ML20125D546
| ML20125D546 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/18/1992 |
| From: | Burski R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3F292-0033, W3F292-33, NUDOCS 9212150264 | |
| Download: ML20125D546 (103) | |
Text
-~ ENTERGY IEIIS """ "*'
K b n LA 70006 Td 504 739 6774 R. F. D urski s , :-t If t_k-f ( ,
WmA c; 3 W3F292-0033 A4.05 QA December 10, 1992 U. S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D. C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 1992 Report of Facility Changes, Tests and Experiments Enclosed is the 1992 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, 1991 through June 18, 1992.
If you have any questions regarding this report, please contact E. D. Linnartz, at (504) 739-6734.
Very truly yours,
, 4W-RFB/EDL/el Enclosure cc: J. L. Milhoan,.NRC Region'IV D. L. Wigginton, NRC-NRR R..B. McGehee N. S. Reynolds NRC Resident Inspectors Office 1
350010
./' I 9212150264 920618 L PDR ADOCK 05000382 L i R .PDR I
Entergy Operations, Inc.
Waterford 3 SES Docket No. 50-382 License No. NPF-38 REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS FOR 1992 PER 10CFR50.59
r ENTERGY OPERATIONS, INC.
WATERFORD 3 10CFR50.59 ANNUAL REPORT FOR 1992 TABLE OF CONTENTS
- 1. FACILITY CHANGES Report Page Number Document / Title No.
SUMMARY
............................................ 1 A. DESIGN CHANGES (DCs)
- 1. DC-3008, Human Factors Enhancement to LCP-42A, 42AX, 42B, 44, 45, 46 ..................... 2 ,
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- 2. DC-3010, Addition of Mass Flow Probes and a Moisturo' Control Unit to the Containment Atmosphere PIG Monitor (PRM-IRE-0100Y) .... 3
- 3. DC-3065, Parish Water to Fire Water Storage Tank (FWST) Fill Connection .................... 4
- 4. ' DC-3073, Reactor Containment Building (RCB) Air Conditioning .............................. 5
- 5. DC-3103, Control Element Drive Mechanism (CEDM)
Cabinet Cooling ........................... 6
- 6. DC-3145, Essential Services Chillers Auxiliary Oil Recovery System ....................... 7
- 7. DC-3162, Reactor Coolant Shutdown Level Measurement System (RCSLMS) ........................... 8
- 8. DC-3175, Essential Services Chillers Dehydrator Addition .................................. 9 i
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TABLE OF CONTENTS (continued)
- 1. FACILITY CHANGES Report Page Number Document / Title No.
A. DESIGN CHANGES (DCs) (continued)
- 9. DC-3176, Essential Services Chiller Hot Gas Bypass Valve Upgrade ............................. 10
- 10. DC-3183, Replacement of Instrument Air Dryer After-Filter and Installation of Instruments ............................... 11
- 11. DC-3198, Relocate ILRT Air Supply .................. 12
- 12. DC-3211, Fuel Handling Building +46 Spent Fuel Pool Valve Gallery Deck Plates ................. 13
- 13. DC-3242, Relocation of RM-80 Power Switch and 120 VAC Outlet on Fuel Handling Building ISO Rad Monitor ................................... 14
- 14. DC-3250, Condensate Polisher Vessel Pre-Service Rinse ..................................... 15
- 15. DC-3257, Quench Tank Level Indication Upgrade ...... 16
- 16. DC-3267, Essential Chillers Oil Cooler Water Valve (OCWV) Actuator Wiring Changes ...... 17
- 17. DC-3298, Refueling Water Level Indication System Range Extension and Indicator Cut-Out............ 18
- 18. DC-3299, Deletion of Pump From Chemical and Volume Control Letdown Radiation Monitor (PRM-IRE-0202)............................ 19 ii l
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m-TABLE OF CONTENTS (continued)
- 1. FACILITY CHANGES Report _ Page Number Document / Title No.
A. DESIGN CHANGES (DCs) (continued)
- 19. DC-3301, RTD Cabl e Temi nati ons . . . . . . . . . . . . . . . . . . . . 20
- 20. DC-3341, Essential Chiller Shunt Resistor Replacement ............................... 21
- 21. DC-3345, Plant Monitoring Computer Storage Disk Drive Module Replacement . . . . . . . . . . . . . 22 B. CONDITIONIDENTIFICATIONS/WORKAUTHORIZATIONS(CI/WAs)
- 22. CI 277626, Repair of Dry Cooling Tower (DCT) 14B Gearbox ................................... 23
- 23. CI 278618/WA 99003176, Acceptance Test for DC-3176, Hot Gas Bypass Upgrade ........... 24
- 24. CI 278676, Deletion cf Supplementary Chiller Motor Winding Temperature Computer Points ....... 25
- 25. CI 279132/WA 01090472, Nanconfomance Report Repair of Steam Generator #1 Cold Leg Manway Cover '
Plate ..................................... 26
- 26. CI 279662, Leak Repair of RC-104 ................... 27-
- 27. CI 279691/WA 99003176, Acceptance Test for-DC-3176, Hot Gas Bypass Upgrade ........... 28
- 28. CI 279829, Removal of CC-102 Surge Tank Overflow Check Valve Internals ..................... 29
- 29. CI 280980/WA 9903176, Acceptance Test for DC-3176, Hot Gas Bypass Upgrade ........... 30 iii l
TABLE Oe CONTENTS (continued)-
- 1. FACILITY CHANGES Report Page Humber Document / Title No.
C. TEMPORARY ALTERATION REQUESTS
- 30. TAR-90-010, Component Cooling Water Bleed and Feed ................................... 31
- 31. TAR-91 234, Removal of Sudden Pressure Relay Inputs to Generator Trip ............... 32
- 32. TAR-91-035, Demineralized Water and Chemical Feed .. 33
- 33. TAR-91-039, Cap Drain Lines Downstream of SI-231A&B and SI-232A&B ......................... 34
- 34. TAR-91-040, Blowdown Piping Reconfiguration ........ 35
- 35. TAR-91-041, Alternate Nitrogen Fill Path for Safety Inj ection Tank 2A . . . . . . . . . . . . . . . . . . . . . . 36
- 37. TAR-91-050, Safety Inj ecti on . . . . . . . . . . . . . . . . . . . . . . . - 38
- 38. TAR-91-054, Jumper of One Cell in Battery 3B-S ..... 39
- 39. TAR-92-003, Temporary Insulation for Steam Generator (Bottom Head) ................ 40
- 40. TAR-92-004, QSPDS Channel 1 Heater 1 & 3 Disconnection .......................... 41
- 42. TAR-92-014, Installation of Auto-Isolation Feature on Vendor Supplied Demineralized-Water .... 43
- 43. TAR-92-020, Mechanical-Jumper from PSL-108 to PSL-205 ................................ 44 iv
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TABLE OF CONTENTS (continued)
I. FACILITY CHANGES Report Page Number Document / Title No.
D. DOCbMENTREVISIONNOTICES(DRNs)
- 44. DRN M8901194, Jacket Water Schematic - Emergency Diesel Generator "B" ................. 45
- 45. DRN M8901195, Lube Oil Schematic - Emergency Diesel Generator "B" ........................ 46
- 46. DRN M9001446, Flow Diagram - Reactor Auxiliary Systems .............................. 47
- 47. DRN 09002428/2429, Site Plan / General Site Grading and Drainage ......................... 48
Emergency Diesel Generator (EDG) ..... 49
W3-DBD-002 ........................... 52 -
- 51. DRN M9102333, flow Diagram - Instrument Air System Turbine Building Elevation +40 .~...... 53
- 52. DRN'C9102637, Nuclear Island and Building Design -
Shield Building ...................... 54
- 53. DRN M9102668, Chemical and Volume Control System ... 55
- 54. DRN C9102893/2894, Site Plan / General Site Grading and Drainage ......................... 56 y
^'r6- M 4 4 TABLE OF CONTENTS (continued)
I. FACILITY CHANGES Report Page Number- Document / Tit 1e No.
E. LICENSEDOCUMENTCHANGEREQUESTS(LDCRs)
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- 55. LDCR-91-0035, Pressurizer Level Setpoints Update .. 57
- 56. LDCR-91-0195, Containment Functional Design: i Containment Analysis Update ......... 58
- 57. LDCR-91-0206, Waterford 3 Organization Changes .... 59
- 58. LDCR-91-0212, Revision of FSAR to Conform to Regulatory Guide 1.97, Revision 3 ... 60
- 59. LDCR-91-0229, Control Room Habitability ........... 61
- 60. LDCR-91-0236, Analysis of Post-Accident Containment Hydrogen Concentration .. 62
- 61. LDCR-92-0004, Offsite Power System Update ......... 63
- 62. LDCR-92-0010, Design Flood Level Change ........... 64
- 63. LDCR-92-0015, FSAR Pipe Break Figures ............. 65
- 64. LDCR-92-0024, Emergency Diesel Generator Air Receiver Starting Capabilities ...... 66
- 65. LDCR-92-0048, Station Service Transformers Impedance Revision .................. 67
- 66. LDCR-92-0246, Chemistry Limits Changes . . . . . . . . . . . . 68
- 67. LDCR-92-0247, Welders Qualification Period ........ 69
- 68. LDCR-92-0328, Emergency Diesel Generator Starting Air ................................. 70
- 69. -LDCR-92-0441, Evangeline Natural Gas Pipeline P roj e c t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 vi i
TABLE OF CONTENTS (continued)-
I. FACILITY CHANGES Report Page Number Document / Title No.
F. MISCELLANEOUS EVALUATIONS
- 70. Implementation of DC-3073 and DC-3198 .............. 72
- 71. Setpoint Change (SPC)91-014, Emergency Diesel Generator (EDG) Turbocharger Lubo Oil Trip ......... 73
- 72. Pump & Valve Inservice Test Plan (Change 1, Revision 7) ........................................ 74
- 73. Radiation Monitoring System (RMS) Database Change 91-50 .............................................. 75
- 74. PEIR-50111, Store Additional Scaffold Sections in the Reactor Containment Building ................... 76
- 75. Spare Parts Equivalency Evaluation Report (SPEER)91-687, Main Transformer A Control Cabinet Preferred Supply ................................... 77
- 76. Spare Parts Equivalency Evaluation Report (SPEER)91-617, 125 VDC Batte ry 3 A-S . . . . . . . . . . . . . . . . . . . . . . . 78
- 77. Setpoint Change (SPC)-92-001, Reactor Coolant Loop 1 & 2 Shutdown Cooling Return Valve Nitrogen Pressure Switch .................................... 79 l
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TABLE OF CONTENTS (continued)
II. PROCEDURES Report Page Number Document / Title No.
A. PLANT PROCEDURES
- 78. UNT-005-021, (Revision 2) - Administrative :
Procedure - Radioactive Material and Radioactive Sources Purchase, Receiving, Surveying, Opening and Disposal .............................. 80
- 79. OP-003-009, (Change B, Revision 7) - Operating Procedure - Fuel Oil Receipt ......... 81
- 80. OP-007-005, (Change A, Revision 7) - Operating Procedure - Resin Waste Management .... 82
- 81. OP-903-046, (Change 2, Revision 9) - Surveillance Procedure - Emergency Feed Pump Operability Check ..................... 83
- 82. CE-002-001, (Revision 8) - Technical Procedure -
Maintaining Steam Generator Chemistry ............................. 84
- 83. NOCP-002, (Revision 0) - Administrative Procedure -
NOC Procedure Classification, Type, Content, Numbering, and Format-........ 85
- 84. N0CP-003, (Revision 0) - Administrative Procedure -
NOC Procedure Initiation, Review, and Approval; Changes and Revision; and.
Deletion .............................. 86 viii
TABLE OF CONTENTS (continued)
II. PROCEDURES Report Page Number Document / Title No.
B. SPECIALTESTPROCEDURES(STPs)
- 85. STP-01089789, (Revision 0) - Nitrogen Addition to Condensers ........................... 87
- 86. STP-01090132, (Revision 0) - Main Turbine Electrical Overspeed Test ..............-......... 88
- 87. STP-01090480, (Revision-0) - Verify Operation of Valve EH ISV0123 A ................... 89
- 88. STP-01091453, (Revision 0) - Nitrogen Sparging of Condensers ........................... 90
- 89. STP-01093279, (Revision.0) - Helium Injection to Condensate System .................... 91
- 90. STP-01093885, (Ravision 0) - Nitrogen Addition to Gland Seel-Leakoff Tank Return to Condenser ............................ 92 l
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- l Waterford 3 SES 1992 Report of Facility Changes, Tests and Experiments StM4ARY This report provides the Waterford 3 facility Changes made .
pursuantto10CFR50.59(a)(1). The report covers the period from June 19, 1991 through June 18, 1992. None of the items ,
in this report represent an unreviewed safety question. No experiments Le tests not_ described in the FSAR were conducted at Waterford 3 during the report period, s The report identifies 77 Facility Changes (21 Design Changes, 8 Condition Identifications / Work Authorizations,-14 Temporary Alterations,.11 Document Revision Notices, 15 License Document Change Requests and 8 Miscellaneous Evaluations) and 13 Procedure Changes (7 Plant Procedures and 6 Special Test Procedures).
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I. A. DESIGN CHANGES (DCs)
- 1. DC-3008. Human factors Enhancements to LCP~42A, 42AX, 428, 44, 45, 46 Description of Change i DC-3008 indicates adding human factors enhancements to local control panels LCP-42A, LCP-42AX, LCP-428, LCP-44, LCP-45 and LCP-46. The systems controlled from the panels are Boron Management, Liquid Waste Management, and Gaseous Waste Management. This change includes wiring changes to these systems.
Reason for Change These enhancements include upgrading mimics and color coding schemes on panels to aid operators in controlling the associated systems.
Safety Evaluation According to the Safety Evaluation, the enhancements do not affect the physical configuration of any systems, nor do they affect the functions or flow paths. Therefore, the design change has no adverse affect on any equipment function, system reliability, or Technical.
Specifications margin of safety.
As documented in the Safety Evaluation, this change affects no equipment which could-be considered important to safety. Neither could this change result in the release of radioactive material to the environment.
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- 2. DC-3010, Addition of Mass Flow Probes and a Moisture Control Unit to the Containment Atmosphere PIG Monitor (PRM-IRE-0100Y)
Description of Change DC-3010 reflects the installation of mass flow probes used for flow control to replace the vacuum switches formerly used. Also addressed was a moisture control unit (MCU) to provide dry sample gas to the detectors on the skid.
Reason for Change The mass flow probes installed by this DC improved the sensitivity _and control of the flow rates, and should result in reduced maintenance requirements.
Water collection in the flowrator glass and the gas detector sample chamber had caused premature detector failure and unnecessary down time for this monitor. The addition of the MCU described by this change should prevent this condition.
Safety Evaluation The Safety Evaluation prepared for this design change indicates these changes are enhancements. As such, the intended function of-the monitor is unaffected, and operation and availability should improve.
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Since the design change affects a gaseous waste system, an additional evaluation for radioactive waste systems was performed, which revealed that the change has no adverse impacts on the potential for radioactive releases.
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- 3. DC-3065, Parish Water to Fire Water Storage Tank (FWST) Fill Connection Description of Change DC-3065 addresses the addition of new piping supports to Fire Water Storage Tank (FWST) piping, assuring long term structural integrity of the system, and establishing a pennanent source to refill the FWSTs.
- Other enhancements included replacing existing angle valves with valves of a superior type, and adding a cold mix asphalt pad near a security perimeter to minimize degradation of the perimeter during Fire Water System testing.
Reason for Change Parish water had been used to fill the FWSTs prior to this time via a temporary connection. Adding the piping supports and making the connection permanent assures a reliable source of makeup water to the FWSTs.
Testing of the Fire Water System frequently resulted in degradation of the perimeter zone near _the test site due to water erosion. Laying the asphalt pad minimizes this problem.
The valves which were replaced were subject to degradation because of prolonged exposure to the weather. The new brass valves exhibit superior resistance to such degradation.
Safety Evaluation According to the 10CFR50.59 Safety Evaluation, none of these enhancements have an adverse affect on any equipment function, system reliability, or Technical Specifications margin of safety. The enhancements are designed to improve reliability of affected systems.
As documented, the design change makes permanent the physical connection between the Fire Water System and the Parish Water Supply, which guarantees a reliable source of makeup water capable of meeting the demands of the system. No equipment which could be' considered important to the nuclear safety of Waterford 3 is affected, according to the Safety Evaluation. Neither could this change result in the release of radioactive material to the environment. The change has no impact on Technical Specifications or their bases.
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- 4. DC-3073. Reactor Containment Building (RCB) Air Conditioning Description of Change DC-3073 was prepared to provide a pennanent means of connecting temporary chillers to the RCB Containment Fan Coolers.
Reason for Change Due to the limitations of the Component Cooling Water (CCW) temperatures, the Containment Fan Coolers limited temperature drops during maintenance outages. The permanent piping addressed by this design change provides connections from temporary chillers to the existing CCW piping for adequate cooling.
Safety Evaluation The 10CFR50.59 Safety Evaluation notes no adverse affect on any equipment function, system reliability, or Technical Specifications margin of safety, since the Containment Fan Coolers are not required to be operational during refueling modes. Therefore none of the changes could affect a system important to safety.
The changes documented do not change the physical configuration of the systems beyond providing a permanent tie-in for the temporary chillers. Nor are the flow paths affected during modes when the Containment Fan Coolers are required to be operational.
The safety evaluation also states that all piping installed by this design change meet the applicable seismic design criteria. Further, the documentation states that penetrations made through the flood wall '
in the implementation of this design change result in no increase of the postulated flood consequences.
The safety evaluation indicates no affect on the potential for a release of radioactive material, nor on the Technical Specifications or their bases.
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- 5. OC-3103, Control Element Drive Mechanism (CEDM) Cabinet Cooling Description of Change DC-3103 addresses replacing existing CEDM Cabinet supply fans with higher flow blowers, as well as certain other modifications in the cabinet bays themselves to enhance cooling of the CEDM cabinets.
Reason for Change The design change documents that CEDM Cabinets located in the +21 Electrical Switchgear Area had suffered from insufficient cooling.
Safety Evaluation Since the CEDM Cabinet fans are not specifically addressed in the FSAR, no FSAR nor Technical Specifications change was required, according to the safety evaluation.
The safety evaluation documentation also states that the CEDM cabinets are not safety related. Thus, the only impact the higher flow coolers had on any systems was to increase life expectancy of components located in the CEDM cabinets by lowering internal temperatures, i
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- 6. OC-3145, Essential Services Chillers Auxiliary 011 Recovery System -
Description of Change This design change was developed to add an oil recovery system to the Essential Chillers.
Reason for Change The design change provides a reliable and economical oil recovery ,
system which enhances chiller operation. The change eliminates the potential for the essential chiller compressors to trip on low oil pressure at low cooling loads due to a lack of oil carryover.
Safety Evaluation The safety evaluation notes that no accidents listed in the FSAR are impacted by the change. Neither is there a potential for radioactive material to be released to the environment, nor is any accident of a different type than those previously evaluated in the SAR created by this change.
The only affect on plant equipment is that the reliability of the Essential Chillers is enhanced. The Essential Chillers are required to provide chilled water to cool essential spaces (those spaces containing equipment required for the safety of the plant) during an accident. However, the malfunction of the Essential Chillers during an accident would not increase the probability of a radioactive release, nor affect the Technical Specifications or their bases.
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- 7. DC-3162, Reactor Coolant Shutdown Level Measurement System (RCSLMS)
Description of Change DC-3162-indicates the addition of a level-transmitter connected to the Reactor Coolant System hot leg drain, and to a RCS high point attached to the Pressurizer.
Reason for Change The change document states the purpose as to add reliable reactor coolant level indication during reduced inventory conditions. This added reliability helps reduce the likelihood of a Loss of Shutdown Cooling event.
These changes were made in response to NRC Generic Letter 88-17.
Safety Evaluation According to the Safety Evaluation, the Reactor Coolant Shutdown Level Measurement System is connected only during non-power operations, using temporary connections. The RCSLMS is a non-safety related sub-system, designed to assure that decay heat removal capability is available during refueling operations. This change does not create any new system interactions, nor does it. alter modes of operation.
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- 8. - DC-3175, Essential Services Chillers Dehy'drator Addition Description of Change DC-3175 provides1for the addition of a commercial' grade' dehydrator unit, seismically supported,-to each of the-Essential Chillers.
Reason for Change.
Previously, none of the Essential Chillers were equipped with.any type of dehydrator. Thus, any entrained water in the refrigerant due to a tube leak or any other type of moisture. intrusion into the. chiller unit could result in the formation of hydroflu' ric o and-hydrochloric-acids, which are extremely damaging to the chillers.. According to the-documentation, the dehydrator will also-remove noncondensible gasses.
-Safety Evaluation The safety evaluation notes that.the only affect on plant equipment is ;
that the reliabilityaof the Essential Chillers-it enhanced. The Essential Chillers are required- to provide.chilleo water; to essential-spaces during an accident. However, the malfunction of the Essential' Chillers during an accident would not increase the-probability of a-radioactive release.
The dehydrators are seismically mounted, such that the integrity of the essential chillers and the seismic qualification is maintained.
l The design change has no impact on plant equipment other than the -
Essential Chillers.
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- 9. DC-3176. Essential Services Chiller Hot Gas Bypass Valve Upgrade Description of Change DC-3176 was developed to replace the existing Hot Gas Bypass Valve (HGBV) on the Essential Chillers with a more reliable and effective HGBV.
Reason for Change The existing HGBV operates strictly in an "open or closed" mode, whereas the new valves modulate on demand. This modulation action results in more efficient operation of the HGBV, and increased efficiency and reliability of the Essential Chillers.
(See also items 23, 27, and 29)
Safety Evaluation The safety evaluation notes that no accidents listed in the FSAR are impacted by the change. Neither is there a potential for radioactive material to be released to the environment, nor is any accident of a different type than those previously evaluated in the SAR created by this change.
According to the safety evaluation, the function and operation of the Essential Chillers is unaffected by this design change. The failure mode of the replacement valve is " closed" rather than "as-is" (which is the failure mode of the previous HGBV), thus resulting in enhanced chiller operation in the case of a failure of the HGBV.
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- 10. 0C-3183, Replacement of Instrument Air Dryer Af ter-filter and ,
Installation of Instruments Description of Change 0C-3183 describes replacing the Instrument Air Dryer After-filter (10 microns filtration) with a dual element 0.3 micron-after-filter, as ,
well as the installation of differential pressure instrumentation on both the pre-filters and the af ter-filter as a means of indicating when the filters require changing.
Reason for Change Prior to this design change, no instrumentation existed to provide indication of when either the pre-filters or af ter-filter required '
changing. Additionally, ANSI Standard ISA-57.3 recommends a maximum particle size of 0.3 microns for instrument air systems.
Safety fvaluation According to the safety evaluation, the Instrument Air System is not required to achieve safe shutdown or to mitigate the consequences of an accident. Thus, the changes to the Instrument Air System represented by this design change can have no affect on safety, nor represent an increased probability of a radioactive release.
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- 11. DC-3198, Relocate ILRT Air Supply Description of Change DC-3198 describes the addition of a new branch connection to the existing Station Air supply line, routed to an area adjacent to the Wet Cooling Tower "B".
Reason for Change According to the design change summary, the trailer mounted air compressors utilized in rsrfonning the 50 month Integrated Leak Rate Test have, in the past, required the displacement of small low level radioactive waste containers 'or the period of the test. Also cited is the fact that these compressors experienced overheating problems due to poor natural air circulation in their previous location.
Safety Evaluation According to the safety evaluation, since the ILRT system is not safety related, and the structural integrity of the Nuclear Plant Island Structure is assured by the use of QC-1 materials, there is no adverse impact on any system considered important to safety. Also documented in the safety evaluation is the fact that no new system interactions are introduced by this design change, nor is there an increase in the potential for a radioactive release.
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- 12. DC-3211. Fuel Handling Building +46 Spent fuel Pool Valve Gallery Deck Plates Description of Change DC-3211 documents the addition of access hatches added to the Spent fuel Pool Valve Gallery deck plates, which allow the use of removable handwheel extension stems to operate fuel pool valves.
Reason for Change According to the design change summary, prior to the installation of this modification, it was necessary to remove the spent fuel pool valve gallery deck plates to gain access to the valves listed above.
Safety Evaluation According to the safety evaluation, this design change is for operator convenience, and has no direct impact on the function or operation of any plant equipment. Therefore, no impact is noted on either the Technical Specifications or their bases.
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- 13. DC-3242, Relocation of RM-80 Power Switch and 120 VAC Outlet on fuel Handling Building 150 Rad Monitor Description of Change According to the design change package, this modification installs protective covers over the RM-80 radiation monitors' high voltage power supply toggle switches.
Reason for Change These switches, as stated in the design change, were subject to inadvertent actuation when technicians were engaged in calibration of the monitors affected. This had resulted in an inadvertent Engineered Safety Feature actuation (reference: LER-89-016-00).
Safety Evaluation According to the safety evaluation, since the design change is only a superficial change to the monitors, there is no adverse impact on _
equipment important to safety, nor on any accidents described in the SAR.
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- 14. DC-3250,-Condensate Polisher Vessel Pre-Service Rinse Description of Change The design change sumary (DCS) for DC-3250 states that this change installs the pre-service rinse feature, which was not installed during initial start-up.
Reason for Change As presented in the DCS, this change will prevent an oxygen spike in the feedwater system experienced when a polisher vessel is returned to service. Further, the DCS documents this change as protecting against the potential intrusion of resin fines into the steam generator.
Safety Evaluation R nfety evaluation concludes that no accidents are affected, nor is any equipment important to safety impacted by this change. Since there is no interaction between the polisher system and any safety-related systems, the safety evaluation states that this change has no affect on the margin of safety defined in the Technical Specifications, or their bases.
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- 15. DC-3257, Quench Tank Level Indication Upgrade Description of Change According to the design change sumary, DC-3257 provides new quench tank level and pressure instrumentation.
Reason for Change The design change sumary cites previously unreliable measurement because of the gradual loss of water in the wet reference leg of the instruments. Also mentioned was the poor availability of the instrumentation because of the reference leg problems.
Safety Evaluation According to the safety evaluation, the instrumentation is being replaced by more reliable equipment. The equipment being installed is documented as in compliance with Regulatory Guide 1.97, and is capable of performing the same functions. The quench tank instrumentation is not safety related, and the new piping and instrumentation are documented as being seismically supported. Therefore, there is no adverse impact on the function or reliability of any equipment important to safety, nor on any potential accidents.
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- 16. DC-3267. Essential Chillers Oil Cooler Water Valve (OCWV) Actuator Wiring Changes Description of Change DC-3267 design change summary (DCS) states that this change is a matter of changing the wiring of the OCWV actuator in a manner to eliminate spurious trips of the essential chillers.
Reason for Change According to the DCS, it was suspected that spurious trips of the essential chillers were occurring because of the charging and discharging of capacitors in the circuitry, resulting in feedback current. The wiring changes proposed by this design change were intended to eliminate this potential cause of. spurious trips.
Safety Evaluation The safety evaluation documents that this design change simply changes the wiring points of the OCWV actuator, and does not affect the function or operation of the essential chillers.
The Essential Chillers are required to provide chilled water to essential spaces during an accident. However, the malfunction of the Essential Chillers during an accident would not increase the probability of a radioactive release.
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- 17. DC-3298.__ Refueling Water level Indication System (RWLIS) Range Extension and Indicator Cut-Out Description of Change DC-3298 states that the intent is to extend the range of the RWLIS narrow range transmitter, provide a means for removing the RWLIS local indicator without affecting the control room indicator, and relocate the control room RWLIS indicator for operator convenience.
Reason for Change According to the design change summary (DCS), the extension of the RWLIS narrow range is to permit. indication prior to entering the hot leg levels. The DCS also states that removal of the local indication prior to the implementation of this design change made the control room indication non-functional. The relocation of the control room indicator was to permit ease of viewing by the full control room.
Safety Evaluation According to the safety evaluation, this design change provides enhancements which ensure that control room operators have reliable, readable indication of reactor water level at an expanded range. The safety evaluation states no accidents or equipment important to safety is adversely affected by the change.
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- 18. DC-3299. Deletion of Pump from CVC Letdown Radiation Monitor (PRM-1RE-0202)
Description of Change DC-3299 addresses removing the section pump on the CVC Letdown Radiation Monitor and replacing it with stainless steel tubing.
Reason for chm 3e According to the design change summary, the suction pump failed, allowing water to leak through the motor and damage some of the electronics associated with the radiation monitor. Testing indicated that the differential pressure.provided sufficient flow through the radiation monitor, therefore the design change package indicates deleting the pump.
Safety Evaluation i
According to the safety evaluation, the radiation monitor affected is not used for any safety purpose, therefore no unresolved safety question exists. The reason cited for performance of the safety evaluation was that the pump is shown on a generic configuration drawing located in the FSAR.
" Since the design change affects the CVC system, an additional evaluation for radioactive waste systems was performed, which revealed that the change has no adverse impacts on the potential for radioactive relet. es.
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- 19. DC-3301, RTD Cable Terminations Description of Change According to DC-3301, this change replaces existing 90*C rated cable for RCS Hot and Cold leg RTDs with 125*C rated cable. The design change summary also states that this replacement will eliminate existing splices in cable seal assemblies. Additionally, the design !
change includes the installation of an isolation card in the PPS PAC i panels to eliminate / minimize noise generation by the Core Protection i Calculator (CPC) into the Plant Protection System (PPS) Resistance Temperature Detector (RTD) signal cables.
Reason for Change, The DCS states that RTD signal transmission through the cables affected had been erratic due to faulty shield grounds. It was suspected that the grounds were due to cable degradation because of high temperatures and exposure to oil film coating. Noise introduced into the RTD signal cables during system testing of the design change prompted the installation of the isolation card.
Safety Evaluation The safety evaluation notes that all analyzed accidents could potentially be affected by the design change, but since the design change simply replaces existing cables with a superior product, no adverse impact is noted. The safety evaluation further indicates that the additional signal isolation cards introduce additional RTD temperature measurement uncertainties, but that calculations perfonned by the NSSS supplier demonstrate that the additional uncertainty is only 10.42 F, which is less than the available conservative value of 1 0.77*F. Therefore, according to the safety evaluation, there is no adverse impact on any accident or the Technical Specifications or bases.
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- 20. DC-3341 Essential Chiller Shunt Resistor Replacement Description of Change The design change summary (DCS) states that this change modifies the compressor guide vane control systems of the Essential Chillers The DCS states that the change replaces the existing shunt resistors with a design less susceptible to resistance drift, as well as adding a new 2 conductor #10 cable to each circuit to reduce the burden on the A phase relaying current transfonners.
Reason for Change The previous shunt resistors, according to the DCS, were susceptible to resistance drift because of their design as variable resistors.
The replacement shunt resistors, which are of a fixed resistance design, are not susceptible to this drift. The resistance drift had resulted in chiller inoperability in the past.
The DCS states that the A phase current transformers would have been driven into saturation should a short circuit occur. The addition of the new 2 conductor #10 cables was intended to prevent this possibility.
Safety Evaluation According to the safety evaluation, both of these changes are enhancements which will improve the reliability of the Essential Chillers. Thus, there is no adverse impact on the Technical Specifications or their bases.
The Essential Chillers are required to provide chilled water to essential spaces during an accident. -However, the malfunction-of the Essential Chillers during an accident would not increase the probability of a radioactive release.
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- 21. DC-3345 Plant Monitorina Computer Storage Disk Drive Module Replacement Description of Change The design change summary states that this change replaces existing Plant Monitoring Computer (PMC) disk drives with new disk drives, and makes necessary repairs to the PMC room floor.
Reason for Change According to the DCS, the new style disk drives are capable of storing more information. The previous drives could store a total of 1080 Mb, ,
whereas the new disk drives are capable of storing 2400 Mb.
The DCS states that the PMC room floor repairs are necessary because ;
the new drives are installed in the PMC cabinets, rather than being-external to the cabinets. ;
Safety Evaluation The safety evaluation notes that these disk drives, while noted in the FSAR, are simply used for data storage, rather than control of the plant. As such, these disk drives have no adverse impact on any systems important to. safety, or on any accident. Therefore, the change has no affect on the Technical Specifications or their bases.
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B. CONDITIONIDENTIFICATIONS/WORKAUTHORIZATIONS(CI/WAs)
- 22. C1 277626 Repair of Dry r alina Tower (DCT) 14B Gearbox Description of Chanae As stated in the Nonconfonnance Report for this C1, the intent of this Cl is to replace the gearbox for DCT 14B with a spare rebuilt gearbox.
Reason for Change According to the CI, trending by the Plant Mechanical Maintenance department indicated increasing vibration in the gearbox. Disassembly-and inspection revealed that the gear box required replacement. -
Safety Evaluation According to the safety evaluation, since the gearbox was replaced with an identical rebuilt spare, there was no impact on any accidents or equipment important to safety. The safety evaluation also states that the ability of the ultimate heat sink to dissipate heat from the reactor and auxiliaries was unaffected, as was the number of fans required by Technical Specification 3.7.4.
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- 23. CI 278618/WA 99003176 Acceptance Test for DC-3176. Hot Gas Bypass
[ggrade Description of Chanae According to the CI documentation, this acceptance test ensures the new Hot Gas Bypass Valve (HGBV) on the "B" Essential Chiller functions per design.
Reason for Change The HGBV installation is documented in DC-3176. This CI/WA documents the steps taken to test the installation of the 4GBV on Essential Chiller "B". -
(Seealsoitems9,27,and29)
Safety Evaluation The safety evaluation notes that certain periphery equipment to the essential chiller will be operated in an abnormal manner. Howevsr, because Essential Chiller "AB" will be available to perform the design function in accordance with Technical Specifications, the safety evaluation concludes that no Unreviewed Safety Question exists.
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- 24. C1 278676, Deletion of Supplementary Chiller Motor Windina Temperature Computer Points Description of Change This Nonconfonnance Condition Identification (NCI) documents the removal of RTD sensors from the motor windings of the Supplementary Chillers.
Reason for Change According to the documentation, the RTD sensors have never operated properly, and are not necessary for the successful operation of the chillers. Further, according to the work package, the holes drilled '
for the installation of the RTDs allow a pathway for leakage.
Safety Evaluation The safety evaluation states that this repair does not represent an Unreviewed Safety Question based on the fact that the functions, operability, and reliability of the Supplementary Chillers will be unaffected. The Supplementary Chillers are not safety related, nor do they impact the Technical Specifications or their bases.
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- 25. CI 279132/WA 01090472, NCR Repair of Steam Generator #1 Cold Lea Manway Cover Plate Description of Change This Nonconformance Condition Identification documents work required to perform an ASME Section XI repair to the Steam Generator (S/G) #1 Cold Leg Manway Cover Plate.
Reason for Change The CI documents corrosion / erosion damage to the manway cover caused by the failure of the manway gasket.
Safety Evaluation The safety evaluation reports no Unreviewed Safety Question based on the final thickness of the manway cover, after repair machining, being-greater than the ASME required minimum thickness of 3.871". Thus, the design function of the manway cover is not affected, and this rep. sir has no affect on the Technical Specifications nor their bases.
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- 26. CI 279662, Leak Repair of RC-104 Description of Change This Condition Identification (CI) deals with the_ repair of the Reactor Coolant System (RCS) Hot Leg Sample Isolation Valve, RC-104.
The repair documented by the CI was made by the use of leak repair sealant.
The CI was later revised to reduce the maximum torque value for the clamp rods from 174 ft-lb to 50 ft-lb to prevent plate deformation.
Reason for Change t This CI documents that RCS leakage was observed from the packing gland of RC-104. This leak repair was necessary to stop the leakage.
Safety Evaluation According to the safety evaluation, this repair does not alter the function of the sampling system, nor involve a change in procedures, tests, or experiments. The safety evaluation documents no new system interactions or modifications to flow paths or functions, therefore no adverse impact on systems important to safety or postulated accidents is noted.
The safety evaluation was revised to evaluate the change in the maximum torque value, with the same result of identifying no unreviewed Safety Question potential.
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- 27. 'CI 279691/WA 99003176 Acceptance Test for DC-3176. Hot Gas Bypass Upgrade Description of Change According to the CI documentation, this acceptance test ensures the new Hot Gas Bypass Valve (HGBV) on the "AB" Essential Chiller functions per design.
R_eason for Change The HGBV installation is documented in DC-3176. This C1/WA documents the steps taken to test the installation of the HGBV on Essential Chiller "AB".
(Seealsoitems9.23,and29)
Safety Evaluation The safety evaluation notes that certain periphery equipment to the essential chiller will be operated in an abnormal manner. However, because Essential Chiller "B" will be available to perfonn the design function in accordance with Technical Specifications, the safety evaluation concludes that no Unreviewed Safety Question exists.
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- 28. CI 279829 Removal of CC-102 Surge Tank Overflow Check Valve Internals Description of Change C1 279829 documents the removal of internals from CC-102, which is a ,
piston check valve with a bolted cover. )
Reason for Change l
Analysis documented in this C1 concludes that CC-102 does not meet the-single failure criteria. The CI further references Revision 5 to the IST plan, which identifies that CC-102 performs a safety function only '
in the "open" position.
Safety Ev luation According to the safety evaluation, the removal of the internals of CC-102 provide: 1 unobstructed overflow path for the CCW Surge Tank. ,
This ensures adequate overpressure protection of the CCW Surge Tank.
Further, the safety evaluation states that the overflow piping from the tank is routed to the Waste Management System, and therefore has no adverse affect on any equipment important to safety or on any postulated accidents. There is no impact noted on the radiological consequences of any accident, nor any impact on the Technical Specifications or their bases.
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- 29. CI 280980/WA 9903176. Acceptance Test for 0C-3176, Hot Gas Bypass Upgrade Description of Change According to the CI documentation, this acceptance test ensures the :
new Hot Gas Bypass Valve (HGBV) on the "A" Essential Chiller functions ;
per design. '
Reason for Change :
The HGBV installation is documented in DC-3176 (see item #9 of this report). Similar acceptance tests are documented in items #23 and #27 of this report. This C1/WA documents the steps taken to test the installation of the HGBV on Essential Chiller "A". l Safety Evaluation The safety evaluation notes that certain periphery equipment to the essential chiller will be operated in an abnormal manner. However, because Essential Chiller "AB" will be available-to perform the design function in accordance with Technical _ Specifications, the safety evaluation concludes'that no Unreviewed Safety Question exists. '
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1 C. TEMPORARY ALTERATI0ll REQUESTS
- 30. TAR-90-010, Component Cooling Water Bleed and Feed Description of Change l
This TAR documents the installation of temporary fittings and hose to valve CC-158A. ,
Reason for Change ,
i The purpose of this 1AR was to perform a bleed-and-feed operation on the Component Cooling Water (CCW) system to reduce CCW chloride content to within specifications.
Safety Evaluation The safety evaluation documents the use of CC-158A as the connection i point for this TAR. CC-158A is designed as a nonnally closed vent valve. The safety evaluation notes that by throttling open the valve to perform the feed-and-bleed, CC-158A was operated outside of its design function. No other system interactions were created. There was no impact on any potential accidents or equipment important to safety, since the CCW makeup system was capable of compensating for inventory loss.
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- 31. TAR-91-034 -Removal of Sudden Pressure Relay Inputs to Generator Trip ,
Description of Change TAR-91-034 documents the removal of the sudden pressure relay trips from the generator lockout relay. This action removes the sudden :
pressure trip from the Main Generator trip circuitry.
Reason for Change According to the TAR documentation, the sudden pressure relays have historically been overly sensitive, and have caused a number of spurious trips at Waterford 3. The trips sense the pressure on-Main Transformers A and B.
Safety Evaluation According to the safety evaluation, the removal of these trips has no 1 affect on the function of the main transformers, as other relays would react to any internal faults in the main transformers and would trip the generator in such cases. As.such, the safety evaluation concludes-that the installation of this TAR has no impact on any accidents or equipment important to safety.
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- 32. TAR-91-035, Demineralized Water and Chemical Feed l Description of Change This TAR documents temporary connections between the Chemical feed system and equipment utilized by a contractor to provide demineralized water to Waterford 3.
Reason for Change TAR-91-035 provides for connecting a sulfuric acid supply to equipment operated by Ecolochem, for the purposes of operating the reverse osmosis filtration skid.
Safety Evaluation The safety evaluation concludes that there is no Unreviewed Safety '
l Question, based on the fact that the Demineralized Water System serves no safety function. Therefore, there is no impact on any postulated i accident or upon equipment important to safety. I 33.
- 33. TAR-91-039. Cap Drain Lines Downstream of SI-231A&B and SI-232A&B Description of Change This TAR documents cutting the drain lines downstream of SI-231A, SI-232A, 51-2318, and SI-232B, and then installing threaded pipe caps on the cut lines. The cuts identified by this TAR were to be located downstream of seismic piping supports on the drain lines.
Reason for Change The TAR states that leakage from these valves results in Safety Injection Tanks A&B level reductions. Capping these lines with threaded pipe caps will stop.further leakage, but will still allow manual operation by removing the caps and then manipulating the valves. The TAR states that the valves will be repaired during Refuel <
- 5. Further, this TAR indicates making the cuts downstream of seismic-supports. The location was chosen to simplify calculations, based on input from the Design Engineering department .
Safety Evaluation The safety evaluation (SE) notes that the piping cuts and the caps added are downstream of any safety-related piping. Therefore, the SE concludes that no accidents are impacted, nor is any equipment important to safety affected by this TAR.
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- 34. TAR-91-040, Blowdown Piping Reconfiguration l Description of Change i i
TAR-91-040 documents cutting out valve 80-112B and eroded portions of piping associated with the valve. BD-1128 is the bypass valve for the Blowdown Control Valve, 80-1098.
Reason for Change According to the TAR, piping erosion in the vicinity of BD-1128 had begun to cause a steam leak, which presented the possibility of '
personnel injury. The valve was noted to be normally closed, with no impact on the Blowdown system. ,
Safety Evaluation According to the safety evaluation (SE), the TAR involves piping which is non-safety and non-quality. The piping is normally not in service, therefore the SE concludes that there is no impact on any accident, nor on any equipment important to safety. ,
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- 35. TAR-91-041. Alternate Nitrogen Fill Path for Safety injection Tank 2A Description of Change TAR-91-041 documents running a mechanical jumper from Nitrogen Header Drain Valve NG-502 to the Safety Injection Tank (SIT) 1A instrument cabinet.
Reason for Change The TAR documents that the normal nitrogen fill valve for SIT 1A, NG-161A, had failed. Therefore, this jumper was proposed to provide a nitrogen supply to SIT 1A.
Safety Evaluation The safety evaluation concludes that no Unreviewed Safoty Question ,
exists. This conclusion is based on the design of the Safety Injection 1 Tanks, each of which are isolated from the Reactor Coolant System by two safety related check valves. Further, the check valve installed in the mechanical jumper is no more likely of failure, resulting in making SIT 1A inoperable, than the instrument plug normally in place.
The safety evaluation documents that the temporary piping of the jumper meets ASME Class 2 requirements, is seismically supported, and meets the same quality level and safety requirements as the original nitrogen path. Finally, since the Technical Specifications-require prompt corrective action taken to repair and repressurize an inoperable Safety Injection Tank, there is no greater risk to the consequences of any accident or to equipment important to safety than the risks already present in the piping and equipment as designed.
Description of Change The TAR documentation deals with cutting and capping the drain line downstream of HPSI Header B drain valve SI-2218. The cut is to be made downstream of the seismic support.
Reason for Change According to the TAR, Safety Injection Tank B level was lowering, and leakage past 51-2218 was suspected. This TAR would eliminate any leakage past 51-2218, and would preserve seismic considerations by making the cut and capping the piping downstream of the seism n --
restraint.
Safety Evaluation According to the Safety Evaluation, the piping cuts and the caps added are downstream of any safety-related piping. Therefore, the SE concludes that no accidents are impe.cted, nor is any equipment important to safety affected by this TAR.
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- 37. TAR-91-050, Safety injection Description of Change The TAR documentation deals with cutting and capping the drain line -
downstream of HPSI Header B drain valve SI-209B. The cut is to be made downstream of the seismic support'.
Reason for Change According to the TAR, Safety Injection Tank B level was lowering, and leakage past SI-209B was suspected. This TAR would eliminate any leakage past SI-2098, and would preserve seismic considerations by.
P.aking the cut and capping the piping downstream of the. seismic ,_
restraint. ;
Safety Evaluation According to the Safety Evaluation, the piping cuts and the caps added are downstream of any safety-related piping. Therefore, the SE concludes that no accidents are impacted, nor is any equipment important to safety affected by this TAR.
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- 38. TAR-91-054, Jumper of One Cell in Battery 38-S Description of Change This TAR documents the installation of a jumper to remove from service one of til 60 cells in the 38-S battery. The TAR as written does not specifically address the jumpering of any particular cell. The cell actually jumpered with the TAR was cell number 59.
Reason for Change The TAR '.s established to permit the operation of battery 3B-S with 59 cells rather than the full complement of 60 cells.
Safety Evaluation According to the safety evaluation, Design Engineering calculations show that the 3B-S battery is still capable of performing its safety-related function with one cell jumpered. Therefore, the safety.
evaluation concludes that no accidents or equipment important to safety is affected.
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- 39. TAR-92-003, Temporary Insulation for Steam Generator (Bottom Head)
Description of Change This TAR documents the installation of thermal blanket insulation-on the bottom head of the steam generator in place of the metallic clad pennanent insulation.
Reason for Change According to the TAR, part of the permanent insulation on the bottom head of #1 Steam Generator was to be removed to investigate / repair a leak on the manway cover. It was anticipated that the insulation might be damaged either in the process of removing it, or by the potential boric acid exposure. This TAR was to install temporary insulation until such time as new permanent insulation could be installed.
Safety Evaluation Since the insulation serves no safety function, and does not affect the operability of tt.e steam generator, the safety evaluation concludes that there is no unreviewed safety question. The evaluation also considers seismic considerations, Regulatory Guide 1.36, and LOCA-induced debris concerns. All of the evaluation results are negative, indicating that there is no adverse impact on any accident or equipment important to safety.
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~ 40. . TAR-92-004, QSPDS Channel i Heater 1 & 3 Disconnection Description of Change This-TAR documents lifting the leads for QSPDS channel one heaters and the heated and unhaated thermocouples of. sensors 1 & 3. The TAR-further states that the heaters were replaced by 25 ohm 75 watt resistors, and the thermocouples were shorted.
Reason for Cha m The safety evaluation summarizes that problems with sensors 1 & 3 in channel one interfered with the-functioning of QSPDS. Jumpering the faulty sensors allowed the proper operation of the remaining 6 sensors for channel one.
Safety Evaluation According to the safety evaluation, since this TAR left one sensor operable in the upper three and five sensors operable in the lower five, all tec.hnical specifications requirements for declaring the channel operable were met. The HJTC system does not affect the ability of the operators to place the plant in a safe condition, nor
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does this TAR _ affect any other equipment important to safety.
Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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The TAR documentation deals with cutting and capping the drain line downstream of HPSI Pump A Discharge Header drain valve SI-2098. The cut is to be made downstream of the seismic support.
Reason for Change According to the TAR, this prevents leakage past SI-209A, which affects Safety Injection Tank level. The TAR would preserve seismic considerations by making the cut and capping the piping downstream of the seismic restraint.
Safety Evaluation According to the Safety Evaluation, the piping cuts and the caps added are downstream of any safety-related piping. Therefore, the SE concludes that no accidents are impacted, nor is any equipment important to safety affected by this TAR.
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- 42. TAR-92-014. Installation of Auto-Isolation Feature on Vendor Supplied Demineralized Water Description of Change This TAR documents the addition of an automatic isolation feature on the effluent from the vendor supplied demineralized water to the demineralized water system. The TAR also provides monitoring conductivity of the water effluent, and diverts water to a waste sump on high conductivity.
Reason for Change The reason stated by the TAR documentation is to divert water to a waste sump when the quality of water being produced by the vendor does not meet specifications.
Safety Evaluation The Safety Evaluation concludes that no unreviewed safety question exists, since the jumper is actually installed on vendor-supplied equipment. Further, since the TAR reduces the possibility of introducing substandard water into various plant systems, there is no adverse affect on any accident, or on any equipment important to safety.
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- 43. TAR-92-020,MerylicalJumperfromPSL-108toPSL-205 Description t.' %ange This TAR documents the installation of tubing between valves PSL-108 and PSL-205. The tubing was field fit and supported in accordance with design engineering specifications.
Reason for Change Due to leakage, RC-104 (the normal flow pcth for hot leg sampling) was out of service, pending repair or replacement during Refuel 5. This jumper was intended to provide a flow path for sampling of primary coolant from the Pressurizer Surge Line. The leak repair performed on .-
RC-104 (see item 26 of this report) failed, therefore RC-104, PSL-103, PSL-105, and PSL-107 were closed to isolate the leak.
Safety Evaluation According to the safety evaluation, the installation of this tubing does not increase the probability of a Primary Sample Line-break.
Therefore it does not increase either the probability or the consequences of any accidents. Neither does the TAR adversely impact any equipment considered important to safety. Since the purpose of this TAR is to allow the Post-Accident Sampling System (PASS) to 1 continue to function in a normal fashion, the safety evaluation '
concludes that no unreviewed safety question exists.
Since the design change affects the PASS system, an additional evaluation for radioactive waste systems was performed, which revealed that the change has no adverse impacts on the potential for radioactive releases.
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- 0. DOCUMENT REVISION NOTICES
- 44. DRN M8901194, Jacket Water Schematic - Emergency Diesel Generator "B" Description of Change This DRN documents the revision of FSAR Figure 9 S-4. It corrects the Jacket Water Schematic for Emergency Diesel Generator (EDG) "B" to reflect the as-built configuration.
Reason for Change The documentation notes that changes were made to the Emergency Diesel --
Generator Jacket Water Cooling and Lubrication systems during startup.
The FSAR figures required revision to show instrument root valves, as-built piping, and additional instrumentation.
Safety Evaluation The safety evaluation notes that the changes made to the physical configuration of EDG "B" during start-up allowed for more convenient and accurate testing of the EDG. The function and operation of the EDG was not altered. Therefore, the safety evaluation concludes that no accidents or consequences are affected by the changes. This DRN corrects the FSAR to reflect the as-built configuration.
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- 45. DRN M8901195, Lube Oil Schematic - Emergency Diesel Generator "B" -
Description of Change This DRN documents the revision of FSAR Figure 9.5.7. It corrects the Lube Oil schematic for Emergency Diesel Generator (EDG) "B" to reflect the as-built configuration.
Reason for Change The documentation notes that changes were made to the Emergency Diesel Generator Jacket Water Cooling and Lubrication systems during startup.
The FSAR figures required revision to show instrument root valves, as-built piping, and additional instrumentation.
Safety Evaluation The safety evaluation notes that the changes made to the physical configuration of EDG "B" during start-up allowed for more convenient and accurate testing of the EDG. The function and operation of the EDG was not altered. Therefore, the safety evaluation concludes that no accidents or consequences are affected by the changes. This DRN corrects the FSAR to reflect the as-built configuration.
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- 46. DRN M9001446, Flow Diagram - Reactor Auxiliary Systems Description of Change The DRN documents correcting the flow diagram 1564-G-164 to show i valves BD-1254-3, BD-1252, and BD-1258 as closed, and to show valve BO-1251 as open.
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Reason for Change The DRN documents that the original design of the Blowdown Booster Pump called for continuous operation. This has been changed to use the Blowdown Booster Pump as a stand-by pump, to be utilized only if system conditions require.
Safety Evaluation According to the safety evaluation, the purpose of the Blowdown Booster Pump is to ensure adequate suction pressure is available to the Blowdown Pumps. Thus, operation of the booster pump in a stand-by mode rather than continuously has no impact on system operation or on any accidents. Therefore the safety evaluation concludes that no unreviewed safety question exists.
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- 47. DRN C9002428/2429, Site Plan / General Site Grading and Drainage Description of Change These DRNs document changes to the Site Plan and the General Site Grading and Drainage drawings to reflect the addition of a new office building. The office building is a three-story structure on the east side of the plant, outside of the Protected Area.
Reason for Change Controlled drawing G-128 required revision to add the three story office building on the east side of the plant.
Controlled drawing G-491 required revision to reflect Site Grading in the vicinity of the new office building.
Safety Evaluation The safety evaluation examines four potential accident scenarios.
These are Tornado Generated Missile, Seismic Events, Probable Maximum Flood, and Fire Hazard. In each case, because of the location and construction of the new building, no increase in probability or consequences of the scenario is documented. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 48. DRN M9100607, Resolution of Discrepancies Between FSAR and Design Basis Document - Ernergency Diesel Generator (EDG)
Description of Change The DRN documentation indicates correcting four discrepancies. These are:
1, FSAR table 3.5-2, Emergency Diesel Generator Overspeed trip was changed from 125% to 110%.
- 2. FSAR paragraph 9.5.7.2, deleted the sentence "In the event of a failure in the instrument air supply, the three-way valve is designed to close the bypass around the coolers."
- 3. FSAR section 9.5.4.5, changed the first sentence to read, "Each fuel oil transfer pump may be controlled manually from a switch at the local control panel, but is nonnally controlled automatically by the oil level of its corresponding feed taak."
- 4. FSAR section 9.5.5.2, changed the description of the 3-way temperature valve to read, "A three-way temperature controlled valve controls the flow either through the heat exchanger or to the bypass around the heat exchanger in order to_ maintain the jacket cooling water at a nominal temperature of 170 F," and deleted the sentence which followed that statement.
Reason for Change The DRN indicates these changes were made to correct the FSAR to reflect the correct design of the Emergency Diesel Generator.
- 1. The mechanical overspeed governor is actually set at 110%
- 2. The 3-way control valve is actually a self contained thermostatic valve, which is independent of instrument air.
- 3. The fuel oil transfer switch was-previously indicated as being located in the control room. .In actuality, the switch is physically located at the engine control panel.
- 4. The EDG technical manual indicates that the expanding wax container inside the 3-way control . valve could be ruptured if excessively overheated. It appears this would prevent any flow to the heat exchanger.
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Basis Document - Emergency Diesel Generator (EDG)
(continued)
Safety Evaluation The-safety evaluation states that the changes are made to correct the FSAR. The Emergency Diesel-Generator equipment is per the EBASCO specification and/or the vendor design. Therefore, the safety evaluation concludes that since no physical changes are made to equipment, no accidents or their consequences are affected. Thus, the safety evaluation concludes that no unreviewed safety question exists.
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- 49. DRN E9100901, Battery 3A-S and 38-S Rack Outline Description of Change This DRN documents the addition of an exception' statement to drawing 1564-1047. The exception states that the cable-to-terminal connection on cells 1 & 60 shall be torqued at 100 inch-ibs.
Reason for Change The corrected torque values are specified by the manufacturer.
Safety Evaluation -
The safety evaluation concludes that no unreviewed safety question exists. This conclusion is based on the fact that the revised torque values are specified by the manufacturer as being the proper values to ensure correct electrical connection. -Therefore, changing these values should reduce the probability of a loss of electrical power from the batteries.
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50.- DRN M9100954 Emergency Diesel Generator (EDG) W3-DBD-002 Description of Chance -
The DRN documents the following three changes made to the FSAR:
- 1. FSAR Table 9.5-3. Total Dynamic Head (TDH) of the engine driven jacket water pump was changed from 62.5 ft to 70 ft;
- 2. FSAR Table 9.5-4, Design Pressure, starting air dryer, was changed from 320 psig to 300 psig;
- 3. FSAR Table 9.5-4, Design Pressure, starting air piping, was changed from 315 psig to 265 psig. -
Reason for Change The DRN notes discrepancies between the FSAR and design data. The DRN documents changing the FSAR to match the design data.
- 1. The pump actually has a TDH of 70 ft. at a flow rate of 1080 GPM and a speed of 1750 RPM;
- 2. The vendor specification sheets and EMDRAC drawings list the .
design pressure of the starting air dryer as 300 PSIG, which is the more conservative number;
- 3. The Line List 5817-758 shows a design pressure of 265 psig, which is again a change in the conservative direction.
Safety Evaluation _
According to the safety evaluation, this DRN only corrects the information in the FSAR. Each of the corrections was made in the conservative direction. Therefore, no accidents or possible consequences are affected. The safety evaluation concludes that no '
unreviewed safety question exists.
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- 51. DRN M9102333, Flow Diagram - Instrument Air-System Turbine Building-Elevation +40- ;
Description of Change This DRN~ indicates that controlled drawing G-152, Sheet 2, was changed to indicate that valve IA-409 serves HVT-104, rather than ANP-104A&B.
Reason for Change According to the DRN, the change was made to reflect the as-built configuration. ANP-104A&B are correctly indicated as being supplied by IA-58121 on controlled drawing G-152, Sheet S.
Safety Evaluation The safety evaluation documents that this change is only to the FSAR figure,_and does not change any physical components in the plant._ As a result, no accidents or possible consequences are affected. Thus, the safety evaluation concludes that_no unreviewed safety question exists.
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- 52. DRN C9102637, Nuclear Island and Building Design - Shield Building-Description of Change The DRN documents changing FSAR Table 3.5-10 to reflect _a p'ostulate'd impact elevation range for the 4,000 lb missile at or below 25 feet above grade level.
Reason for Change According to the DRN, page 18 of Design Specification LOU-1564.461 specifies an impact height at "a height no greater than 25 ft above grade...". However, Table 3.5-10 indicated an impact height at grade level. Therefore, to align the values associated with the 4,000 lb.
missile in the FSAR, the postulated height needed to be set "at or below a height of 25 feet above grade level."
Safety Evaluation The safety evaluation notes that the change only affects information pertinent to an external missile hazard. Changing the referenced impact height has no impact on the-probability of occurrence of the external missile striking the Shield Building. Further, the safety evaluation concludes that the ability of the Shield Building to resist the missile is unchanged, therefore there is no affect on the consequences of the postulated accident. Thus, the safety evaluation indicates ro unreviewed safety question exists.
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- 53. DRN M9102668, Chemical and Voluine Control Systa Description of Change According to the DRN, FSAR Table 9.3-9-previously indicated the Chemical Addition Tank volume to be_4 gallons. This DRN documents correcting that volume to 18 gallons.
Reason for Change The DRN notes that Purchase Order #9770797, and the Chemical Addition Metering System technical manual both indicate a volume of 18 gallons.
Therefore, the DRN was generated to correct the FSAR to the as-built condition.
Safety Evaluation The safety evaluation indicates that the tank size is sufficient for the purposes for which it was designed, which is to deliver a certain quantity of chemical to the Reactor Coolant System (RCS).:Since the actual volume of the tank is greater than the previously indicated volume, this function is not impaired in any way. Thus, the safety evaluation concludes that the change has no adverse impact on the probability or consequences of any accident or radiological event.
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- 54. DRN C9102893/2894 Site Plan / General Site Gradino and Orainage Description of Change These ORNs indicate the addition of the Metal Waste Pond to both the Site Plan and the General Site Grading and Drainage drawings.
Reason for Change According to the DRN, the metal waste pond was added to the west side of the plant, outside of the protec^ed area, by Station Modification Package (SMP) #178. Thus, the Site Plan and General Site Grading and '
Drainage drawings required change to reflect the as-built configuration.
Safety Evaluation According to the safety evaluation, the only accident which could be affected by the construction of the metal waste pond is the Probable Maximum flood. However, since the Probable Maximum Flood is postulated based upon a water elevation of +27.00 feet, and the surface of the metal waste pond is at +21.00 feet,_there is no increase in either the probability of the occurrence or the consequences of the Probable Maximum Flood. Thus, the safety evaluation concludes that no unreviewed safety question exists.
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E. LICENSEDOCUMENTCHANGEREQUESTS(LDCRs)
- 55. LDCR-91-0035, Pressurizer Level Setpoints Update Description of Change The LDCR package documents changing FSAR Figure 5.4-9 to agree with setpoints as found in the setpoint document.
Reason for Change According to the LDCR, the setpoints shown in the setpoint document were determined to be accurate. This change incorporates the correct-setpoints into the FSAR figure.
Safety Evaluation According to the safety evaluation, the correction of the FSAR figure will not impact the pressurizer instrumentation. The change is purely administrative, in that it brings the FSAR into agreement with the approved setpoint document. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 56. LDCR-91-0195, Containment functional Design: Containment Analysis Update Description of Change According to the LDCR, FSAR section 6.2.1 was changed to account for a Shutdown Cooling Heat Exchanger (SDCHX) overall heat transfer value which was determined to be incorrect. Also, the containment heat sink and free volume data contained in the FSAR were updated.
Reason for Change According to the LDCR documentation, a detailed engineering evaluation revealed that the correct SDCHX overall heat transfer value.is 216.0 BTU /hr ft* F rather than the 244 BTV/hr ft' F previously shown.
Also, a calculation (EC-C90-056) revealed changes in the containment heat sink and free volume data which resulted from the new SDCHX overall heat transfer value.
Safety Evaluation According to the safety evaluation, the results of the analyses were very similar to the previous results, except that the margins to the containment design peak pressure of 44.0 psig were increased slightly.
Thus, the safety evaluation concludes that no unreviewed safety question exists.
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- 57. LDCR-91-0206, Waterford 3 Organization Changes-Description of Change This LOCR documents changes made to FSAR Chapter 13, " Organizational Structure ~of_ Applicant". The major changes included the creation of a
" Director Plant Hodification and Construction" and associated changes in responsibilities and reporting structure of the " Director Design Engineering" organization.
Reason for Change The LDCR documents these changes as resulting from corporate structure changes made by Entergy Operations, Incorporated.
Safety Evaluation l
According to the safety evaluation, these changes will have ro impact on administrative controls already in place. The only impact on procedures will be to revise them as necessary to reflect the new organizational structure. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- S8. LDCR-91-0212, Revision of FSAR to Conform to Regulatory Guide 1.97, Revision 3.
Description of Change The LDCR documents changes to FSAR Sections and Tables to' conform with' Revision 3 of Regulatory Guide 1.97. Specifically, the changes deal with Accident Monitoring Instrumentation. i i
Reason for Change According to the LDCR, the changes are necessary to update the FSAR accident monitoring information. These changes identify all accident monitoring instrumentation in accordance with Regulatory Guide 1.97, Revision 3.
Safety Evaluation The safety evaluation notes that this change simply lists all accident monitoring instrumentation used in compliance with Regulatory Guide 1.97, Revision 3. No physical changes to the plant or the method in which equipment is operated were-made. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 59. LDCR-91-0229, Control Room Habitability Description of Cliance The LDCR documents changes to FSAR Sections 6.4.2.3, 2.2.3.3.3, as well as to Tables 6.4-1 and 2.2-4. The changes involve Control Room Habitability.
Reason for Change The LDCR documents the changes as resulting from discrepancies noted during Control Room Habitability Analysis, which was performed as a result of PRE-90-074.
Safety Evaluation According to the safety evaluation, the revisions were made to account for information contained in the original design calculations regarding Control Room habitability, and leakage rate calculations.
The safety evaluation notes that there was no change in the design leakage rate, nor were any physical changes made to the Control Room envelope. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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-60.- LDCR-91-0236, Analysis of Post-Accident Containment Hydrogen
-Concentration
--Description of~ Change The LDCR documents an update-to_FSAR Section 6.2.5,-regarding_tne licensing basis. calculation for hydrogen concentration in containment.
following a Loss'0f Coolaat Accident (LOCA).
Reason for Change The analysis involving hydrogen generation were perfonned to include [
hydrogen generation from zinc-based paint corrosion,-rather than the :
previous figures which graphed-post-accident hydrogen generation with and without zinc-hased paint corrosion.
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Safety Evaluation According to the safety evaluation, the_ changes in the FSAR have no- <
affect on the-probability or consequences of an accident, or upon-any.- .
safety related equipment. The changes. simply document an updated-analysis of the post-LOCA hydrogen concentration _in containment.
Since that concentration is still-calculated to~ remain less-than 3.5%
by volume,_which is below the'4% ignition limit, the. safety evaluation concludes that no unreviewed safety question exists.
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- 61. LDCR-92-0004, Offsite Power System Update Description of Change The LDCR documents' changes to various portions of FSAR Chapter 8,-
which update system information and practices. The changes also reflect present transmission system and switchyard configurations.-
Reason for Change According to the LDCR documentation, these changes resulted from a system _ stability study. During the study, certain obsolete practices and outdated information were noted.
Safety Evaluation According to the safety evaluation, the changes documented by this LDCR are improvements in availability, equipment design, and maintenance practices associated with the offsite power equipment.
Therefore, there are no adverse impacts on either accidents or safety related equipment. The safety evaluation concludes that no unreviewed safety question exists.
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- 62. LDCR-92-0010, Design Flood Level Change Description of' Change This LDCR documents changing the design flood level from El. +30 ft.
to El. +29.25 ft.
Reason for Change LDCR-92-0010 states that the top of the floodwall was surveyed at El. +29.27 feet, which was nine inches lower than the design' level.
The survey also indicated that the entire plant is uniformly 9 inches lower than originally surveyed.
Safety Evaluation The safety evaluation notes that the highest postulated water level in the event of a flood is El. +27.6 ft. Therefore, since the change lowers the design height of the floodwall to a level which is still above the highest postulated flood level, there is no adverse impact on any accident. As a result, the safety evaluation concludes that no unreviewed safety question exists.
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- 63. LDCR-92-0015, FSAR Pipe Break Figures Description of Change This LDCR corrects the pipe break locations for the Blowdown System, and documents the addition of Break Point Designators to the Main Steam and Feed Water systems. FSAR Figures 3.6A-2, 3.6A-4, 3.6A-38e, and 3.6A-38f are revised.
Reason for Change The documentation notes that these changes were made to bring the FSAR into agreement with the as-built configuration.
Safety Evaluation 1 This LDCR simply correct; and clarifies information relative to the actual location of pipe break locations. The information added to the j FSAR Figures has been previously analyzed and documented.- Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 64. LDCR-92-0024 Emergency Diesel Generator Air Receiver Startin1 Capabilities Description of Change This LDCR-documents the revision of TSAR Section 9.5.6 to correctly-describe the Diesel Engine start system and tests and inspections.
Reason for Change According to the LDCR, these changes were made as a result of input from an Emergency Diesel Generator System functional Inspection.
Safety Evaluation According to the safety r. valuation, the changes described by the LDCR
- do not affect the operation of the Emergency Diesel Generator (EDG).
Therefore, the refety evaluation concludes that no unreviewed safety question exists.
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- 65. LDCR-92-0048 Station Service Transfonners Impedance Revision ,
pescriptionofChange LDCR-92-0048 documents changing the impedance of the Station Service Transformers (3A31-S, 3B31-S, 3A315-S, and 38315-5) as shown in FSAR i Tables 8.2-5 and 8.3-2 in accordance with calculation EE3-19-05. l l
Reason for Change According to the documentation, the impedance listed originally in Tables 8.2-5 and 8.3-2 reflect the initial impedances provided by calculation EE2-16-06. EE3-19-05 is a more recent calculation, which results in different values.
Safety Evaluation The safety evaluat'on notes that the impedance values are not associated with any accidents in the FSAR. However, since the impedance value3 do affect the voltage drop values in the system, the safety evaluation-references calculation EC-E91-050, which deternines-that equipment will perform the required safety functions during specified low voltages. Therefore, the safety evaluation concludes no .
unreviewed safety question exists. !
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- 66. l.DCR-92-0246. Chemistry Limits Changes Description of Change The LDCR documents changes to FSAR Tables 10.3-2, 10.3-3, 10.4-13, and i 10.4-16. The changes affect the Chemistry Limits for Steam Generator Blowdown, feedwater, Condensate Storage Pool, and Condensate Demineralizer Effluent.
Reason for Change ;
This change was made to be consistent with Chemistry procedures.
Safety Evaluation The safety evaluation states that the changes in Steam Generator Chemistry limits are not significant enough to increase the probability or consequences of any postulated accident. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 67. LDCR-92-0247 Welders Qualification Period Description of Change L This LD.R documents changing the qualification period for welders from
- three months to six months.
?) Reason for Change According to the LDCR, the December 1990 addenda to ASME Boiler &
Pressure Vessel Code Section IX changed the qualification period.
This LDCR was generated to make the FSAR section consistent with the ASME code, as required by 10CFR50.55.
Safety Evaluation According to the safety evaluation, the time limits of welder qualifications is the only change made. Thus, the safety evaluation concludes this change has no adverse affects on the probability or consequences of any accident.
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- 68. LDCR-92-0328 Emergency Diesel Generator Startina Air Description of Chance The LDCR documents the addition of an instrument and the complete revision of FSAR Figure 9.5-5.
Reason for Change According to the documentation, this LDCR resulted from discrepancies noted by plant personnel between the field and controlled drawings.
Safety Evaluation According to the safety evaluation, no physical changes were made in the plant as a result of these discrepancies. The FSAR figure contained some errors, and therefore was revised. The safety evaluation states that there is no adverse affect on safety related equipment or on the probability or consequences of any accident, since no physical changes are made.
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- 69. _LDCR-92-0441 Evangeline Natural Gas Pipeline Project Description of Change This LDCR documents an update to the FSAR to incorporate the Evangeline Pipeline project, and corrects existing errors associated with the related calculations.
Reason for Change According to the LDCR, a new natural gas pipeline crosses the Louisiana Power and Light Company property in the vicinity of Waterford 3. This pipeline supplies natural gas to Waterford 1 & 2 and Little Gypsy electric stations.
Safety Evaluation According to the safety evaluation, the new pipeline does not alter the function or ability to perform the function of Waterford 3 safety related structures. Therefore the safety evaluation concludes that no unreviewed safety question exists.
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F. MISCELLANEOUS EVALVATIONS
- 70. Implementation of DC-3073 and DC-3198 Description of Change According to the safety evaluation, DC-3073 and DC-3198 required chipping a temporary hole in the Nuclear Plant Island Structure (NPIS), while installing a new connection for an Integrated Leak Rate Test (ILRT) air compressor. This feature of the implementation of the subject design changes was not covered by the individual safety evaluations (see items #4 and #11 of this report.)
Reason for Change According to the safety evaluation, the temporary hole was required both as a part of DC-3198 and DC-3073, " Reactor Containment Building (RCB) Air Conditioning". The hole was cut to allow routing of piping through the wall of the NPIS, and was then sealed with new concrr.ie once the piping had been run.
Safety Evaluation According to the safety evaluation, the only accident which could h0ve been affected by this temporary hole in the Nuclear Island flood wall is the Maximum Probable Flood of +27 feet. Since this temporary hole was cut and scaled back at a time when the Mississippi River level was less than +21 feet, there was no impact on either the probability or conseq ences of the flood accident. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 71. Setpoint Change (SPC)91-014. Emergency Diesel Generator (EDG)
Turbocharger Lube Oil Trip.
Description of Change According to the SPC documentation, this change lowers the EDG Turbocharger Lube Oil Low Pressure Trip setpoint from 4.0 psig to 3.0 psig.
Reason for Change The SPC states that this reduction in the pressure setpoint lowers the time for the trip to reset during an EDG start, which will help prevent spurious trips when starting the EDG.
Safety Evaluation According to the safety evaluation, this SPC results in reducing the possibility of spurious trips, and still provides adequate protection to the EDG Turbocharger during maintenance testing. The safety evaluation notes that the trip is locked out when the EDG is operating in the emergency mode. Therefore, the safety evaluation concludes that there is no adverse impact on accidents or equipment important to safety, j
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- 72. Pump & Valve Inservice Test Plan (Change 1. Revision 7)
Description of Change This documents Change 1 to Revision 7 of the Pump & Valve Inservice Test Plan. These changes are sumarized as follows:
Changed valve testing legends to be more consistent with the required testing; Included in the general coments section reference to Technical Specification 4.0.2, which allows extending the time interval between tests!
Included a listing of the subject components in each relief request and clarification; Revised relief requests 2.1.4, 2.1.5, and 3.1.56 as approved by the NRC for Revision 6, Change 1.
Reason for Change According to the documents, the majority of the changes arose from a review of previously approved relief requests. This review resulted from a Notice of Violation issued by the NRC. Other changes resulted from changes in the FSAR or are the result of continuing review of the Inservice Testing (IST) program.
Safety Evaluatior, According to the safety evaluation, the level of testing assumed in the SAR is maintained by this change. Requests for relief are based on system design limitations, and are therefore consistent with the assumptions in the SAR. Thus, the safety evaluation concludes that no unreviewed safety question exists.
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- 73. Radiation Monitoring System (RMS) Database Change 91-50 Description of Change According to the documentation, this change revises the flow control deadband and the particulate, iodine, and gas radiation alarms on the Reactor Auxiliary Building (RAB) Heating Ventilation and Cooling ,
(HVAC) and mobile Particulate, lodine, and Gas (PIG) radiation l monitors. The flow control deadband is changed from 10% to 20%, and the radiation alarms are revised based on 10 CFR 20, Appendix 8. Table l 1, Column 1 values.
Reason for Change '
The RMS Database Change Request states that the purpose of this change is to reduce the number of alams from these monitors.
Safety Evaluation According to the safety evaluation, the changes in the flow control deadband have no impact on the function c the equipment. This is explained by the fact that the isokinetic nozzles used in some of the ,
monitors are designed for 1 25% of the design flow rate. The remainder of the monitors do not use isokinetic nozzles, and flow rate is not critical in these monitors. The safety evaluation also states that the revised radiation alarm setpoints are based on the Maximum Pemissible Concentration (HPC) values found in 10 CFR 20 Appendix B, Table 1, Column 1. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 74. PEIR-50111. Store Additional Scaffold Sections in the Reactor Containment Building Description of Change ThisProblemEvaluation/InformationRequest(PEIR)evaluatesthe storage of fourteen additional sections of scaffold in the Reactor
-Containment Building (RCB) on or near the existing scaffold storage racks.
Reason for Change According to the PElR, the additional sections of scaffold are constantly needed in the RCB for maintenance. To promote safety and cost-effectiveness, the PEIR was generated to obtain engineering determination on the feasibility of storing the additional materials inside the RCB rather than removing them each time they are used.
Safety Evaluation According to the safety evaluation, a-review of the RCB Passive Heat Sink Capacity Analysis indicates that sufficient margin exists to allow the storage of the additional scaffold sections in the RCB.
Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 75. Spare Parts Equivalency Evaluation Report (SPEER)91-687, Main Transformer A Control Cabinet Preferred Supply Description of Change l This SPEER documents the utilization of a replacement breaker that exceeds the original specifications.
Reason for Change The documentation states that this change allows the use of a replacement breaker in the Main Transformer control cabinet. ,
Safety Evaluation According to the safety evaluation, the use of this replacement breaker will not affect the function or operation of the Main-Transformer A or the Motor Control Center. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 76. Spare Parts Equivalency Evaluation Report (SPEER)91-617. 125 VDC Battery 3A-S Description of Change According to the documentation, this SPEER allows the replacement of cell #26 of battery 3A-S with an equivalent cell.
Reason for Change The SPEER notes that the manufacturer's part number changed for the affected cell.
Safety Evaluation .
l The safety evaluation states that the replacement of cell #26 with an i equivalent cell has no affect on the battery or its ability to perform safety related functions. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 77. Setpoint Change (SPC)92-001 Reactor Coolant Loop 1 & 2 Shutdown Cooling Return Valve Nitrogen Pressure Switch.
Description of Change The SPC documents changing the pressure switch setpoints for Safety Injection (SI) valves51-405 A & B. The setpoints are changed from 1435 1 25 psig to 1275 1 25 psig.
Reason for Change The SPC package states that the change is made to provide additional margin with regard to valve opening capabilities.
Safety Evaluation According to the safety evaluation, the change has no adverse impact on the function of equipment important to safety, nor on any postulated accidents. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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II. A. PLANT PROCEDURES
- 78. UNT-005-021, (Revision 2) - Administrative Procedure - Radioactive Material and Radioactive Sources Purchase Receiving, Surveying, Opening and Disposal Description of Change This revision clarifies the responsibilities of all personnel who order, possess, or use radioactive sources at Waterford 3. Specific instructions were added to package surveys and disposal to ensure that systematic inspection and documentation was perfonned by Health Physics technicians.
Reason for Change According to the documentation, this revision was generated in part as response to a Quality Notice, QA-91-191.
Safety Evaluation According to the safety evaluation, this revision deals with administrative duties of personnel, and does not deal with opertaion of any safety systems or with any potential release paths for radioactive materials. As a result, the safety evaluation concludes that no unreviewed safety question exists.
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- 79. OP-003-009, (Chance B. Revision 7) - Operatino Procedure - Fuel Oil Receipt Description of Change According to the documentation, this change added two sections to the procedure. The first section details Emergency Diesel Generator (EDG)
Fuel Oil Storage Tank cleanup. The second section added deals with EDG Fuel Oil Storage Tank emergency restoration.
Reason for Change The change documentation states that chemistry samples indicated the need for processing the EDG Fuel Oil Storage Tank contents through a vendor-supplied filter truck.
The second section added allowed for restoration of the EDG to service if required while the cleanup was in progress.
Safety Evaluation The safety evaluation notes that the change does affect the operability of the EDGs, in that the EDG associated with the tank being cleaned is inoperable during the cleaning process. The cleanup is to be performed only in Mode 5 or Mode 6, when Technical Specification 3.8.1.2 requires only one EDG to be operable.
Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 80. OP-007-005, (Change A. Revision 7) - Operatino Procedure - Resin Waste i Management l Description of Change This change added sections for changing and restoring the Spent Resin l Transfer Pump discharge pressure trip setpoints. J Reason for Change ;
This change was processed because of intermittent difficulties experienced in operating the Spent Resin Transfer Pump.
Safety Evaluation According to the Safety Evaluation, the temporary setpoint changes could result in operating the Spent Resin Transfer Pump outside of design pressures for short periods of time. However, an engineering evaluation included in the documentation indicates that this would not result in any damage to the pump.
The Spent Resin Transfer Pump is a part of the Resin Waste Management' System, which has no affect on the probability or consequences of any accident. Therefore the safety evaluation concludes that no unreviewed safety question exists.
Because the change affects a radioactive waste system, a " Radioactive Waste Systems Additional Safety Evaluation" was performed. The additional evaluation concluded that this change has no impact on the system to perform its intended function, and so has no radiological consegur.nces.
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- 81. OP-903-046, (Chance 2, Revision 9) - Surveillance Procedure -
Emergency Feer ? ump Operability Check Description of Chance According to the documentation, this change corrected the minimum recirculation flow requirements of Emergency Feedwater (EFW) Pump AB from 80 gpm to 72 gpm.
Reason for Change This change was made to conform the minimum recirculation flow requirements with UNT-006-021, Pump and Valve Inservice Testing.
Safety Evaluation According to the safety evaluation, the chan0e in minimum recirculation flow requirements for EFW pump AB does not alter the function or ability of the pump to perform its designed purpose. The changes were made to comply with ASME Section XI Pump and Valves Testing Requirements. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 82. CE-002-001 (Revision 8) - Technical Procedure - Maintaining Steam Generator Chemistry Description of Change According to the documentation, this revision:
- 1. Adds a note stating that the plant should not voluntarily enter action level conditions above 30% power, and that if such conditions are entered, then power escalation should be postponed until the conditions have been cleared;
- 2. Made changes to the definition of operating conditions for steam generators;
- 3. Changed the Sulfate limit for Steam Generator Blowdown from 0.015 ppm to 0.020 ppm;
- 4. Clarified the corrective measures to be taken if an Action Level One condition occurs.
- 5. Made other editorial changes.
Reason for Change The revision documentation states that these changes were made due to INPO recommendations, and to be consistent with EPRI guidelines.
Safety Evaluation The safety evaluation documents that the changes made in Sulfate limits are within EPRI guidelines, and do not increase either the probability or consequences of any accidents. Further, since the function of steam generator chemistry control is to maintain the integrity of the steam generator, and has no operational implications, there is no affect on equipment important to safety. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 83. NOCP-002. (Revision 0) - Administrative Procedure - NOC Procedure ~~
Classification. Type. Content. Numbering, and Format Description of Change This new administrative procedure provides specific requirements and methods to be used to prepare Nuclear Operations Construction Procedures (NOCPs).
Reason for Change This procedure was developed to control those NOCPs which require PORC review.
Safety Evaluation According to the safety evaluation, this procedure is strictly administrative in nature, and has no affect on any postulated accidents or their consequences. Because of this, and the fact that each procedure generated or changed under the control of this procedure will be subject to evaluation per 10CFR50.59, the safety evaluation concludes that no unreviewed safety question exists.
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- 84. NOCP-003 (Revision 0) - Administrative Procedure - NOC Procedure i Initiation Review, and Approvalt Chances and Revisions and Deletion Description of Change According to the documentation, this procedure provides instructions for the initiation, change, revision, or deletion of Nuclear Operations Construction (NOC) procedures.
Reason for Change This procedure was developed to control NOC procedures which require-either PORC or PORC Subcomittee review.
Safety Evaluation According to the safety evaluation, this procedure is strictly administrative in nature, and has no affect on any postulated accidents or their consequences. Because of this, and the fact that each procedure generated or changed under the control of this procedure will be subject to evaluation per 10CFR50.59, the safety evaluation concludes that no unreviewed safety question exists.
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B. SPECIAL TEST PROCEDURES (STPs)
- 85. STP-01089783, (Revision 0) - Nitrogen Addition to Condensers Description of Change According to the documentation, this special test implements the addition of nitrogen to the shell side of the condensers.
Reason for Change The purpose for this test is stated as an attempt to reduce condensate dissolved oxygen.
(See also items 88 and 90)
Safety Evaluation The safety evaluation notes that the rate of nitrogen addition to the condensers by this special test will be limited to a combined flow rate for all three condensers of 6 standard cubic feet per minute (SCFM). Since the capacity of each condenser evacuation pump is 25 SCFM, this addition rate is well within the abilities of the condenser evacuation pumps. Thus, there is no impact on condenser vacuum, and no impact on the probability or consequences of any accident. The safety evaluation therefore concludes that no unreviewed safety question exists.
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- 86. STP-01090132, (Revision 0) - Main Turbine Electrical Overspeed Test Description of Change This test independently verifies proper operation of the main turbine electrical overspeed protection devices by disconnecting electrical speed indications and connecting a function generator to the circuitry. The function generator is then used to simulate turbine speed and test the electrical overspeed trip functions.
Reaso_n for Change The special test procedure indicates that this is done to independently verify the proper operation of the main turbine overspeed functions.
Safety Evaluation The safety evaluation notes that this test is to be performed while the plant is in mode 5, or with the Main Steam Isolation Valves closed and the steam header to the main turbine depressurized. Thus, there is no impact to the turbine or operating systems, since the systems involved will be inoperable during the test. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 87. STP-01090480, (Revision 0) - Verify Operation of Valve EH ISV0123 A Description of Change According to the documentation, this test disables portions of the turbine overspeed sensing circuitry, and utilizes a function generator ,
to cause solenoid valve EH ISV0123 A to actuate, i Reason for Change The stated purpose is to test valve EH ISV0123 A.
Safety Evaluation According to the safety evaluation, this test is to be performed while the plant is in mode 5, or with the Main Steam Isolation Valves closed and the steam header to the main turbine depressurized. Thus, there is no impact to the turbine or operating systems, since the systems involved will be inoperable during the test. Therefore, the safety evaluation concludes that no unreviewed safety question exists.
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- 88. STP-01091453. (Revision 0) - Nitrogen Sparaina of Condensers Description of Change According to the documentation, this special test implements the addition of nitrogen to the shell side of the condensers.
Reason for Chance The purpose for this test as stated is to determine the effectiveness of nitrogen sparging of the main condensers for contrnlling dissolved oxygen in the condensate system.
(Seealsoitems85and90)
Safety Evaluation The safety evaluation notes that the rate of nitrogen addition to the condensers by this special test will be limited to a combined flow rate for all three condensers of 18 standard cubic feet per minute (SCFM). Since the capacity of each condenser evacuation pump is 25 SCFM, this addition rate is well within the abilities of the condenser evacuation pumps. Thus, there is no impact on condenser vacuum, and no impact on-the probability or consequences of any accident. The safety evaluation therefore concludes that no unreviewed safety question exists.
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- 89. STP-01093279, (Revision 0) - Helium Injection into Condensate System Description of Change STP-01093279 states its purpose is to inject helium into the main condenser and Condensate Pump "A" suction to measure condenser efficiency and locate sources of condenser dissolved oxygen.
Reason for Change The document package states that STP-01093279 was developed and implemented in order to troubleshoot the sources and amounts of dissolved oxygen in feedwater.
Safety Evaluation According to the safety evaluation, this special test procedure does not represent a potential unreviewed safety question because the amount of helium to be injected was within the capabilities of the Air Evacuation system. Therefore, neither the possibility of nor the consequences from a loss of condenser vacuum do not increase as a result of this test. None of the specific components affected by this special test are considered important to safety, nor does the injection of helium into the condensate system represent any potential for radioactive release.
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- 90. STP-01093885, (Revision 0) - Nitrocen Addition to Gland Seal Leakoff TE k Return to Condenser Description of Change The test documentation states that this test introduces nitrogen to the gland seal leakoff tank return line to the condenser.
Reason for Change The purpose for this test as stated is to determine the effectiveness of nitrogen addition to the gland seal leakoff. tank return line for controlling dissolved oxygen in the condensate system.
(See also items 85 and 88)
Safety Evaluation The safety evaluation notes that the rate of nitrogen addition to the condensers by this special test will be limited to 10 standard cubic feet per minute (SCFM). Since the capacity of each condenser evacuation pump is 25 SCFM, this addition rate is well within the abilities of the condenser evacuation pumps. Thus, there is no impact on condenser vacuum, and no impact on the probability or consequences of any accident. The safety evaluation therefore concludes that no unreviewed safety question exists.
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