ML20076B260

Rept of Facility Changes,Tests & Experiments

ML20076B260
Person / Time
Site:	Beaver Valley
Issue date:	12/31/1990
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20076B255	List: ML20076B253 ML20076B258 ML20076B260
References
NUDOCS 9107110143
Download: ML20076B260 (160)

Text

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h5 Muesm @ Comparty BEAVER VALLEY POWER STATION UNIT NO. 2 DOCKET NO. 50-412 LICENSE NO. NPF-73 t

l ATTACHMENT 2 l

1990 REPORT OF FACILITY CHANGES, TESTS, AND EXPERIMENTS l

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Bsaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments

-Igble of contents Testina Procedures Pace Cycle 2 Extension 1 Health Physics ProceglitLqn Alarm Sotpoint Chr.nge To 2SSR-RQIl00 and Change To UFSAR 10.4.8.3 and 11.E.2.5.9 -2

-Operatino Procedures OM 2.19.4.I " Clearing GWS Overhead Gas Coopressor" '3 Temporary. Modification - Installation of Temporary Pressure Gauge at (PDI-1RW-109) 4 Normal System Arrangement Change'for [2SAS-41 AND 43] 5 OST 2.24.9 "Overspeed Trip Test of Turbine Driven AFW Pump [2FWE*P22)" 6 TOP 2-90-1 " Transferring Resin from 55 Gallon Drums to a HIC-at Unit 2. Waste Handling Building Truck Bay" 7 TOP i-90-2 " Charging Pump Lines Hydrogen Accumulation Test" 8 TOP 2-88-26, Revicion 1 " Head Capacity Curve and Base Line Data Collection for CCP. Pump [2CCP*P21A,B,C) Test" 9 Temporary Modification - Gagged (2CNM-RV-100A] 10 OST 2.15.3 and 4 "CCP Valve Position Verification" 11 Isolation of [2GNS-PCV101A,B] by Changing (2GNS-310,311,320 and 321] to NSA Closed 12 TOP 2-90-4' "D" Moisture Separator Reheater Reheat Steam' Isolation 13 Isolation of Reheater Drain Receaver Level Gauges

[2HDH-LG100A,B,C,D) and Moisture Seperator Drain Receiver Level Gauges [2HDH-LG115A,B,C,D}- 14

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimonts Table of Contente Operatino Procedures (Continued) Eag2 Isolation of First, Third, Fourth, Fifth Point Feedwater Heater Level Gauges (2FWS-LG-103A,B, 120A,B, 122A,B, AND 124A,B) 15 TOP 2-89-37 " Temporary Filter Installation Across Startup Feed Pump Lube Oil Coolers" 16 OM 2.30.4.L "SWS Silt and Corbiculate Control Part A, Flushing SWS Pumps Seal Water Supply Line 17 TOP 2-89-36 " Flushing the SW Pumps Seal Water Supply Lines" 18 OM 2.30.4L "SWS Silt and Corbiculate control Part B Flushing Rod Control Area ACU 19 Failed Relay 69-FWSNB 20 TOP 2-89-38 "SWS Back Flushing of CCP HX'S" 21 OST 2.30.20 "CCP Hx Performance Trending and RSS Hx Dry Layup Check - Part B" 22 TOP 2-90-7 " Recirculation of (2WTD-TK23) Through a Temporary Dimineralizer for Cleanup" 23 Temporary Modification - Defeat of (2HVP-ACU211A & B)

Low Temperature Trip 24 TOP 2-90-8 " Removal of Reheat Steam to (2MS!-H21B]" 25 Temporary Modification - Jumper Across Failed open Limit Switch on (2CCP*AOV107C) 26 OM 2.9.4H " Aligning the Main Steam Valve Room and Instrument Air Compressor Floor Drains to catch Basin 23" 27 TOP 2-90-11 " Bypass of Instrument Air Dryer and Receiver" 28 TOP 2-90-17 "(2FPD-TK22) CO2 Purge to Allow Valve Maintenance" 29 TOP 2-90-14 2 "t FPD-TK23) CO2 Purge to Allow Valva Maintenance" 30 OM 2.6.4W - " Isolating a Reactor Coolant Loop" 31

Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Ipble of contents Operatina Proceduren (Continued) Eagg 2 TOP 90-15 "[2 CSS-E21A, B} Turbine Plant CCS Hx Turbine Side Performance Verification" 32 2 TOP 90-25 " Containment Backup Fire Protection During Type "C" Testing" 33 2 TOP-90-23 " Supplying both SWS Headers with One SWS Pump" 34 OM 2.11.4.C - " Tilling Reactor Refueling Cavity" 35 2 TOP-90-21 "High-Volume Air Supply to Containment" 36 OM 2.6.4W - " Isolating a Reactor Coolant Loop" 37 2 TOP-90-10 " Moisture separator Rehoater Tube Leak Test" 38 2 TOP-90-13 " Temporary Aux Steam Supply to Turbine Lube Oil Coolers 39 OM 2.6.4D - Fill of an Isolated RCS Loop, 2.6.4.N - Reactor Coolant System Isolated Loop Rocovery 40 Temporary Modification - Install Proximity Limit Switchen in Series with the Electrical and Mechanical Torque Switches for the Fuel Transfer Cart 41 20M-51.4.D - " Station Shutdown - Cooldown form Hot Shutdown (Mode 4) to Cold Shutdown (Mode 5) 42 Raintenance Procedures Se'urity computer System Multiplexor Board Modification 44 Disable CCW Hi Flow Closure Logic to 2RCS-P21C Thermal Barrier Flow Valve 2CCP-AOV107C for Corrective Maintenance 45 Defeating Hi CCW Flow Close Signal to RCS-P21A Thermal Barrier Isolation Valve 2CCP-AOV107A for Performance of 2LCP-15-F107A During Mode 5 or 6 or for Corrective Maintenance in an Operating Mode 46 Defeating Hi CCW Flow Close Signal to RCS-P21B Thermal Barrier Isolation Valve 2CCP-AOV107B for Performance of 2LCP-15-F107B During Mode 5 or 6 or for Corrective Maintenance in an Operating Mode 47

B aver Valley Powsr Station Unit 2 l 1990 Report of Facility Changes, Tests, and Experiments l Table of Contente l

Onoratina Procedures (Continued) Eagg Defeating Hi CCW Flow Close Signal to RCS-P21C Thermal Barrier Isolation Valve 2CCP-AOV107C for Performance of l

2LCP-15-F107C_During Mode 5 or 6 or for Corrective l

Maintenance in an Operating Mode 48 480V Bus 2H Ground Alarm Relay 49 480V Bus 2F Ground Alarm Relay 50 Temporary Installation of Gag on 2CNM-RV151 51 Temporary Installation of Gag on 2CNM-RV115 52 Temporary Repair of 2SVS-PCV101B Shaft coupler 53 I

Temporary Modification of 2GWS-315 54 I

Temporary Modification to Turbine Turning Gear and i Bearing Lift Oil Pump Aux Relay Circuit THJND 55 I I

480V Bus 2D Ground Alarm Relay 56 l-f i

Temporary Power for Turbine Generator overhaul 57 Connecting Recording Equipment to Operating Equipment 58 Vendor Procedures for Hydrolasing 1" & 1 1/2" SWS y Supply and-Return Lines to Sample Coolers for

! 2SWS-RQIl00 A, B, C, & D in Emergency Diesel Generator Buildings 59 Supplying Temporary Trailers Sprinkler System from Turbine Building Fire Suppression System- 60 i

l Cross Tie Between MCC-2-E05 and MCC-2-E06 during 2R Only 61 Temporary Hodification to Hook Up a Flush Rig Between l

2HVC-ACV201A(B) SWS Cooling coils and Associated SWS

Supply and Discharge Piping 62 I.

l Temporary Power for Cooling Tower Work During 2R 63 i

Source Range High Flux at Shutdown Alarm Modification 64 Temporary Modi fication for Flush of 2HVC-ACV207A & L 65 l Temporary Steam Supply from Aux Steam Valve 2 ASS-646 to Temporary Flush Rig in PAB 66 l

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Boaver Vallsy Pow 2r Station Unit 2 1990 Report of Facility Changes, Testo, and Experiments Table of Contents Facility Chances Page Desian Chance DCP-851, Elimination of Vibration on MSR Scavenging Vent Condenser Piping 67 DCP-801, Oland Steam Exhauster and Filtration System 69-DCP-899, Diesel Generator Backup Phase Fault Detection 71 DCP-905, Rev. O, Steam Generator Blowdown Sample sodium Analyzer Replacement 73 DCP-907, Rev.-1, computer Status Valve Position Indication 74 DCP-963, Enhancement of the Turbine Plant Sample System 76 ,

DCP-979, Rework Vendor Supplied "PVC" Jacketed Cables for Radiation Monitor Printers 78 DCP-1046, Provide Oil Fill Capability to Reactor Coolant Pump Motors 80 DCP-1049, Auxiliary Feed Pump Steam Drain Valves 82 DCP-1063, Relocate Fire Hose Rack No. HR-243 and Provide Supports 83 l~ DCP-1078, Disabling the Auto Dispatching System (ADS Function) 84

'DCP-1130,, Rev. O, Temporary Reactor Vessel Head Shielding for Unit No. 2 86 DCP-1135, Rev. O, BV-2 Small Bore Snubber optimization 89 DCP-1160, Rev. O, Provide Heat Tracing for the Pump / Heater Assembly of WTD-TK23 and the Sensing Line for 2WTD-LT107 91 DCP-1195, Rev. O, Instrument Air Dryer Coalescent Pre-Filters 93 DCP-1237, Rev. 2, DBMS Modifications to Prevent Bad Data Quality Indication on ERFCS/SPDS Displays. 95 DCP-1265, Rev. O, Relocation of Level Switches 2FWS-LS?O4A/B 97

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Baaver Valley Powar Station Unit 2 l 1990 Report of Facility Changes, Tests, and Experiments Table of Contents l l

Facility Chances (Continued) EA_q2 l Desion Cnance DCP-1275, Rev. 1, S/G Primary Manways Bolt to Stud conversion 99 DCP-1286, Rev. O, 2 MSS *AOV101A,B,C C/NC Contact Modification 101 DCP-1313, Rev. O, Sprinklers for Turbine Building Pedestals and Amertap-Condenser Pit 103 DCP-1314, Rev. O, Ground Test Circuit for Bus 2-5 and Bus 2-6 105 DCP-1321, Rev. O, Annunciator Window A8-6B, " Heat Tracing System Trouble" 107 DCP-1328, Rev. O, Inadvertent Actuation of SSPS Outputs 109 DCP-1336, Rev. O, Nuclear Instrurvntation System Source Range High Voltage Cutoff 111 DCP-1339, Rev. O, C/NC Contract Indication 114 I DCP-1356, Rev. O, 2CHS*P21A and C Vents 116 DCP-1360, Rev. O, Installation of New Liners Under 2HDH-P21A&B 118 DCP-1364, Rev. 1, Rod Position Deviation Alarm 120 LDCP-3366, Rev. 0, _ Removal of Brace from Duct Support 2HVP-DSA684N 122 DCP-1371, Rev. O, Replace Existing 2DGS-300 Globe Valve with a 2" Gate Valve 124 DCP-1390, Rev. O, Containment Air Recirc Fans 2HVR-FN201A and 2HVR-FN210B Status 126 DCP-1403, Rev. O, 2RCS*MPV535 and 536 OPEN/NOT OPEN Contact Indication 128 DCP-1440, Rev. 0, Main Unit' Generator' Seal Oil System Drain and Vent Valve Addition 131 DCP-1456, Rev. 1, DRMS, RM-11 Software Revisions 133 DCP-1469, Rev. O, RTD Bypass Manifold Elimination 136

Bsevor Valloy Powcr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Inble of contente Facility Chances (Cot.tinued) EA2R gggion chanos DCP-1498, Rev. 2, RWST LO-Lo Bistables 138 DCP-1500, Rev. O, Steam Cenorator Blowdown Drain Line 140 DCP-1503, Rev. 1, Unit 2 Hodifications for float Exchangers Performance Monitoring 142 DCP-1545, Rev. O, Replacement of Diesel Generator 2-1 and 2-2. Automatic Loading Sequence Timer Relays 144 DCP-1567, Rev. O, Replacement of Recirculation Spray Pump Timer / Relays 146 DCP-1576, Rev. O, Fuel Transfer Tube Blind Flange Modification 148 DCP-1589, Rev. O, Steam Generator Level 150 Technical Evaluation Reoort (TER)

TER-1144, Rav. O, Change of QA Category for PAL ,

Ilydraulic System 152 l

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Deaver Valley Power station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 1 of 152 CitANGE TITLE Cycle 2 Extension filANGE DEECRIPTION An extension of Cycle 2 was desired to permit a September 1, 1990 shutdown date for 2R.

SAFETY EVAL MTION SUff2@X A W?ttanghoupe evaluation of the Cycle 2 extension was performed which endorsed the results of the evaluation previously performed for the Cycle 1.

Additionally, Duquesne Light performed a review of FSAR and Technical specification changes that had been issued subsequent to the Cycle 1 cafety evaluation. It was determined that no changes were made that would affect the validity of the previous evaluation. Thus, the Cycle 2 extension did not pose an unreviewed safety question. No Technical Specification changen are required. No FSAR changes are required.

Usevor Valley Powar Station Unit 2 l 1990 Report of Fac411ty_ Changes, Tests, and Experiments I Page 2 of 1$2 CHANGE TULE Alarm Setpoint change To 2SSR-RQIl00 and Change To UFSAR 10.4.8.3 and 11.6.2.5.9 CfdEEE_DIECRIFTION In order to minimite spurious steam generator blowdown isolation (n.b., ESF activation), the alarm setpoints for monitor 2SSR-RQ1100 were revised. The previous HIGH setpoint was set _at 1.0E-5 uci/ml. This level in Close to the observed background. The 1.0E-6 uC1/ml value is now used for the ALERT alarm

.(no automatic actions), and a value based on T/S effluent releases used for the HICH alarm. The UFSAR changes involve correcting references of sampling on a HIGH alarm to refer to the ALERT alarm instead.

SAFETY EVALUATION

SUMMARY

The safety evaluation concluded that the operative criterion was the 1.0E-5 uci/ml value, not that this value was assigned as a HIGH alarm. The existing alarm response procedures and the EOPs ar. keyed to sample results that indicate 1.0E-$ uCi/ml. The sampling-is triggered by either an ALERT or HIGH alarm. The Turbine Building sump is isolated, by procedure, at activity levels exceeding-1.10E-5 uCi/ml. Blowdown isolation occu- automatically at a level equivalent to effluent' T/ Sis. Based on this, che safety evaluation concluded that there would be no increase in the offsite radiologicci consequences of an SGTR. Since the alarm setpoint is a software database item, there le no feasible failure mode associated with this change and, therefore, no increase in probability of such events.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 3 of 152 g1Migt TITLE OM 2.19.4.I " CLEAR 1NG CWS OVERHEAD CAS COMPRESSOR" Q1M1QE.DEECRIPTION A new procedure was developed to purge the Overhead Gas compressor (CWS-21A(D)) of hydrogen gas (using nitrogen) prior to maintenance and to purge the compressor of oxygen (uoing nitrogen) after maintenance.

SAFETY EVALUATION SUMMAFY The cas Compressor [GWS-C21A(D)) is isolated from the gaseoun waste system and placed on clearance so that purging the compressor will not cause a malfunction of the gaseous w6ste system. The compressor is non-cafety related and no other safety related equipment is located in the cubicle. Any radioactive gas released in the PAD will be processed by PAD ventilation and the SLCR system. The amount of radioactive gas purged in the procedure is a small fraction of the volume assumed in UFSAR 15.7.1. The purging operation is performed to ensure the oxygen and hydrogen concentrationo are within Technical Specification 3.11.2.6 limits. No unreviewed safety questions exist.

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B2 aver Valloy Power Station Unit 2 1990 Report of racility Changes, Tests, and Expts. ants Page 4 of 152 CilANGE TITLE TEMPORARY MODIFICATION - INSTALLATION OF TEMPORARY PRES $URE CAUGE AT

[PDI-1RW-109) gilANGE DESCRIPTIO1{

A ternporary pressure gauge was installed at (PDI-1RW-109) to perform procedure 2.30.4L, SWS Silt and Corbicula control. The installed pressure gauge PSID range is to large for the procedure.

SAFETY EVALUATION

SUMMARY

The temporary gauge is isolated by.[1WR-220) and [2SWE-220]. This will ensure all minimum flow requirements per UFSAR 14.3 (Unit 1) and 9.22 (Unit 2) are met. All equipment important to safety will continue to be supplied by minimum flow requirements while (1WR-220) and (25WE-220) are closed. No ,

unreviewed safety questions exist.  ;

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Boavor Valley Power Station Unit 2 1990 Report of racility Changen, Tests, and Experiments Page 5 of 152 CHANGE TITLE NORMAL SYSTEM ARRANGEMENT CHANGE FOR (2SAS-41 AND 43)

CHANGE DESCRIPTIM The NSA for valves (2SAS-41 and 43),

• Station Air Compressor Hanifold to Backup supply to condensate Polishing Building Air", and " Station Air Backup Supply Isolation", where changed from closed to open.

SAFETY EVALUbTION

SUMMARY

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.UTSAR 9.3.1.1.3- and 9.3.1.2.3 state that station air and condensate polishing air are not required for safe plant shutdown. If control air is lost, valves will fail in the safe position (per UFSAR 6.2.4.1), therefore the consequences of an accident will not be increased. Per UFSAR 9.3.1-.1.2,-instrument air isolates from station air on low pressure; opening (2sAs-41 and 43) allows condensate polishing air to be quickly supplied to the station air header. No unreviewed safety questions exact.-

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Bsavor Valle,v Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 6 of 152 CHANGE TITLE OST 2.24.9 "OVERSPEED TRIP TEST OF TURBINE DRIVEN AFW PUMP (2FWE*P22)"

CHANGE DESCRIPTION A new test was developed to demonstrate the operaitlity of the overspeed trip mechanism of the turbine driven AFW pump (2FWE*P22). This test was generated in response to IE Notice 88-67 and INPO's SOER 89-1.

SAFETY EVALUATION

SUMMARY

The AFW system will not be affected by this test since (2FWE*P22) is closed for the _ test, and the two other trains of AFW pumps remain operable; therefore, the accidents described by UFSAR 15.2.6, 15.2.7, 15.2.8, 15.6.3, 15.6.4 and 15.6.5 are not affected. The_ turbine driven AFW pump (2FWE*P22) is intentionally _ made inoperable by the test, however,_this is allowed by UFSAR 10.4.9. The probability loss of auxiliary feed water accidents discussed in UFSAR 15.9.3 and 15.9.4 is not_ increased since these accidents assume (2FWE*P22) is not available. No unreviewed safety questions exist.

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Deaver Valley Power Station Unit 2 1990 Report of Facility Chenges, Tests, and Experiments Page 7 of 152 CHANGE TITLE TOP 2-90-1 " TRANSFERRING RESIN FROM 55 GALLON DRUMS TO A IIIC AT UNIT 2 WASTE ilANDLING BUILDING TitUCK BAY" CHANG M fqRIPTION A new temporary procedure was developed to charge and dewaste 55 gallon drums of spent resin at Unit 2 Waste llandling Duilding Truck Day Area.

SAFETY EVALUATION SUMMAE1 The portable pumps, 55 gallon drums and the HIC do not connect to the RWST and hence will not affect the assumptions used in UFSAR 15.7.3. Any spill would be enveloped by UFSAR 15.7.3. No safety related equipment is used or located in the area of reoin transfer and aewatering operation. Transferring resin from drums to a llIC will not violate Technical Specification 3.11.3.1. No unreviewed safety questions exist.

B3 aver Valley Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 8 of 152 CHANGE TITLE TOP 2-90-2 " CHARGING PUMP LINES HYDROCEN ACCUMULATION TEST" '

l CHANCE DESCRIPTION A new temporary procedure to isolate the operating charging pump miniflow isolation valve was written to that Engineering can perform a test to detect the accumulation'of hydrogen.

SAFETY EVALUATION

SUMMARY

The charging pump will perform its ESF and ECCS function upon SIS actuation

-and satisfy the assumptions of UFSAR 15.1.4, 15.1.5, 15.2.8, 15.4, 15.5 and 15.6- since the miniflow valve is closed in its SIS actuation position. In the event the charging pump wi.1 automatically start upon an SI signal as desigrmd and provide adequate core- cooling (UFSAR 6.3.2). The procedure requires an operator to monitor pump and motor paramoters to detect the onset of a malfunction- while the miniflow valve is closed. Closing the miniflow valve will not render the charging pump incapable of performing its intended function in accordance with Technical Specification 3.1.2.4 nor will it violate the boration flow path requirements of Technical Specification 3.1.2.2. No unreviewed safety question exist.

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B3ever Valley Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 9 of 152 CHANGE TITLE TOP 2-88-26, Revision 1 " HEAD CAPACITY CURVE AND BASE LINE DATA COLLECTION FOR CCP PUMP (2CCP*P21A,B,C) TEST" CHANGE DESCRIPTlQH i An existing TOP has been revised to operate and test the CCP system while (2CCP*DCV100-1,2] Unit 2- CCP pump recirculation valves are closed or I isolated. (The valves are currently out of service). The valves help maintain minimum CCP pump discharge flow by controlling differential pressure across the CCP pumps.

{AEETY EVALUATION

SUMMARY

i UFSAR 15 accident analysis states that following an analyzed event initiation, a reactor trip occurs and possibly an SIS or CIB. The RCS is then cooled to

' hot standby by AFW or Safety Injection. The function of CCP system (UFSAR 9.2.2.1) is to supply sufficient cooling water to enable a cold shutdown by RHR. .This is a long term function and not short term following the event.

Although UFSAR Table 3.9B-19 places these valves in a safety category, UFSAR ,

9.22 mentions -the function of the valves, but does not indicate that they are ,

safety related. The CCP system operation will occur as described-in UFSAR 9.2 except for isolation of the recirculation line. the line servos no other function to the CCP pumps (i.e., bearing cooling or real water). No unreviewed safety questions exist.

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B3ever Valloy Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 10 of 152 CHANGE TITLE TEMPORARY MODIFICATION - CAGGED (2CNM-RV-100A)

CHANGE DESCRIPTION (2CNH-J21A) Air Ejector Inlet Relief Valve, [2CNH-RV-100A) has excessive seat leakage. This temporary modification installe a gag in the relief valve until repairs can be made.

SAFETY EVALUATION

SUMMARY

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(2CNM-HV-100A) protects condensate piping on "A" air ejector. When it is teolated, failure of this piping will not affect any accident. The condensate system overall integrity will not be challenged because other overpressurization protection exists. No unroviewed safety questions exist.

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Bsaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 11 of 152 CHANGE TITLE OST 2.15.3 and 4 "CCP VALVE POSITION VERIFICATION" CHANGE DESCRIPTION The CCP valve position verification OSTs were revised to eliminato  ;

(2CCP*DV100-1,2) inlet, outlet and instrument root valves. (2CCP*DV100-1,2) i (CCP Pump Recirculation Valves) are closed or isolated. The summary of safety evaluation evaluates operation with [2CCP*DCV100-1,2) closed or isolated.

SAFETY EVALUATION SUMHARY UFSAR 15 accident analysin states that following an analyzed event initiation, a . reactor trip occurs and possibly an SIS or CIB. The RCS is then cooled to het standby by AFW or Safety injection. The function of CCP system (UFSAR 9.2.2.1) is to supply sufficient cooling water to enable a cold shutdown by ,

RHR. This is a long term function and not short term followir.g the event. l Although UFSAR Table 3.9B-19 places these valves in a safety category, UFSAR 9.22 mentions the function of the valves, but does not indicate that they are safety related. -The CCP system operation will occur as described in UFSAR 9.2 _

l except for isolation of the recirculation line. The line serves no other l function ~ to the CCP pumps (i.e., bearing cooling or soal water). No unreviewed questions exist.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 12 of 152 l

CHANGE TITLL j ISOLATION OF (2GNS-PCV101A,D) DY CHANCING (2CNS-310,311,320 AND 321) 70 NSA CLOSED CHANGE DESCRIPIlOJJ Degasifier Hydrogen Supply Headers Pressure Regulators [2GNS PCV100A and B) leak by and cannot be repaired. As a means of isolation, Degasifier Hydrogen Supply Isolation Valves [2GNS-310,311,320 r.nd 321) were changed from NSA open to closed.

AbFETY EVALUATION

SUMMARY

! solation of hydrogen purge will not effect UFSAR 11.3.1 design requirements l of 10CFR50 Pnnex to Appendix 1. Isolation of H2 purge is also bound by i UFSAR- 15.7.1. Caseous Waste compressors will still shut down on high oxygen j concentration as described in Uv5AR 11.3.1. Purging will take place by manually opening of the isolation valve. Isolating hydrogen purge will not effect oxygen monitor described by Technical Specification 3.3.3.10. No unreviewed safety questions exist. i l

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B3avor Valloy Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 13 of 152 CHANGE TITLE TOP 2-90-4 *D" MOISTURE SE?ARATOR REHEATER REHEAT STEAM ISOLATION CHANGE DESCRIPTION

. A new temporary procedure was generated to isolate reheat steam to the "D" HSR to reduce MSR tube damage and to reduce the probability of turbine blado damage from tubing debris.

SAFETY EVALUATION

SUMMARY

loolating reheat steam will not impact any turbine trip function and will not ,

increase the consequences should the turbine fail to trip as described in l UFSAR 15.0, 15.1 and 15.2. Steam entering the low pressure turbine will be i superheated from the 'C" HSR and does not create an unanalyzed accident as described in UFSAR 10. No unreviewed safety questions exiet.

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D3Avor Valley Pow 3r Station Unit 2 1990 Report of Tacility Changes, Tests, and Experiments Page 14 of 152 CHANGE TITLE ISOLATION OF REHEATER DRAIN RECEIVER LEVEL CAUCES (2flDil-LG100A, B,C,D) AND

- HOISTURE 6EPARATOR DRAIN RECEIVER LEVEL CAUCES [ 2ilDH-LO115 A, B, C, D )

CHANCE DESCRIPTION Isolation valves for (2HDH-LC100A,B,C,D) and [2HDH-LG115A,B,C,D), Reheat Drain Receiver and Moisture separator Drain Receiver local gauges were changed from l NSA open to closed. '

SAFETY EVALUATION

SUMMARY

Control' Room indications and alarms will remain operable while these local indicators are isolated. All equipment upstream of the feedwater isolation valves are not safety related per UFSAR 10.4.7.2. The local gauges will be available to be valved in on an as-needed . basis. Hoisture separator l performance is unaffected with these gauges isolated. No unroviewed safety questions exist.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimento l

Page 15 of 152 CHANGF TITLE ISOLATION OF FIRST, THIRD, FOURTH, FIFTH POINT TEEDWATER HEATER LEVEL GAUGES (2FWS-LG-103A,B, 120A,B, 122A,B, AND 124A,D)

CHANGE DESCRIPTIOfi Isolation valves for (2FWS-LG-100A,B, 120A,B, 122A,B, and 124A,B), lot, 3rd, 4th and 5th point feedwater heaters were changed from NSA open to closed.

SAFETY EVALUATION

SUMMARY

UFSAR 10.4.11.3 states that extraction steam has no safety function and its failure will have no effect on the equipmant. Contrni Room indications and alarms will remain unaffected while these local indientors are isolated. The local gauges will be available to be valved in on an as-needed basis.

Feedwater heater performance will be unaffected with the gaugen isolated. No unreviewed safety questions exist.

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Bsavor Valloy Power Station Unit 2 1990 Report of Pacility Changes, Tests, and Experiments Page 16 of 152 CHANCE TITLE TOP 2-09-37 ' TEMPORARY PILTER INSTALLATION ACROSS STARTUP FEED PUMP LUBE OIL J COOLERS" CifANGE DESCRIPTION A new temporary procedure was generated to place in service, operate and remove from porvice a temporary sidestream filter.- The filter is necessary to assist in cleanup of the Turbine Plant component cooling Water System. l t

S AFE1 Y_ ggybLMATION SUHMARY l The components -used will have the same or greater rating as design, therefore f it would be equivalent to that previously evaluated in UFSAR 3.4 and 9.2.7.

The temporary filter will provide for cleanup of the system to reduce any possibility of a malfunction of equipment. The filter is installed in parallel with the permanent filter, maintaining that component's availability. No unreviewed safety questions exist.

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B3avor Valley Power Station Unit 2 1990 Report of Facility changes, Tests, and Experiments Page 17 of 162 c!!ANCE TITLE OM 2.30.4.L "SWS SILT AND CORBICULATE CONTROL PART A, FLUSilING SWS PUMPS SEAL WA1ER SUPPLY LINE CllANGE DESCRIPTION A new procedure was developed to flush the bypass seal water line using-filtered water. The filtered water will be supplied from a temporary hose connected from the sand filter pump discharge to the inlet of (2SWS*83A(D)).

The flush is to be performed to remove silt and clam embryos. This procedure is based in part on TOP 2-89-36, " Flushing the SW Pumps seal Water Supply Lines".

RAFETY 2 VALUATION SUHMARY A review of UFSAR 15 determined that no accident was evaluated that would be initiated by a loss of service water pumps. UFSAR 14.1.14 (Unit 1) takes credit for- the flood proof doors of intake structure. The procedure contains i a caution that if the river elevation approaches 730 ft., the procedure for flushing should be terminated so the jumper used for flushing will not block the flood door. The $W pump may be operated during the flushing operation is described in UFSAR 9.2 with the exception that seal water will be provided by a temporary hose. The procedure requires the opposite train to be operable prior to starting the fluoh to satisfy Technical Specification 3.7.4.1. No unreviewed safety questions exist.

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Daavor Valloy pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 18 of 152 CHANGE TITLE TOP 2-89-36 " FLUSHING THE SW PUMPS SEAL WATER SUPPLY LINES" CHANGE DESCRIPTION t

A new temporary procedure was generated to flush the servico water pumps seal water supply lines using filtered water to remove silt and clam ombryou.

SAFETY EVALUATION

SUMMARY

A review of UFSAR 15 determined that no accident was evaluated that would be ir.itiated by .a loss of service water. UFSAR 14.1.1.4 (Unit 1) takes credit for flood proof doors of the intako structure- cubicles. The procedure ,

contains a caution to terminate the procedure if river water level approachos 730 feet and to remove the flush hose. The procedure uses non-radioactive water to flush non-radioactive SWS pump seals. A temporary block will be installed on (2SWS-A0V118A(118B)) to block open the valve to ensure seal water will be supplied when [2SWS*SOV130A (1308)) is de-energized to open. This will ensure the SWS pump will receive normal seal water during various flushes. The procedure requires the opposite train to be operable prior to the flush to satisfy Technical Specification 3.7.4.1. No unroviewed safety questions exist.

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Doaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments l Page 19 of 152 GHhHQE TITLE OM 2.30.4L *SWS SILT AND CORDICULATE CONTROL PART U FLUSilING ROD CONTROL AREA ACU GlihE9E. DESCHI PTIOff A now procedure was developed to flush the rod control area air conditioning unit cooling coils supplied from service water system using domineralized water. The coils are normally isolated and a*e fluched to remove any silt or clam embryos.

SAFETY EVhkUAT_ON I SUHMARY In the event the rod control area ACU failed during the procedure, the assumptions used in UFSAR 15.7.1 would remain valid and the accident would be enveloped by UFSAR 15.7.1. In the evont the temporary hose failed, the demineralized water would be collected by floor drains. The rod control area ACU are 100% capacity units. In the event an omorgoney occurrod, the non-flushed ACU would automatically start and perform its function as described in UFSAR 9.4.12.2. Tl.e service water system will be capable of performing ite function as required by Technical Specification 3.7.4.1.

Deaver Valley Power Station Unit 2

. 1990 Report of Facility Changes, Tests, and Experinnent s I

Page 20 of 152 CilANGE TITil FAILED RELAY 69-FWOND pHANGE DESCRIPTION Plant desires centinued operation with relay 69-FWSHD, on the main foodwater pumps low suction / oil pressure trips, failed until relay can be repaired.

SAFETY EVALUATION SUMMABX Loss of feedwater flow / reduced flow will not effect reactor trip function assumed in UFSAR 15.2.7.1. All of the main FWS safety functions described by UFSAR 10.4.7.3 are unaffected. The AFWS ensures a sufficient supply of cooling water for safe shutdown. The condensate and non-safety related portions of FWS are automatically isolated from the steam generator within 5 seconde following a FWI signed. This signal is unaffected and will continue to trip the FW pumps. The Aux FW Pump automatic actions described by UFSAR 15.2.6.1 are also unchanged. The Aux FW pumps and flow path described by Technical Specification 3.7.1.2 are unaffected. Also, the FWI signal described by Technical Specification 3.3.2.1 is unaffected. No unreviewed safety questions exist.

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Bsavor Valley Power Station Unit 2 1990 Report of racility changes, Tests,-and Experiments Page 21 of 152 CHANGE TITLE

-TOP 2-09-38 'SWS DACK TLUSHING or CCP HX's" CHANGE DESCRIPTION A new TOP was generated to backflush any CCP heat exchanger on the SWS side to  :

l dislodge clams or debrio.

SAFETY EVALUATION

SUMMARY

The proceduts does not alter or reduce performance of the CCP and SWS systems that could increase probility of accidents described in UFSAR 15. only_one CCP heat exchanges- can be isolated by procedure (while in modes 1-4). This ensures that the_other two heat exchangers are inservice to remove heat loads descrihed by- UFSAR- 9.2.1 and 9.2.2. In modes 5 & 6 at least one CCP. heat exchanger will remain in service. In the event any activity is detected by CCP or SWS radiation monitors, a step in the procedure instructs an operator to close the drain salve of-the heat exchanger being flushed. The margin of safety described by Technical Specifications 3.7.3.1 and 3.7.4.1 are maintained during this procedure. No unreviewed safety questions exiet. ,

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 22 of 152 CHANCE TITLE OST 2.30.20 "CCP Hx PERFORMANCE TRENDING AND RSS Hx DRY LAYUP CHECK - PART B" CHANGE DESCRIPTIOH A new procedure was generated to check RSS heat exchanger service water inlet piping for accumulation of water.

SAFETY EVALUATION

SUMMARY

Opening drain isolation supply to RSS Coolers valve (2SWS*906) to check for drainage cannot physically effect the tanks considered for radioactive release by UFSAR 15.7.3. In the vent [25WS*906) could not be closed and water continued to drain, the water would be collected in floor draine and directed to safeguards sumps. The service water system is operated as dosetAbed by UFSAR 9.2, except for the opening of (2SWS*906, 104). General Design Criteria 46 permits functional testing of componenta to ensure leak tightness. The procedure contains a caution to close (2SWS*104) prior to starting a RSS pump, should an ESF actuation occur, so the concerne for entry into Technical Specification 3.03 can be 6 voided. No unreviewed safety questions exist.

Banvor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 23 of 152 GilANGE TITLE TOP 2-90-7 " RECIRCULATION OF (2WTD-TK23] THROUGH A TEMPORARY DEMINERALIZER FOR CLEANUP" QE NCE DESCRIPTION A new temporary procedure was developed to cleanup and recirculate  ;

domineralised water storage tank (2WTD-TK23] through a temporary dominocallter I train and supply domineralized water to system demand.

SAFETY EVALUATION

SUMMARY

Should. a temporary home rupture, the hose is isolable. Pumps will 1 automatically trip on low discharge pressure to prevent rapid pump down of

( 2 WTD-TK2 3 ) . . A hose rupture is enveloped by flood analysis. . Safety related systems are not involved and the temporary hose is not routed through any safety related area. No Technical Specification are involved. No unreviewed safety questions exist.

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Boaver Valloy Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 24 of 152 pHANGE TITLE TEMPORARY HODIFICATION - DEFEAT OF (2HVP-ACU211A & B) LOW TEMPERATURE TRIP gilANGE DESCRIPTION A defective volume booster (2HVP-TX21] prevent proper oporation of temperature switches (2HVP-TS21A & D). This prevents the Auxiliary Building and Waste Handling Building Air Handling Units [2HVF-ACU211A & B) from operating. Leads were lifted. on temperature switches (2HVP-TS21A & B) to allow operation of

[2HVP-ACU211A & B).

3AEETY EVALUATION

SUMMARY

-Defeating the low temperature trip on (74VP-ACU211A & B) may increase the possibility of a failure of the air conditioning - heating colla. However,

[2HVP-ACU211A & B) are not safety related and are not required for accidents as described by UFSAR 9.4.3, 9.4.3.1, 9.4.3.2. Additionally no Technical Specification are involved. No unreviewed safety questione exist.

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Boaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 25 of 152 C}lANGE TITLE TOP 2-90-8

• REMOVAL OF REHEAT STEAM TO [2 MSS-H218)"

CHANGE DESCRIPTION A new temporary procedure was developed to isolate the "D" MSR by removing high pressure reheat steam. Itolation of the MSR is necessary due to leaking tube bundles which could result in increased tube damage and turbine blade damage.

SAFETY EVALUATION

SUMMARY

Isolating reheat steam will not impact any turbine trip function and will not l increase the consequences should the turbine fail to trip as describe in UFSAR.

15.0, 15.1 and 15.2. Steam entering the low pressure- turbine will be

- superheated from the "A" MSR -and doos not create an unanalyzed accident as i deperibed - in UFSAR 10. No Technical Specifications are affected by this procedure. -No unroviewed safety questions exist.

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Bsaver Valloy Powsr Station Unit 2 l 1990 Report of Facility Changes, Tests, and Experiments Page 26 of 152 CHANGE TITLE TEMPORARY MODIFICATION - JUMPER ACROSS FAILED OPEN LIMIT SWITCH ON

[2CCP*AOV107C)

CHANGE DESCRIPTION Reactor Coolant Pump Thermal Barrier Supply Valve [2CCP*AOV107C) has dual I indication and will not stay opon. A jumper was placed across the open limit switch so the valve will stay open.

SAFETY EVALUATION

SUMMARY

(2CCP*AOV107C) is required to close on high pressure or high flow to safely contain high reactor coolant pressure as described by UFSAR 9.2.2.1.3. The automatic closure of (2CCP*AOV107C) is unaffected by the temporary modification and will still close on high flow or preocure. No Technical ,

Specifications are affected since the valve will function to isolate CCP 4 system - from high pressure reactor coolant as designed. No unreviewed safety questions exist.

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B3 aver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 27 of 152 CHANGE TITLE OH 2.9.4H " ALIGNING THE MAIN STEAM VALVE ROOH AND INSTRUMENT AIR COMPRESSOR FLOOR DRAINS TO CATCH BASIN 23" CHANGE DESCRIPTION A new procedure was. generated to align the Main Steam Valve Room (HSVR) and containment Instrument Air compressor (IAC) floor drains to catch basin 23 )

[CD-23). This will allow non-radioactive floor drains to be directed to the 1 catch basin instead of being processed in the liquid waste system.

SAFETY EVALUATION

SUMMARY

1 Aligning floor drains to cat.cn Lesin 23 does not modify plant systems. The alignment uses installed' valves and piping provided for this purpose.

Aligning the floor drains-to catch basin 23 vill not affect the performance of any safety related system. After the floor drains are aligned, Radeon will sample at least shiftly for radioactivity in -the drains. If-radioactive content of the drains exceeds limits, then stepo are provided to return the floor drains to the tunnel sump. No unreviewed safety questions exist.

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Bsavor Valloy Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments )

Page 28 of 152_  ;

I Cl[ANGE TITLE TOP 2-90-11 " BYPASS OF INSTRUMENT AIR DRYER AND RECEIVER" CHANGE DESCRIPTION A new temporary procedure was generated to bypass instrument air dryer anti receiver in- order to perform maintenance work on valves and to install a new filter under DCP 1354. Reduced capacity will be available to both instrument air and station air. Instrument air quality will be degraded.

3

-SAFETY EVALUATION-

SUMMARY

No safety systems are involved. All components are in the Turbine Building where missile generation from a rupture type failure will not cause damage to any safety related equipment. The procedure is not to be performed in Modes 1, 2, 3 or 4 so that loss of air would not cause a loss of feedwater control or closing of the main steam stop valves. No Technical Specifications are affected by this precedure. No unreviewed safety questions exist.

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Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 29 of 152 CHANGE TITLE TOP 2-90-17 "[2FPD-TK22] CO2 PURGE TO ALLOW VALVE MAINTENANCE" QH&HgE DESCRIPTION A temporary procedure was generated to prepare 30 Ton CO2 Unit (2FPD-TK22) for valve maintenance. Temporary piping will be installed to depressurize the unit to a CO 2 truck. Blank flanges will be installed so that the 24 Ton Backup CO 2 Unit can be placed in service while maintenance is being performed.

SAFETY EVALUATION SUMHARY

%he y tank is isolated from the oystem prior to depretsurization through temporary piping. Failure of the piping will not affect the remaining system. The flange is installed so that the backup 24 ton CO2 unit can function as its designed to protect the areas served by the system. Fire watches will be implemented when the flange is being installed. Fire protection is not Technical Specification related. No unreviewed safety questions exist.

Beaver Va?. ley Power Station Unit 2 1990 Report of Facility Chan;es, Toute, and Experiments Page 30 of 152 CilANGE TITLE TOP 2-90-14 "[2FPD-TK23) CO2 PURGE TO ALLOW VALVE MAINTENANCE" CHANGE DESCRIPT1QH A temporary procedure was generated to prepare 10 Ton CO2 Unit (2FPO-TK23) for valve maintenance. Temporary piping will be installed to depressurize the unit to a Co p truck. Blank flanges will be installed so that the 24 Ton Backup CO 2 Unit can be placed in service while maintenance le being performed.

SAFETY EVALMA. TION

SUMMARY

The CO 2 tank is loolated from the system prior to depressurization through temporary piping. Failure of the piping will not affect the remaining system. The flange is installed so that the Backup 24 Ton CO2 Unit can function aa designed to protect the areas served by the system. Fire watches will be implemented when the flange la being installed. Fire protection 10 not Technical Specification related. No unreviewed safety questione exist.

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Baaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page'31 of 152 QHAUGE TITLE OH-2.6.4W " ISOLATING A REACTOR COOLANT LOOP" CHANGE DESCRIPTION An initial condition in procedure 2.6.4W, " Isolating a Reactor Coolant Loop",

which requires the primary grade low pressure. alarm to be operable, was eliminated. This procedure requires all dilution water supply to the boric acid blender to be closed and administratively controlled. This lineup eliminates the possibility of the primary grade water alarm from being an indicator of a dilution accident.

SAFETY EVALUATION

SUMMARY

This change reduces the probability of a dilution accident doscribed in UFSAR 15.4.6. The procedure requires. [2CHS-27 and 91) to be isolated and administratively controlled: as a result there is no effect on safety. system perforrance and places the plant in a more conservative condition. This change places the plant in a configuration which is required by Technical Specification 3.9.1, 3.1.9.2 and 3.3.11. No unreviewed safety questions exist.

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--Bsavor Valley.Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 32 of 152 CHANGE TITLE 2 TOP 90-15 "[2 CSS-E21A, B) TURDINE PLANT CCS Hx TURDINE SIDE PERFORMANCE VERIFICATION" CHANGE DESCRIPTION A new temporary procedure was generated to throttle the CCS heat exchanger SWS outlet . valve to obtain several D/P's and corresponding flows. This information. provides data to the Engineering Department to evaluate heat

- exchanger performance. The procedure installs temporary pressure indicators to allow for more accurate data.

SAFETY EVALUATION SUMMART-No_ systems that_ are safety related are affected by this procedure. Any failure of temporary pressure indicators can be easily isolated if necessary.

Failure of the temporary pressure indicators cannot initiate any design bauis I No accident. No technical Specifications are affected by thin new procedure.

unreviewed safety questions exist.

Baaver Valley Powar Station Unit 2 1990 Report of Facility changes, Tests,. and Experiments Page 33 of 152 CHANGE TITLE 2 TOP 90-25 " CONTAINMENT BACKUP FIRE PROTECTION DURING TYPE "C" TESTING" CHANGE DESCRIPTION A temporary procedure was developed to connect a temporary hose from fire hydrant H-12 to a temporary valve manifold attached to Containment Home Rack (HR-268). This . will provide backup fire protection to the containment hose j racks when normal supply of protection water is not available due to type "C" l testing.

l SAFETY EVALUATION

SUMMARY

l No safety systems are affected by this change. Fire protection water is still ]

available to containment hoce racks through the temporary modification.

Failure of the temporary hose is enveloped by UFSAR 9.5.1.2.3.2. The RHR/ Penetration and CNMT Iodine Filtor Fire Protection Systems are still available and unaffected by this procedure. Failure of the temporary modification will not affect plant response since the plant is shutdown during the performance of this procedure and backup fire protection is available.

Fire protection- is not Technical Specification related. No unroviewed safety questions exist.

Beover Valley Power Station Unit 2 l 1990 Report of Facility Changes, Tests, and Experiments I

Page 34 of 152 QRANGE TITLE 2 TOP-90-23 " SUPPLYING BOTH SWS HEADERS WITH ONE SWS PUMP" CHANGE DESCRIPTION A new temporary procedure was developed to allow one operable SWS header to supply water to the opposite header in order to facilitate partial SWS header clearances of the priority train header during Modes 5 and 6. Tts priority train SWS or SWE pump will supply the operable diesel generator, charging pump and inservice CCP heat exchanger.

,AFETY EVALUATION

SUMMARY

A loss of service water is not probable since standby SWS is available or another SWS pump is available to supply SWS flow. The vales used in the procedure are administratively controlled to maintain a " hardened" train of SWS flow for operable components. SWS and CCP cooling to RHR systems are continuously maintained by priority train SWS header, hence no increase of occurrence of DBA will occur. Priority trait SWS flow is maintained to the operable diesel generator and charging pump cooler to met Technical Specification 3.8.1.2 and 3.8.2.2. No unreviewed safety questions exist.

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 35 of 152 QHANGE TITLE OM 2.11.4.C " FILLING REACTOR REFUELING CAVITY" CHANGE DESCRIPTION Procedure OM 2.ll.4.C, " Filling Reactor Refueling Cavity" was revised to allow opening a loop stop valve while the reactor vessel is defueled and the water level in the vessel is less than loop elevation. This will allow water to enter a leap and fill the loop as the water level is being raised in the reactor vessel.

SAFETY EVALUATION

SUMMARY

Filling the vessel, loop and cavity will not affect assumptions and radiological consequences of the fuel handling accident described by UFSAR 15.7.4, since all of the fuel is in the spent fuel pool with no fuel in the reactor vessel. Filling the loop while filling the cavity will not affect the fuel i the spent fuel pool. Loop stop valve vendor stated that opening the vales when no water exists in the loop will not damage the valves and is permitted. Seal injection flow will be provided to the RCP in the loop to be filled when water level is less than 3 feet below the flange. Technical Specification 3.4.9.3 is not af fected by this ; rocedure. No unreviewed safety questions exist.

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Bsavor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 36 of 152 CHANGE TITLE 2 TOP-90-21 "HIGH-VOLUME AIR SUPPLY TO CONTAINMENT" gitANGE DESCRIPTION A new temporary procedure was written to install, operate and remove a temporary modification which supplies high volume compresoed air to containment during periods in which containment integrity may be established for refueling operations. This air supply is provided for normal station air uses during refueling outage.

SAFETY EVALUATION SUM M During. refueling operations only single isolation in requited. Any single ,

failure of this modification will not prevent containment isolation in the  !

event of a refueling accident. The temporary air compressor configuration is considered a closed pressurized system, thus not providing a pathway from containment to outside and satisfying the basis of Technical Specification 3.9.4. No unreviewed safety _ questions. exist.

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Baavor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 37 of 152 CHANGE TITLE OH 2.6.4W " ISOLATING A REACTOR COOLANT LOOP" CHANGE DESCRIPTIOJ,{

Part. B was added to procedure OH 2.6.4W, " Isolating a Reactor Coolant Loop" to provide instructions to allow the isolation of all three reactor coolant loops to allow flexibility in plant configuration for maintenance during refueling otuages.

SAFETY EVALUATION

SUMMARY

This procedure change closes the third reactor coolant loop to the reactor vessel. This reduces the water volume available to mitigate a dilution-accident, however, the procedure requires that the refueling cavity be filled to greater than 23 feet above the reactor vessel flange to provide a large volume of borated- water to mitigate a dilution. Site Engineering.has determined that RCS loop volume is not required if the reactor cavity contains greater than 23 feet of water, and is bound within the accident analysis of UFSAR 15.4.6 and the standard review plan. No Technical Specifications are affected. No unreviewed safety questions exist.

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Boaver Valley Power. Station Unit 2

. 1990 Report of Facility changes, Tests, and Experiments Page 38 of 152-CHANGE TlILE 2 TOP-90-10 " MOISTURE SEPARATOR REHEATER TUBE LEAK TEST" CHANGE DESCRIPTION A new temporary procodure was generated to locate small leaks in the Moisture Separator Reheater (MSR) tube bundle after large leaks have been found and repaired by an air _ test. The MSR head chamber is filled with domineralized

. water. Air will be drawn by vacuum into the leaking valves and visually observed.

SAFETY EVALUATION

SUMMARY

A rupture of water supply hose or failure of temporary MSR manway cover would be enveloped by assumptions of the flood analysis. The Test (demineralized) water is not potentially radioactive. No Technical Specifications are affected by this change. The MSR is not inservice during test and will not be overpressurized. No unreviewed safety questions exist.  ;

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B3 aver Valley-Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 39 of 152 CHANGE TITLE 2 TOP-90-13 " TEMPORARY AUX STEAM SUPPLY TO TURBINE LUBE OIL COOLERS CHANGE DESCRIPTION A new temporary procedure was written to install and remove temporary tubing to supply auxiliary steam to the turbine lube oil coolers. The-steam will be used for warming and agitation of the cleaning fluid being used on the lube oil side of the coolers. The steam condensate and cleaning fluid will be removed from the coolers and lube oil reservoir by subsequent cleaning and oil flushing.

SAFETY EVALUATION

SUMMARY

The teamporary-line does not' connect to or is located in close proximity to any safety related euqipment. The temporary line and failure of the line will not effect the steam jet-air ejector radiation monitor. Therefore the radiation monitor will function to alert the operator of a steam generator tube failure. Auxiliary steam -is normally non-radioactive, if the line ruptured steam would be collected by turbine building sumps. The radiation monitors and HELB trip valves will function an described by UFSAR 10.4.10. No unreviewed safety questions exist.

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-Baaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 40 of 152 j

EllANGE TITLE OM 2.6.4D " FILL OF AN ISOLATED RCS LOOP, 2.6.4.N REACTOR COOLANT SYSTEM '!

ISOLATED LOOP RECOVERY" l l

CHANGE DESCRIPTION  !

Procedures OM 2.6.4.D and 2.6.4N were revised to recover the RCS loop if all thret loops were isolated. A recent change has been made to OM 2.6.4.W,

" Isolating a Reactor Coolant Loop", to allow isolating all three loops.

SAFETY EVALUATION

SUMMARY

All dilution paths are isolated by procedure and Technical Spectication 3.9.1, therefore a dilution described by UFSAR 15.4.6 is not assumed to-occur. The-procedures mout be performed prior to draining down the. reactor cavity to less than 23 feet above the reactor vessel flange. An engineering analysis determined that plant conditions fall within the bounds of dilution accidents analyzed by U FS A'A 15.4.6 and the Standard Review Plan. This change does not impact Technical Specification 3.9.1 or 3.1.2.9. No unreviewed safety questions exist.

Baaver. Valley Powar Station Unit 2 1990 Report of Facility Changes, Tes*;s, and Experiments Page 41 of 152 CHANGE TITLE TEMPORARY MODIFICATION - INSTALL PROXIMITY LIMIT SWITCHES IN SERIES WITH THE ELECTRICAL AND MECHANICAL TORQUE SWITCHES FOR THE FUEL TRANSFER CART CHANGE DESCRIPTION Due to the extreme difficulty in adjusting the electrical torque switch for the fuel transfer cart during operation (which is not functioning correctly) a new control interlock was installed in series with the proximity switch.

SAFETY EVALUATION

SUMMARY

Failure of the fuel transfer cart to travel vill not change any assumptions or consequences of a fuel handling accident. Failure of the cart to travel will not impact the mainpulator crane, fuel fullding crane or cause the fuel pool to drain to uncover spent fuel. No Technical Specifications are affected. No unreviewed safety questions exist.

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Bsaver Valley Power Station Unit 2 1990 Report of, Facility Changes, Tests, and Experiments Page 42 of 152 CHANGE TITLE 20M-51.4.D " STATION SHUTDOWN .COOLDOWN FROM HOT SHUTDOWN (MODE 4) TO COLD SHUTDOWN (MODE 5)

CHANGE DESCRIPTION The Unit 2 Gaseous Waste- Disposal System (GWDS) alignment per procedure 20M-51.4.D Attachment I will purge bydrogen and RCS fission gasses, following plant shutdown to Mode 5 conditions, using permanent installed Nitrogen supply equipment, permanent plant GWDS piping and normal flow paths as described in UFSAR 11.3.2, from the PRZR, PRT, VCT, degasifiers to the GWDS delay beds, then through the GWDS surge tank to either Unit 1 GWS decay tanks or Unit 2 GWS storage tanks for processing. When hydrogen (H p ) and fission gasses levels are less than requirements of approved Radeon calculation package ERS-ATL-90-019 (Rev. 0), then the delay bede exhaust is rerouted to the sweep gas system for continuous exhausting to Unit 1 GWS exhaust daring the platn outage. The overhead gas compressors and GWS surge tak will be isolated and j pressurized with a temporary nitrogen (N2) gas bottle (with proper rigging). l l

This alignment of the delay beds to the sweep gas system will reduce waste gas l storage and processing using the gaseous waste disposal system during plant outages. This_ new technique of removal of RCS nitrogen gasses and fission )

gasnes during outages will significantly reduce waste gas volume to the GWDS 1 storage tanks and allows utiliting the sweep gas system to dispose nitrogen gasses from the RCS, in compliance with radiological requirements of the BV-2 Technical Specifications.

I SAFETY EVALUATION

SUMMARY

When__ waste gas through the above listed GWDS flow path to the delay bede haa

-hydrogen (H2 ) levels less than 4% and gaseous activity less than requirements of -the1 approved Radeon calculation package, then the delay beds exhaust is rerouted to sweep gas system for continuous exhausting to Unit I

~CWS exhaust, until the.H 7blanket is to be restored to the VCT, PRT and PRZR l in _- accordance with procedure OH-2.6.4.E " Fill and Venting of RCS".- This

continuous' purge will also be terminated it Unit I rad monitor (RM-1GW-108A,.

108B or 109) s9nsing Unit 2 sweep gas stream alarm at high or high-high setpoint. . This continuous GWDS discharge path is different from that described- in UFSAR 11.3.2.1, which describes GWDS discharge being directed to either' Unit 1 GWS decay tanks or Unit 2 GWDS storage tanks. No significant increase of radioactivity to- the environment will occur, since proceduro l 2OH-51.4.D attachment 1 ensures the requiremetne and assumptions of approved Radeon calculation package are met. The approved Radeon calculation package l

ensures all UFSAR 11.3.1 and 11.3.2 bases and design analysis are met, and ensures radiological effluent releases are less than 0.001% of the Tech Spec effluent release requiremetna. Therefore the results of this calculation package -are within- the boundaries set forth in the BV-2 Environmental Protection Report and BV-2 Tech Specs. No significant change in effluents activity levels wil- occur while discharging GWDS gasses to the sweep gas system in accordance with 20M-51.4.D attachment 1.

L The oxygen (02) m nitors for the surge tank will not be capable of performing their automatic function of stopping the gas compressors on detection of high 0 2 in the surge tank, as described in UFSAR section 11.3.1. However according to Attachment 1 of procedure 2OM-51.4.D.

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Beaver Valley Power Station Unit 2 i

1990 Report of Facility Changes, Tests, and Experiments i Page 43 of 152 The probability of the event described in UFSAR 15.0.2, 15.7.1 and 11.3 will not be increased for the following reasons. Malfunction of ad moni. tor (2GWS-RQIl03) is normally addressed by Operating Manual alarm response procedure and Radeon procedures. To ensure radioactive effluent releases to atmosphere via sweep gas system are within Tech Spec limits, any increase of gaseous activity occurring in Unit 2 GWDS flow path will be detected by Unit I rad monitors (RM-lGW-108A, 108B or 109) and procedure 20M-51.4.D attachment 1 and the caution tage affixed to those rad monitor indications will require this GWDS alignment to be promptly directed back to the GWDS surge tank to the Unit 2 GWDS storage tanka or Unit 1 GWS decay tanks for normal processing until the activity levels are returned to their previous levels required by Attachment 1 in the Unit 2 GWDS.

The changes to this procedure ensure the RCS and GWDS are pruged of H2 and radioactive gasses, so that the safety margin of basis of Tech Specs 3.11.2.6 and 3.11.2.1 are maintained. The overhead gas compressors and GWS surge tank will be isolated and pressurized with a temporary nitrogen (Np) gas bottle (with proper rigging), thus avoiding a potential explosive environment in the surge tank, when H 2 gas is restored to the GWDS during platn startup in accordance with procedure OM-2.6.4.E. The tmeporary jumper is installed and maintained ina ccordance with the requiremetna of approved station administrative procedure NGAP 7.4 " Temporary Modifications". The temporary modification will not challenge the design of the GWDS piping or surge tank, therefore no failure of the GWDS should occur.

The procedure ensures the PRT rupture disks will remain intact by ensuring PRT pressure stays between < 65 peig and >0 psig during the purging and drain down of the PRT, VCT and degasifiers. In addition the PRZR is purged when PRZR temperature is less than 200F, so that the PRZR-RCS 200F delta-T limit is not exceeded, since this purging process will cause PRZR surge line insurges and outeurges. All of the GWDS equipment utilized by this procedure is operated within its design functions as described in UFSAR 11.3.2.

In summary, it was determined that no unreviewed safety questions are involved with this safety evaluation.

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Bsaver' Valley Powsr Station Unit 2 1990 Report'of Facility Changes, Tests, and Experiments Page 44 of 152 CHANGE TITLE Security Computer System Multiplexor Board Modification

~~ CHANGE DESCRIPTIOji L Changes were made to the DI and CR boards in each multiplexor.. This was.to correct an outstanding deficiency and to make the system perform as originally designed.

SAFETY EVALUATION

SUMMARY

This change affects only the security system computer. There is ru) interaction with~any system or component important to-safety. No unreviewed safety question is involved._ No Technical Specification bases are affected by

-this change. No change to'the FSAR is required.

I Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 45 of 152 CHANGE TITLE Disable CCW Hi Flow closure Logic to-2RCS-P21C Thermal Barrier Flow Valve 2CCP-AOV107C for corrective Maintenance CHANGE DESCRIPTLQH

.This was a one time _only procedure change to 2LCP-15-F107C which is Component Cooling Water Reactor Coolant Pump (2RCS-P21C) Thermal Barrier Flow Loop 2CCP-FlO7C Calibration. This change prevented the C Reactor Coolant Pump Thermal Barrier Flow Valve'2CCP-AOV107C automatic closure due to HIGH flow signal from 2CCP-F107C loop which was out of service for corrective

-maintenance. The system was returned to NSA after maintenance was completed.

SAFETY EVALUATION

SUMMARY

The probability or consequences.of an accident described in the FSAR will not be increased or created since,- although the CCW loop Hi Flow instrumentation 11e inoperable during maintenance, the CCW flow is maintained to the thermal barrier. .Hi pressure and hi temperature signals to the valve are maintained.

The possibility of an accident or malfunction of a different type than any

-previously evaluated in_the-FSAR is not created since the system function is maintained by hi pressure and hi temperature signals and process radiation.

monitoring. No Technical Specification bases are affected since the component 3' cooling water system remains in operation. Since this is a temporary

' modification,-no changes _to the FSAR are required.

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B3 aver Valley Powar Station Unit 2 1990 Report of Facility changes, Tests, and Experiments Page 46 of 152 CHANGE TITLE Defeating Hi CCW Flow Close Signal to RCS-P21A Thermal Barrier Isolation Valve 2CCP-AoV107A for Performance of 2LCP-15-F107A During Mode 5 or 6 or for Corrective Maintenance in an operating Mode CHANGE DESCRIPTION This change was made to allow the procedure to be performed for corrective maintenance purposes only during plant modes 1, 2, 3, or 4. The loop would be considered out of service and.not operable. These same changes also allow performance in modes 5 or 6 without-cycling isolation valve 2CCP-AoV107A.

Double verification is used to connect and disconnect the jumper wire.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created since, although the CCW loop Hi Flow instrumentation is inoperable during maintenance, the CCW flow is maintained to the thermal barrier. Hi pressure and hi temperatura signals to the valve are maintained.

The possibility of an accident or malfunction of a different type than any

-previously evaluated in the FSAR is not created since the system function is maintained by hi pressure and hi temperature signals.and process radiation monitoring. No Technical Specification bases are affected since the component cooling water system remains in operation. Since this is a temporary modification, no changes to the FSAR are required.

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Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 47 of 152 CHANGE TITLE Defeating Hi CCW Flow Close Signal to RCS-P21B Thermal Barrier Isolation Valve 2CCP-AOV1078 for Performance of 2LCP-15-FlO7B During Mode 5 or 6 or for Corrective Maintenance in an Operating Mode CHANGE DESCRIPTION This change was made to allow the procedure to be performed for corrective maintenance purpose' only during plant modes 1, 2, 3, or 4. The loop would be considered out of service and not operable. These same changes also allow performance in modes 5 or 6 without cycling isolation valve 2CCP-AOV107B.

Double verification is used to connect and disconnect the jumper wire.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created since, although the CCW loop Hi Flow instrumentation is inoperable during maintenance, the CCW flow is maintained to the thermal barrier. Hi pressure and hi temperature signals to the valve are maintained.

The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created since the system function is maintained by hi pressure and hi temperature signals and process radiation monitoring. No Technical Specification bases are affected since the component cooling water system remains in operation. Since this is a temporary modification, no changes to the FSAR are required.

Baevsr valley Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 48 of 152 CHANGE TITLE Defeating Hi CCW Flow Close Signal to RCS-P21C Thermal Barrier Isolation Valve 2CCP-AOV107C-for Performance of 2LCP-15-F107C During Mode 5 or 6 or for Corrective maintenance in an Operating Mode CHANGE DESCRIPTION This change was made to allow the procedure to be performed for correceive maintenance purposes only during plant modes 1, 2, 3, or 4. The loop would be considered out of service and not operable. These same chnnges also allow performance in modes 5 or 6 without cycling isolation valve 2CCP-AOV107C.

Double verification is used to ecnnect and aisconnect the jumper wire.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created since, although the CCW loop Hi Flow instrumentation is inoperable during maintenance, the CCW flow is maintained to the thermal barrier. Hi pressure and hi temperature signals to the valve are maintained.

-The possibility of anLaccident or malfunction of a different type than any previously evaluated in the.FSAR is not created since the system function is L -maintained by hi pressure and hi temperature signals and process radiation

! -monitoring. No Technical Specification bases are affected since the component

,- cooling water system remains in operation. Since this is a temporary modification, no changes to the FSAR are required, l

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Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Toots, Snd Experiments Page 49 of 152 CHANGE TITLE 480V Bus 2H Ground Alarm Relay CHANGE DESCRIPTION This temporary modification was made to restore redundant dual feed capacity to the substation as no relays were available to replace ground alarm relay (74-RH200).

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created because no systems important to safety are affected by this temporary modification. For the same reason, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created. No Technical Specification bases are affected by this change. No changes to the FSAR are required.

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimonts Page 50 of 152 CHANGE TITLE 480V Bus 2F Ground Alarm ' lay CHANGE DESCRIPTION This temporary modification was made te restore redundant dual feed capacity to the substation as no relays were available to replace ground alarm relay (74-RF200).

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created because no systems important to safety are affected by this temporary modification. For the same reason, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created. No Technical Specification bases are affected by this change. No changes to the FSAR are required.

Bsavor Valloy Powar Station Unit 2 1990 Repor* of Facility Changes, Tests, and Experiments Page 51 of 152

' CHANGE TITLE Temporary Installation of Gag on 2CNM-RV151 CHANGE DESCRIPTION Relief valve 2CNM-RV151 lifted during plant trip and would not reseat. _A gag was installed to coat the valve. Over pressure protection of the feedwater auction piping was provided by relief valves 2CNM-RV150A and 2CNM-RV1508.

SAFETY EVALUATION

SUMMARY

Duo to' redundant relief values on the system, this temporary modification does not create an unreviewed safety question. The operation of systems important to safety are unaffected. No Technical Specification bases are affected by this change. No change to the FSAR is required.

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Baavar Valley Powar Station' Unit 2 1990 Report of Facility Changes, Tests, and Experimente l Page 52 of 152 CHANGE TITLE 1

Temporary Installation of Gag on 2CNM-RVll5 j CHANGE DESCRIPTION Relief _ valve 2CNM-RVll5 lifted during plant trip and would not reseat. A gag was installed to seat the valve. This is a bonnet relief valve for 2CNM-23 which remained NSA and was manipulated as needed. System protectio 1 was provided by 2CNM-RV150B.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created since this bonnet relief valve is not required by code and the gagging of the valve will not affect the operation of 2CNM-23. The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created because neither system

_ parameters nor operation are not affected by this modification. No Technical Specification bases are affected by this change. No changes to the FSAR are requ ired. -

-Banver Valley Power Station Unit 2-1990 Report of Facility-Changes, Teste, and Experiments

-Page 53 of 152 i

- G12d%UL TITLE i Temporary Repair of 2SVS-PCV101b Shaft coupler CHANGE DESCRIPTION Due to unavailability of spare coupler and long delivery lead time, a temporary coupler was fabricated per EM 65213 so that the valve could be returned to service.

EA.FETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not

- be increased or created since the installation of the temporary coupler allows equipment to be restored to normal operating conditions. The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created since the temporary coupler is structurally and '

functionally equivalent'to the permanent coupler. No Technical Specification bases are affected by this change. No FSAR changes are requirad.

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Beaver Valley Powar Station Unit 2 ,

1990 Report of Facility Changes, Tests, and Experiments Page 54 of 152 CHANGE TITLE Temporary' Modification of 2GWS-315 CHANGE DESCRIPTION The disc spring was removed from 2GWS-315, a Volan 3/4" piston lift check valve. This was done because normal system flowrate and velocity could not overcome spring force causing trap 2GWS-TRP21 to overflow and 2GWS-LS101 to alarm in the Control Room.

SAFETY EVALUATION

SUMMARY

The probability or-consequences of an accident described in the FSAR will not be increased or created since no safety systems are affected by this change.

The possibility of an accident or malfunction of a different type than previously evaluated in the FSAR is not created since the only credible failure mode of this modification would not prevent normal system operation.

No Technical Specification bases are affected by this change. No FSAR changes are required.

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Doaver Valley Power Station Unit 2 1990 Report of Facility Changos, Tests, and Experiments PLge 55 of 152 QHANGE TITLE Temporary Hodification to Turbine Turning Gear and Bearing Lift Oil Pump Aux Relay Circuit THJND CHANGE OflCRIPTION Pressure switch contact 2TMT-PS, 63BLD-2 "2idJ-P212 Norm. Dioch. Prene." in turning gear aux relay circuit TMJND was jumpered out. This was to prevent trip of turning gear on low discharge pressure on Lift Oil Pump 2THJ-P212.

Normal setpoint for low discharge pressure of this pump is 600 psig. This temporary modification allowed continued turning gaar operation at any discharge prossure. Operations monitored Control Room indication of operating pump every two hours and local indication was monitored once a shift to verify pump operating and discharge pressure of at least 600 psig.

SAFETY EVALUATION

SUMMARY

This change affects the turbine turning gear subuystem only. There is no interaction with any system or component important to safety. No unreviewed safety question is involved. No Technical dpecification bases are af fected by this change. No change to the FSAR is required.

Denver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and I:xperiments Page 56 of 152 l

ClihlM L IIILE 480V Bus 2D Ground Alerm Helay C11!dGEEECRIET10H This temporary modification was made to rostore redundant dual food capacity to the substation as no rolays were available to replaco ground i.larm relay (74-RD200).

SAFETY EVALMhTION $UMMAH1 The probability or consequences of an accident described in the FSAR will net be increased or created becauso no systems important to safety are affected by this +emporary modification. For the same reason, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created. No Technical Speci'ication bases are affected by this change. No changes to the FSAR are required.

15eaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimonts Page $7 of 152 CithNG EJ1TLE Temporary Power for Turbine Generator overhaul CilANGE DESCElPTIOP Temporary power was supplied f rom 480VUS 2-6, Compt 3D to power equipment en the turbino deck for 2R.

1. ATETY lVALVhTIO1LSit&M This change providos temporary non-1E power to the turbine dock ouritig the refueling outage. There is no interaction with any system or component important to safety. No unreviewed safety question is involved. No Technical Specification bases are affacted by this chango. No change to the FSAR is required.

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Daavor Valley Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 58 of l';2 CHANGE TITLE-Connecting Recording Equipment to Operating Equipment CHANGE DESCRIPTION This safety evaluation was for a new procedure to install, control, evaluate, ,

and remove recording equipment to monitor plant parameters while equipment is in service. . The procedure is generic.

SAFETY EVALUATION

SUMMARY

l As stated above, this procedure is generic. Controle within the proceduro require that~a-failure analysis be performed for each application of the i l

. procedure. These controls will ensure that no unreviewed safety question is

= created by the use of this procedure. No Technical Specification bases are affected by this procedure. No change to the FSAR is required.

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Deaver Valley Powar Station Unit 2 1990 Report of Facility Changes, Testo, and Esperiments Page 59 of 152 CHANCE TITLE Vendor Procedures for Hydrolasing 1" & 1 1/2" SWS Supply and Return Lines to Sample coolers for 2SWS-RQIl00 A, D, C, & D in Emergency Diesel Generator Buildings ,

CHANGE DESCRIPTION This safety evaluation allowed the cleaning of these fouled lines during operation (Mode 1) to lessen the impact on the outage schedule.

SAFETY EVALUATION

SUMMARY

The probability-or consequences'of an accident described in the PSAR will not be increased or created because the-diesel generators will remain operable during the cleaning procedure. The possibility of an accident or malfunction of a different type than any previoonly evaluated in the FSAR is not created since any credible failure during tho operation would not result in the loss of more than one diesel generator. No Technical Specification basen are affected by this procedure. No FSAR changes are required.

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i' 11eavor Valloy Power Station Unit 2 1990 Report of Facility changes, Tests, and Experiments  ;

Pagt 60 of 152 -

gilANGE TITLE supplying Temporary Trailors Sprinkler System from Turbine Building Firn l Suppression System gilAHOE DESCRIPTION A 1" valve (2FPN-96) was removed and replaced with a 2* temporary valve which l facilitated the connection for supplying fire suppression water to the Westinghouse trailers sprinkler system. The trailert were located on the Unit 2 Turbine pock.

SAFETY EVALUATION __

SUMMARY

r This chango affects only the fire suppression system. There arn no systems or components important to saiety affected by this change. Failure of the

-temporary system would not result in the loss of the entire system.- flo unreviewed safety question is involved.. No Technical SpecifiJation bases are '

affected by this change. No change to the FSAR 19 required.

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l Beavor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experin.ents l Page 61 of 152 Cl!ANGE TITLE Cross Tie Detween MCC-2-E05 and MCC-2-E06 during 2R Only ClMNGE DESCRIPTION A cross tie was made between HCC-2-E05 and MCC-2-E06 for power to battery

hargers 1 and 2 and the UPS 1, 2, 3, and 4 so PH activities could be perf ornied on SN and 9P transformers during 2R.

EAEgly EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be incroaned or craated since this temporary modification maintains redundant backup power to the chargers and UPSs. The possibility of an accident or malfunction of a different type than any previoanly evaluated in the FSAR is not created since any credible failure would not result in the loss of more than one train. No Technical Specification bases are affected by this change. No change to the FSAR is required.

haaver Valley Powsr station Unit 2 1990 Report of Facility changes, Tests, and Experiments Page 62 of 152 QihligE TITLE l

Temporary Hodification to Hook Up a Flush Rig Detween 2HVC-ACV201A(n) sWS Cooling coils and Associated SWS Supply and Discharge Piping. >

l GLANCE DESCRIPTION The inlet and outlet noaales were removed on 2HVC-ACV202A(D) and a temporary flush rig was connected up between ACV and $WS system supply and discharge piping. Flush was performed to a TOP, an HWR, and EMa 6$132 & 65254.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created because partial failure of-the SWS is analyzed and no credit is taken for any of the involved equipment during a design basio accident. The possibility of an accident or malfunction-of_a different type than any previously evaluated in the FSAR is not created since partial failure of $WS piping is already evaluated. No Technical Specification bases are -

affected by this change. No change to the FSAR is required.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 63 of 152 i

CHAtJGr TITLE Temporary Power for Cooling Tower Work During 2R CHAf1GE DESCRIPTipJi Temporary power was supplied from HCC-2-24 spare breakers 3r and Sr to power equipment on or near the cooling tower during 2R.

SMRILEVALUATION

SUMMARY

This change provides non-1E power for cooling tower work during the refueling outage. There are no systems or components important to safety affected by this change. No unreviewed safety question is involved. Ilo Technical Specification bases are affected by this change. No change to the FSAR is required.

Denver Valley Power Station Unit 2 1990 Report of Facili'y Changen, Tests, ari Experianents Pcge 64 of 152 C11hl10E TITLE Source Range Hign Flux at Shutdown Alarm Hodification pHAf4GE DEECRitTJpli A solid state timer was added to relays NC31CX and NC32CX to provide a variable time delay in order to suppress spurious containnient evacuation alarms caused by the upender operation.

SMETY EVALVATlpN SUMMARE The probability or consequences of an accident described in the FSAR will not iso increased or created because this change af f tet s only the non-saf ety related portion of the nuclear instrumentati)n system. For the same reason, the possibility of an accident or malfunction of e difforent type than any previously evaluated in the FSAR is not created. Ito Technical Specification bases are affected by this modification. No changea to the FSAR are required.

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 65 of 152 CHANGE TIILE Temporary Modification for Flush of 2HVC-ACV207A & D

[11bHQE DESCRIPTION The inlet and outlet SWS nozzles were: removed on 2HVC-ACV207A & B and a temporary flush rig was connected up between ACV and SWS system cupply and discharge piping. Flush was performed to 2 TOP-90-18, an HWR, and EMo 65556 &

65254.

$AFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created because partial failure of the SWS is analyzed and no credit is taken for any of the involved equipment during a design basic accident. The possibility of an accident or malfunction ot a different type than any previously evaluated in the FSAR is not created since partial failure of SWS piping is already evaluated. No rechntcal Specification basen are affected by this change. No change to the FSAR is required.

B3 aver Valloy Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 66 of 152 CHANGE TITLE Temporary Steam Supply from Aux Steam Valve 2 ASS-646 to Temporary Flush Rig in PAB CHANGE DESCRIPTION A temporary hose with a regulator was hooked up to Aux Steam valve 2 ASS-646 to supply steam to the temporary flush rig heat exchanger to heat MCC cooler chemical flush water.

SAFETY EVALUATION

SUMMARY

The probability or consequences of an accident described in the FSAR will not be increased or created because f ailure of this non-safety related equipment is bounded by the -analysis of the ability for equipment in the area to withstand a steam leak due to failure of auxiliary steam system piping. For the same reason, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created. No Technical Specification bases are affected by their change. No changes to the FSAR are required.

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Boavor Valley Power Station Unit 2 j 1990 Re, port of Facility Changes, Tests, and Experiments j Page 67 of 152 CHANGE TITLI DCP-851, Elimination of Vibration on MSH Scavonging Vent Condunner Piping QH&HQF DESCRIPTION The Moisture Separator Reheators (HSRs) condoneo main steam to superheat the steam discharged from the high pressure turbine prior to its entering the low pressure turbine. In order to maintain offectivo host transfor capability, the roheater must be vented to removo noncondensable gases. The Steam Vont System (SVS) piping, which vonts tho MSRs to tho first point hoatora or the condonner, vibrates excessively. Thio excessivo vibration could cause damage to system components.

A temporary jumper (3-21-1B) was issued on 11/30/90 to temporarily restrain the piping system. This vibration problem was recognized by the HED in 1987, and the SWEC was contacted for engineering assistance. The SWEC then determined tnat the piping was in violation of the vibration critoria 2BVM-158, and subsequently issued E&DCR D-5490-501 to correct the problom.

Part of the vibration hao been corrocted por E&DCR D-5490-501, but part of the work scope was never incorporated into the final donign drawings or stress reports.

This DCP shall incorporata the work dono under ELDCR D-5490-501, and modify or install additional supports to reduco the vibration lovol on the ocavenging steam vont piping from tho MCRo. The piping will moet the applicablo design stress limits.

This DCP will maintain the reliability, integrity and operability of the Steam Vent System (SVS) and will have no adverse effecto on any other equipment, abFETY EVALUATION SUMMAR1 No Unit 2 Chapter 15 accidents will be affected by this DCP because thic DCP does not adversely affect any oafety or non-cafety systems, doon not exacerbate any existing accidents, and does nct introduco any now hazard beyond that already conoidered in the UFSAR.

This DCP will not adversoly affect the safoty function of any system. This DCP will only affect non-safoty related supports and piping (indirectly). The reliability, integrity, and operability of the SVS will be maintained and no other systems will be affected.

The probability of an occurrence of an accident previously evaluated in the UFSAR will not be increased. This DCP will maintain the rollability, integrity, and operability of the SVS, and it will have no offect on any other equipment; thereforo, no probabilition of occurrences of any accidents will be increased.

Deover Volley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 68 of 152 The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP is minor and the change will have no effect on any other equipment. This DCP will not affect any parameter which would increase the ce : 7tences of an accident beyond that previously considered in the UFSAR.

The reliJbility, integrity, and operability of the SVS is being maintained.

The probability of a malfunction of equipment important to safety ao previously evaluated in the UFSAR will not be increased. This DCP is minor and the changes will not adversely affect any equipment, including the SVS.

The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP will not adversely affect any parameter which would increase the consequences of a malfunction. This DCP will not adversely affect any safety used to mitigate all accident. Therefore, there will be no effect on the consequences of a malfunction of equipment important to safety.

This DCP will not cause any new credible failure modes because the fundamental design features and functions of the equipment have not been significantly altered. This is a relatively minor change.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created. This DCP is minor, and the reliability, integrity, and operability of the SVS will be maintained. Nothing is being added or altered in a way which creates the possibility of a different type of accident.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR' will not be created. This DCP is minor and the reliability, integrity, and operability of the SVS will be maintained. The fundamental design features and functions will not be changed in a way that creates the possibility of a malfunction of a different type. This DCP only affects non-safety related supports and piping (indirectly).

This DCP will not change any parameter which affects the course of any accident analysis supporting Technical Specifications bases.

The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this DCP will not adversely affect the margin of safety as defined in the bases for any Technical Specifications because tha reliability, integrity, and operability 'of the SVS will be maintained, and no other equipment will be affected.

This DCP will not require any changes to the technical specifications.

This DCP will not require any changes to the UFSAR.

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B3avor Valley Pow 3r Station Unit 2 1990 Report of Pacility Changes, Tests, and Experiments Page 69 of 152 CHANGE TITLE DCP-881, Oland Steam Exhauster and Filtration System pflMOE DESCRETIQB The purpose of this modification is to reduce the amount of moisture from the Turbine 01and- Steam condenser that collects in the oland Steam Exhaust Filters. This is . to be accomplished by installation of a passive moisture separator upstream of the filter inlets and by insulating the piping between the moisture separator outlet and the filter banks' inlet to prevent further condensation of moisture in the system. The drains collected by the moisture separator will be returned to the existing drains system of the oland Steam condenser. The charcoal filters work best when the relative humidity of the air passing through them is losc than 70s; however, the filter banks have been fi?oding due to the high moisture content. This moisture reduces the effectiveness of the charcoal filters in-removing radioactive affluents from the air. This modification will serve to remove this moisture and return it i to the condenser.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the i DVPS-2 Final Safety- Analysis Report (FSAR) is not increased. This modification does not alter- the flow path of the Gland Seal Steam Exhaust system other than the installation of the moisture separator. No alternations to any instrumentation and control or- radiation monitoring equipment are planned. -This modification will allow the system to operate as described in Section 9.4.15 of- the BVPS-2 FSAR. This modification will also allow for isolation of the non-operating train by installation of isolation dampers at the inlet and outlet of each train to allow for safer maintenance of the unit.

The possibility for an accident or malfunction of a different type than Lpreviously evaluated in the BVPS-2 PSAR is not created. The modification will improve the margin of safety by removal of moisture from the Gland Steam Exhaust -air to allow more officient operation of the charcoal filters. Also, u the drains collected _ by the moisture separator will be returned to the cland Steam condenser drains, which will drain bach to the main condenset, thus preserving the original intent of the system. 01and Steam Exhaust Fan startup procedures will ensure that the isolation dampers modification will be open h

during startup of the fans and prevent damage to the filter being placed in service and thus prevent an accidental release of radioactive materials to the environment. Thorofore, no new accident situations are created.

No Technical Specifications are affected by this modification. The margin of safety of any Technical Specification is not decreased, and no changen to the Technical Specifications are required.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 70 of 152 A change to the BVPS-2 UFSAR is required by this change. The description of the Gland Seal Steam Exhaust System in Section 9.4.15 will need to be revised to reflect the addition of the moisture separator. Figure 9.4-16 noodo to be revised to show the location of the moisture separator, the moisture separator

. drain line, and the isolation dampers. The moisture separator and the isolation dampers need to be added to Table 9.4-25.

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Doaver Valloy Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments l

Page 71 of 152 CllANGE TITLE DCP-899, Diesel Cenerator Backup Phase Fault Detection ElihlMl_EIAEBlPIlEH This modification is to install four Westinghouse type SKD-T distance relays to provide backup fault protection for diesel generators in compliance with Reg. Guide 1.9/1979 and to replace the installed SP relays which are not qualified and presently bypassed. To make the SKD-T relays functional to satisfy the coincident logic on a per train basis, (2) ITE-50D overcurrent relays and (2) AR auxiliary relays are also required to be installed.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because this modification will provide coincident logic to provent spurious diesel trips. Consequently, the reliability of the onsite emergency power supply will be increased. When the preferred power source is not available, the onsito emergency power supply feeds power to reactor protection instrumentation and control systems and to other Class lE components and systems essential for safe roactor operation and shutdown.

The consequence of an accident previously evaluated in the safety analysis report wilt not be increased because all relays to be installed will bo designed to maintain the physical and electrical independonce required of safety-related systems. The failure mode and effects analysis (FMEA) will be considered when designing the phase fault protection circuitry to ensure the single failure criteria is mot.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety report will not be increased because this modification will provide backup phase fault protection and ensure that spurious signals will not prevent the diesel generator from performing its functions.

The consequencu of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not bo increased because this modification meets the requiroment of Rog. Guido 1.9, Rev. 2, 1979, Section 7, that a diesel generator trip shall be implemented with two or more independent measurements for each trip parameter under all plant conditions. Any single failure does not disable both dieuel generator units or their auxiliary systems.

The design basis accidents which were toviewed for potential impact by the proposed design change included loss of offsite power.

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Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Expuriments Page 72 of 152 The safety systems which will be affacted by the proposed design change include onsite power systems.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because the original diesel protection scheme failed to include (2) relays per train to satisfy the coincident logic. This modification will correct that problem and adhere to the required separation critoria of Reg. Guide 1.76.

The possibility for a malfunction of equivalent important to safety of a different type. than previously evaluated in the safety analyais report will not. be created because this modification will increase the reliability of the diesel generator. All relays will be QA Category I and seismic analysis will be performed to reassures the added weights to PNL*REL243/253 and 4KV switchgear cubicles 2E9 and 2F9.

The margin of safety as defined in the basta for any Technical specification will not be reduced because the installation of backup phase fault protection will provido a more reliable emergency power supply system.

The proposed design change will require change to the technical specifications. Amendment to the Technical specification Section _l 4.8.1.1.2.b.4 has been submitted to the NRC on 6-30-88 for approvnl. I i

The proposed change will not require changes to the Updated Final Safety Analysis Report.

ansver Valley Pow 3r Stetion Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 73 of 152 CHANGE TITI&

DCP-905, Rev. O, Steam Generator Blowdown Sample Sodium Analy.er Replacement CHANGE DESCRIPTION Existing Leeds & Northrup Sodium Analyzers for the Steam Generator Blowdown samples require a supply of anhydrous ammonia to each analyzer for calibration and propur operation during normal sampling. However, the ammonia gas escapes the sample chanber of the sodium units, fills the casing of the analyzer and fills the sample cubicle. This condition constitutes a safety hazard to personnel within the cubicle. In addition, the aramonia gas being removed by the ventilation within the cubieles will accelerate charcoal degradation of the SLCRS filter beds. Thin modification will replace the existing LLN sodium analyzers mounted to the Roactor Plant Sample Panel 2SSR-PHL21 with new Orion Research Model 1811LL Anelyzore locate external to the sample panel.

EhfrTY EVALUATION SUMMAPY The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the final safety analyses report are not increased. Replacing the existing L&N sodium analyzers with Orion Research Analyzers and relocating them outside the sanple panel will not increase the probability of occurrence or the consequences of an accident or malfunction as evaluated in the safety analysis report. This modification will ensure the original design intent is met for the primary sample system steam Generator Blowdown samples without introducing any additional safety concerns or hazardt.

The possibility for an accident or malfunction of a different type than any evaluated is previously in the final safety analysis report is not created.

The new equipment being installed will not affect any safety related system regardless of system or component failure. This modification will not create or increase the possibility for ar, accident or malfunction different than previously evaluated in the FSAR. Implementing this modification will enhance syrtem operability of the sampling system without compromising system design or increase pctential safety hazards to personnel operating the equipment.

The margin of safety as defined in the basis for any technical specification is not reduced. The SGB samples are not defined in the bases of the Technical Specification and hence are not a consideration. This design change will not affect any components or systems that are defined in the Tech Specs. No Technical Specification changes will be required.

This modification will require a change to the Updated Final Safety Analysis Re port , Figure 9.3-7.

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Boavor Valloy Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 74 of 152 CHANGE TITLE DCP-907, Rev. 1, Computer Status Valve Position Indication CHANGE DESCRIPTIDH The Safety Parameter Display System (SPDS) provides a concentrated display of a minimum set of parameters which define the safety status of the plant. The SPDS is a function of the Emergency Response Facility (ERF) computer system.

The ERF computer system obtains its database from the Plant Computer System (PCS) database.

The SPDS has been subjected to a Verification and Validation (V&V) program to assure its adequacy. A plant-specific evaluation of BVPS-2 has resulted in the determination of the specific parameters to be displayed and they are

'tisted in WCAP-10710. The WCAP-10170 is referenced in UFSAR Section 18.2.1 and is part of the V&V requirement.

The SPDS currently receives an open/not open indication for the following valves: 2FPW'AOV204, 205, 206, and 221. The WCAP requires a closed /not closed indication for these valves, and there is a licensii.g commitment to correct the indication prior to restart from the 2R outage. This DCP will eliminate the open/not open indication (which is not required) and provide the closed /not closed indication.

This will be accomplished by rewiring in the limit switches and junction boxes, adding cable where needed (2FPW*AOV206 already has a spare), removing the BISCO seals, and changing the PCS and ERT computer system databases. The Brady wire markers in the T imit Switch Open (LMO) will also be corracted.

One additional change will be made to the PCS and ERP computer system databases to show correct indication for 2CCP-MOV112A. Field conditions provide an open/not open signal for this valve, the. databases will be corrected to indicate the same, the current indication is incorrect (closed /not closed). Indication for this valve is not required by SPDS per WCAP-10170.

No operating parameters of the Fire Protection System-(FPW) or the Primary component- Cooling Water System (CCP) will be affected by this change. The above changes will only affect the indication for these valves. There will be no effect on valve operability or reliability. Removal of the BISCO seals requires _ a change to the environmental qualification design parameters for the FPW valve limit switches.

SAFETY EVALUATION SUMMABI No Unit 2 Chapter 15 accidents will be affected by this DCP. ]

No BVPS Unit 2 safety systems will be affected by this DCP. This DCP will not adversely affect the performance of the limit switchen and as deberibed -j in the 'EQR. Compliance with 10CPR50.49 Guidelines are maintained. This is BVPS Unit 2 commitment for EG after initial license.

B3nvor Vc11ey Pow r Station Unit i 1990 Report of Facility Changes, Tests, and Experiments Page 75 of 152 The probability of an occurrence of an accident previously evaluated in the UFSAR will not be increased. This DCP will maintain the reliability and performance of the si is, and it will not alter equipment in a way such that accidents are more credible; therefore, no probabilities of occurrence of any accidents will be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR.

The reliability and performance of the SPDS is baing maintained.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. The probability is not increased by following the present station commitment understood by NRC and Licensing. That commitment is 10CFR50.49. The EQR referenced in FSAR Section 3.11 was only for initial licenso submittal. The present methodology maintains required conservatism. The present methodology does not require the original submitted EQR commitment to protect public health and safety or any NRC commitment Accepting 10CFR50.49 methodology, the OSC agrees the probability of a malfunction of equipment important to safety is not increased. Reduction in previous conservatism is acceptable to the NRC if 10CFR50.49 is correctly implemenced and documented.

The cono6quences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP will not adversely affect any parameter which would increase the consequences of a malfunction. Therefore, there will be no effect on the consequences of a malfunction of equipment important to safety.

No new credible failure modes are created by this design change in accordance with 10CFR50.49.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created. This DCP is minor, and the reliability and performance of the SPDS will be maintained. Nothing is being added or altered in a way which creates the possibility of a different type of accident.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR is not created. Although the limit switches are no longer EQ'ed per the original EQR, the malfunction is analyred per present BVPS-2 commitment to 10CFR50.49.

This DCP will not change any parameter which affects the course of any accident analysis supporting Technical Specification bases.

The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this DCP will not adversely affect the margin of safety as defined in the bases for any Technical Specifiestions, including 3/4.6.3.1, because the reliability, integrity and operability of the SPDS will be maintained. No Technical Specification changes are required.

The UFSAR Section 3.11 will be updated to state the new basis for the methodology of determining environmental service conditions.

Daevor Valley Powor Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 76 of 152 CHANGE TITLE DCP-963, Enhancement of the Turbine Plant sample System gjlANGE DE$CRIPTION The Chemistry Department requested that several modifications be performed oil the Turbine Plant Sample System to improvo the performance and sampling capability. The modifications include the following:

Addition of a grab sample port on the effluent of the cation columns. l

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This is necessary to perform cation conductivity, chlorides and sulfates analyses.

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2) An addition of a grab sample port for chilled water to provide the capability of easily obtaining samples.

3) Replacement of the existing sodium analyzer. This new analyzer will-be identical to the sodium analyzers installed in the Primary Plant Sample System. Additionally, it will not release any ammonia gas or any other hazardous material to the environment.

3AFETY EVALUATION SUHMARY l

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased. The Turbine Plant sample System  ;

performs no safety-related functions and its failure will not result in any previously evaluated accidents.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because the failure of the Turbine Plant sample System will not result in any accidents previously evaluated.

l The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. As stated in UFSAR Section 9.3.2.2.3, any failure of the Turbine Plant sampling

-System equipment will not affect the safety functions of other equipment.

l The consequence of a malfunction of equipment important to safety as L previously evaluated in the safety analysis report will not be increased because the failure of the equipment being installed undar this dasign change will not affect the safety function of other equipment.

No design basis accidents were found to be potentially impacted by the proposed design change.

No safety systems will be affected by the proposed design change.

Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 77 of 152 The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because all of the modifications being performed will be within the turbine plant cample panel located in the Turbine Building, and will not perform any safety-related functions.

The possibility for a malfunction of equipment important to safety of a different typa than previously evaluated in the safety analysis report will not be created because this design change does not involve any safety-related equ ipment . Moreover, the equipment being installed is located in a non-seismic area and will meet the design temperaturo and presouro requirements.

Failure modes of the proposed design change which were reviewed included piping and valve failures, and earthquake generated missiles.

The margir. of safety as defined in the basis for any Technical Specification will not be reduced because the Turbine Plant sample System is not addressed.

No Technical Specification changes are requiJed.

UFSAR Figure 9.3-9 will require a revision to show the proposed modifications.

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B2evor Valley Powor Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 70 of 152 CHANGE TITLE l

DCP-979, Rewcrk Vendor Supplied "PVC" Jacketed Cables for Radiation Monitor Printers CHANGE DESCRIPTIQH The objective of this design change is to add protective cable wrap, Sil-Temp, to two vendor supplied PVC jacketed cables. The cables to be wrapped are in the control room between radiation monitor printer 2RMS-PNTR-1 and interf ace terminal cabinet 2RMS-TD-1, and radiation monitor printer 2RMS-PNTR-2 and j interface terminal cabinet 2RMS-TD-1. DVPS-2 UFSAR Section 8.3.1.4.2 requires  !

minimum separation distances between todundant Class 1E cables and between l Class 1E cables and non-Class 1E cables. The problem is that the above i mentioned non-Class 1E cables pass through a trough under both the orange and purple train radiation -monitor cabinets. If, for example, the orange train cables caught fire and ignited one of the above neutral train cablas, it is conceivable that the flames could spread to the purple train or vice-versa.

To preclude such a scenario, a protective wrap of woven silicon dioxide (trade names Sil-Temp) and glass tape see to be installed on the neutral cables to protect the adjacent Class 1E cables from electrically induced problems. Item 10 of UPSAR Section 8.3.1.4.2 considero caoles protected in the manner the ,

same as cables in an enclosed raceway. Figure 8.3-52 allows a 1" minimum acceptable separation when the non-Class IE cables are wrapped in this fashion within a. cable spreading area.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysis. report will not be increased because this design change is being made to preclude the possibility of an accident. It will reduce the  ;

chance of a fire batard that could render both radiation monitoring trains inoperable.

The consequence of an accident previously evaluated in the safety analysis report' will not be inercased because the design change will reduce the consequence of a fire in one train of the radiation monitoring system. A fire  !

in one train will now. be .solated from the redundant train, thus ensuring reliable redundancy.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the design change is passive in nature. It will not affect the operability of any equipment nor will -it impact safety. In fact, the probability of malfunction due to a redundant train failure will be reduced.

The . consequence of a malfunction of equipment important to- safety as previously evaluated in the safety analysis report will not be increased because again the design change is passive. The consequence of a malfunction that results in ignition of the cabics and flame spread in reduced.

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Doaver Valley Power Station Unit 2 1990 Report of Facility Changos, Touts, and Experiments Page 79 of 152 l

The design basis accident which were reviewed f or potential impact by the proposed design change included radioactive release from a syetem or component.

The safety systems which will be affected bi the proposed design change include the Radiation Monitoring System.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because the design change is using UFSAR approved material to provida train separation.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the design change does not impact the operability of any equipment, and therefore will not lead to malfunction.

Failure modes of the proposed design change which woro reviewod included unraveling of Sil-Temp and fire in the area of application.

The margin of safety as definod in the basis for any Technical Specification will not be reduced because the design change will increase the reliability of radiation monitoring for channels required by the Technical Specifications.

The proposed design chango will not require chango to thu technical specifications. The Technical Specifications require various radiation monittre and electric power sources to be operable, yet they do not go into the detail of explaining individual cable characteristics and methods of separation.

The proposed chango will not require changos to the Updated Final Safety Analysis Report. The UFSAR does not specify the method of separation used for individual cable assemblies, it on describes the acceptsble generic methods that cables shall be separated, the minimum distancos of those separations, and any cables using unique methods of separation. No chango would be required unless methods of separation are spalled out on Tablo 8.3-10, and this design change does not fall into that category.

Banver Valley Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 80 of 152 CHANGE TITLE DCP-1046, Provide 011 Fill Capability to Reactor Coolant Pump Motors gjihNGE DESCRIPTIpd The present design of the reactor coolant pump motors (2RCS*P21A,B,C) does not allow gravity oil fill of lower bearing oil reservoirs. The existing one-inch valve serves as both the fill valve and the drain valve, and is located below the normal oil level. It was originally intended that replacement oil be pumped into the oil ressrvoir through this velve. Gravity fill capability is desired.

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The proposed solution is to install a one-inch standpipe assembly to provide a fill connection at a location above the normal oil level. The addition of the ,

standpipe will require a modification of the oil rollection pan to extend it under the added standpipe and valve.

SAFETY EVALUATION

SUMMARY

There are no BVPS-2 UFSAR Chapter 15 design basis cecidents that are affected by the. proposed design change. All of the accidert anal;ises review do not depend upon the operation of the reactor coolant pump n otors, but use the pump.

casing and seals as a pressure boundary.

This modification will have no adverse effects on any Unit . safety systems.

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because no changes are being made to the motor lube oil- reservoir or circulation systems that affects pump operation. The existing till/ drain valve will remain 'normally closed' thus isolating the added piping from the lube oil. In addition, the added piping and the addition to the oil collection pan wil1~ be seismic, so their structural integrity is assured. This modification will simplify maintenance work on the motor by requiring less e7uipment to fill the bearing oil reservoirs.

The consequences of an accident previously evaluated in the safety analysis report will not be increased because no previously evaluated accidents are affected by this modification as described in Section 5.4.1 of the BV-2 UFSAR.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will n't be increased because this modification does not affect any safety functio.s performed by the reactor coolant- pump motors as described in Section 5.4.1 of the BVPS-2 UFSAR. The integrity of the lube oil reservoirs will be maintained because the existing fill / drain valve will remain 'normally closed'.

The consequence of a malfunction of equipment important to safety as previously evaluated in. the safety snalysis report will not be increased

t B3cver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments  !

Page 81 of 152 becauce rupture or failure of thin modification will have no effect on any safety-related equipment. The existing isolation valve will prevent loss of oil from the bearing reservoirs and preclude reactor coolant pump motcr failure; therefore, this modification will have no effect on any safety-related equipment.

The possibility for an accident of a different type than previously evaluated in the safety . analysis report will not be created becauce this design change does. not affect- the operation of the reactor coolant pump motors; it will simplify maintenance by requiring fewer tools to add oil to the bearings. The

• oil collection pans will be extended under the added piping to prevent spillage of oil into the containment during oil fill operation and to collect any small oil leaks. Rupture or . failure of the new piping is of no consequence to safe operation of the reactor coolant pump motor.

The. possibility 'or a. malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the safety functions of the reactor coolant pump motors are not being changed by this design change. This modification will create no new failure modes and will not add any new hazards.

There are no changes that affect the course of any accident analysio supporting Technical Specification bases that exceed the acceptance critoria for fuel cladding, RCS boundary, or containment integrity.

The margin of safety ao defined in the basis for any Technical specification will not be reduced because no Technical specification bases are affected by this modification. The proposed design change will not require change to the technical specifications.

This. change is not subject to the review requirements of the Environmental Protection Plan because this change does not affect the environment.

The proposed change will require a change to the Updated Final safety Analysis Report, Figuro D.4-9.

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Daavor Valley Powar btotion Unit 2 1990 Heport of Facility Changes, Toots, and Experiments Pago 82 of 152  ;

1 Cl4ANGE TITLE DCP-1049, Auxiliary Feed pump Steam Drain Valves citANGE DESCRIPTION During normal surveillance start-up of tha Turbinw Driven Auxiliary Foodwater pump, an operator must o en the drain valves (2SDS*212 and 213) on the trip throttle valve (2FWE*TTV22) in order to drain water from it. After the turbine has been. started, an operator must throttle or close those drain valves. This presents a serious safety hazard and is inconvenience to the operator, who must stand on the pump skid and lie ovat the pump shaft guard in order to reach. the drain valves. The purpose of this modification is to relocate the drain valves to a more accessible location on-the upatream vartical section of the drain lines. The drain line piping will also be slightly modified so that the vertical section is in front of the trip throttle valve and an additional pipe support will be installed to maintain .

the seismic design of the lines, gA1 TTY EVALUATION

SUMMARY

The probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the Updated Final Safety Analysis Report (FSAR) is not increased. This design change is relocating the drain valves for the trip throttle volve to a new ~

location on the upstream vertical section of the drain lines and rerouting the piping so that the vertical section of pipo la in front of the trip throttle valve. The design function of the drain valves will not be altered by this modification and their safety class boundaries will still be maintained. The seismic design of the piping will be maintained with the addition of a new pipe support. Therefore, the consequence'of an accident or malfunction of equipment important to safety as previously evaluated in UFSAR Sections 10.3.

10.4.9, 15.1.5, 15.2.7, 15.2.8, and 15.6.5 will not be increased.

The possibility for an accident or malfunction of a different type than-previously evaluated in the Updated Final Safety Analysis Hoport is not created. This modification will not alter the function of the drain valvos for= the turbine driven auxiliary feedwater pump trip throttle valve, nor will it degrade the. reliability of the auxiliary foodwater pump to provide high pressure foodwater to the secondary side of the steam generators, when

l. required. Additionally, since the seismic design and the piping safety class E breaks will be maintained, the poonibuity of an accident or malfunction of a .

L different type than any previously evaluated in UFSAR Sections 10.3, 10.4.9, i 15.1, 15.2 and 15.6.5 will not be created.

The. margin of safety as defined in the basis for any Technical Specification is not reduced. The operability of the Auxiliary Feedwater System will not be affected by this design change since the design function of the drain valves for the trip throttle valves will romain unchangod after relocating them to a more accessiblo location. Therefore, the margin of safety as defined in Technical Specification Basis 3/4.7.1.2 will not be reduced. This design change will not require a change to the Technical Specifications.

This design change will not require a change to the Updated Final Safety Analysis Report.

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Bsavor Valley Powsr Station Unit 2 1990 Report of Pacility Changes, Toots, and Experiments Page 83 of 152 CHANGE TITLE DCP-1063, Relocato Fire Hoso Rack No. HR-243 and Provido Supports pHANGJ DESCRIPTIpft{

Fire hose rack number HR-243 has becomo virtually inaccessible due to tubing and supports located in front of it. Since NFPA requires hose stations to bo located where they are not likely to be obstructed, this radification proposes to relocate the rack to the front of the tubing and suppurts. This relocation will constitute the need for a new piping configuration along with additional piping supports.

SAFETY EVALUATION SUlHQE1 The probability or the consequences of any accident previously evaluated in the FSAR will not be increased. This modification serves to enhance the reliability of the fire protection system by providing easier access to an existing home rack. This change will not affect the system's intended function and the new piping and supports will be designed to the appropriate specifications.

The probability or consequencos of a malfunction of equipment important to safety will not be increased. As stated before, this modification will actually enhance the Fire Protection System's reliability. No safety-related equipment will be affected by this modification since the hose tacks are not the primary fire suppression device for any safety-related equipment.

The possibility for an accident or malfunction of a different typo than previously evaluated in the FSAR will not be created. No new failure modos or potential hazards will be created by the implementation of this modification.

During the modification's installation phase, when portions of the hose rack system are inoperable, appropriate means of backup fire protection must be made available (Ref. FSAR Section 9.5.1.2.3.2.9). This will be ensured by the adherence to Site Administrative Proceduro 9D, " Fire Protection".

The margin of safety as defined in the basis for any Technical Specification is not reduced because no Technical Specifications or their bases are affected by this design chango. This change does not require a chango to the Technical Specifications.

This change does not require a chango to the Updated Final Analysis Report.

Baevar Valley Pow 3r Station Unit 2 1990 Report of /acility Changes, Tests, and Experiments Page 84 of 152 CHANGE TITLE DCP-1078, Disabling the Auto Dispatching System (ADS Function)

CHANGE DESCRIPTION The System Operator is not required to control the BVPS Unit #2 Generator Load utiliting the Automatic Dispatching System (A7S). The ADS is a subsystem of the Terbine Generator control System. The design of this system includes an annunciator for when the talemetering signal between the station and the system operator is lost. Since auto regulation of output power la not desired for Unit #2, the annunciator signal will be climinated so that the " trouble alarm will not be constantly alarmed". The ADS will also be rendered inoperable.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety enalysis report will not be increased because the Automatic Dispatching System is not used, is not safety-related, and in no way would its inoperability affect plant operation. The trouble alarm that is to be disconnected also has no affect on any accident scenarios.

The consequence of an accident previously evaluated in the safety analysis report will not be increaeed because the inoperability of the ADS and its associated trouble alarm input do not impact safety-related equipment and therefore affect no accident consequences.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the proposed changes will not affect any safety-related equipment.

The conscquence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will net be increased because the proposed changes will not affect the consequences of any malfunction of safety-related equipment.

No dasign basis accidents are impacted by thn proposed design change.

No safety systems are affected oy the proposed design change.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because these changes do not impact the operability of any equipment required for plant safety.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the proposed changes to the ADS will not impact any safety-related equipment.

Deaver Valley Power Station Unit 2 1990 Report of Facility Changes, Testa, and Experimente Page 85 of 152 The margin of safety as defined in the basis for any Technical Specification will not be reduced because the equipment affected by the changes do not impact any Technical Specifications. The proposed denign change will not require change to the technical epecifications.

The propooed change will not require changes to the Updated Final Safety Analysis Report. The UFSAR does not address the ADS, nor do any of the

eident scenarios.

Baaver Valley Powar Station Unit 2

-1990 Report of-Facility Changes, Tests, and Experiments Page 86 of 152 CHANGE TITLE DCP-1130, Rev. O, Temporary Reactor Vessel Head Shielding for Unit No. 2

-CHANGE DESCRIPTION Experience at BVPS-1 has provea that a major portion of refueling outage exposure occurs during. the reactor vessel head removal and replacement, primarily during stud detensioning and tensioning operations. A temporary reactor head shielding system was installed at BVPb-1 during SR (DCP-701) and has resulted in the reduction of exposure rates by a factor of four (4).

This modification proposes to provide BVPS-2 with the necessary permanent equipment to enable the support tubes and shielding (purchased under DCP-701) to be used during BVPS-2 refueling outages. Three (3) support brackets will be installed, (one [1] on each of the reactor vessel head lifting rig rods).

These- brackets will remain permanently affixed. Prior to detensioning the reactor vessel head stude, the support tubes and shielding will be lowered into the cavity using the Polar Crane. The stud tensioner hoists will then be used' to insert the tubes into the support brackets and to install the shielding onto the support tubes. The shielding will remain in place until after the reactor vessel head is placed back on the vessel and the studs are tensioned. The support tubes and shielding will then be removed and placed in

'their storage boxes until the next refueling (most probably at BVPS-1).

SAFETY EVALUATION

SUMMARY

BVPS-2 UFSAR, Chapter 15, and DVPS-1 UFSAR, Chapter 14, were reviewed to identify what design basis accidente could be impacted by the proposed modification. BVPS-2 UFSAR, Section 15.7.4, discusses the consequences of a fuel handling accident. Since this modification is not directly involved in the actual- handling of fuel, it will not have any adverse affect on this type of. accident nor will this modification affect any systems required to mitigate this type of accident.- The impact of this modification on BVPS-1 is limited to radiological concerne during the transport of the shared equipment between the two (2) units. Any required decontamination and equipment transportation will be performed under approved radiological procedures. None of the BVPS-1 UFSAR Chapter 14 safety analyses will be affected by this modification.

No safety systems will be adversely affected by the proposed design change.

The shielding will be placed on the reactor vessel head prior to stud detensioning and will remain there until the head is placed back on the vessel and the stude are tensioned. This will in no way affect the integrity of the reactor vessel head.

The proposed design change will not increase the probability of occurrence of an _ accident previously evaluated in the Safety Analysis Report. The functions of all safety related equipment and systems will be unaffected by this modification.

The proposed design change will not increase the <onsequences of an accident previously evaluated in the Safety Analysis Report. No equipment or systems required to mitigate the consequences of any previously analyzed accident will be affected by this modificatien.

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 87 of 152 The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report. The loads that will be lifted by the Polar Crane and the Stud Tensioner Hoist are within their rated capacities. Analyses specific to BVPS-2 must be performed to verify that the weight that will be added due to this modification can be safely handled by the Polar Crane, Stud Tensioner Holst, Reactor Vessel Head Lifting Rig, and Reactor Vessel Head Storage Stand. BVPS-1 analyses were reviewed. Assuming that the BVPS-2 analysen have similar results, the extra load added will be safe. The total weight of the shielding assembly including the lead blankets, support tubes, brackets, chains and lifting gear is approximately 9000 pounds or 4.5 tons. This weight will be temporarily added to the reactor vessel head and must be lifted by the Polar Crane. From Unit 2 UFSAR, Table 9.1-4, the reactor vessel head and attachments presently weigh 130 tons. The addition of the shielding and support components will increase this weight to approximately 134.5 tons.

This is within the capacity of the Polar Crano, which is rated at 334 tons for the bridge, with each of the two (2) main hoists rated at 167 tone. Table 9.1-4 should be revised to include a new Heavy Load Identification of " Reactor Yessel head, attachments, and temporary shielding (used during refueling",

with a Load Weight (tons) of "134.5". (or as determined by Unit 2 Specific Analysis.)

The heaviest individual components that will be lifted with the Stud Tensioner Holst will be the long shield blankets. These blankets weigh approximately 850 pounds each, which is within the 4000 pound capacity of the hoist.

The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report. No safety related equipment, other than previously discussed, will be affected by this modification. No new failure modes or potential har.ards will be created.

No new credible failure modes will be created by the implementation of this modification. The permanently affixed brackets will meet applicable seismic requirements. The weights that will be lifted by the Polar Crane and Stud Tensioner Holst are all within their rated capacities. And, there will be no functional changes to any safety related equipment, systems or structures.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the Safety Analysis Report. All equipment and system functions will remain unchanged. The reactor vessel head lifts will be performed in the same manner as before and will follow the applicable Heavy Load Handling procedures.

Daaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 88 of 152 The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated it. the Safety Analysis Report. No new failure modes or potential hazards will be created by the implementation of this modification.

There are no changes in parameters which affect the course of any accident l analys's supporting Technical Specification bases and that result in exceeding j the acceptance criteria for fuel cladding, RCS boundary, or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification (T.S.).

No T.S. basis will be a. . ced, in any way, by this modification. The proposed design change will not require any changes to the technical i specifications.

This design change is not subject _to the review requirements of the Environmental Protection Plan since this change will not affect the environment in any way.

The proposed change will require a change to the Up'ated d Final Safety Analysis Report. All heavy loads currently identified in Table 9.1-4 remain unchanged; however, a new heavy load, similar to the following should be added: " Reactor Vessel head, attachments, and temporary shielding (used during refueling)" -

weight of "134.5" tons. (or the appropriate weight as determined by a Unit 2 analysis). Also, Figeros 9.1-13 must be reviewed and revised as necessary to include the temporary reactor vessel head storage boxes as a heavy load lifted by tho Polar Crane.

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B3svor Valloy Powar Stetion Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 89 of 152 CHANGE TITLE DCP-1135, Rev. O, DV-2 Small Bore Snubber optimization CHANGE DESCRIPTION The purpose of this design change in to eliminate snubbers on small bore piping due to their poor reliability, and failures which lead to increased testing, radiation exposure and analyses to qualify failures. Due to analytical and economic considerations, as well as operational considerations, this design change will target the optimization of the subject piping listed in the Design Input Index section 1.0. By making use of the ASME Code Case N-411 damping values, it is expected that a minimum of 60 parcent of the 400 or so snubbers associated with the target piping can be eliminated. This will result in approximately 250 onubbers being removed. The modifications will be specifically defined once the analyses are complete, then Engineering will issue those blocks of snubbers or modifications to a line which must be performed in ita entirety in order to ensure code compliance. By performing these modifications, financial benefits of decreased down time, reduced manREM exposure, and reduced maintenance efforts will be provided.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because seismic analyses will be performed on the piping using ASME Code Case N-411 damping values to ensure that the seismic design of the piping is maintained. In addition, the existing lines analyzed by WHIPJET, as discussed in UFSAR Section 3.68.2.1, will be reviewed to ensure that the Leak-Defore-Break conclusions are still valid. This will demoastrate that the fluid leakage from a postulated defect at the highest stress location concurrent with minimum msterial proportion can still be detected well before the rupture of the pipe by using the DJ-2 Leak Detection System.

The consequence of an accident previously evaluated in the safety analysis report will not be increased since the minimum requirements of ensential safety-related systems and structures are not compromised, the plant can be safety shut down, and offsite doses in excess of applicable guidelines will not occur. This design change will ensure that adequate clearance still exists between piping, components, and adjacent structures due to increased pipe movement which could occur. Protection against the dynumic effects of postulated pipe ruptures, where applicable, will be maintained, and the criteria of UFSAR Sections 3.6N.2.3.2 and 3.6N.2.3.2.2 will be met. This design change does not affect the design of the ESP systems.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because any increases in valve accelerations will be evaluated to ensure operability and c ompli ance with existing equipment qualification standards.

Nozzle load evaluations will also be performed to justify the increased nozzle loads, where applicable.

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Beaver Valley. Power St ation Unit 2 1990 Report of Facility _ Changes,. Tests, and Experiments Page 90 of 152 The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased.

Pin-to-pin replacement struts will be installed, where needed, in place of the eliminated snubbers,- and some of the spring hangers may either require upgrading or be replaced with struts in order to meet the design objectives.

The PSA 1/4 and PSA 1/2 anubbers which'cannot be eliminated per the snubber reduction techniques will be replaced due to their poor reliability and high failure rates. In- addition, those snubbers which remain on an optimized system may be required to be replaced with larger snubbers in order to ensure compliance with the analyses; however, this is not expected.

The design bdsis accident which were reviewed for potential impact by the proposed design change included small break LOCAs, main steam line breaks, and feedwater line breaks.

The safety systems which will be affected by the proposed design change include PCS, SIS, FWS, SWS, CHS, CCP, EDG, FNC, QSS, RHS, DCS, BDG, BRS, MSS and SVS.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created. By using the increased damping values allowed by ASME Code Case N-411, snubber elitaination will be possible py showing that the pipe stress values or cumulative usage factors are still within the allowable limite specified in UFSAR Section 6.3, such that no new intermediate break locations are postulated.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the cafety analysis report will not be created. If the analyses performed for this design change show that thm support loads are increased, a reevaluation of the pipe supports and welded attachments will be performed to ensure that they remain within their

-design allowables. Additicnally, a thermal reanalysis will be performed whenever struts are installed, to ensurn compliance with the code.

Failure modes of the proposed design change which were reviewed included pipe breaks,_ pipe whip, jet impingcment, and earthquakes.

The margin of safety as defined in the bases of Technical Specification 3/4.4.10' and 3/4.7.12 will not be reduced because the analyses performed and reviewed will demonstrate that safety-related components and systems have not been adversely affected by the removal of snubbers, and that the structural integrity of the Reactor' Coolant System and branch connections is maintained during and following seismic or similar events initiating dynamic loads. The.

proposed design change will not require change to- the technical specifications.

The proposed change will require changes to the Updated Final Safety Analysis Report. UFSAR Table 3.9B-13 will have to be updated to reflect any enubbers which are eliminated.

Baavor Valley Power Station Unit 2 1990 Report of Facility Changes, Tente, and Experimente Page 91 of 152 CHABQE TITLE DCP-ll60, Rev. O, Provido Heat Tracing for the Pump / Heater Assembly of WTD-TK23 and the Sensing Line for 2WTD-LT107 CHANGE DESCRIPTION During cold weather, Domineralized Water Storage Tank (2WTD-TK23) heaters (2WTD-H23A,B) will freeze when not in operation since the linen in and out are not insulated; also, the senoing line for level transmitter (2WTD-LT107) is susceptible to freezing even though the lino is heat traced and insulated.

This design change will provido for heat tracing and insulation on the tank heater lines and will provide for improved heat tracing and insulation on the sensing line for (2WTD-LT107).

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysin report will not be increased because this design change is correcting a deficient situation by adding insulation and increasing the Kilowatt per foot rating of the heat tracing to prevent frooting in the Domineralized Water System Storage Tank heatore and level instrumentation.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because failure of the new heat tracing equipment will be of no greator consequence than f ailure of the existing heat tracing.

As otated in section 9.2.3.3 of the BVPS-2 UFSAR, no failure or malfunction of the domineralized water makeup system will adversely affect equipment or systems needed for safe shutdown of the reactor.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysie report will not be increased because no safety-related equipment is affected by thin donign change.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because no safety-related equipment is affected by thin design change.

The design basis accident which were reviewed for potontial impact by the proposed design chango included freezing of domineralized water system lines due to failure of hoat tracing.

No safety systems will be affected by the proposed design change.

The possibility for an accident of a differont type than previously evalu ted in the safety analyelo report will not be created because this design chasge is raplacing existing heat tracing and insulation with higher rated equipment of the same type, and therefore no new accidents or malfunctionn can be postulated.

Deaver Valley Power Station Unit 2 1990 Report of Pacility Changes, Tests, and Experiments i Page 92 of 152 The possibility for a malfui:c* ion of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because no safety-related equipment is affected by this modification.

Failure modes of the proposed design change which were reviewed includea freezing of domineralized water system lines due to heat tracing failure.

The margin of safety as defined in the basis for any Technical Specification will not be reduced because no Technical Specification bases are affected by this design change. The proposed design change will not require change to the technical specifications.

The proposed change will require a change to the Updated Final Safety Analysis Report, Figure 9.2-22.

Beavor Valloy Pow 3r Station Unit 2 1990 Report of Facility Changes, Teate, and Experiments Page 93 of 152 CHANGE TITLE DCP-1195, Rev. O, Instrument Air Dryer Coalescent Pre-Filters CHANCE DESCRIPTIQH Probleme have been identified with the plant instrument air system maintaining an acceptable dew point. Similar problems were encountered at Unit 1 and resolved. The resolution called for improved maintenance and the installation of coalescing filters as a profilter system prior to the air dryers. Unit 2 is going to follow this action by replacing the existing air dryer particulate profilter [2IAS-FLT21] with a coaleucing filter system.

SAFETY EVALUATION SUMMARE The probability of an occurrence of an accident previously evaluated in the safety analyste report will not be increased because the new coalencing filter assembly will provide additional moisture removal capacity that will result in drier air in the Instrument Air System.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because as stated in Table 9.3-2 of the UFSAR, if the coalcocing profilter clogs, cartridge-type bypass filters are available to be placed in service until the clogging in corrected. No changes are being made to the bypass filters under this modification. If the filter pressure boundary container fails, this is bounded by failure of the instrument air dryer, which would result in the bypass air filters being placed in service.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased

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because no safety-related equipment is affected by this modification.

The consequence of a malfunction of equipment important to safety no previously evaluated in the safety analysis report will not be increased because no safety-related equipment is affected by this modification.

The design basis accidents which were reviewed for potential impact by the proposed design change included clogging of the new coalencing prefilter and air dryer rupture.

No safety systems which will be affected by the proposed design change.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because no credible failure modes other than those previously analyzed can be postulated.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because no safety systems will be affected by this modification.

Failure modes of the propooed design change which were reviewed included clogging of prefilter and air dryar rupture.

B2avor Valley'Powsr Stction Unit 2 1990 Report of Facility changes, Tests, and Experiments Page 94 of 152 The margin of safety as defined in the basis for any Technical Specification will not be reduced because no Technical Specifications bases are affected by this modification'. The proposed design change will not require change to the technical specifications.

The prcposed change will require changes to the Updated Final Safety Analysis Report. . Table 9.3-1 should be revised to. reflect the new pre-filter paramotors. Figure 9.3-3 should be revised 1 to reflect the new system ,

configuration.

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i Beaver Velley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments ,

Page 95 of 152 CHANGE TITLE l 1

DCP-1237, Rev. 2, DRMS Modifi ations to prevent Bad Data Quality Indication on ERFCS/SPDS Displays CHANGE DESqBIPTION Depending- on the application, certain rad monitors remain in a standby j condition until process flow exists. During this condition, a loss of sample I flow is set in the DRMS Monitors. This logic results in a " Poor" data quality d indication on' the SPDS. This poor data quality is a misleading indication on the SPDS displays because the_ DRMS monitor 15 operating-as designed. This

. design change will modify the logic of the RM-80 (microprocessor based signal

• processor and controller) for each of the DRMS monitors involved. The logic will be changed such that a normal condition will be generated at the DRMS monitors if there is no process flow input as sensed by the RM-80s.

Communications with the vendor during the design of the firmware revealed that a process flow element failure would develop into an unmonitored release. If the process. flow element was to fail, the sample pumps .ould shut off after 50 seconds. and1the. DRMS would not alarm. Radiation Monitor [2 SCC-RQIl00) automatically closes [2SGC-HCV100) on a high radiation signal of the Liquid Waste Process Ef fluen?. Release path. -Radiation Monitors

[2SWS*RQIl00A(B)(c)(D)] require operator actions on a-high radiation signal to mitigate the amount of leakage from the RSS Heat Exchangers. Radiation monitors [2HVS*RQIl09C(D)) are Elevated Release Detectors for gaseous effluents and are required operable by Technical Specifications. [2CNA-RQIl00) and [2CNA-RQIl01) are Cat. II monitors with no automatic actions.

SAFETY EVALUATION QU4 MARY The ' design basis accident which were reviewed for potential impact by the proposed design change included Section 15.7, Radioactive Release From A System or Component.

The safety systems which will be affected by the proposed design change

. include Service Water and SLCRS.

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because this logic change does not inhibit nor affect rad monitor safety functions that are taken credit for i in the accident analysis.

The consequence of an-accident previously evaluated in the safety analysis report will not be increased because the rad monitors will still perform as designed and as analyzed in'the UFSAR.

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The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because- the design change is not changing the operation of the equipment, it is only changing the state of the sample alarm during the absence of process flow. Administrative controls will' ensure a loss of process flow signal is alarmed for [2SGC-RQIl00).

Beaver Valley Power Station Unit 2

-1990 Report of Facility Changes, Tests, and Experiments-Page 96 of_152 The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased

'because a malfunction of equipment important to safety will yield consequences to those already analyzed.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because the safety functions of the affected monitors is unchanged.

The possibility for a malfunction of equipment important to safety of a different. type than previously evaluated in the safety analysis report will not be created because the design change does not affect the possibility for a malfunction of safety-related equipment.

The margin- of safety as defined in the basis for any Technical Specification will not be reduced because sample flow ierification is only required when

. process flow is available. The proposed design change will not require' change to the technical specifications.

The proposed change will' require changes to the Updated Final Safety Analysis Report. Page' 11.5-16, 5th paragraph should be changed to describe that administrative controls are uced to ensure the low process flow alarm on (2SGC-RQIl00] performs automatically.

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Bacvsr Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 97 of 152 CHANGE TITLE DCP-1265, Rev. O, Relocation of Level Switches 2FWS-LS104A/B CHANGE DESCRIPTION The setpoints of the Level Switches (2FWS-LS104A and B) on the second point heators are too restrictive. The switches are presently mounted on standpipes Ao and AP at Elevation 776'-11-1/2", which is la above the heater's normal level. while operating at 100% power, small turbine load swings cause small level changes in the second point heaters, which causes undesired closure of 2ESS-MOV102, thus turbine " runback" and annunciation of Window A6-8D.

To prevent turbine " runback" and to keep with the dark board philosophy, the switches will be elevated to 777'-2-1/2", which increases the setpointe by 3".

This modificatio7 is limited to the second Point Heaters Standpipes AO and AP, Level Switches (2FWS-LS104A and B) and associated cables (2ESSNNCO24, 2ESSNNCO25, 2FWSANC810 and 2FWSBNC810). By elevating the switches, the cables will be short; therefore two (2) terminal boxes must be added :long with four (4) new cables.

This DCP is not considered to cause any increased turbine water induction concerne because: (1) the second point heater drains by gravity via a 16" line (with no valves) to the heater drain tank in the turbine basement; (2) the extraction steam connectica is from the crossunder piping upstream of the moisture separators and reheaters; and (3) the existing setpoint calculation, SP-2FWS-5, Rev. O, makes no mention of turbine water induction concerns or requirements.

Thim DCP will maintain the reliability, integrity and operability of the Feed Water System (FWS) and will have no adverse effects on any other equipment.

SAFETY EVALUATION

SUMMARY

No Unit 2 Chapter 15 accidents will be affected by this DCP because this DCP does not adversely affect any safety or non-safety systems, does not exacerbate any existing accidents, and does not introduce any now hazard beyond that already considered in the UFSAR.

This DCP will not adversely affect the safety function of any system. This DCP will only affect the non-safety related parts of the FWS. The reliability, integrity, and operability of the FWS will be maintained and no other systems will be affected.

The probability of an occurrence of an accident previously evaluated in the UFSAR will not be increased. This DCP will maintain the reliability, integrity, and operability of the FWS, and it will have no effect on any other equipment; therefore, no probabilities of occurrence of any accidents will be increased.

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Beaver Valley Power Station Unit 2 1990 Report.of Facility Changes, Testo, and Experiments Page 98 of 152 The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP is minor and the change will have no effect on any other equipment. This DCP will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR.

The reliability, integrity, and operability of the FWS is being maintained.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will r.ot be increased. This DCP is minor and the changes will not adversely affect any equipment, including the FWS.

The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP will not adversely- affect- any parameter which would increase the consequences of a malfunction. This DCP will not-adversely affect any safety system used to mitigate an accident. ~%erefore, there will be no effect on the consequences ,

of a malfunction of equipment important to safety.

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This DCP will not cause any new credible failure modes because the fundamental design features and functions of the equipment have not been significantly altered. This is a relatively minor change.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created. This DCP is minor, and the reliability, integrity, and operability of the. FWS will be maintained Nothing is being added or altered in a way which ..eates the possibility of a different type of accident.

The . possibility for a malfunction of a different type than any previously evaluated in- the UFSAR will not be created. This DCP is minor and the reliability, integrity, and operability of the FWS will oe maintained. The fundamental design features and . functions will not be changed in a way-that creates the possibility of a malfunction of a different type. The safety-related parts of the FWS will not be affected.

This DCP will not change any parameter which affects the course of any accident analysis supporting Technical Specification bases.

The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this DCP will not adversely affect the margin of safety as definsd in the bases for any Technical Specifications because the reliability, integrity, and operability of the FWS will be maintained, and no other equipment will be affected. This DCP will not require-any changes to the technical specifications.

This DCP will not require any changes to the UFSAR.

l. D3cvor Valloy Pow 3r Stction Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 99 of 152 CHANGE TITLE DCP-1275, Rev. 1, S/G Primary Manways Bolt to Stud Conss 31on QHANGE DESCRIPTION The purpose of this design change is to replace the steam generator primary manway bolting with stude and nuts which will be used with a stud tensicner device to provide the correct preload. This modification is being made in accordance with Westinghouse Technical Dulletin NSID-TB-87-01, Rev. 1 to provide a high degree of preload accuracy and to reduce fastener damage. New manway insulation will also be installed due to studs protruding farther from the manway than the bolts. The new insulation will be reflective type.

SAFETY EVALUATION 3UMMARY The design basis accidents which were reviewed for potential impact by the proposed design change included small and large break LocAs. Neither accident will be impacted by this design change.

The safety system which will be affectod by the proposed design change is the reactor coolant system. No adverse effects on the RCS safety functions will occur due to this design change.

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased. This modification will provide a higher degree of preload accuracy and help to reduce fastener damage, thereby not increasing the probability of sn accident occurring.

The consequence of an accident previously evaluated in the safety analysis report will not be increased. This modification will provide an enclosure equal to the one currently provided and therefore will not increase the consequence of any accidents.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this design change will help to reduce fastener damage.

The consequence of a malfunction of equipment important to safety au previously evaluated in the safety analysis report will not be increaseo because the new stude and nuts will be designed to meet the pressure and temperature design conditions, and any operational requirements.

Failure modes of the proposed design change which were reviewed included reactor coolant leaking from the steam generator primary manway closure joint. LOCAs in various fashions are already analyzed.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because this modification will provide a closure joint equal to or better than the existing.

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baaver Valley Powor Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 100 of 252 The possibility for a malfunction of. equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the new studs and nuts, when used in conjunction with' the tensioner device, will help to provide a more accurate proload. This anould provide a more evan compression of the manway closure joint.

The margin of safety as defined in the basis,for any Technical specification will not be reduced since this design change will not affect the operability of the steam generators and does not impact the steam generator-tube sample selection / inspection. The proposed design change.will not require change to the technical specifications.

The propocod change will not require changes to the Updated Final Safety-Analysis Report.

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B:cvor Vallsy Powsr Station Unit 2 1990 Rep 3rt of Facility Changes, Tests, and Experiments Pago 101 of 152 CHANGE TITLE DCP-1286, Rev. O, 2 MSS *AOV101A,B,C C/NC Contact Modification CHANGC DESCRIPTIQH Westinghouse document, WCAP-10170, " Key Safety Parameter Selection for the Beaver Valley Unit 2 Safety Parameter Display System (SPDS)", identifies the Main Steam Isolation Valve (MSIV) positions as mapped parameters. The Plant Computer System (PCS) should have the appropriate inputs available to it so that the MSIV's positione, at any time, can be correctly identified (OPEN/ INTER / CLOSED). Currently only OPEN/NOT OPEN positions are displayed.

This mcdification proposes to add CLOSED /NOT CLOSED indications to the PCS and the ERFCS. The points that correctly indicate OPEN/NOT OPEN, (Y6104D),

(Y6105D), and (Y6106D), will not be altered. However, points (Y5858D),

(Y5859D), and (Y5860D) will be modified to provide CLOSUD/NOT CLOSED indication. These points are currently incorrect in that they provida redundant OPEN/NOT OPEN indication.

SAFETY EVALUATION

SUMMARY

The MSIV's help provide the necessary protection against an unwanted depressurization of the Main Steam System piping. The MSIVs are used to help mitigate such accidents as steam piping failures, feedwater piping failures, and steam generator tube failures (UFSAR Sections 15.1, 15.2.8 and 15.6-3).

However, none of these accidents will be affected by this modification since the function of the MSIVs will remain unchanged.

No safety-related systems will be adversely affected by this modification.

All physical work will be performed on QA Category I wiring in junction boxes (2JB*3740, 1, and 2), (Rod Control Bldg., El. 773'). However, impact of this modification, on any system, is limited to the correction of the MSIV CLOSED /NOT CLOSED indications on the PCS and ERFCS, which are QA Category III and II, respectively, and are isolated from the QA Category I portions of the system as discussed in the QA Category Conclusion.

The probability of an occurrence of ar. accident prevdously evaluated in the safety analysis report will not be increased. No safety-related components or functions will be affected by this modification.

The consequence of an accident previously evaluated in the saf rey analysis report will not be increased. No systems required for mitigation of any of the previously analyzed accidents will be affected by this modification.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. The functions of all safety-related equipment, including the MSIVs, will remain unaffected by this modification. No new failure modes sa potential hazards will be created by the implementation of this modification.

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BOavar Valley Power Station Unit 2 j 1990 Report of Facility Changes,. Tests, and Experiments Page 102 of 152

.The - consequence of a malfunction of equipment important to eafety as previously evaluated in the safety analysis report will not-be increased.

This modification will enhance the ability- of plant personnel to retrieve information. regarding MSTV positions. All MSIV control functions, indicating lights, applicable OPEN/NOT OPEN computer points, and annunciator windows will remain unchanged. l Failure modes of the proposed design change which were reviewed included failure of the CLOSED /NOT CLOSED computer points to indicate properly. This would not degrade the system to any great extent since theos points do not currently exist at all. All other indications associated with the HSIVs will remain- unaffected by this modification; therefore, their functioning capability will remain unchanged.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created. This modification serves only to provide more complete MSIV indication to the PCS and ERFCS. No equipment functions will be changed and nc new failure modes will be introduced.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created. No safety-related equir, ment will be adversely af fected by this modification.

The margin of safety as defined in the basis for any Technical Specification (T.S.) will not be reduced. No T.S. bases, including that of T.S. 3.7.1.5,

" Main Steam Line Isolation Valves", will be affected by-this modification.

The proposed design -change -will noc require any changes to the technical specifications.

The proposed change will not require any changes to the Updated Final Safety l

Analysis Report.

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• Deever Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimento Page 103 of 152 CHANCE TITLE DCP-1313, Rev. O, Sprinklere for Turbine Building Pedestals and Amertap Condenser Pit CHANGE DESCRIPTION Additional automatic sprinklern fire protection capability has boon requested for the areas directly below the concrete pedoctais along the north and south side of the turbine condensers and below the numerous obstructions directly above the Amertap Condensate Pit area.

The proposed solution is to tap off existing pipes in the Turbine Building Mezzanine Elevation 752', and provide a closed head sprinkler system for these areas.

The basis for the proposed solution was an ANI 1974 recommendation. ANI requested an automatic sprinkler system be installed in all areas to which oil may spread in the event of a break in an oil line above the concrete pedestals along either side of the turbine condensero. Also, there is a large number of obstructions between the present oprinklers and the pit (i.e. piping and cable trays) that would not allow adequate sprinkler spray pattern coverage.

The mechanical boundaries of this change include the north and south side of the concrete pedesta19 along either side of the turbine condensors and another eet of oprinklera running cant and weet directly over the Amertap Condenser Pit Area.

Proposed system additiono to the current mechanical boundary consist oft

- A sprinkler header network extending from an existing 3" pipe (AstoA Dug. 16-1781 Sh. 9C) for the north side of the turbine pedestals, southwest of Columns Z and 7. (Elevation 771'4")

- A second sprinkler header network extending from an existing 4" pipe (ASCOA Dwg. 16-1781 Sh. 105) for the south side of the turbine pedestals east of Columnu 7 and X. (Elevation 769'-7 3/4")

The third oprinkler header network extends from an existing 2 1/2" pipe (ASCOA Dwg. 16-1781 Sh. 4A) for the center of the Amortap Condenser Pit Area south of Columns C and 9. (Elevati.on 741'-3")

A minimum of 12, 165*F, sprinkler heado. (Two under each pedestal arch for a total or four on the north side and four on the couth side and four heado directly over the Amertap Condenser Pit Area.)

EhFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previounty evaluated in the safety analysis report will not be increased because this modification does not change the probability that a fire will start in the area. Since no safety-related equipment is located in or near the turbine building banoment, electrical conduit flooding of safety-related componento due to a discharge of water from this modification is not increased. Also, no equipment that could {

cause a release of radioactivity to the environnient is affected.

B3 ver Valley Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 104 of 152 The consequence of an accident previously evaluated in the safety analysis report will not-be increased because this modification will provide add.cional fire suppression capability. Any failure of this modification can have consequences no worse than the present configuration, which is analyzed in Section 9.5.1 of the BVPS-2 UFSAR.

1The -probability of a malfunction of equipm it important to safety as previously evaluated in the safety analysis report will not be increased because no safety-related equipment is affected by this modification.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because no safety-related equipment is affected by this modification.

No design basis accidents from BVPS-2 UFSAR Chapter 15 are affected. This modification conforms to the basis that no single fire will cause unacceptable risk - to public health and safety, will not adversely affect the ability to l safely shut down the reactor, and will not significantly increase the risk of I radioactive release to the environment.

No safety systems will be affected by the proposed design change.

The. possibility for an accident of a different type than previously evaluated

'in the safety analysis report will not be created because the limiting accidents have been previously analyzed in Ses cion 9.5.1 of the UFSAR. This modification will serve to reduce the consequences of a turbine building fire.

The -possibility for a malfunction of equipment important to safety of a different . type than previously evaluated in the safety analysis report will not be -created because no safety-related equipment is affected by this modification.

Failure modes of the proposed design change which were reviewed included piping rupture or leak.

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!- The margin of safety as defined in the basis for any Technical Specification will not. be reduced because no Technical Spocification bases are affected by this modification.

! No changes in parameters that affect the course of any accident analysis l supporting Technical Specification bases and result in exceeding the acceptance -criteria for fuel cladding, RCS boundary, or containment boundary are affected by this modification. The proposed design change will not require change to-the technical specifications.

l The proposed change will require a change to the Updated Final Safety Analysis Report, Figure 9.5-2.

I Dsaver Valley Power Station Unit 2 1990 Report of Facility Changen, Tests, and Experiments Page 105 of 152 CHANGE TITLE DCP-1314, Rev. O, Ground Test Circuit for Due 2-5 and Bus 2-6 gj!ANGE DESCRIPTION At the present time a ground on either DC Bus 2-5 or 2-6 is annunciated in the control room as a ground on both DC Buses. The two DC bus ground detection circuits are interconnected through the isolation power diodes in logic cabinet 2IHA-LOG-CAB. There presently is no easy way to determine which bus has the ground fault. The proposed solution is to install three ground test circuit breakers within the annunciator logic cabinet to isolate the common

" Ground Trouble" annunciator when a ground appears on either DC Bus 2-5 or 2-6. These breakers will be used for troublechooting to determine upon which bus the ground fault is located and to isolate the faulted bus from the nonfaulted bus while repairs are being made.

SAFETY EVALUATION

SUMMARY

No. Chapter 14 or 15 design basis accidents are impacted by this design change.

No safety systems are affected by this design change.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report because the system configuration is not changed by this design change. The breakers will be added to assist operators in searching for ground faults.

The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report because the analysis in Section 8.3.2.1.2 of the UFSAR is not affected by this modification. The Vital Puses described in this noction are not affected by this design change.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because this circuit performs an indication function only to detect ground faults, thus it performs no control or protective functions. No changoo are being made to the ground fault circuit that affects how it operates. The addition of these breakers will provide the additional capability to troubleshoot to determine the location (Bus 2-5 or 2-6) of the ground fault. No equipment important to safety is affected by this design change.

The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because no radiological systems are affected and thus the consequences of malfunctions of equipment important to safety are not increased.

The only credible failure mode is that one or both of the breakers could trip open. If this happens, it will have no effect on the operation of the essential bus system; only ground detection capability will be impaired. )

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 106 of 152 The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safetyinnalysis report because this -design change performa no protective, indicating, or control functions that affect the operation of any system. These breakers will be used to track ground faults in the Essential DC System.

The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report. The system configuration of the circuits are not changed. The added breakers are to be used for troubleshooting and perform no protective functions. No new malfunctions are created.

There are no _ changes in parameters that affect the course of any accident analysis supporting Technical Specification bases.

No Technical Specification bases are affected by this design change. The proposed desiqn change will not require change to the technical specifications.

The proposed change will not require changes to the Updated Final Safety Analysis Report.

Daavor Valley Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 107 of 152 Q}{ANGE TITLE DCP-1321, Rev. O, Annunciator Window AB-60, " Heat Tracing System Trouble" pHANGE DESCRIPTION This design -change- is being made to comply with the NRC's " dark board" l concept. Currently, annunciator AB-6B receives inpute from QA Category II heat trace panels 2 HTS-PNLNICP and 2 HTS-PNLN2CP on any undertemperature, overtemperature, or loss of power (LOP) alarm. The temperature alarms occur frequently due- to normal system operation and flow paths of the Steam Generator Blowdown, Solid' Waste Bisposal, and Condensate Systems. This-modification proposes to remove the undertemperature and overtemperature-alarm inputs to the plant computer and annunciator A8-6B. The LOP inputs will remain. All cf the inputs (undertemperature, overtemperature, and LOP) to j panels 2 HTS-PNLNICP and 2 HTS-PNLN2CP will also remain.

CHAN7E DESCRIPTION There .are no previously. analyzed accidents that will be affected by this modification. These- temperature alarms are not associated with any-safety-related systems nor are they needed to assure operability of any system tequired for accident mitigation.

No safety systems will be affected by the proposed design change.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report since none of the accidents will be affected in any way by this modification.

The- proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report since no safety-related systems or systems -required to help mitigato any of these accidents will be-affected by this modification.:

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'The design change will not increase the probability of. occurrence of a malproposed

• unction ~ of equipment important to safety:as previously evaluated in the safety analysis report. There - is no safety-related equipment that-will be affected by this modification either directly or indirectly.

The proposed design change will not increase the consequences of a malfunction of equipment important 'to- safety as previously 'ovaluated in the safety analysis report. No safety-related equipment will be affected by this modification. The components involved in this change are all non-safety related. The alarm inputs -to local heat trace panels 2 HTS-PNLNICP and 2HTSPNLN2CP will' be unaffected by this modification. Operations personnel will continue to monitor these~ panels.

No new credible failure modes will be introduced by the implementation of this modification. The temperature alarms will continue to provide input to the-local panels in- the same manner as before. Therefore, the probsbility of failure will not be increased and no new failure modes or potential hazards will be introduced.

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Beaver Volley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments ,

Page 108 of 152 The proposed design change will not e ste the possibility of an accident of a

'different type than previously evalv ced in the safety analysis report. All system functions will remain unchanged. l The proposed design change will not create the possibility of a malfunction of ,

a different type than previously evaluated in the safety analysis report. All component functions will remain unchanged.

There are no changes in parameters which affect the cource of any accident anclysis supporting Technical Specification bases and that result in exceeding the acceptance criteria for fuel -cladding, RCS boundary, or containment integrity.

The proposed design change does not reduce the margin of sat.cy as defined in the basis for any Technical Specification (T.S.). No T.S. or its bases will be affected by this modification including T.S.s in Section 3/4.3,

" Instrumentation." The proposed design change will not require any changes to the technical specifications.

The proposed change will not require any changes to the Updated Final Safety Analysis Report. ,

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Banvar Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 109 of 152 CHANGE TITLE DCP-1328, Rev. O, Inadvertent Actuation of SSPS Outputs CHANGE DESCRIPTION At present, some of the output relays of the Solid State Protection System (SSPS) can be activated by two signals, one being the logic of the SSPS and the other by means of a test panel. The test panel permits periodic testing of the Engineered Safeguards Teatures (ESP) actuation scheme of the QS pumps in compliance with Technical Specification (T.S.) surveillance requirement 4.6.2.1.

SSPS output relays K643 and K644 actuate co provide start signals to the Quench Spray (QS) Pumps (2QSS-P21A(B)) and auto open signals to the pumps, suction and discharge valves, (2QSS-MOV100A(B)] and [2QSS-MOV101A(B)]. On a QS demand signal the SSPS logic causes the suction and discharge valvec to open and the pumps to auto start. Two redundant master relays in each circuit (K505 & K519), located on the neutral side of the associated output relay, energize to close. This action completes the circuits to open the valves and to start the pumps.

To anable testing of the K643 and K644 output relays without spraying down containment,, the relays' test circuits are interlocked with each other and with field contacts (valve limit switch contacte and 4KV breaker auxiliary switch contacts) so that both relays cannot be actuated at the same time during testing. In the test mode of operation, the neutral sides of the relay circuits are completed through test switches S842 and S843 via those interlocks. Since the output relays are controlled by their neutrals, the relays can, and have been, accidentally energized by simply grounding any part of their test circuits between the relayo s. the test switches.

This modification proposes to prevent accidental activation of the output relays by moving the test switches to a location between the output relays and the associated field permissives. This will offectively isolate the field test interlock circuits from the output relays; therefore it will prevent accidental activation of the relays should the field circuits be grounded.

Both trains will be modified as described above. The changes will be restricted to the internal wiring of the Train A and a safeguards test cabinets. the functions of the QS System and the ESP System will remain unchanged.

SAFETY EVALUATION SyMMARY The Quench Spray System is used for containment depressurization following such accidents as a steam line or feed line break inside containment or a loss of coolant accident (LOCA). These accidents are discussou in UFSAR Sections 15.1, 15.2 and 15.6.

The safety systems that will be impacted by the proposed design change include the Quench Spray System and the Engineered Safeguards Features System.

However, this change will not affect the design function of any system components or of either system as a whole. .

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Denver Valley Powar Station Unit 2 1990 Report of Faci'ity Changes, Tests, and Experiments Page 110 of 152 The proposed design change will nc' increase the probability of occurrence of an accident previously evaluated ti the safety analysis report. This modification will not affect the design function of any safety-related components. It is a change only to the test circuitry, located in the Safeguards Test cabinets, and for the QS pumps and their associated suction and discharge valves. The change will not affect any existing test or operating procedures; therefore no procedure changes will be required.

Thefproposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report. No components or systems required to -mitigate any of the previously analyzed accidents will be adversely affected.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. No safety-related equipment, including that of the QS and ESP Systems, will be adversely affected.

The proposed design change will not increase the consequences of a malfunction ol equipment important to safety as previously evaluated in the safety analysis report. All safety-related equipment functions will remain unchanged.

No new credible failure modes will be introduced by this modification. The physical changes associated with this modification involve the rewiring of existing test switches. The rewiring will be restricted to the internal wiring of the Safeguards Test cabinets and will comply with QA Category I, seismic requirements.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report. The functions of all systems will remain unaffected by this modification.

The proposed design change will not create the possibility of a malfunction of a different type than.previously evaluated in the safety analysis report. The functions -of all components will N unaffected by this modification. Even the method of testing these components will be unchanged.

There are no changes in parameters which affect the course of any accident analysis supporting Technical Specification bases and that result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification (T.S.). None of the T.S. bases will be affected by this modification including that of T.S. 3/4.3.2 and 3/4.6.2.1. The proposed design change will not require any changes to the technical specifications.

The proposed changa will not require any changes to the Updated f inal Safety Analysis Report.

Bacvor Valley powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 111 of 152 CHANGE TITLE DCP-1336, Rev. O, Nuclear Instrumentation System Source Range High Voltage cutoff CHANGE DESCRIPTIOH problems have been experienced with the Source Range (SR) Drawer high voltage cutoff circuit, namely SR high voltage has boon turned on and off unaxpectedly. This DCP will modify the circuitry to eliminate the unexpectec, spurious turn-on and turn-off of the SR high voltage. The function of the circuit will remain the same, but it will becomo more reliable.

The more reliable operation of the automatic high voltage cutoff circuit will eliminate the need to remove fusos during power operation which de-energizes the source range flux doubling circuit. Removing the funes activates the source range flux doubling trip annunciator light which does not support the NUREG-0700 " Dark Board" Concept. This DCP, therefore, is the response to licensing commitment 6.A4-3D (Licence Condition No. 2.C.8, Detailed Control Room Dosign Review, Annunciator " Dark Board" item A,6.

A High Voltage manual control switch, S104, vith three positions (On, Normal, and Off) will also be added to the drawor. The switch will be positioned to

" Normal" except as described in the following:

With the addition of the High Voltage Manual Switch S104, the operator can manually turn on the Source Rango high voltage when approaching the Source Range levels to determino if the SR channel (detector, electronics, etc.) in operable. If it to found to be inoperable, then the operator can take appropriate action.

During maintenance. and surveillance testing of the Source Range channel, the technician can place the HV manual control S104 in the "Of f" position to avoid installing jumpero in the rear of the NIS console in order to deenergize the SR high voltage.

With the high voltage manual switch S104 in the "On" position, the input error inhibit is in the inhibit position ooprator can keep the cource ranges energized when shutdown and the SSPS input error inhibit is in the inhibit position.

The switch will only be placed in the "Off" position for maintenance and testing, and these activities are not expocted when the protection system must be operable. Therefore, per item 3C of FSAR Section 7.5.5, automatic Dypass and Inoperable Status Indication (BISI) is not required for this switch. BISI may of courne be performed administratively whenever necessary.

A light on the SR drawer will illuminate if the switch is not in the " Normal" position. Thio DCP will meet all requirements of the applicablo code IEEE 279 per the lead engineer.

Annunciator A4-4E, "NIS Detector / Compensator Loos of Voltage" currently alarma for a loss of high voltage to olther SR detector or Intermediate Range detector loss of voltage. The SR eignal will be rewired at 2CES-SLTCAB to

Decvor Volloy Pow r Stction Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 112 of 152 alarm if the high voltage fa on above P-6 or off below P-6. This is considered an accurate and acceptable way to indicate trouble for the SR high voltage. Annunciator Window A4-4E's nomenclature will be modified to indicat6 HIS Detector / Compensator Trouble.

Source range drawers and Annunciator Window A4-4E are located in the Unit 2 Control Room Status Light Cabinet, 2CES-SLTCAB is located in the Control Building, Elevation 70, Sout t.we st Corner. There is one spare source range drawer in Storeroom 22 which will also be modified.

Tris DCP will maintain the reliability, integrity and operability of the Source Range Nuclear Instrumentation System (SRNIS) and associated alarms and sill have no advSrse effects on any other equipment.

Eh[?TY IV LUATION

SUMMARY

No Unit . Chapter 15 accidente vill be affected by this DCP because this DCP does not sdversely affect an/ safety or non-safety systems, does not exacerbate an/ cxisting accidents, and does not introduce any new hazard beyond that already considered in the UFSAR.

This DCP will not adversely affset the safety function of any system. The reliability, integrity, and operability of the SRNIS wi?.1 be maintained and no other systems will be affected.

The probability of an occurrence of an accident previously evalucted in the UFSAR will not be increased. This DCP will maintain the reliability, integrity, and operability of the SRNIS, and it will have no effect on any other equipment; therefore, no probabilities of occurrence of any accidents will be increased.

The consequences of an accident previously evaluat.ed in the UFSAR will not be increased. This DCP is minor and the change will nave no effect on any other equipment. This DCP will not affect aay paramster which would increase the consequences of an accident beyond that previously considered in the UFSAR.

The reliability, integrity, and operability of the SRMIS is being maintained.

The SRNIS is not credited in any FSAR accident analysis as indicated by its absenca from FSAR Table 15.0-4.

The probability of a malfunct! m 3f eq'. ipment important to safety as previously evaluated in the UFSAR wt;l ne; be increased. This DCP is minor and the changes will not adversely affect any equipment, including the SRNIS.

The conraquencea of a malfunctien of equipment importrnt to safety as previously evaluated in the UFSAR will not be increased. This DCP will not adversely affect any parameter which wcuAd increase the consequences of a malfunction. This DCP will not adversely affect any safety system used to mitigate an Accidant. Therefore, there will be no effect on the consequences of a malfunction of equipment important to safety.

This LCP vi); not cause any new credible failure modes because the fundamental deeign features and functions of the equipment have not been significantly altered. This is a relatively minor change.

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Bsavor Valley Powsr Station Unit 2 1990 Report of Facil.cy Changes, Tests, and Pwportments i Page 113 of 152 The possibility for an accident of a different type than previously ev.luated in the UFSAR will not be created. This DCP is minor, and the reliability, integrity, and operability of the SRNIS will be maintained. Nothing is being added or altered in a way which creates the possibility of a dif ferent type of accident.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR will not be created. This DCP is minor and the reliability, integrity, and operability of the SRNIS will be maintained. The fundamental design features and functions will not be changed in a way that creates the possibility of a malfunction of a different type.

This DCP will not change any parameter which affects the course of any accident analysis supporting Technical Specification bases.

The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this DCP will not adversely affect the margin of safety as defined in the bases for any Technical Specifications because the reliability, integrity, and operability of the SRNIS will be mal;ttained, and no other equipment will be affected.

This DCP will rot require any changes to the Technical Specifications. As long as the manual switch, S104, is only pinced in the "off" position for r maintenance and testing, there is no change required to the Technical Specifications or bases. If the manual switch, S104, is in the "off" position and a reactor t:1p occurs, the SR trip and high voltage will not be automatically restornd below the P-6 setpoint as stated in the bases of T.S.

2.2.1. under. " Reactor Trip System Interlocks". However, as long as the manual switch, S104, is only "off" for maintenance and testing, the current bases remain- correct and do not need to be changed. When the switch S104 is in the "off" position the source range channel shall be considered inoperable per the applicable Tech. Specs.

This DCP will require changes to the UFSAR. UFSAR Sections 7.2.1.1.2, 7.2.1.1.3 and Figure 7.2-1 (Sh. 3 of 18) need to be updated to describe the added manual switch, S104.

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B3avor Valley Powsr Station Unit 2 1990 Report of racility Changes, Tests, and Experiments Page 114 of 152 CHANGE TITLE DCP-1339, Rev. O, C/NC Contract Indication CHANCE DESCRIPTIOR Closed /Not Closed (C/NC) contact indication is expected by the Plant Computer System (PCS) and the safety Parameter Display System (SPDS) function of the Emergency Response Computer System (ERFCS) in accordance with WCAP 10170 for the valves listed below.

Currently a- contact to provide the C/NC indication for these valves is taken off a Limitorque rotor which is set to provide an electrical bypass of the j open torque switch at the beginning of the opening stroke. On a closing -

stroke, this switch changes state before the valve is completely closed. As a result, the computers show that the valves are closed slightly before the flow path is completely shut off. If a valve were to stop between the change of.

state point and the full shut off position, a flow path through the valve could exist even though a CLOSED indication had been achioved. Concerning human factors it is undesirabis to have a mismatch between the bench board indicating lighta and the computer displays especially during an accident as the above valves' positior. is used to evaluate safety status. .

Accurate C/NC position indication to the computers can be provided for each '

valve as follows:

.Y.911E E9h.t.1911 2 SIS *MOV863A&B Rewire Use spara relay to multiply contacts Run wire in MCC Run wire from MCC to Limitorque 2 SIS *MOV8887A&D Lift and land 2 wires: Use contact 14 2 SIS *MOV8811A&B Lift and land 2 wires: Use contact 14 SAFETY EVALUATION

SUMMARY

A rev:ow of the accident scenarios in Chapter 15 of the Unit 2 UFSAR was pe r f or _.sd . None of these accidents are affected by modifications being made per this design change.

The only safety . systema _affected are the Safety Parameter Display System (SPDS) (Sections 7.5.6 and.18.2 of the BVPS-2 UFSAR) and the Safety Injection dystem (Section 6.3 of the BVPS-2.UFSAR). The operation and descriptions of the Safety Injection System are not affected by this design change. The operation of the SPDS is not changedt however, this design change will correct deficiencies in . inputs to the SPDS to provide more accurate information to the i control room operators concerning the status of certain motor operated valves in the Safety Injection System.

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l Beaver Valley Power Station Unit 2 1990 koport of racility Changes, Tests, and Experiments Page 115 of 152 The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report because the SPDS and the ERFCS perforr no control or protective functions that could lead to an accident.

The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report because this design change will provide more accurate SIS flow path information to the operators during an accident than presently exists. This will increase the confidence level of the operatnr in the information provided to him via the SPDS or the ERFCS and will reduce the likelihood of an operator error that could increase the consequences of an accident.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because the SPDS and the ERFC6 perform no control or l protective functions over any plant equipment and thus the probability of-occurrence o f- a malfunction of equipment important to safety is not increased for any Chapter 15 accident. )

i The proposed design change will not increase the consequences of a malfunction i of equipment important to safety as- previously evaluated in the safety analysis report because this design change will correct existing deficiencies in the C/NC circuits to provide more accurate information to the plant operators.

-No new credible failure modes can be postulated. No existing failure modes, L such as wiring damage or contact switch malfunctions, can lead to.an accident

( or malfunction of equipment important to safety, although the confidence level i .of the operator in the validity of this information may be affected.

i The proposed design change will not create the possibility of an accident of a different type- than previously evaluated in the safety analysis report no new accidents are created by this design change because it will correct existing deficiencies in the C/NC indication for the SIS HOVs affected.

The proposed design change will not create the- possibility of a malfunction of a different type than previously evaluated in the safety analysis report i because it _will correct exieting deficiencies in the C/NC indication for the SIS HOVs affected.

No changes are being made. that affect the course of accident analyses that support Technical specification bases, l

l No Technical Specifications are affected by this change. The proposed design change will not require change to the technical specifications.

The proposed change will not require changes to the Updated Final Safety Analysis Report.

The proposed change will not involve an Unreviewed Environmental Question.

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B3Svar Valloy power Station Unit 2 1990 9eport of Facility Changes, Tests, and Experiments Page 116 of 152 C3ANCE TITLE U- 1355, Rev. O, 2CHS*P21A and C Vents CHANGE DESCRIPTION Hydrogen gas sometimes collects in the piping of Charging Pumps (2CHS*P21ACC) when they are not running, and it to recognized that starting a sharging pump with hydrogen voids is not desirable. Presently, the gas is removed through the use of a manual venting procedure that employs the use of tygon tubing.

This has resulted in equipment and personnel contamination.

l This modificatien proposes to maintain the ALARA concept by installing permanent piping for a manually operated vent system for pumps A and C. (The  ;

I piping arrangement does noc allow accumulation of hydrogen gas in pump D).

SAFETY EVALUATION

SUMMARY

The charging pumps, used as High Head Safety Injection (HilS!) pumps, are required to operate following the occurrence of several analyzed accidents, including: Accidental Depressurization of the Main Steam System, Feedwater Syssem Piping Break, Inadv rtant Opening of a PORV, Steam Generator Tube Rupture, and Loss of Coolant Accident.

l The safety systems that will be affected by tha proposed design change are the Chemical and Volume control and the Safety Injection Systems. However, no i adverse effects will be -created by this modification. ihis modification serves only to more easily vent the HHSI pumps in order that their reliability may be maintained.

The proposed design change will not increase the probability of occurrence of

. an accident previously evaluated in the safety analysis report. No changes will be made that could affect any of the initiating events for any of the previously analyzed accidents. Thorefore, the probability of occurrence of any of the accidents will not be increased.

The pt; posed design char.ge will not increase the consequences of an accident previously evaluated in the safety analysis report. The function of all safety-related equipment, including the HHSI pumps and the Volume Control Tank (VCT) will. remain unchanged.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety Le proviously evaluated in the safety analysis report. No component functions will be adversely affected by this modification.- This change will-improve the method of venting the HHSI pumps by replacing tygon tubing with permanently installed vent piping. This will help enhance the ability to maintain the HH3I pumps and piping in an acceptable condition.

The proposed design change will not increase the consequ6nces of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. No safety-related system or component functions will be adversely affected by this modification.

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Deover Valley power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments page 117 of 152 No new failure modes or potential harards will be introduced by the implementation of this modification. An existing, temporary arrangement will be replaced with permanent piping, valves, and sight flow indicators. The components will meet the required temperature and pressure ratings and will be installed as QA Category I, Safoty class 2, except for the sight flow indicators, which will be installed as QA Category II, Safety Class 4. All of the components will meet seismic requirements.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report. The HHSI pumps will continue to be manually vented using permanently installed components rather than temporary rigging. No safety-related cystem functions will be adversely affected by this change.

The proposed design change will not create the possibility of a malfenction of a different type than previously evaluated in the nafety analysis report. No safety-related component functions will be adversely affected by this change.

. valves- installed _ as part of this modification will be normally closed.

Operating procedures will be revised to provide instructions on the proper method of venting the pumps using the new installation.

This modification will- not produce changes in paremeters which affect the course of any accident analyses supporting Technical Specification bases and that result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or containment integrity.

The proposed design change will not reduce the margin of safety as defined in the basis for any Technical Specification (T.S.). No T.S. or its bases, including that of T.S. 3/4.1.2.3, 3/4.1.2.4, 3/4.5.2, and 3/4.5.3, will be affected by this modification. The proposed design change does not require any changes to the technical specifications.

The proposed change does require a change to the Updated Final Safety Analysis Report. Figures 9.3-21 and 9.3-25 must be revised to include the new vent paths.

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Boavor Valley Pow 2r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 118 of 152 CHANGE TITLE DCP-1360, Rev. O, Installation of How Liners Under 2HDH-P21A&B CHANGE DESCRIPTION During plant operatio.1, steam blows out from beneath the solo plate of the heater drain pumps, [2HDH-P21A&D). The steam is a safety hazard to personnel and is corrosive to the pump components and the concrete.

This modification proposes to eliminate the steam by eliminating sources of ~

water entering the pump pits. A bottom plate will be fabricated for the pump pits. It will be installed over the existing concrete plug at the bottom of

-the pits -and- will be welded to the steel easing lining of the pits. Also, a-curb will be constructed around the pumpa' sole plates to prevent water from running into the pump pit.

SAFETY EVALUATION

SUMMARY

No design basis accidents will be impacted by this modification. Because of the . relationship of the Heater Drain System to the Feedwater System, UFSAR Sections 15.2.7 and -15.2.8 were reviewed but found to be unaffected by this modification. There are no design basis accidents associated with the Hector Drain System. ,

No safety systems will be affected by the proposed design change.

The -proposed- design change will not increano the probability of occurrence of an accident previously- evaluated in the safety analysis report. Since these pumps supply water to the suction of the Main Feedwater Pumps, the loss of there pumps could initiate a loss of normal feedwater flow. In this case, the

-Auxiliary- Feedwater would then supply sufficient flow to the steam generators. However, this modification will not affect this design function of the Heater Drain Pumps, and therefore will not increase the probability of l

this type of failure.

I c The- proposed design change will not increase the consequences of an accident

! -previously evaluated in the safety analysis report. None of the systems required to mitigate a previously analyzed accident (including the Anxiliary Feedwater System) will be affected by this tuodification.

The proposed design ahange will not increase the probability of occurrence of a ' malfunction of equipment important to safety as previously evaluated in the safety analysis report. No safety-related components will be affected by this modification. The Heater Drain Pumps themselves are non-safety related, QA Category II.

. The proposed design change will not ! crease the consequences of a malfunction of equipment important .to safety is previously evaluated in. the safety analysis report. No components required to operate due to the fallare of snother safety-related component will be affected by this modification.

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Bsavor Valley Powar Station Unit 2 1990 Report of racility changes, Tests, and Experiments Page 119 of 152 The proposed design change will not create the possibility of an accident of a different type than previously evaluated ir the safety analysis report. The functions of all systems (both safety and non-safety related) will remain unchanged.

The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report. The functions of all equipment and the ability to perform those functions will remain unchanged.

There are no changes in paramotors which affect the course of any accident analysis supporting Technical Specification bases and that result in exconding ,

I the acceptance critoria for fuel cladding, RCS boundary, _or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification (T.S.). None of the T.S.s or their bases will be affected by this modification. The proposed design change will l not require any changes to thu technical specifications. l

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The proposed change will not require any changes to the Updated Final Safety Analysis Report.

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• i B2cvar Valley Pow 3r Station Unit 2 1990 Report of Facility Changes, Tes>s, and cxperiments Page 120 of 152 CHANGE TITLE DCP-1364, Rev. 1, Dsd Position Deviation Alarm CHANGE DESCRIPTION Software changen have been implemented on Unit #2 PCS to provide a computer contact output for rod position deviation. Engineering is required to provide wiring from the computer contact output to a spare annunciator window.

This design change is to provide an audible e.larm utilizing an existing cable between the computer output cabinet to the multiple input cabinet. This alarm will annunciate on annunciator window box A4-80. Annunciator window A4-80 will be engraved to read " Rod Position Deviation". The difference of Rev. 1 to Rev. O is that the cabinet has been changed from 2IHA-HI-EXP, Bay 33 to 21HA-LOG-CAB, Bay 22.

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because tne plant computer system does not have a safety function and its failure will not affect the safety functions of other equipment.

The consequence of an accident previously evaluated in the safety analysis report vill not be increased because this modification on the plant computer system has no effect on consequences of an accident previously evaluated in the UFSAR Chapter 15.

The probability of a malfunction of equipment importent to safety as previously evaluated in the safety analysis report will not be increased because the modification of the plant computer system does not create a situation which would increase the probability of a malfunction.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the plant computer system is not used to pr6 vent or mitigate the consequences of an accident. There should be no effect on the consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR.

The safety systems which will be affected by the proposed design change include reactor control and protection system.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because there is no configuration change such that an accident of a different type is created.

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i Beaver Valley Power Station Unit 2 1 1990 lleport of Facility Changes, Tests, and Experiments Page 121 of 152 The possibility for a malfunction of equipment important to safety of a different type than previously evalusted in the safety analysis report will not be created because the failure of this equipment has no effect on safe plant operation and shutdown of the plant, and the failure of this equipment i will not affect the safety function of other equipment. i Failuro modes of the proposed design change which were reviewed included equipmont failure of the plant computer system.

The margin of safety as defined in the basis for any Technical Specification will not be reduced because the change does not affect the basis of the i Technical specification, Section 3/4.1.3.2. The proposed design change will not require change to the technical specifications.

l The proposed change will not require changes to the Updated Final Safety Analysis Report.

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D3 aver Valley Pow 3r Station Unit 2 1990 Report of Facility changes, Tests, and Experiments Page 122 of 152 CllANGE TITLE DCP-1366, Rev. O, Removal of Brace from Duct Support 2HVP-DSA684N CHANGE DESCRIPTION The aisleway to' and from the elevator at elevation 773'-6" in the Auxiliary Building (Aux. B1dg.) is impeded by the presonee of a seismic support brace.

This brace is used as a lateral restraint for seismic support (2HVP-DSA684N),

which is on the exhaust duct for Aux. Bldg. Air conditioning Unit (ACU),

(2HVP-ACO211B).

This modification proposes to increawe the usable width of the aisleway by either relocating the brace or by removing it entirely. This modification will not degrade the seismic capacity of the support.

This' safety evaluation assumos that an adequate reanalysis of the seismic support is performed and that it concurs with the conclusions of the preliminary reanalysis. If this analysis indicates that the brace must be relocated, the safety evaluation will have to be re-evaluated to include possible effecto of the brace on the now location and on the method of installation.

SAFETY EVALUATION

SUMMARY

None of the previously analyzed accidents, including that of UFSAR Section 15.7.4, " Radiological Consequence of Fuel Handling Accidents", will be affected by this modification. In the case of a fuel handling accident in the Fuel Building, the atmosphere would be filtered by the Supplementary Leak Collection and Release System (SLCRS), UFSAR 6.5.3. The operation of the normal Aux. Bldg. Ventilation System, UFSAR 9.4.3, would not be required.

No safety-related systems or functions will be affected by this modification.

The brace and support are located on a QA Category II portion of the Ventilation System. The Auxiliary Building -Ventilation System is.used to provide a- suitable environment for personnel and equipment operation and minimizes the potential for the spread of airborne radioactive material within the building during normal operation. The normal operating portions of this system, including (2HVP-ACU211D), are considered as QA Category II and are not requirsJ to operata during accident conditions. A QA, category I backup system -(UFSAR 6.5.3) is provided for those segments of the building with equipment required - during accident conditions. one exhaust path of (2HVP-ACU211B),. downstream of support (2HVP-DSA684N), provides the main purgo supply for containment during periods of occupancy such no during retualing.

Although this. ductwork is seismically su pport ed , it doen not provide a QA Category I -function except for the portion between and including the inside and outside isolation valves (UFSAR 9.4.7.3.1, paragraph 4). This modification will have no effect on this safety-related portion of the system.

The probability of an occurrence of an accident previously evaluated in the '

safety analysis report will not be increased. No safety-related components or functions will be affected by this modification.

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B3 aver Valley Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 123 of 152 The consequence of an accident previously evaluated in the safety analysis report will not be increased. No systems required for the mitigation of any of the previously analyzed accidents will be affected by this modification.

The probability of a malfunction of equipn.ent important to safety as previously evaluated in the safety analysis report will not be increased. The functions of all safety-related equipment will remain unaffected by this modification. Even though the part of the ventilation system in close proximity to this modification is not safety-related, it is required to be seismically supported, and it will remain seismically supported. Therefore, no new failure modes or potential hazards will be created by the implementation of this modification.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased.

This modification will have no effect on any equipment important to safety.

Failure modes of the proposed design-change which were reviewed included the failure of the seismic support, and the associated ductwork, because of the brace removal. However, since the ductwork will remain in compliance with seismic requirements, this is not a credible failure mode.

Tra possibility for an accident of a different type than previously evaluated in the safety analysis repart will not be created. No equipment or system functions will be affected by this modifteation.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created. This modification will not affect the safety-related function of any equipment including that of the containment purge line isolation valves (dampers).

This modification will not adversely affect any parameters pertaining to fuel cladding, RCS boundary, or containment integrity. Containment integrity is further discussed below.

The margin of safety as defined in the basis for'any Technical Specification (T.S.) will not be reduced.- T.S. 3.9.9, "00ntainment Purge and Exhaust Isolation- System"- assures the ability of the sentainment penetrations of this

-s ystem to isolate in the event of a fuel nandling accident. The isolation dampers on the containment purge line are located downstream of the ACU in one of its discharge flow paths. However, they are located sufficiently downstream and are designed such that this brace removal will not' adversely affect their isolation function. These dampers and the piping between them l are safety-related and are designed, installed, and_ tested as QA Category I,

( Seismic Category I components. Even if the unlikely failure of the equipment

l. or ductwork anywhere upstream of these dampers were to occur, both dampers l would retain their ability to close. Therefore, the margin of safety is not l reduced. The proposed design change will not require any changes to the technical specifications.

The proposed change will not require any changes to the Updated Final Safety Analysis Report.

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DCP-1371, Roy. O, Replace Existing 200S-300 Globe Valve with a 2" cate Valve CHANGE DESCRIPTION The. primary drains header presently has only one isolation valve, [200S-300),

as the boundary between safety class 1 and non-safety-related piping. This modification proposes to replace 2005-300 with a now 2* gate valve. The existing valve is leaking bl and cannot be repaired.

SAFETY EVALUATION

SUMMARY

l None of the previously analyzed accidents of UFSAR Section 15 will be affected by the addition of this valvo. The Reactor Plant Vent and Drain system is discussed in Section 9.3.3.

The portion of the Reactor Plant Drain System where the new valve will be added is considered as safety-related piping. However, the addition of this manual isolation gato valve will not have any effect on system function. It will serve only to provide better isolation capability for the drain header.

The probability of an occurrence of an accident previously evaluated in the safety analysis _ report will not be increhsed. This modification serves only to enhance the isolation capability of .the primary drains header. No previously analyzed accidents will be affected by this chango.

The consequence of an accident previously evaluated in the safety analysis report will not be increased. This modification will not change any system design functions nor will it provido any active safety functions.

The probability of a malfunction of equipment important to safety as previonely evaluated in the safety analysis report will not be increased. No safety-related- equipment or functions will be advorsely affected by this change. Because the now valvo will be purchased and installed to QA Category I and Seismic category I requiremonta, it will be doomed to be as reliable as the original design.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis roport will not be increased. The valve added by this- modification will replace the existing isolation valve.

This modification servos only to enhance isolation capabilities. No system functions will be affected.

l Failure modos of the proposed design change which were reviewed included the i leaking by of this valvo.- Existing valvo [2DGS-300) will be replaced _with a now valvo, which will provide isolation as tne existing valve currently does.

I By the replacement of [200S-300}, the primary drains header reliability and maintainability will actually be enhanced.

The . possibility for an accident of a different type than previously evaluated

'in the safety analysis report will not be created. No system functions will be changed. This modification will replace the existing valve.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 125 of 152 The possibility for a malfunction of equipment important to safety of a different type than previously evsluated in the safety analysis report will not be created. No other equipment will be affected by this modification.

This modification will not adversely affect any parameters pertaining to fuel i cladding, RCS boundary, or containment integrity. It will, if fact, provide isolation between- the RCS's excess letdown lines, loop drain lines, the spray line scoop drain line, and the Primary Drains Transfer Tank (inside containment). ,

i The margin of safety as defined in the basis for any Technical Specification (T.S.) will not be reduced because no T.S. bases, including that of T.S.

3/4.4.6.2, will be affected by this modification. The proposed design-change will not require any changes to the technical specifications.

The proposed change will require a change to the Updated Final Safety Analysis Report. Figure 9.3-13 must be revised to include the new valve.

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Bsavor Valley Power Station Unit 2 1990 Report of Facility ChangeL, Tests, and Experiments Page 126 of 152 CHANGE TITLE DCP-1390, Rev. O, Containment Air Recirc Fans 2RVR-FN201A and 2HVR-FN210B Status EllM(QE DESCRIPTIOff currently, when the breakers for the containment air recirculation fans 2HVR-FN201A and B are racked in and closed, the fans are on and tha F0S and ERFCS computers indicate that the fans are on. However, when the breakers are racked out and the circuits are de-energized (fans are off), the PCS and ERFCS computers also indicate that the fans are on. This modification will correct the situation by rewiring the breakers to existing normally open spare contacts and set the ' INVERSION function in both of the computers. By performing this design change, the proper on/off status for the containment air recirculation fans will be provided whether the breakers are racked in or out.

SAFETY EVALUATION

SUMMARY

There are no SVPS Unit #2 UFSAR Chapter 15 design basis accidents which are impacted by the proposed design change. This DCP will only correct the fan status information provided to the computers; therefore, there will be no effect on any accidents analyzed in UFSAR Section 15. In addition, accidents due to electrical failures should be avoided by disconnecting the power supply to the circuits prior to changing the contacts.

The BVPS Unit #2 safety systems which will be affected by the proposed design change include the containment air recirculation fans, which are non-safety related; 'however, tney are powered from an emergoney power supply and are connected to safety-related circuit breakers. There will be no adverse effect on any system.

The proposed design change increase will not the probability of occurrence of an accident previously evaluated in the safety analysis report because this design change will not affect the operability of the c-Itainment atmosphere recirculation ayetem as described in UFSAR Section 9.4.7.1.

-The proposed design change will not increase the onsequences of an accident previously evaluated in the safety analysis report because UFSAR Section 9.4.7.1.3 states that the containment atmosphere recirculation system is not required to operate during accident conditions. Since this modification will not affect the operation of the fans, consequences of previously evaluated accidents will not increase.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because the containment air-recirculation fan controls will remain unchanged, and therefore will still automatically atop the fans on an SI signal, a high containment water level, er a fan high-high vibration signal to protact the integrity of the emergency power source, as stated in UFSAR Section 9.4.7.1.3.

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Daevor Valloy Powsr Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 127 of 152 The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluatad in the safety analysis report because this modification will be performed to provide the proper PCS and ERFCS computer indications for the on/off status of the containment air recirculation fans and will not increase the consequences of a malfunction of safety equipment.

There is no change in failure modes associated with this proposed design change since nothing will be added or deleted, just a change of terminals to un existing spare set of normally open contacts and setting the INVERSION function in the PCS and ERFCS for points Y35990 and Y3600D.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report the containment atmosphere recirculation system is not required to operato during accident conditions. Implementation of this design change will not create the possibility of a different type of accident than previously evaluated in the UFSAR.

The proposed design chang 6 will not create the possibility of a malfuaction of a different type than previously evaluated in the safety analysis report this design change will provide the correct on/off status for the containmet air recirculation fans e.t the PCS and ERFCS computers and will involve work on safety-related circuits; however, the emergency power supply to these circuits will be disconnected prior to making the proposed modifications. Therefore, the possibili'v of a malfunction of a different type than previously evaluated in the UFSAR wila not be created.

The proposed design change will not reduce the margin of safety as defined in the basis for any Technical Specification. This design change will not affect the operability of the containment atmosphere recirculation system. Itn ability to maintain the bulk air temperature in contai.tment to 105'F during normal plant operation will not be degraded. Therefore, the margin of safety as defined in Technical Specification Basis 3/4.6.1.5 will not be reduced.

The proposed design change will not require any changes to the technical specifications.

The proposed change will not require a change to the Updated Final Safety Analysis Report.

Banver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 128 of 152 CHANGE TITLE DCP-1403, Rev. O, 2RCS*MOV535 and 536 OPEN/NOT OPEN Contact Indication CHANGE DESCRIPTION  !

The Safety Parameter Display System (SPDS) provides a concentrated display of a minimum set of parameters which define the safety status of the plant. The SPDS is a function of the Emergency Response Facility computer System (ERFCS). The ERFCS obtains its database from the Plant Computer System (PCS) database.

The SPDS has been subjected to a Verification and Validation (V&V) program to assure its adequacy. A plant-specific evaluation of DVPS-2 has resulted in the determination of the specific parameters to be displayed and they are listed in WCAP - 1070. The WCAP - 1070 is referenced in UFSAR Section 18.2.1 l

and is part of the V&V requirement.

The WCAP - 1070 calls for the OPEN, INTERMEDIATE, and CLOSED positions of the Pressurizer Relief Block Valves (2RCS*MOV535 and 2RCS*MOV536). To realize this, both OPEN/NOT OPEN and CLOSED /NOT. CLOSED indication is needed.

CurrLntly, the OPEN/NOT OPEN Andication is available when the Reactor Coolant System (RCS) cold overpressure mitigation system is in the ARM mode, but it is not available when in the BLOCK mode. (2RCS*MOV537 is already-correct because it is not part of the RCS cold overpressure protection system.)

LThis DCP proposes to modify the indication circuitry to provide the OPEN/NOT OPEN indication no matter which mode the cold overpresnure mitigation system is in. Prasently, the circuit consists of two (2) contacts, ir, series, that sense first, the overpressurization mode (ARM or DLOCK) and second, the valve position (OPEN or NOT OPEN). The contact from the overpressurization mode will be removed so that, the CPEN/NOT OPEN signal will be sensed, and sent to the computer system, independent of the mode selection. Both train A-and train B will be modified in this manner.

The OPEN/NOT OPEN signal will be rerouted such that it will go to one of the alarm screens of the PCS. An audible alarm associated with these screens will sound in the Control Room when a valve is in the NOT OPEN position. This will

-essentially replace the alarm function of annunciator window A4-C2 which presently alarms when a valve is NOT OPEN and the cold overpressurization ,

mitigation -system is -in ARM.- Therefore, this annunciator window will be spared.

Both the annunciator and the computer are non-safety related. These portions-of 'the circuitry will remain properly separated from safety related portions via the signal isolators in the annunciator isolation cabinets.

This DCP will affect only the computer and annunciator indications for these valves. The annunciator alarm will be replaced by the equally reliable computer system alarm. The associated computer point nomenclaturce will be changed as applicable, since the computer will now tense the Ove positions regardless of the overpressurization mode. All other indication and controls associated with these valves, including control board indicating lights and control logic, will remain unchanged. Therefore, the reliability and design function of the valves will be maintained.

I B3Sver Valley Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments page 129 of 152 FAFETY EVALUATION

SUMMARY

No Unit 2 Chapter 15 accidents will be affected by this DCP because this DCP

- does not adversely affect any safety or non-safety systems, does not exacerbate any existing accidents, and does not introduce any new hazard beyond that already considered in the UFSAR.

This DCP will not adversely affect the safety function of any system. The reliability and performance of the Block Valves will be maintained and no other systems will be affected.

The probability of an occurrence of an accident previously evaluated in the UFSAR will not ba increased. This DCP will maintain the reliability and performance of the Block Valves, and it will have no offect on any other equipment; therefore, no probabilities of occurrence of any accidents will be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR.

The reliabilicy and performance of the Block Valves and of the cold overpressure mitigation system is being maintained.

! The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP will not adversely affect any safety related system or component functions.

The consequences of- a malfunction of equipment important to safety as l previously evaluated in the UFSAR will not be increased. This DCP will not I adversely affect any parameter which would increase the consequences of a malfunction. This DCP will not adversely affect any safety system used to mitigate an accident. Therefore, there will be no effect on the consequences of a malfunction of equipment important to safety.

This DCP will not cause any new credible failure moden because the design features and functions of the equipment have not been significantly altered.

The possibility for an accident of a different type than previously evaluated

- in the UFSAR will not be created. The reliability and performance of the Block Valves and cold overpressure mitigation system will be maintained.

Nothing is being added or altered in a way which creates the possibility of a different type of accident.

The possibility- for a malfunction of a different type than any previously-evaluated in the UFSAR will not be created. The reliability and performance of all systems- will be maintained. The fundamental design features and functions will not be changed in a way that creates the possibility of a

_ malfunction of a different type. The Annunciator System alarm will be replaced with an equally reliable PCS alarm. Proper signal isolation between safety and non-safety related portions of the circuitry will be maintained.

i

! This DCP will not change any parameter which affects the course of any i accident analysis supporting Technical Specification bases.

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I Bsaver Valley Power Station Unit 2 l 1990 Report of Facility Changen, Tests, and Experiments l Page 130 of 152  ;

The Technical Specification index was reviewed to determine if any bases might .i be affected. It was determined that this DCP will not adversely affect the .

margin of safety as defined in the bases for any Technical Specifications, I including 3/4.3.3.8, 3/4.4.9.3, 3/4.4.11. This DCP will not require any changes to the technical specifications.

This DCP will require change to the UFSAR. Figures 7.2-1 (Sheets 17 and 18),

7.2-22 and 7.2-23 must be revloed to show that the control Room A1.n'inciation will now be via the PCS rather than by the annunciator system and that the cold _ overpressure mitigation system does not input a signal to this part of the-Block valve logic.

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Dsaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimente Page 131 of 152 pHANGE TITLE DCP-1440, Rev. O, Main Unit cenorator seal oil system Drain and Vent Valve Addition CHANGE DESCRIPTION The objective of this design change is to add s'x (6) 3/4" threaded end globe valves will allow periodic blowdown of the two (2) system Cuno filters to remove accumulated sludge and will provide an efficient method of draining and filling the two (2) system heat exchangers. Presently, pipe plugs exist at the valve addition locations. A capped drain valve will be added to the hydrogen side- Cuno Filter [2 CMC-FLT204) and the air side Cuno Filter

[2CMo-FLT203). A capped drain and vent valve will be added to the hydrogen sido 011 Cooler (2GMO-E21) and the hir side 011 Cooler (2GMO-E22).

SAFETY EVALUATION SUHMARY The BVPS-2 UFSAR, Chapter 15 design basis accidents were reviewed to iaentify what design basis accident could be impacted by the proposed modification.

There is- no safety related equipment in tne turbine buildup. None of the BVPS-2 Chapter 15 Safety Analyses will be affected by this modification.

No safety system will be adversely affected by the proposed design chango.

The addition of the valves will not affect the hydrogen seal oil system to prevent hydrogen leakage or air leakage through the generator shaf* seals. It is not safety related.

The proposed design change will not increase the probnbility of occurrence of

- an accident previously evaluated in the safety analysis report because the functions of all safety related equipments and systems sill be unaffected by this modification. The addition of valves has no way to cause turbine overspeed turbine trip events. It has no safety function.

The proposed dec)gn change will not increase the consequences of an accident previously evaluated in the safety analysis report because no equipment or systems required to- mitigate the consequences of any previously analyzed accident will be affected by this modification.

No new credible failure modes will be created by the implementation of this modification. There will be no functional change to any safety related equipment, systems or structure.

The proposed design change will not create the possibility of an accident of a different' type than previously evaluated in the safety analysis report because all equipment and system functions will be performed in the same manner as before. There is no configuration change such than an accident'of different type is created.

The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report because no new failure modes or potential hazards will be created by the implementation of this mooitication. The failure of this equipment has no effect on safe plant operation and shutdown o! the plant, and the failure of this equipment will not affect the safety function of other equipment.

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Beavor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 132 of 152 The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because the hydrogen seal oil system is not used to prevent or mitigate the consequences of an accident. There should be no effect on the consequences of a malfunction _of equipment important to safety as previously l evaluated in the UFSAR.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report because all- equipment ~and. system functions will be pe'.iuc.~ad 1*. the same manner as before. There is no configuration change sutn that an accident of different type is created.

The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report j because no new failure modes or potential hazards will be created by the '

implementation of this modification. The failure of this equipment has no effect on safe plant operation and shutdowie of the plant, and the failure of thin equipment will not affect the safety function of other equipment.

There are no changes in parameters which affect the course oJ any accident analysis supporting Technical specification bases and result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or containment integrity.

The proposed design changt does not reduce the margin of safety as defined in the basis for any Technical Specification. No T.S. basis will be affected in any way in this modification. The proposed design change will not require change to the technical specifications, The proposed change will require changes to the Updated Final Safety Analysis Report. Additional valves should be added on Figure 10.2-9.

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Banvor Valloy Pow 3r Station Unit 2 1990 Report of Facility changes, Tests, and I<3eriments Page 133 of 152 CHANGE TITLE DCP-1456, Hev. 1, DRMS, RM-11 Software Revisions CHANGE DESCRIPTION the Digital Radiation Monitoring System (DRMS) computers (RM-11) require modifications to existing software to correct deficiencies identified during the testing and operation of this syst em. The deficiencies are summarized as follows:

1) A cold start of the DRMS cannot be completed with only one of the two RM-11 system computers operational. Excessive computer communications during the initial startup process is believed to inhibit the single computer startup capability.

2) The DRMS/ERFCS communication over the DRMS/ERFCS datalink is experiencing intermittent failures. -The cause of theos failures is believed to be caused by the RM-11 computers taking excessive time to respond to a data request from the Emergency Response Facilities Computer System-(ERFCS).

3) The Digital Equipment Corporation (DEC) crror log function cannot be activated on the existing DRMS system.

4) During the startup, acceptance testing and initial operation of the DRMS site specific modifications to the RM-11 computer software were initiated.- This design change will incorporate and permanently install these revisions.

This DCP will involve DRMS software changes to solve the above problems. In addition, ._ the ERFCS and ARERAS sof tware that directly interf aces with the DRMS RM-11 data link software will be reviewed and modified as required to operate efficiently with the revised DRMS software.

The Atmospheric Radioactive Effluent Release Assessment System (ARERAS) computer uses- radiation data from the Unit 2 DRMS (or the Unit 1 SPING systems) and meteorological information from the meteorological- tower to calculate offsite dose projections in support of the DVPS Emergency Preparedness Plan. The ARERAS computer is physically located in the ERF Building. UFSAR Figure 11.5-8 schematically shows, the DRMS to ARERAS data link.

This DCP will maintain the reliability and performance of the DRMS, ARERAS and

.ERFCS and will not adversely af fect any oti.er equipment. The DRMS will remain

. properly isolated from the safety-related parts of the radiation monitoring system.

SAFETY EVALUATION

SUMMARY

No Unit 2 Chapter 15 accidents will be affected by this DCP because this DCP does not adversely affect any safety or non-safety systems, does not exacerbate any existing accidents, and does not introduce any new hazard beyond that already considered in the UFSAR.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 134 of 152 This DCP will not adversely affect the safety function of any system. The reliability and performance of the DRMS, ARERAS and ERFCS will be maintained and no other systems will be affected.

The probability of an occurrence of an accident previously evaluated in the UFSAR will not be increased. This DCP will maintain the reliability and i performance of the DRMS, ARERAS and ERFCS and it will have no ef fect on any other equipment; therefore, no probabilities of occurrence of any accidents will be increased. j The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP is minor and the change will have no effect on any other equipment. -This DCP will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR.

The reliability and performance of the DRMS, ARERAS and ERFCS is being maintained.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP is minor l and the changes will not' adversely affect any equipment, including the DRMS, 1 ARERAS and ERFCS. i The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increassi This DCP will not adversely affect any parameter which would increase the cor.=equences of a malfunction. This DCP will not adversely affect any safety matem used to mitigate an accident. Therefore, there will be no effect on the consequences of a malfunction of equipment important to safety.

This DCP will not cause any new credible failure modes because the fundamental design features and functinns of the_ equipment have not been significantly altered. This is a relatively minor change.

The possibility for an accident of a different type than previously evaluated in the UFSAR *ill not be created. This DCP is minor, and the reliability and I

performance or the DRMS, ARERAS and ERFCS will be maintained. Nothing is being. added or altered in a way which creates the possibility of a dif ferent type of accident.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR will not be created. This DCP is minor and the reliability and performance of the DRMS, ARERAS and ERFCS will be maintained.

The fundamental design features and functions will not be changed in a way that creates _ _the possibility of_a malfunction of a different type. The DRMS will remain properly _ isolated from the safety-related parts of the radiation monitoring system.

This DCP will not change any parameter Phich affects the course of any accident analysis supporting Technical Specitication bases.

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B3 aver Va11cy Pow 3r Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 135 of 152 The Technical specification index was reviewed to determine if any bases might be affected. It was determined that this DCP will not adversely affect the margin of safety as defined in the bases for any Technical specifications because the reliability and operability of the DRMS, ARERAS and ERFCS will be i maintained, and no other equipment will be affected. This DCP will not l require any changes to the technical specifications.

i This DCP will not require any changes to the UFSAR.

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Boaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 136 of 152 CHANGE TITLE DCP-1469, Rev. O, RTD Bypass Hanifold Elimination CHANGE DESCRIPTION i l

The Resistance Temperature Detector (RTD) Bypass System during plant operation has been a major contributor to plant outages as well as to an increase in Occupational Radiation Exposure. . Worker radiation exposure has primarily-been associated with mait:tenance of the RTD Bypass Manifold piping and to crud traps of the Reactor System Loop Compartment. Plant shutdowns on Unit #1 have been associated with excessive priniary leakage in valves and through I interruptions of bypass flow due to valve stem failures. I l

The objective of this modification is to replace the RTD Bypass System with a temperature measurement system using RTDs in thermowells in the reactor coolant loop piping. This modification is intended to resolve the following problems associated with the RTD Bypass mrnifolds:

1) The elimination of the Kerotest <alves, along with the RTD bypass manit,ld and associated piping, will satisfy the NRC and INPO concerns relative to misapplication of packless metal diaphragm valves.

-2) The installation of new. RTDs, which will be required to meet the time response requirements of the Technical Specifications, will satisfy the BVPS-2 time response issue and. eliminate the need for JCOs.

3) The elimination of the RTD bypass manifold and associated piping will significantly reduce the radiation exposure in the reactor coolant pump cubicles.

4) This modification will enhance the general reliability of the plant and eliminate forced outages as a result of problems associated with maintenance and/or operability of the RTD bypass manifold.

SAFETY EVALUATION

SUMMARY

The design. basis accidents which were reviewed for potential impact by the proposed design change included core thermal. limit protection, loss of l

electrical load / turbine trip, uncontrolled RCCA bank withdrawal at power, CVCS

malfunction that results in decrease in the RCS boron concentration.(Hode 1),

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partial loss of forced reactor coolant flow, reactor coolant pump locked rotor,. steamline break outside containment, and reactor core response to excessive secondary steam release (at power). WCAP-12478 describeb the j extensive analyses, evaluation, and testing performed to ensure the new design l- ~ meets all safety, licensing, and control requirements necessary for the safe

( operation of the plant. It will not introduce any new hazard beyor.d that already evaluated in the UFSAR.

l l The BVPS-2 safety cystems that will be affected by the proposed design change are the reactor coolant system, main steam system and containment system.

WCAP-12478 evaluated the safety function of the affected systems. The reliability, integrity and operability of these safety systems will be maintained.

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Bosvor Valloy Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experimente l

Pago 137 of 152 The proposed design change will not increase the probability of occurrence of an accident previously evaluated in Safety Analysis Report because the new thermowell mounted RTDs have a response timo equal to or better than the old bypass piping transport, thormal lag, and direct immersion RTD. This allows the total RCS temperature measuromont responso timo to remain unchanged.

The proposed design chango will not increase the consequences of an accident previously evaluated in the safety analysis report because Westinghouse WCAp-12478 evaluated thu effects of the RTD Dypass Elimination on the uncertaintion associated with reactor sotpoints that supported the continuing validity of the current LOCA and non-LOCA safety analysis assumptions for the transients in UFSAR Chaptor 15.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report because the elimination of the RTD bypass piping will not affect the input of LOCA and non-LOCA transient analysis previously evaluated in the safoty analysis report.

The consequences of a malfunction of equipment important to safoty as previously evaluated in the safety analysis report will not bo increased because the change does not affect the safety function of the components.

Failure modes of the proposed design change which woro reviewed included the resistance temperaturo detector failure causing OTDT/OPDT protection failure.

The possibility for an accident of a different type than previously ovaluated in the safety analysis report will not be created because the magnitude of the uncertainties affected by this change ascociated with RCS temporature and flow measurement, which are used in the LOCA and non-LOCA transiont analysis, will not be affected.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the chango does not affect the safety function of the components. These transients protected by OTDT/OPDT are not affected.

There are no changos in parameters which affect the courso of any accident analysis supporting Technical Specification bases and result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or containment integrity.

The margin of safety as definod in the basis for any Technical Specification will not be reduced because the change does not affect the basis of Technical Specification Section 3/4.3.1.

The proposed deofgn change will requiro change to the technical specifications. Reactor notpoints on loss of flow, overtemperaturo delta T, and overpower delta T on Technical Specification Table 2.2-1 will be changed.

Tables 3.2-1, 3.3-1 and 3.3-2 will be revisod.

The proposed change will require changes to the Updated Final Safety Analysis Report on Sections 5.4.3.1, 5.4.3.2, 5.4.3.4, 7.2.1.1.4, 7.2.2.3.2, 7.7.1.1, 7.7.1.2.1, 7.7.1.3.3, 7.7.1.8, 7.7.1.8.1, 7.7.1.8.2, 7.7.2.1, Table 5.1-2, and Figures 5.1-1, 5.1-2, 5.1-3, 5.1-4, 5.1-5, 5.1-6, 5.1-7, 7.2-1, 7.7-1, 7.7-2, 7.7-5, 7.7-7, 7.3-14, 7.3-15, 7.3-16.

Banvor Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 138 of 152 GHANGE TITLE DCP-1498, Rev. 2, RWST LO-LO Bistables CHANGE DESCRIPTIQH No Control Room (CR) indication presently exists to alert CR operators that the RWST LO-LO bistables, for safety injection transfer to recirculation, are placed in the bypass position (2Q$$-LSLL-104A, B, C and DJ. IEEE-279 requires that continuous indication is provided in the CR if the protective action of some part of a system is bypassed or deliberately rendered inoperative for any purpose.

This modification proposes to provide the required indication via four (4) existing spare status lights and provide the existing containment high-high bistable circuitry with approved signal isolators. .All of the necessary wiring

-changes will occur inside of the Primary Process Racks [RK*2PRI-PROC-1, 2, 3, i and 4). The changes will include wiring from contacts, located in the 7300 comparator trip switch, through an annunciator interface card (qualified isolation device) to a termination panel. One annunciator interface card will be added to (RK*2PRI-PROC-4).

The installation will adhere to the proper methods of isolation that are required between all safety-related and non-safety related portions of electrical systems. The 4 status light nomenclatures will be identified as RWST LO-LO Level TEST.

i ELEETY EVALUATION

SUMMARY

These bistables provide a signal to initiate transfer from the safety injection mode to the recirculation mode. Therefore, they are required to help mitigate any of the previously analyzed accidents whore safety injection is required. ,

Numerous accidents were reviewed for any possible impact caused by this '

modification, including those discussed in UFSAR Sections 15.1.4, 15.1.5, ,

15.2.8, 15.6.1, 15.6.3, and 15.6.5. The possibility of any adverse effects who not found.

The safety systems that will be affected by the proposed design change include the Quench Spray System, Safety Injection System, and the Plant Process Control System. However, this modification will have no adverse effects on any of these systems. No safety-related functions will be- changed, and the installation will meet the isolation requirements that must be implemented when-electrical work involves both safety-related and non-safety reltted portions.

L The proposed' design change will not increase the probability of occurrence of an accident prevleusly evaluated in the safety analysis report. These bistables and 9tatus lights _ are not involved in the initiation of any of the previously ana'.yzed -accidents. Therefore, the possibility of an increase of occurrence will not be created.

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1 Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 139 of 152 The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report. The protective function of these bistables is to provide a signal on RWST LO-LO level to transfer from safety injection to recirculation. This function will be unaffected by this modification.

The proposed design nhange will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. No safety-related systems or their functions will be adversely affected by this modification.

The preposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. No safety-related components or their functions will be adversely affected by this modification.

No new failure will be created by the implementation of this modification.

Signal isolators will preclude the possibility of a failure in a non-safety related portion of a circuit from affecting any of the safety-related portions. the modifiention will be installed as safety-related (QA Category I) from the bistables, to and including the signal isolators in the Primary Process Racks and will be installed as non-safety related (QA Category II) from the process racks to the status lights. All of the modification will be installed as seistnic.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report. No new potential hazards will be created by the implementation of this modification.

The proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report. No safety-related or non-safety related system or component functions will be adversely affected.

There are no changes in parameters which affect the course of any accident analysis supporting Technical Specification bases and that result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or conthinment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification (T.S.). No T.S. or its bases will be adversely affected by this modification, including those in T.S. Sections 3/4.1.2 and 3/4.3.2. This modification will provide indication to the CR operators that a bistable has been removed from service for maintenance or testing. The requirements of T.S. 3/4.3.2.1 will continue to be met. The proposed design change will not require a change to the technical specifications.

The proposed change will require a change to the Updated Final Safety Analysis Report. Section 6.3.5.4 should mention the status lights (both the LO-LO LEVEL status lights and these bistable in TEST status lights).

Baaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, anf Experiments Page 140 of 152 CHANGE TITLE l DCP-1500, Rev. O, steam Generator Blowdown Drain Line CHANGE DESCRIPTION {

During outage situations, no existing drain paths to the secondary side of the i steam generators wero available with the steam generators at atmospheric ]

i pressure, except to the steam generator blowdown tank. The gravity drain path to the' steam generator blowdown hold tanks is undesirable due to these tanks being allized for liquid waste processing. Presently, 3/4 inch vent paths are  !

utilized to drain the secondary sidw of the steam ganerators after flushing; i draining via the vent path is very time-consuming.

The- purpose of this design change is to install a 3-inch diameter drain valve and- associated piping, utilizing the existing 3-inch diamnter future wet layup I

connection available in the Terbire Building. The piping connection after the l

valve- shall be. threaded to allow for connectior of a 2-1/2 inch diametor fire hoee, which will be field routed to an acceptable catch basin.

SAFETY EVALUATION

SUMMARY

'The BVPS-2 UFSAR, chapter 15 design basis accidents were reviewed to identify what design basis accidents could be impacted by the proposed mod!fication.

None of 'the previously analyzed accidents of UFSAR Chapter 15 will be af fected by this design change. Steam generator blowdown system is discussed in Section J.4.8.

No safety system till be adversely affected by the proposed design change. The L

portion of the steam generator blowdown system where the new valve and piping will_,be 3dded is considered non~ safety related. It will serve only to provide a fastar mechod of draining the secondary side of the steam generators during outage situations.

The probability- of an occurrence of an accident previously evaluated in the safety analysis report will not be increased. This modification serves only to enhance the drain paths to the occondary side of the steam generators after flushing. No previously analyzed accidents will be affected by this change.

l The consequence of an accident previously evaluated in the safety analysis report will not be it.reased. This modification will not enange any syncem l

design functions nor will it provide any active safety functions, g The probabilicy of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. No safety-related equipment ar function will be adversely affected by this change because the affected piping and the new valve is considered non-safety related.

j .The consequence of a malfunction cf equipment important to safety as previously

evalueted in the sarety a~alysis report will not be increased. The new piping I and valve added by this modification serves only to enhance the deain paths to tne accondary side of the steam generators after flushing. No system functions will be affected.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experi.nents Page 141 of 152 Failure modes of the proposed design change which were reviewed included the failure of +1a newly added piping and valve. Existing drain path to the new piping is normally isolated by twa locked closed valves in series. Failure of the new piping and valve will have no functional changes to any safety-relt.ted equipment system and structures.

The possibility for an accident of a different type than previously evaluated in ~ the safety analysis report will not be created. No system functions will be changed. This modification will perform the same function as tt.e 3/4 inch drain path, but allow faster drain down the steam generatore during outtue situations.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created. No other equipment will be affected by this modification.

There are no parameter changee associated with this design change which could affect the cource of any accident analysis supporting technical specification bases.

The margin of safety as defined in the basis for any Technical Specification (T.S.) will not be reduced because no T.S. bases will be affected by this modification. The proposed design change will not require change to the techniual specifications.

Th* proposed change will require changes to the Updated Final Safety Analysis Report. UFSAR Figure 10.4-23 must be revised to include the added steam generator blowdown system piping and valve.

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Boavor Valley Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 142 of 152

-CHANGE TITLE DCP-1502, Rev. 1, Unit 2 Modifications for Heat "xchangers Performance

. Monitoring CHANGE DESCRIPTION NRC Generic' Letter 89-13 requires each licensee to establish an acceptable Surveillance Monitoring Program to ensure that safety-related heat exchangers that are cooled by Open Loop Cooling Water Systems (such as Service Wator) are i not allowed to 'degrado excess!vely without being detected and corrective actions taken. This modification will add thermowells, pressure taps, and _i ultrasonic flow instrumer.ts to provide for performance monitoring of heat  !

exchangers. Where ultrasonic flow esters are being installed, short sections of ca bon steel SWS piping wil- be replaced with stainless steel seccions.

This will provide an added benefit of an access way for inspection and cleaning of piping. Use of ultrasonic flow meters will allow for flow monitoring l without degrading flow through the system. i E

l l This design change will be installed in two phases.

Phase I (which includes L

modifications -near 2HVP*CLC265A and B; LRSS*E21A,B,C, and D; and 2HVR*ACU20*lA L and L) will be installed 2R while Phase II (which includes modifications near 2CCP*E21A,B,C; 2HVR*ACU208A,B; 2CHS*E25A,D,C; 2ECS*E21A,D; and 2EGS*E22A and B)

- will be installed 3R.

SAFETY EVALUATION SUKMARY None- of the d Jign basis accidents in Chapter 15 of the Unit UFSAR.are affected by this. design change.

l Modifications will be made to the Service Water System piping and the Charging L Pump Lube oil piping, however, the current design and operating parameterw will

i. not J be enanged and no new control or protective functions will be added; I therefore, no safety systems wil? Se adversely affected by this design change.

The . proposed design change will not increase the probability of occurrence of an accident previously evaluated in the the safety analysis report because the l proposed modifications will ' add monitoring instrumentation only. This

. instrument will not provide any control or protective functions. The instrumentation will be installed to the appropriate design and seismic-L requirements. The existing system flowpaths and functions will not be L changed. Each' task . in this design change is. Independent of the others.

Providing that each task is completed once started, partial completion of this l

design change does not increase the probability of a previously evaluated j accident.

The proposed design change does not increase the consequonees of an arcident previously evaluated in the safety tr.alysis report because use of this modification will provide early warning of service water flow degradation due to . fouling and will not affect the consequences of any accidentn previously evaluated in the UFSAR.

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Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 143 of 152 The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report this desiga change will not alter any parameters of the affected systems and will not change system operations in any way.

The proposed desigt. change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report Because of the passive nature of this design change and that the system will remain equivalent to the original design. Because the instrumentation is of a passive nature (does not provide protective or control functions) and does not alter the existing operation of the affected systems, no new failures are created.

The design and cperating pat' meters of the affected systems are not changed by this design chsnge. This design change will be installed to the appropriate quality and seismic requirements.

No new control or protective functions ill oJ performed and no changes to the existing design and operation are being made; also, this design change will be installed to the appropriate quality and seismic requirements.

No parameters that affect the course of any accident analysis supporting Technical Specification bases and result in exceeding the acceptance criteria for fuel claddiag, RCS boundary or Containment integrity are affected by this design change.

No Technical Specifications or bases are affected by this design change. The proposed design change will not require cbange, to the technical specifications.

The proposed change will require changee to the Uudated Final Safety Analysis Report. UFSAR Figures 9.2-1 and 9.2-4 should be revised to show the locations of the performance monitoring instrumentation as a minimum.

Usaver Valloy Power. Station Unit 2 1990 Report of Facility ~ Changes, Testa, and Experiments Page 144 of 152.

CilANGE TITLE DCP-1545, Rev. O, Replacement of Diosol Conerator 2-1 and 2-2 Automatic Loading sequence Timer Relays CHANGE DESCRIPTION The motor driven time delay relays that allow electrical loads to be soquentially added to BVPS-2 Emorgoney Due 2ht (2DF) are not operating at the predatormined time interval required in relay calibration procedure 1/2 RCP-30.

Station ' Electrical Maintenance has requested a suitable replacement for 16 ATC Type 3053 motor driven time delay relays.

Engineering shall replace all ATC motor driven Type 305E timo dolay relays with ATC's highly reliable solid state digital timo delay relays.

Modification of each sequencer panol (PNL* SED-244(254)) ic necessary to make possible; the replacement. of the round-style panel mounted 30$E relays with ATC's newer 72 MM. aquare DIN size enclosure.

EhETTY EVALUATION

SUMMARY

Ur t #2 UFSAR Chapter 15 design basis accidento that nre impacted by the proposed design change include the Loss of Offsite Power accident described in Section 15.2.6 of the UFSAR which is not affected by this design change. The new solid-state relays will be a direct one-for-ono exchange with the existing electromochanical relays with only minor modifications to the sequencer panels for mounting purposes.

Unit 7 #2 cafety ayatems which will be affected by the proposed dosign chango xnclude the operation of the Emergency 4KV Power Supply System, which in not changed by this modification. The system will continue to perform the same

= function in the manner as it presently does.

The proposed design change will not increane the probability of occurrence of an accident previously evaluated in the cafety analysis report because the probability of the occurrence cf a Loss of Offnito Power (LOOP) accident as described.in Section 15.2.6 of the UFSAR is not changed by this design change.

The proposed design change will not increase the consequencos of an accident previously evaluated in the safety analysis report the consequences of a LOOP accidant are not increased because the now sequencing will perform identically to the original relays, and will meet the same quality requirements as the original relays. The new relays will be at least reliable as the existing relays (UFSAR, Section 8.3).

B*avor Valloy Power Stttion Unit 2 1990 Report of Faciaity Changes, Tests, and Experimenta Page 145 ef 152 The proposed design en6 age will not increase the probability of occurrence of a malfunction of equiprent important to safety as previously evaluated in the safety analysis report because the new re'ays will be purchased to the quality and EQ requirements of che existing relays, and exe deemed to be at lest as reliable. Thus, they will not increase the probability of a malfunction of eqaipment important to safety.

The proposed design change will not increase the canaequences of a malfunction of equipmer.t important to safety cs previously evaluated in the safety analysis report because malfunction of a relay will not increase the consequences of a Lose Of Offsite Power accident because a redundant Emergency Diesel Generator, capable of 100% of the power, needs to achieve safety shutdown exists on the opposite train, an evaluated in UFSAR Section 15.2.6.

The only credible failure mode is that the relay will not start the pump or other safety-related equipment as required by the sequencer. This le not a new concern because these relayu replace older versions that perform the same function, also, a redundant train diesel generator exists to provide additional protection against a single failure.

The proposed change will not create the possibility of an accident type than previously evaluated in the Safety Analysis Report because the new relays perform the same function as the existing relays. No changes to any circuits are required for this installation. Panel modifications to fit the new relays wl31 be seismically qualified.

The proposed change will not create the possibility of a n.alfunction of a different type than previously evaluated ' n the Safety Analysis Report. The new relays will perform the same function as the exieting relays without any modifications to the existing circuits (a direut one-for-one swap).

There are no changes in parameters which affect the course of any accident analysic supporting Technical Specification bases and result in exceeding the acceptance criteria for fuel cladding, RCS boundary, or containment integrity that are being made by this change.

The proposed design change will not reduce the margin of safety as defined in the basis for any Technical Specification because the provisions of Technical Specifications 3/4.8.1 and 3/4.8.2 are not affected. The proposed design change will not require change to the technical specifications.

The proposed change will not require changes to the Updated Final Safety Analysis Report.

Beaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 146 of 152 GHAHQ3 TITLE DCP-1567, Rev. O, Replacement of Recirculation Spray Pump Timer /Pelays CHANGE DESCRIPTIQH The motor driven time delay relays that allow the Recirculation Spray Pumps to be energized from BVPS-2 Emergency Zus 2AE (2DF) are not operating at tha predetermined time interval required in the Technical Specifications.

Station Electrical Maintenance has requested a suitable replacement for ATC Type 305E motor driven time delay relava.

Engineering shall replace all four (4) ATC motor driven Type 305E time delay relays with ATC's highly reliable solid state digital timer / relays, type 365A.

Mooification of each Recirculation Spray Pump 4KV cubicle is necessary to make possible the replacement of the round-style panel mounted 30'E relays with ATC's newer 72 MM. square DIN size enclosure, i

SAFETY EVALUATION

SUMMARY

j Unit #2 UFSAR Chapter 35 design basis accider's that are impacted by the ,

proposed design change, including the Main Steam Line Break accidents described in Section 15.1.5 and Loss of Coolant Accidents in Section 15.6.5 of the UFSAR l were reviewed; they are not affected by this design change. The new solid-state relays will be a . direct one-for-one exchange with the existing electromechanical relays with only minor modifications to the Recirculation Spray Pump cubicles for mounting purposes. No wire changes are needed.

There are no BVPS Unit #1 or Unit #2 safety systems that will be affected by the proposed design change. The operation of the Recirculation Spray = System is not changed by this modification. The system will continue to perform the same functions in the same manner as it does presently.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report. The

-probability of occurrence of any of the accidents listed above is not changed by this design change. The new so .d state relays are direcc one-for-one exchange with the existing electromechanical relays with only minor modifications to the panels for mounting purposes.

The pra aosed design change will not increase the consequences of an accident previov+1y evaluated in the safety analysis report. The consequences of the above 1Astod accidents are not increased because the new relays will perform ido.wically to the original relays and will meet the same quality requirements. The new relays will be at least as rollable as the existing ones. (UFSAR i.2.2.2.2)

D:ever Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 147 of 152 The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. The new relays will be purchased to the same quality and EQ requirements as the existing relays and are considered to be at least as reliable. Thus, they will not increase the probability c,- a malfunction of equipment impor+ ant to safety.

The proposed .spign change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. Malfunction of a relay will not increase the consequences of any of I the- above accidents because a redundant independently powered train of l l

Recirculation Spray equipment capable of 100% of the design requirements needed to achieve safety shutdown exists on the opposite train.

The only credible failure mode is that the relay will not start the J Recirculation Spray putop as required. This is not a new concern becauro these relays replace older versions that perform the same function; also, a rrJundant train of RSS equipment exists to provide additional protection against a single failure.

The proposed design change will not create the pooribility of an accident of a different type than previously evaluated in the safety analysis report the new relays- perform the -same function as the existing relays. No changes to any circuits- or wiring era required for this installetion. Panel modifications to fit ~the new relays will be seismically qualified.

The proposed design change will not create the possibility of a malfunction of a different. type than previously evaluated in the safety analysis report. The new relsys will perform the same function as tra existing relays without any modifications to the existing circuits (a direct one-for-one swap).

No changes in parameters which affect the course of any accident analysis supporting Technical Specification bases and result in exceeding the acceptance criteria for fuel cladding, RCS boundary, er containment integrity are being l made.

The proposed desit' change will not reduce the margin of safety au defined in

. the basis for any Technical Specification. The provisions of Technical

! Specification 3/4.6.2.2 are not affected by this design change. The proposed design change will not require chat.ge to the technical specifications.

The proposed change will not require changes to the Updated Final Safety

, Analysis Report.

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Bsaver Valley Powar Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 148 of 152 CHANGE TITLE DCP-3576, Rev. O, Fuel Transfer Tube Blind Flange Modification CHANCE DESCRIPTION This design change proposes to install a fitting on the fuel transfer tube blind flange separate from the testport.

During 7R at Unit 1, the fuel transfer tube blind flange pas installed such that the gaskets were mispositioned. Following the installation, the blind flange was successfully Type B tested. The Type B test was successful because the mispositioned gaskets effectively covered the testport (i.e. the gasket interspace was never pressure tested). This new fitting will be installed approximately 90' from the testport and will allow a pressure gage to be attached to ensure the interspace between the two gaskets is pressurized.

SAFETY EVALUATION

SUMMARY

There are no BVPS Unit 2 Chapter 15 design basis accidents impacted by this DCP because test connections are not specifically addressed in the accident analysis.

The fuel transfer tube blind flange is part of the containment system; however, the proposed new fitting will not adversely affect safety because the potential leakage path will be sealed when not in use.

The probability of an occurrence of an accident previously evaluated in the safety analysis will not be increased. This change does not affect any Chapter 15 analysis, and does not, therefore, affect the probability of those accidents.

The consequences of an accident previously evaluated in the UFSAR will not be increased. This DCP is minor and the change to the blind flange will not affect any other safety systems or components.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP only modifies the blind flange; this type of a malfunction was not previously evaluated.

The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This DCP will not ..

I adversely affect any parameter which would increase the consequences of a malfunction. This DCP will not advereely affect any safety system used to mitigate an accident. Therefore, there snould be no effect on the consequences of a malfunction of equipment important to safety.

This DCP will not cause any new credible failure modes because the fundamental design features and functions of the equipment have not been altered. This DCP adds a test fitting.

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Beaver Valley Power Station Unit 2 l-- 1990 Report of Facility Changes, Tests, and Experiments Page 149 of 152 The possibility for an accident of a different type than previously evaluated In the UFSAR will not be created. This DCP is minor, and the existing gasket interspace will be drilled for a fitting for a pressure gage; therefore, the change is not significant enough to create the possibility for an accident of a different type than analyzed in the UFSAR.

I The possibility for a malfunction of a different type than-any previously evaluated in the UFSAR will not be created. Again, because this DCP is so minor, and the proposed new fitting would aave the same type of malfunctions as the original testport, the possibility of malfunction of a different type is not created.

This DCP will not change any parameter which affects the course of any accident analysis supporting Technical Specification bases.

The Technical Specification index and Specification 3/4.6.1 and J/4.6.3 were reviewed to -determine if any bases might be affected. It was determined that' this DCP will not adversely affect the margin of safety as defined in the bases for any Technical Specification because the reliability of the flange seals will be maintained, and no other equipment will be_affected.

I l The proposed design change will not require a change to the technical i specifications. The blind flange is indicated for Penetration No. SS in Tech i l Spec 3/4.6.3. Blind flanges in general are addressed in surveillance 4.6.1.1.a.

The propoced change will not require a change to the Updated FSAR. Section 9.1.4.3.3 identifies that leak-check provisions for the flange are provided.

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i Beaver Valley Power Station Unit 2 i 1990 Report of Facility changen, Testo, and Experiments page 150 of 152 CllAHEE TITLE DCP-1589, Rev. O, Steam Generator Level gilANCE DESCRIPTION The BVPS Unit 2 narrow range steam generator level instrumenta are cuttantly experiencing level shifts of 6-20% during power ascension. This level shift occurs on all three steam generators and begins at approximately 45% power.

Engineering believes the changes in steam generator pressure resulting from.the plant power level changes creates-a siphoning effect on the reference leg from the condensing pot back into the impulse lines. The impulse lines contain Kerotost metal diaphragm globe valves which are installed on a 45 degree upward slope toward the- condensing pot. When the imrJ1ce linen fill with water, c negative pressure in drawn on the reference leg cauning the instrumentation to indicate an increase in level in the steam generator. Duquesne Light in currently operating under a justification for continued operation (JCO) with reduced steam generator narrow range low and low-low setpointo provided and approved by Westinghouse Electric.

Engineering recommends rotnting the Kerotect isolation valves by placing the etem in the horizontal clane to allow condensate and steam to pass through the valve more efficiently. In additinn, the condensato poto will be lowered maintaining a one inch per foot slope toward the steam generator thereby removing the current 45 degree incline. Thia will eliminate the negative pressure effect on the reference logs should the impulse lines fill the water.

Lowering of the reference legs will require replacing the narrow range Barton transmitters and rencaling both the narrow and wide range level tranumitters to reflect the shorter reference logs.

Completion of- this design change will allow BVps Unit 2 to operate without a JCO from Westinghouse and remove the requirements for reducing the low and low-low steam generator setpoint prior to increasing reactor power >45%.

Components associated with this change are located in; the Reacter Containment Building Elevations 790' and 760' near Steam Ganerators A, B and C. This design change in QA Category I, poismic and electrical Class IE.

SAFETY EVALUATION

SUMMARY

The following Design Basis Accidento from UFSAR Chapter 15 (Unit 2) were reviewed: Execesive Increase in Secondary Steam Flow (15.13), Inadvertent Opening of S/G Rollef or Safety Valve (15.1.4), Spectrum of Steam Piping Failures (15.2.7), Loan of Offalto Power (15.2.6), Loss of Normal Foodwater

'(15.2.7), and Feedwater Line Break (15.2.8). None of these accident scenarios are affected by this design change.

Feedwater IsolEtion, Main Steam Isolation, and Safety Injection receive signal from Feedwater Level Control. These cafety systems will not be affected, adversely or otherwise, by this design change. The operation of theer systems

! will not be changed and will continue to operate as described in UFSAR.

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Deaver Valley Power Station Unit 2 1990 Report of-Facility Changre, Tests, and Experiments Page 151 of 152 The proposed design change will not increase the consequences of an accident previously evaluated Ein the safety analysis report. This design change will correct deficiencies in the steam Generator Level control System to allow it to work as described in the UFSAR sections 10.3, 10.4, 15.1, and 15.2.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment important to safety as previously evaluated in the  ;

safety analysis report . The possibility of spurious operation of Feedwater l Isolation,. Main Steam Isolation, or safety injection resulting from a l malfunction of S/G Level Control will not be increased. Sufficient redundancy exists in the logic for initiation of these signals that spurious operation or failure of any part of this design change will not in itself initiate any of these events.

The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report. The consequences remain bounded by the descriptions of the accidents listed above as described in the UFSAR.

This design change introduces no new failure modes. Most of the work involves relocating, roorienting, and recalibrating existing equipment or equivalent roplacements while maintaining the exiating flow paths and function.

The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the safety analysis report. This

, design change will -enhance the performance of the existing system but correct deficiencies in the current design. The system configuration and operation will not be changed and'will require no changes to any existing safety limits.

~The. proposed design change will not create the possibility of a malfunction of a different type than previously evaluated in the safety analysis report. All replacement parts (transmitters, valves, etc.) will meet the design requirements of the existing equipment for QA category, seismic, EQ and other categories as required.

No changen -in parameters that affect the course of any accident analyses supporting Technical Specification bases are being made.

~The proposed' design change does not reduce the margin.of safety an defined in L the basis for any Technical Specification, tsone of the Technical L Specifications in Section 3/4.5 or 3/4.7 are affected by this design change.

The proposed design change will require change to the-technical specifications.

l The- change. will not require changes to the Updated Final Safety Analysis Report.

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-Bsaver Valley Power Station Unit 2 1990 Report of Facility Changes, Tests, and Experiments Page 152 of 152 CHANGE TITLE TER-ll44, Rev. O, Change of QA Category for PAL Hydraulic System CHANGE DESCRIPTION l

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.This TER proposed to reclassify the hydraulic system for the Personnel Air Lock (PAL) aa Q.A. Category III, Seismic Category II. The system was criginally purchased along with the Containraent Liner via 2BVC-65, and was classified as Q.A. Category I,. Seismic Category I. However, the system dogs not perform any safety related function and it does not contain radioactivity thereof, a reclassification to Q.A. Category III, Seismic Category II is in order.

RAFETY EVALUATION

SUMMARY

The ~ probability of an accident previously evaluaied in the safety Analysis Report will not be increased. None of the previvauly analyzed accidents of UFSAn adction 15, " Accident Analyses" will be affected, in any way, by the PAL hydraulic system reclassification, l

, The consequences of an accident previously evaluated in the Safety Analysis l

( Raport will not be increased. This reclassification will not affect any systems or component required for the mitLgation- of any of the previously analyzed accidents.

The probability of a malfunction of equipment important to safety will not be increased.- _

The hydraulic system does not perform a safety related function.

It' is not vital to the safe shutdown of the station, it is not required for the

-removal of decay or sensible heat, and it is not necessary for the prevention or mitigation of any postulated accident.

The consequences of a malfunction of equipment important to safety will be increased. The function of the hydraulic system will remain unchanged. It is used only as a convenient way to latch the PAL doors. It is not used to-actually open or close the doors since this is-done manually. Latching of the

-doors can_ also be accomplished manually in a more time consuming manner.

However, the hydraulic system will remain as the primary means of latching the doors.

-The possibility of an accident of a different type than previously analyzed in the safety Analysis report be created. No system or component functions will be changed by this reclassification.

The . possibility of a malfunction of a different type than any previously l

evaluated in the Safety Analysis Report be created. No new failure modes or L potential hazards will be. created by this reclassification.

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The margin of safety as defined in the basis for any Technical Specification will not be reduced. This classification will not atfect tho' basis of any of

! the Technical Specifications (T.S.) including that of T.S. 3/4.6.1.3, l " Containment Air Locks." This chan-a will not require a change to the technical soecifictions.

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This change will require a change to the Updated FSAR, Figure 3.8-23.

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