCSLMS)
Description of Change The work under CI #274033/WA #9900404 was to add Process Analog control (PAC) Cards, wiring and perform transmitter calibration of the new Reactor Coolant Shutdown Level Loop installed by Design change, DC-3162. The Design Change, DC-3162, was implemented to install a second, independent, continuous, redundant Reactor Coolant Shutdown Level Measurement System (RCSLMS) to fulfill a commitment made in response to NRC Ceneric Letter No. 88-17.
Document Revision Notice, DRN #I-9100825 incorporated the vendnr manual for the level monitoring system into the design change package, DCP-3162. The vendor manual is: Fluid Components Inc.
(FCI) Document No. 703185 (Rev. A), entitled, Installation and Operation Manual for FCI's CL86 Level Monitoring System. The manual explaina the operating principles of FCI's CL86 Level Monitoring System and it contains information on installation (and testing), operation, and maintenance of the device.
Reason for Change The facility change was made in response to the NRC Generic Letter No. 88-17. STP-99000404 provided for the installation acceptance testing (including level transmitter calibration) for the vendor supplied level monitoring system which was installed via the design change process as the RCSLMS.
Safety Evaluation The DRN #I-91000825 incorporated the vendor-technical manual into the design change package. The DRN referenced the existing ,
approved safety evaluation for DC-3162 which addressed the installation and acceptance testing of the level monitoring system.
STp-99000404 did not affect the function of existing equipment and did not require a system to be operated in an abnormal manner.
The RCSLMS is a non-safety related subsystem. The installation testing did not increase the probability of occurrence or consequences of an-accident or equipment malfunction previously evaluated. The special testing did not create any new system interactions, connections or modes of operation. The possibility of an accident or equipment malfunction of a different type than previously evaluated was not wreated by STP-99000404. Margin of safety was not impacted. A Technical Specification change was not
-needed for STP-99000404.
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125. STP-99003147, Emergency Diesel Generators A and B Start and
-Governor Control Description of Change The Special Test Procedure provides instructions for acceptance testing of the criteria for the design changes implemented by DC-3147-for replacement of relays and engine control logic changes for Emergency Diesel Generators A and B.
The acceptance testing was inplemented in operational Modes 5 and 6 (i.e., Cold 6hutdown and Refueling) during Refueling 4 Outage.
Reason for Change STP-99003147 was implemented for acceptance testing of the Design Change (see_this Report Item #18).
DC-3147 was considered a design enhancement. The reliability of the EDO control scheme is enhanced and surveillance testing of the EDC can be performed using the one start concept which reduces EDG engine start stress.
Safety Evaluation The acceptance test did not change the function of the EDO and did not compromise the safety of thf> EDCs or the licensing basis of operation. FSAR Figure 8.3-1 is affected by the DC-3147 as the operational methods of the EDGs were altered.
Thv Emergency Diesel Cenerators are the standby power supplies for the nnsite power distribution system. The design change was ir 'ed tr enhance control performance and provide redundancy feacurwa. The EDG syctem is classified safety class 3 and Seismic Category I.
The acceptance testing did not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated.
STP-99003147 did not creatu the possibility of accident'or equipment malfunction of a different type than previously evaluated.
Margin of safetyLis not impacted by the acceptance testing of the design changes. A Technical Specification change was not needed for STP-99003147.
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