ML20116L789

1991 Rept of Facility Changes,Test & Experiments for Beaver Valley Power Station

ML20116L789
Person / Time
Site:	Beaver Valley
Issue date:	12/31/1991
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9211190130
Download: ML20116L789 (197)

Text

T V d Den f Va W Power $tation Shtemgpuri. PA 150fNlt04 141f) 393-!?$$

JCaed D fitBER vice Proodent . Nacieet Group November 6, 1992 U. S. Nuclear Regulatory Commission Attn Document Control Desk Washington, DC 20565

Subject:

lloavor Valley Power Station, Unit No. 1 Docket No. 50-334, Licenso No. DPH-66 Report of Facility Changen, Tests and Experimento In accordance with 10 CFR 50.59, the Annual Roport of Facility Changos, Tests and Experiments for the Beaver Valley Power Station, Unit No. 1, is attached. A brief description and safety ovaluation summary is provided for each facility and proceduro chango. The annual report consists of two attachments as described below.

Attachment 1 is the 1991 Report of Facility Changes, Tests, and Experiments. The report covers the period of January 22, 1991 through January 21, 1992. Attachment 2 presents a safety evaluation summary from the 1990 report marked to correct an editorial error.

Sincerely, stery W J. D. Sieber cc: Mr. L. W. Rossbach, Sr. Resident Inspector Mr. T. T. Martin, NRC Region I Administrator Mr. A. W. De Agazio, Project Manager .

Mr. T. M. Gerusky, Director BHP / DER Mr. R. Janati, BRP/ DER l Mr. M. L. Bowling (VEPCO)

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BEAVER VALLEY POWER STATION UNIT NO.1 DOCKET NO. 50-334 LICENSE NO. DPR-66 l

ATTACHMENT 1 l 1991 REPORT OF FACILITY l CHANGES, TESTS, AND EXPERIMENTS l

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lleaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Tab' ' Contents Telhte Procedures 111VT 1.6.1, Reactor Coolant Flow hfeasurement . .I lilVT 2 34.4, Accumulator and Check Valve Verincation Test . .2 311VT 2.49.5, Core Physics hionitoring During Refueling . .3 F1OP-918, River Water Flow hicasurement Test. .4 Operatine Procedure.1 OST I 39.l A, Weekly Station llattery Check, llattery Number i .5 Operating hianual Change Notice to ITOP 90-11, Domestic Water Supply to WT-TK-13 and the 6 Way Flow Splitting Ilox . . . . .6 ITOP-91-2, Fuel Pool Draining To Support Transfer Canal Drain Down . .7 OSTs 1.30.13 A And 13.11, Reactor Plant River Water System A & 11 lleaders Valve Position Verincation. . . .. .S ITOP 91-3, Compressed Air System Temporary hiodifications For DCP 1546 Installation . . . .. . . .9 Operating hianual Procedure 1.6 4 AO, Isolating A kcactor Coolant Loop. . . . .10 Operating hianual Procedure 1.51.4.D, Station Shutdown - Cooldown From 1101 Shutdown (hiode 4) To Cold Shutdown (hiode 5) . . .. . . ...,,,11 ITOP 91 il, Supplying Containment Station Air From The Containment Instrument Air lleader.. . .. . . .12 Temporary hiodification 1-91 011, Temporary hiodification Of Fuel Transfer System . . . . 13 ITOP 91-15, Ilydrostatic Test For IC Steam Generator Feedwater NozzJe Welds.. . . . . . . . . .14 Oh1CN 191316, Emergency Nitrogen Pressurization Of The Control Room Damper System. . . . . 15 Alarm Response Procedure A3 60, " Loop Fill lleader Access liigh" Annunciator . . . . . . .. . . . . . .16 Temporary hiodification To Allow Testing To Determine If Flow Restrictions Exist At The River Water Outlet Of Control Room Air Conditioning River Water Cooling Coils (IVS-E-1411). . .. . . 17 Temporary hiodification To Provide A Flow Path From The Charging Pump Lubrication Oil Cooler [lCll E-713] To River Water Valve (IRW-5 /8).. . . . . 18 1 TOP-91-24, Temporary Power To Fuel Transfer Pump From Opposite Tram Power Supply For Either Emergency Diesel Generator. .. . .19 1 TOP-92-01, Temporary Safety Injection Reset Train A Jumper. . . .20 TOP l-91-18, Temporary llypass Of River Wate Return lleader. .21 i

Besver Valley Power St: tion Unit 1 1991 Report of Facility Changes, Tests, and Experiments Table of Contents Annunciators A3-47 Boric Acid llatching Tank Temperature liigh Low, A3 48 Iloric Acid Batching Tank Level Low.. .... .. .. ... .., .. .................... . . . . . . . . ............22 Operating hianual 1.6.4.D, Reactor Coolant System Loop Clearance And Drain Temporary Change,.... . .... . .. .. ... . ........................................................23 ITOP 9121. Test Of Feedwater Ilyonss bgulating Valves For DCP 1484. .... . ....... . .. .24 Procedure ITOP 90-07 (Revision 2), *BV 1 Asiatic Clam Chemical Treatment Program" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . .. . . .2$

Deslun Chnnue Packnuc5 DCP-405, Rev. O, Replacement of fleat Ciretdts ET-28, 250, 513................. .. . . . . .. .. 27 DCP 504, Rev.1, improve Response of Feedwater Flow Control Valves.. , . . . . . .. 28 DCP 683, Rev. 0, Renovate Unit l Warehouse. . . . . . . . . . . . . . . . . .. . .. . . 29 DCP 775, Rev. O, Chlorine Analyzer Replacement in Outfall Analyzer Structure . .. ... . . .. ..m. 30 -

DCP-784, Rev. O, Automatic Wet Sprinkler Protection for Anti-contamination Clothing (Anti C) Storage Area and Potentially Contaminated Area (PCA)

S h o p 1 I a l l w ay . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 DCP 804, Rev.1, Program to Correct the Control Room Iluman Engineering Diserepancies.. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . . . . .........34-DCP-814, Rev.1, Control Room Chlorine Detectors...... .. . ... .. . . . . . . ................... . 35 DCP 820, Rev. O. Computer Temperature Alarms for the Emergency Response Facility (ERF) ... . .. ............................................................................,,37 DCPM/4, Rev. 0, Pressure Surge Protection For CO2 System Fire Damper -

Relcases .. . .. . ....... . . . . . . . . . . . . . . . . . . . . . . . . . . . ....................................................38' DCP 1179, Rev.1, Replacement of hiission Check Valves..... .. . . .. .. ..... ..... . .. . .. . 40 DCP 1210, Rev. O, Deletion of Evaporator Bottoms Conductivity Loop (CC-BR.101) ... . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . , .42 DCP 1213, Rev. O Fuel Transfer System Gemco Resolver Unit .. . ..... ..... ..... . . . . .. .. . .... 44 DCP 1227, Rev. O, Enhancement of hiain Steam isolation Valve Response Time.......................,...........................,............................... .........46 .

I DCP 1254, Rev. O, Deposit Control of Reactor Plant and T'irbine Plant River Wat e r Syst em s . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 4 8 DCP 1302, Rev. O, Solid State Protection System AC Power . .. .. .. . .. . ....... .., . ... . .. . . 50-L DCP 1319, Rev. O, Fuel Pool Weir Gate Air Supply hiodification-......... . ........ . . . . . . . . .52 t

DCP 1391, Rev. O, hieteorological. Computer System (hiET), Plant Variable

/ Computer System (PVS), & Safety Parameter _ Display- System (SPDS)

Elimination From The Alter. tate Technical Support Center (ATSC). ... . ....... . . . ..........54

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DCP 1392; Rev. O, Instmment Air Filter Changes.... .. ..........,... .. .. ..,.. .. . .. ,. . . . . . , . . 55.

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Ileaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Table of Contents DCP 1393, Rev. O, Replace intake Structure Instrument Air System . .. . . . . . . . . . . . . . 57 DCP 1401, Rev.1, General Electric AK Air Circuit lireaker Modi 0 cation... . ...............59 DCP 1405, Rev. O, Solid State Outp: t Relay Testing .. . . . . . . . . . . . . . . . . . . . . .. 61 DCP 1431, Rev.1, Surge Line Thermal Strati 0 cation.. . .. .. . ............. ... ... . .... .. . .. . . .. 64 DCP 1433, Rev. O, Invert Reactor Vessel Level Instrumentation System llead

. Sensors .. . . . . . . . . ... . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......66 DCP 1453, Rev. O, Closed Position Indication on hiotor Operated Valve's. . . .. . . .... 68 DCP 1464, Rev. O, Transfer of flattery 15 Loads to 13attery 12 and 1 1. .. ...... .. .. ..... 70 DCP 1467, Rev. O, Moisture Separator Reheater Replacement Tube llundles. ...... ... .. . 73 DCP 1470, Rev. 0, Unit 1 instrument & Control Shop Ventilation llood.. .. .. . . . .. . 76 DCP 1476, Rev. O, Color Separation Deliciencies Resolution. ... . . . . . . . . . . . . ........78-OCP id82, Rev.1, Group 1 Fire Damper Replacement , .... . . . . . . . .........................79 DCP 1484, Rev. O, Feedwater Ilypass Valve Control Modi 0 cation. . . . . . . . . . . . . . . . . 81 DCP 1514, Rev.1, Regulatory Guide 1.97 Limit Switch Upgrade Outside of Containment.. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 83 DCP 1521, Rev. O, Unit i Modification for licat Exchanger Performance Monitoring. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 85 DCP 1522, Rev. O, Seismic Supports for Pressure Indicator (PI MS-501, 502, 503) Tubing .. . . . . . . . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 DCP 1526, Rev. O, Removal of Out Of Service Equipment . . . .. . .. , . .. ......, . . . . .90 DCP 1527, Rev. O, Pressurizer Surge Line Rupture Restraint Modi 0 cations. .. . . . . . . . . . . . . 92 ]

DCP 1530, Rev. O, installation of Diesel Engine Mounted Day Tank Drain Valves.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . .. . . ... .. 94 DCP 1 s2, Rev. O, Permanent Vibration Proximity Probes on Safety Injection Pumps SI P 1 A,111.... ..... , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ............96 DCP 1543, Rev. O, Removal of the Unit 1 Auxiliary 11 oilers... ............ .. . .................98-DCP 1546, Rev. O, Replacement of Station Air Compressors ... . .. .. ... .. .... . ...... . .. .....100 DCP 1557, Rev. O, Suction Pressure Gauge for Auxilia:y Feedwater Pump

( P I FW- 1 5 6, 1 5 6 A , 1 5 613 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , .. ... 103 DCP 1578, Rev. O, Fuel Transfer Tube Blind isange Modi 0 cation.. ..... . . .. . . . .. .. .. . . . .104:

DCP 1598, Rev.- 0, CO 8 to C0-11 Relay Replacement on IE12-lF12 Air Circuit B reak er Cubicles.. . . ... . ..., .. . ..... . . ...... . . .. . . .. .. . . . ,, . . . . . . . . . . . . ........ 106 DCP 1622, Rev. O, Control Room IIVAC Instrument Air Filter Change...... .. . . . . .. 108 DCP 1546, Rev. O, installation of Fire Protective Wrap on Conduit IFC439002., , . , , . .I10 DCP 1684 Rev. O Feedwater Modifications . . . . . . . . . . .. . . . . . . . . . . . . ....I12 iii

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1. Dc:ver Valley Power Station Unit I j 1991 Report of Facility Changes, Tests, and Experiments Table of Contents DCP 1702, Rev. O, Permanent installation of Output hieterine Test Circuitry.. . .. . . .. .... I14  !

DCP 1712, Rev. O, Removal of Pilot Wire W-2 Test Switch.. . ... .... ... . ......... . .. . . .115 DCP 1726, Rev. O, Remove Internals ofCheck Valve RW-197.... ....... 4. . . ..... . .. . .. ... I16 DCP 1730, Rev 0, Solenoid Vahe Changeout. ....... . . . . . . . ..... ... . . . .. ... . ... I18 DCP 1731, Rev. O, Replacement of the Control Room Air Conditioning j '

Condensing Unit Compressor VS E-4A .. . .. . . .. .. ... .. .. .. .... . . . . .119 DCP 1736, Rev. O, Splices in Terminal Boxes TB-132 and TB 119... .... . ... . . .... .. . . . 121 i l

DCP 1737, Rev. O, Pilot Valve Pressure Switch Seat Leak By, Sensing Line Tee Addition. . . . . . . . . . . . . . . . . . . . .... . . . . ... . . . . . .... . .. . .. .... .. .. . .. . . 122 DCP 1738, Rev. O, Replacement of Relays 27 RP100 and 27 RPl100... .. . .. . . ... . . .. 123 DCP 1740, Rev. O, Diesel Generator Strip licaters.. ., .. .. . ...... ... . . . . . .... ... . .. .. .. ... .. 124 DCP 1741, Rev. 0, Replacement of fischer and Porter Transmitter LT-WT- 104 A 1.... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......., ...... ...... , , . . , , 126 DCP 1743, Rev. O Electro flydraulic Fluid Reservoir Low-Low Level Lockout Switch.. ..... ..... .... . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . .... . . .... . . . .. 128 DCP 1746, Rev. O, Modification of Steam Dump Conti01 System . .. .... .. .. . . .... . . ,, , . 129 DCP 1747, Rev.- 0, Relocation of Terminal Boxes From Flow Indicating Switches FIS-FW-151 A,151B,152... . .. , , , .. .. . . . .. .. . .. ......... .>.. ... . .... .. 130 j.

DCP 1748, Rev. O, Permanent Controls for Installation of Westinghouse L1FETIME Temperature Monitors. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ..,. 132 DCP 1750, Rev. O, Pressurizer lleater Controller Modification... . .... .. .... ., . . . .133 DCP 1751, Rev. O, Source Range Monitor Nuclear instrumentation System

- (N1S) Rack Grounding.c.. . . , . .. ... . . . . . . . . . . . . . . . . . . . . ... . .... . . .. . ...... . .. . . .. 134 DCP 1752, Rev. O, Source Range Monitor Penetration Wiring Change... . .. .. .. . ... . . .135 DCP 1753, Rev. O Removal of Check Val /e Internals for Valves IRW-133, 134,- 135, and 136. ............................................... . .. .. .. .. ... .136 -

-1 DCP 1755, Rev. O Disconnecting Two Pressurizer lieaters ........ ...... . . ..... . . ....... . ., 138 DCP 1756, Rev_0, Removt,1 of Check Valve Internals s.t 1RW-158 and 159. . .. ..,. .. . . .U9 DCP 1757, Rev. O, Ilydrogen Recombier Wiring Modification . .. .. . .., . . . M 0; DCP 1780, Rev. J, Modification to Control Room Air Conditioner VS-E Support ... . . .... .. .. . .... . . ... .. . . . . . . . . . . . . . , . . . . . . . , . . . . . . . .. . . ., .. . .141-DCP 1800, Rev. O, Reloccte Reactor Vessel Overpressure Protection Relays.. . . . 142-DCP 1816, Rev,0, Permar.ent Removal of Exciter Bearing Oil Drain..... .. . ..... . . .. 143 iv

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Beaver Valle; iower Station Unit 1 1991 heport of Facility Changes, Tests,' and Experiments -

Table of Contents Technical Evaluntion Reports TER 48, Rev. O, Pressurizer Sealed Reference Leg Level System - UFSAR Figure 7.2-4. . .. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . ... . . . . . . . .144 TER 1101, Rev. O, Resolution of SSFE Observation WRS-ME-001 for the River Water System VOND/ FLOW Discrepancies.. .... .. . ...... .. .... ...,. . , . , .145 TER 4964, Rev. O, Resolution of SSFE Observation RSS-ME-004 and RSS-ME-005 . . . . . . . . . . . . . . .. ,. .. . . . . . . . . . . .. . . . . . . .. .147 TER 5293, Rev. O, Fire Protection System Modifications.. . . . . . . . .149 TER 5580, Rev. O, Replace River Water Valve RW-580 with a 3/4" Stainless Steel Contromatic Ball Valve . .. .. . . . . .... . . . 151 TER 5795, Rev. O, A3-47 Boric Acid Batching Tank Hi Lo and A3-48 Boric Acid Batching Tank Level Low., . . . . ., ,. .. .... .153 TER 5797, Rev. O, Loop Fill Header Press Iligh.. . . .. . . .. .155 TER 5951, Rev. O, Drawing No. 8700-RM-43 A Discrepancy... .. . . . 156 TEk 6035, Rev. O, Pipe Section Replacement in River Water System Due to Cavitation Damages .. .. .. . . . . . . . . . . . . . . . . .15' TER 6038, Rev. O, Baron Injection Surge Tank Level Indication.. . . . . . . . .158 TER 6249, Rev. O, Reclassify component LG-CC-100 from Quality Assurance Category I to Non-Nuclear Safety Class. . . . . . . . . . . .. ,, ,,.160 TER 6403, Rev. O, Replace River Water System Valves RW-357 and RW-359. .. .162 Updated Final Safety Annivsis Rgnort Chances UFSAR l-9-4, Bypass Valve Around Main Feedwater Control Valve... . . . . . . . .163 UFSAR l-9-27, Recalculation of Minimum River Water Flow Requirements... . . . . .165-UFSAR l-9-38, Emergency Diesel Generator Fuel Suction Strainer Description... .;167 UFSAR l-9-46, Add Reference to Electrical Calculation Program In Place of -

Battery Duty Cycles and DC Load List.. . . .. . . , . . . .. . . ,168 UFSAR l-9-47, Fuel Transfer Tube Containment Isolation Test Requirements.. . .169 UFSA.R l-9 48, Control Room X/Q Values. . . .. .. .. .. ... , .. . 171 UFSAR l-9-54, Charging Pump Flow Requirements. ... . .. . . . . . . . . . .. .173-UFSAR l-9-56, Addition of Supplementary Leak Collection and Release System Heat Removal Function to the System Description.. . . . .. .. . . . . .175 UFSAR l-9-57, Remove The East and West Cable Vaults From The List Of Areas Served By The Supplementary Leak Collection and Release System To Maintain Negative Pressure. . . . . . , , . , i178 UFSAR l-9-58, WCAP 12966, 20 Percent Steam Generator Tube Plugging Analysis Program Engineering and Licensing Report . .. . . .. . .. . ;181

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1 Beover Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experimenta Table of Contents Maintenance Procedures

- Temporary Modification For Blocking Open Two Power Operated Relief Valves (PORVs).. .. . . ... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . .., , . . . . ..183 Sump Pump Installation For Clam Study Program Discharge Sampling... .. ... ,.. .184 Temporary Replacement of PCV-RW-130A With A Substitute Part... . . . . . .185 Temporary Modification To Install A Temporary Safety Injection System Reset / Override Train A Push-button.. . . . . . . . . . . . . . . . . . . . . . . , ... .. ..I86 TV BD-101 Al, A2, B1, B2, Cl, C2 Mechanically Blocked Open . . . .. .. . . . .187 k

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Beaver Valley Power Station Unit 1 -

1991 Report of Facility Changes, Tests, and Experiments Page1of187-CII ANGE TITI E IBVT 16.1, Reactor Coolant Flow Me-asurement Cil4NGE DESCRIPTION This test measures the Reactor Coolant System total flow rate in accordance with Unit i Technical Specification 3.2.5. The change to the procedure reduced the assunied flow measurement uncertainty from 5 percent to the Westinghouse standard of 3.5 percent.

SAFETY EVAI UATION

SUMMARY

The safety evaluation wa: written because the new flow measurement uncertainty value is not specified in either Technical Specific ons or the Updated Final Safety Analysis Report. The change did not affect the actual thermal design flow of 265,000 gpm. It only reduced the amount of measurement uncertainty needed to ensure that Unit I measured Reactor Coolant System flow is above the required thermal design flow upon which all accident analyses are based. Based on the above, the safety evaluation concluded that this test procedure posed no unreviewed safety questions.

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c Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 2 of I87 CII ANGE TITI,E IBVT 2.34A, Accumulator and Check Valve Verification Test CIIANGE DESCRil' TION 5

This test verifies that the check valves and pneumatic actuators associated with Power Operated Relief Valves (PORVs) [PCV-lRC-455C and D], and feedwater bypass valves [FCV-lFW-479,-

489, and 499] perform their intended function. Verification that nitrogen leakage through Instrument Air System check valves [IlA-ll7 and 119] is within design limits is measured by temporarily isolating instmment air to PORVs [PCV-lRC-455A, B, C, D, and 456] and valves

[TV-lSS-108 and 110] while the plant is in Modes 5 or 6. Verification that leakage through internal check valves associated with the air supply to feedwater bypass valves [FCV-lFW-479, 489, and 499] is within design limits is measured by temporarily isolating instrument air to the valves while the plant is in Modes 5 or 6.

SAFETY EVAI,11ATION SUMM ARY This test is not described in the Updated Final Safety Analysis Report. The Safety Evaluation was written because of the reduced operability of the PORVs and feedwater bypass valves ca sed by isolating the primary source ofinstrument air, There is no cl.ange to the margin of safety because the PORVs will remain operable with at least 600 psig of nitrogen maintained in the associated nitrogen accumulators which are designed to replace a loss ofinstrument air to the PORVs. The feedwater bypass valves are not required to be operable in Modes 5 or 6. Their accumulators are designed to close the feedwater bypass valves upon loss ofinstrument air. Based on the above, the safety evaluation concluded that this test procedure posed no unreviewed safety questions.

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Beaver Valley Powcr Station Unit i 1991 Report of Facility Changes, Tests, and Experiments -

Page 3 of 187 Cil ANGE TITI E 3BVT 2.49.5, Core Physics Monitoring During Refueling CII ANGE DESCRIPTION This test monitors the core during refueling to ensure that unanticipated criticality does not occur.

The change to the procedure allowed for the monitoring of a portable source range detector (or

" Dunker")in place of an inopereble N31 detector.

SAFETY EVAI,UATION SUMM ARV A safety evaluation was not required for this test since the change was made due to a Temporary Waiver of Compliance for Technical Specification 3. ).2 which allowed the use of a " Dunker" The function of the source sange detectors is to provide direct n-ntron flux manitoring of the core to detect positive reactivity additions which would result in a loss of the required shutdown margin for Mode 6. The Temporary Waiver of Compliance stated that the portable detector would provide neutron flux monitoring in place of the inoperable source range detector during fuel movement thus assuring core monitoring at a level consistent with Technical Specification 3 9.2.

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g Be:ver Vt.lley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 4 of187 OIANGE TITII ITOP-91-8, River Water Flow Measurement Test CII ANGE DESCRIPTION This Temporary Operating Procedure (TOP) was written to gather flow data on the "A" River Water System header through various flow paths to ascertain the condition of the River Water System prior to the 8th refueling outage. The flow paths / alignments used in the temporary test are also used in current "at power" pump operations surveillance tests or. during normal plant operation.

SAFETY EVAttIATION SITMM ARY The safety evaluation was written to address the concern that with flow directed through the auxiliary intake recirculation line, the "A" River Water System header would not be able to supply the design flow to all necessary loads _ including the diesel generator heat exchangers during accident conditions. The TOP stations operators at the auxiliary intake structure and auxiliary building, and requires constant communications between these locations and tL: control room.

This was done so that immediate operator action could be taken to restore the system to normal alignment if a design basis accident occurred during the performance of the test. In addition, the length of time permitted for pump flow to be directed through the intake structure recirculation line was restricted to 45 minutes. Based on the above, the safety evaluation concluded that this temporary operating test procedure posed no unreviewed safety questions.

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. Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 5 of 187 CII ANGE TITI.E OST 1.39.l A, Weekly S*.ation Dattery Check, Battery Number 1 Cil ANGE DESCRIPTION Operations Surveillance Test 1.39.l A, Weekly Battery Test, has been rewritten to be consistent with the equivalent Unit 2 battery test. The new test method will place reliance on measuring charging amperage to verify Technical Specification battery operability rather than measuring specific gravity. ,

SAFETY EVAL.IIATION SIIMM ARY Failure of a batte:y as described in Updated Final Safety Analysis Report Sections 8.5.3, 14.1,7 and 14.1.11 is not increased due to the new testing method. Making measurement of charging current the primary method of verifying battery operability is allowed by the Technical Specifications and is the primary method used at Unit 2. No unreviewed safety questions exist.

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' Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 6 of I87 CII A NGE TITI.E Operating Manual Change Notice to ITOP-90-11, Domestic Water Supply to WT-TK-13 and the 6-Way Flow Splitting Box CIIANGE DESCitIPTION Temporary procedure ITOP-90-II, " Domestic Water Supply to WT-TK-13 and the 6-Way Flow Splitting Box," was changed to provide an option to utilize the water softener instead of supplying the Domestic Water Storage Tank directly.

SAFETY EVAttlATION StiMMAltY The change to the procedure did not deviate from the original safety evaluation prepared for ITOP-90-11. Any leakage in the temporarf ose h system would be small compared to the flood analysis described in Updated Final Safety Analysis Report (UFSAR) Section 2.3. Domestic Water does not serve safety related equipment as described by UFSAR Section 9.11. Technical Specifications are not affected by this change. No unreviewed safety questions exist.

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Beaver Vallcy Power Station Unit 1 -

1991 Report of Facility Ch*nges, Tests, and Experiments Page 7 of187 CilANGE TITI,E-ITOP-91-2, Fuel Pool Draining To Support Transfer Canal Drain Down CIIANGE DESCRIPTION ,

A new temporary procedure has been created to drain the fuel pool to the coolant recovery tanka through a temporary hose connceting the Fuel Pool Purification System and the coolant recovery tank. Draining of the fuel pool is necessary to allow draining of the transfer canal. The water in the transfer canal will be pumped to the fuel pool, and the excess water in the fuel pool will then be drained to the coolant recovery tank.

SAFETY EVALIIATION SilMM ARY Should the temporary hose fail, water released would be collected by the Liquid Waste System and would be enveloped by Updated Final Safety Analysis Report Section 14.2.2, Accidental Release of Waste Liquid accident analysis- Failure of the temporary modification will not drain the fuel pool Perfonnance of the Fuel Pool Purification System is unaltered since the purification system will be operated within normal operating parameters. Technical Specification equipment is not involved in this new procedure. No unreviewed safety questions exist.

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1991 Report of Facility Changes, Tests, and Experiments Page 8 of 187 CII A NGE TITLE -

OSTs 1.30.13A And 13.B, Reactor Plant River Water System A & B Headers Valve Position Verification Cil ANGE DESCRIPTION Operations Surveillance Tests (OSTs) for monthly verification of River Water Systern safety -

related valve positions are being revised to allow for swing pumps and other safety related River Water System supplied components to be out of service while still complying with Technical Specifications, in addition, valves servicing the non-safety related Reactor Plant River Water System pump seal lines through strainer (LWR-S-2) have been relaxed from' the acceptance criteria.

SAFETY EVAL UATION SUMM ARY Updated Final Safety Analysis Report (UFSAR) Section 9.9.2 states that a third Reactor Plant River Water System (RPRW) pump cannot be racked on the same emergency bus with another .

RPRW pump; therefore, the third pump is always inoperable. UFSAR.Section 9.9.3, states that all components, pumps and heat exchangers can be individually isolated to provide continual-operation during equipment repair and maintenance; hence, the probability of failure is not j increased by have one charging pump cooler, one Reactor Plant Component Cooling Water System heat exchanger and one RPRW pump and its supporting equipment isolated during plant operation. Also, valve alignment revisions in these OSTs will not. affect the UFSAR Section 14.1.11 and-14.3.4 analysis since these are normal valve alignments described in UFSAR Section 9.9.3. The margin of safety defined in the basis of the Technical Specifications is not reduced from this change. No unreviewed safety questions exist.

- Beaver Valley Power Station Unit i 1991 Repo'rt of Facility Changes, Tests, and Experiments -

Page 9 of 187 CIIANGE TITI.E

'lTOP-91-3, Compressed Air System Temporary Modifications Fo_r DCP 1546 Installation CII ANGE DESCitIPTION A new temporary procedure was written to provide controls for_the installation and amoval of -

four temporary hoses on Station Air and Instrument Air Systems to maintain compressed air supply to the station while modifications to the air systems are being performed.

SAFETY EVAI.tTATION StiMMARY In accordance with Updated Final Safety Analysis Report Section 9.8.1 and 10.2, no safety related systems are involved nor are any located in the area of the temporary hoses Air operated valves are not required to respond during or aller a shutdown. The portions of the procedure-which have an increased potential foi causing a loss of instrument air are not permitted to be performed in modes 1 or 2. Technical Specifications are not afTected by this procedure. No unreviewed safety questions exist.

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Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Enperiments Page 10 of 187 CII ANGE TITLE Operating Manual Procedure 1.6.4.AO, Isolating A Reactor Coolant Loop CIIANGE DESCRIPTION A new section was added to an existing procedure to allow isolating all three Reactor Coolant System loops. "his prwides flexibility in plant configuration during maintenance. .When all three loops are isc 1, the plant must be in Mode 6 with no reactor coolant pumps running. The refueling cavity water level raust be greater than 23 feet above the flange, and all charging pumps -

are in PULL-TO LOCK.

SAFETY EVAL,UATION

SUMMARY

To meet the concerns of Updated Final Safety Analysis Report Section 14.1,4, Uncontrolled -

Boron Dilution, the procedure requires the refueling cavity water level to be greater than 23 feet above the vessel flange to provide a large volume of borated water to mitigate a dilution event..

The charging pumps are in PULL-TO-LOCK to ensure the pumps du not supply unborated water to the Reactor Coolant System. Site Engineering performed a Three-Loop Isolated Baron Dilution Analysis, which demonstrated that loosing 5% shutdown margin will not occur sooner than 30 minutes after accident initiation per standard review plan criteria. The p:acedure meets or exceeds all Technical Specification requirements during loop isolation and restoration. No unreviewed safety questions exist.

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Deaver Valley Power Statica Unit 1 1991 Reporbf Facility Changes, Tests, and Experiments c Page 11 of187 CII ANGE TITIE Opercing Manual Piocedure 1.51.4.D, Station Shutdown Cooldown_ From Hot Shutdown (Mode 4) To Cold Shutdown (Mode 5)

CilANGE DESCRIPTION The shutdown procedo.e is being revised to pyrge hydrogen and Reactor Coolant System fission gasses using nitrogen from the pressurizer to either the Unit 2 Gaseous Waste System (GWS) storage tanks or Unit 2 GWS decay tanks for processing. When hydrogen and fission gasses are less than Health Physics Department approved levels, then the delay beds exhaust will be re.

routed to the Unit 1 gas waste disposal blowers for continuous exhausting to Unit 1 GWS.

exhaust. This alignment of the delay beds will reduce waste gas volume. The GWS surge tank will be isolated and pressurized with a temporary nitrogen gas bottle.

SAFETY EVAI,UATION

SUMMARY

If an increase of gaseous activity is injected into the Unit 1 GWS flow path, the radiation monitors

[RM-lGW-108A,B and 109] would detect the increase and albw operators to initiate corrective actions. If the temporary hose from the temporary nitrogen bottle to the surge tank would fail, the release of waste gas into the primary auxiliary building is enveloped under the Updated Final' ,

Safety Analysis Report (UFSAR) Section 14.2.3 analysis. All of the GWS equipment utilized by-this procedure is operated within its design functions as described in UFSAR Section 11.2.3.2.

No unreviewed safety questions exist.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments

! Page 12 of 187 Cil ANGE TITLE ITOP 91 11, Supplying Containment Station Air From The Containment Instrument Air Header Cil ANGE DESCRIPTION A new temporary procedure was generated to cross connect the Containment Instrument Air and Containment Station Air headers. This arrangemem was required due to a clearance on the nonnal station air supply which isolated air that is used during outages for air tools.

SAFETY EVAL.liATION SITMM ARY No safety systems are affected by this new procedure. The source of air to fuel handling equipment is not significantly less reliable than the normal supply. Interlocks prevent the release of a fuel assembly if the gripper is loaded as described by Updated Final Safety Analysis Report section 14.2.1. No Technical Specifications are alTected by this new procedure. No unreviewed safety questions exist.

Beaver Valley Power Station Unit 1

- 1991 Report of Facility Changes, Tests, and Experiments Page 13 of I87 Cll ANGE TITIJ -- 3 Temporary ModiGeation 1-91-011, Temporary Modification Of Fuel Transfer System -

CII ANGE DESCRIPTION The fuel transfer conveyor car drive chaidsprocket slips during operation of the upender, causing -

the car to move from its end limit. The temporary modification will place a fabricated sta nless -

steel bridging piece across the rail ends and two stainless steel wedges through slots on the conveyor car to prevent movement of the car during operation of the upender.

SAFETY EVAL,tf ATION Stim M ARY Installation of ;he temporaiy modification will physically restrain car travel during upender operation in the spent fuel pool. This will reduce the chances of a fuel handling accident as described in Updated Final Safety Analysis Report (UFSAR) Section 14.2.1. No failure modes of equipment may result in a fuel assembly being stuck in the Fuel Transfer System or dropped.

Both accidents are bounded by UFSAR Section 14.2. No Technical Specifications are afTected by this change. No unreviewed safety questions exist.

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Deaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 14 of 187 C11ANGE TITI E

- ITOP-91-15, Hydrostatic Test For 1C Steam Generator Feedwater Nozzle Welds CII ANGE DESCIllPTION A new teinporary procedure was developed to perform an ASME Section XI hydrostatic pressure test of the IC steam generator feedwater nozzle welds. The procedure provides steps for documenting setup and removal of a hydro-test pump and temporary rigging. The procedure will utilize a hydro. test pump injecting into the IC steam ger.erator main feedwater lines to fill the feedwater lines, the steam generator, and the assceiated main steam lines up to the main steam isolation valve, using water from the turbine plant demineralized water storage tank (lWT-TK-11). The new temporary procedure will only be performed in Mode 5 or 6.

SAFETY EVALUATION SUMMAltY All systems and components within the hydro-test boundary are designed and capable of withstanding test pressures. Water chemistry and cleanliness of system / components will be controlled and monitored to meet Chemistry Department requirements Per Updated Final Safety Analysis Report Section 14.2.2, all portions of Steam Generator Blowdown System within the test boundary are in areas designed to handle radioactive waste. A steam generator tube rupture may cause an uncontrolled boron dilution; however, the temporary procedure provides for immediate boration in the event of a dilution accident along with monitoring source range monitors N31 and N32, pressurizer levels or temporary Reactor Coolant System level indication.

The margin of safety as defined in the basis of the Technical Specifications is not reduced. No unreviewed safety questions exist.

q Deaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 15 of 187l Cil ANGE TITLE OMCN 1-91-316, Emergency Nitrogen Pressurization Of The Control Room Damper System CilANGE DESCRIPTION A procedure change was made to previously existing temporary procedure ITOP 8711. The-change installed a' temporary modification to supply nitrogen to the Control Room Venti'ation System control air system in support of system modification work for Design Change 1622.

SAFETY EyAI,UATION SIIMMARY Should the temporary air supply hose fail, resulting in a loss of nitrogen / air supply to control area zones 1.through 5 pressure boundary smoke dampers, the dampers will fail in the safe position as described by Updated Final Safety Analysis Report Section 9.10.2.3. Scaling bladders for control room dampers [lVS-D-40-1 A, B, C, and D] will be fitted with a temporary check valve to prevent the bladders from deflating on a loss of air supply. If both Unit I and Unit 2 control room ventilation outside dampers are all closed during this procedure, then the control roomsLwill be considered a " confined space" and only compressed instrument air should be used instead of nitrogen. In this case, all "Confmed Space" requirements will be followed. The temporary change does not reduce the margin of safety as defined in the basis for any Technical Specifications. No -

unreviewed safety questions exist.

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m Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 16 of 187 Cll ANGE TITI.E Alarm Respcase Procedure A3-60, " Loop Fill Header Access Iligh" Annunciator CII ANGE DESCRIPTIOff This procedure will disable the Control Room annunciator by maintaining its knife switch open.

The alarm is normally lit (green) when the plant is operating at power. This change will incorporate the Dark Board Concept of NUREG 0700, Guidelin:s for Control Room Design.

The alann is normally lit due to leakage through fill header flow control valve (FCV-lCli-160) nr through reactor coolant loop fill valves (MOV-lRC-556A,B,C).

SAFETY EVALITATION SiiMMAllY System failure probability is not increased since the control room and backup indicating panel preswre indication will remain unchanged. The loop fill header is non safety related. DisabFng the alann will not change the performance of the Chemical and Volume Control System (CVCS).

The entire Updated Final Safety Analysis Report Chapter 14 safuy analysis was reviewed for applicability. Since the fill header is non-safety-related, there was no mention of this header in the accident analysis. No failure modes associated with the change can be an initiating event since the line affected is normally isolated at power. Plant response is not changed since the line is not used at power and is not used to mitigate the consequences of an accident. No new unanalyzed types of malftmetions are created since a control system is not affected. The change will not an'ect any limits of the Technical Specifications. Disabling the alann will not afTect the stroke time of[FCV-ICH-160]. No unreviewed safety question was identined.

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y Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments -

Page 17 of 187 -

CII ANGE TITI E Temporary Modification To Allow Testing To Determine If Flow Restrictions Exist At 'g The River Water Outlet Of Control Room Air Conditioning River Water Cooling Coils (iVS-E-14B)

CIIANGE DESCRIPTION A temporary modification consisting of a piping Tee added downstream of control room air conditioning river water cooling coils (IVS-E-14B) and a temporary hose to the drain (IRD-29) ,3 for the emergency flood control discharge line. The temporary line will allow bypassing the g normal discharge line to determine if flow restrictions exist in the discharge line.

SAFETY EVAL,UATION SUM M ARY -

Unit 1 Updated Final Safety Analysis Report (UFSAR) Chapter 14, Section 9,13.4, and Unit 2 -

UFSAR Sections 6.4.2.3 and 9.4.1.2.2 were reviewed to prepare the safety evaluation. River water cooling coils (1VS-E-14A,B) are backups to the norma' %ntrol room air conditioning units (100% cach). The modification is used during Mooe 5. In this mode, the control room air conditioning units are not required. If the temporary hose broke, it would not damage the reactor core. River water cooling coil (IVS-E-14B) will be operated as normal, except its discharge will be directed to the roof drains instead o.f the River Water System. River water cooling coil (IVS-E-14 A) will be unattached. No design basis accidents are credible due to the failure of the hose.

No radiologica! consequences exist since no radioactive fluids or gases are involved. No design bases accidents are afTected because failure of the hose will not affect any primary systems. No new type of accident will be created. 'The hose will be rated at greater than 100 psi, firmly attached to supports, installed for a short duration, and an operator will be present to isolate river water cooling coil (IVS-E-14B) in the event the hose failed. The floor drains will collect any-leakage. Unit I and Unit 2 Control Room ventilation will be unaffected by the change. Technical-Specification 3.7.7.1 is unaffected since Unit 1 is in Mode 5. Technical Specification 3.9.15.1 is unafTected since Unit 2 is not in Modes 5 or 6. No unreviewed safety question was identified.

Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 18 of 187 CIIANGE TITI E Temporary Modification To Provide A Flow Peh Fr^m Tla,Garging Pump hibrication Oil Cooler [ICll E 7B) To River Water Valve [lRW-578)

CII ANGE DESCRIPTION To provide a temporary flow path from the outlet of charging pump lube oil cooler [lCII-E-7B]

to River Water valve [lRW-578), connection on the common discharge header of the primary component cooling water heat exchangers using a fire hose. This will provide a temporary river water flow path for the inservice charging pump while the river water return header is on clearance for cleaning.

SAFETY EVAL,tIATION SIIMM ARY River water flow through the charging puinp cooler may be afTecied; however, a flow Controlatron will be installed at the cooler inlet to measure flow to ensure the minimum flow

.through the cooler is maintained. The temporary hose connects to the cooler discharge, therefore, it has no afTect on the cooler. The flow will be measured prior to the use of the temporary hose.

Flooding by the 1 1/2" line can be handled by the floor drains. The minimum flow through the cooler will be verified. The affect on the performance of the river water system is minimal and '

acceptable. The temporary hose does not bypass any normal radiation monitor sample points.

The temporary hose does not introduce a new type of accident. The hose is rtm in areas which contain permanent river water piping where potential for flooding has previously existed. The plant is in Mode 5 where loss of a charging pump is within the Technical Specification Bases.

There is no change to the acceptance limits of Technical Specifications. Updated Final Safety Analysis Report Sections 9.1,9.9,14, Figure 9.9-1 A,lB and Technical Specifications 3/4.7.7,4.1 were reviewed. No unreviewed safety question was identified.

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Beavor Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, und Experiments Page 19 of 187 Cil ANGE TITI4 ITOP-91-24, Temporary Power To Fuel Transfer Pump From Opposite Train Power Supply For Either Emergency Diesel Generator C81ANGE DESCillPTION This modification will provide a power supply to the opposite train emergency diesel generator (EDG) fuel oil transfer pump [EE-P-1 A,lB or IC,lD) during an emergency. Updated Final Safety a..alysis Report (UFSAR) Section 8.5.2.3 refers to having seven days of fuel oil supply for fully loaded EDG operation. Presently, the associated fuel tank only provides 3.5 days supply of oil. By supplying power to the opposite train fuel oil transfer pump, fuel oil can be delivered to the operating EDG for seven days. This modification will be performed during emergency operation of one EDG when the other EDG failed to operate.

SAFETY EVALUATION SUMM ARY No additional credible failure modes are associated with the change beyond those already analyzed for the EDG in UFSAR Sections 8.5.2.4, and 8.5.2.6. A transfer failure would be the same as that of the existing pump, redundant pumps are available (UFSAR Section 9.14.4.1).

The additional EDG loading is insignificant (Calculation 8700-E-48). N . new failure modes were identified associated with this change. UFSAR Sections 14.1.11, " Loss of OtTsite Power," and 14.3 A.1, "Large Dreak LOCA," assumes EDG operability as the sole method of powering safety related equipment, No radiological consequences are identified with the change. The modification will ensure EDG operation for the required time to power safety related equipment and mitigate accident consequences. This modification atTects accident mitigating equipment and cannot itself cause a design basis accident. Technical Specification 3.8.1.1 requires 17,500 gallons of fuel oil available in each tank and does not require a seven day supply of fuel oil, but UFSAR Section 8.5.2.2 requires seven days supply for full load operation of an EDG. UFSAR Section 9.14.4.1 states that both tanks together supply the required amount. No unrev wed safety question was identified. This modification has been reviewed by the NRC and found to be acceptable (Reference NRC Inspection Report 50-334/91-22 and related NRC letter dated November 25,1991, TAC No. 81766).

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7 Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 20 of I87 CII A NG E TITI.E ITOP-92-01, Temporary Safety Injection Reset Train A Jumper CilANGE DESCRIPTION The procedure places a temporaryjumper in control room bench board Section A to allow for a Safety Injection (SI) Train A Reset in the event reset cannot be accomplished using the installed push-button. Intermittent, spurious failure of the Train A SI reset push-button has been noted.

The Temporary Operating Procedure (TOP) will be in effect until the push-button operation is corrected.

SAFETY EVALUATION SU5th1 ARY The temporary change will not increase the probability of system failure since the jumper installation will be double verified to ensure the proper terminals are used. The system will function as designed. The jumper affects Train A SI reset only. Operation of SI equipmer.t will-not be affected by SI reset. Other Engineered Safety Features functions are unaffected.- Updated Final Safety Analysis Report Section 14.2.4, Steam Generator Tube Rupture, was identified for potential impact. Train A reset will take approximately five minutes longer with the jumper.

Train B SI reset will remain available from the control room and can be used to stop the Train'B charging /high head safety injection pump during a steam generator tube mpture. If the jumper failed, the open circuit would not cause any equipment operation. Resetting SI does not stop any equipment needed for safe shutdown. SI reset is only used when plant parameters are controllable -

and manual operator control is desired. Si reset is not discussed in Technical Specifications.

Technical Specification 3.3.2.1 is not affected by this change. SI still functions as designed and does not affect the ability of SI components to perfbrm their design function. No unreviewed safety question was identified.

o Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments -

Page 21 of 187 CIIANGE TITI E TOP l-91-18, Temporary Bypass Of River Water Return lleader Cil ANGE DESCRIPTION The procedure will provide an alternate river water flow path which bypasses the normal return lines to allow evaluation and cleaning of river water piping. The procedure will use the following four temporary lines to accomplish the task:

a. A temporary hose from the outlet of a control room air conditioner condenser to the river water return header.

b. A temporary line from the outlet of the B charging pump cooler to the discharge header of the primary component cooling water heat exchanger.

c. A temporary hose to supply water from the domestic water tank to the river water line.

The cleaning rig will force a PIG (cleaning device) through the river water line to clean the line.

d. A temporary hose from the Control Room iedundant cooling coil to valve [TCV-RW.

101 A] in the event the PIG needs back flushed.

SAFETY EVAI,UATION

SUMMARY

The following steps were taken to ensure cooling water flow: temporary installation of a Controlatron flow indicator, reinforced hoses and cooling water throttling. Any potential flooding is mitigated by hourly tours to verify the integrity of the hoses, monitoring sump level annunciators, and floor drains are available in the areas. All fire doors that are requircJ to be-blocked open have a fire watch, and Health Physics Department and Security notification is required. prior to blocking doors. The temporary hose connections do ~ not affect cooler performance. The plant is in Mode 5, and only one river water header is required by Technical Specifications. The radiological consequences of an accident are not afTected since the hoses do not bypass any normal radiation monitors. The emergency diesel generator will start on a loss of-_

offsite power. All hoses are located in areas protected from floods. This change does not introduce a new type of accident. No new failure modes of installed equipment important to-safety are created by the changes. Updated Final Safety Analysis Report Sections 9.1, 9.9, 9.13.4, 14,- Figures 9.9-1 A,lB and 9.13-2 -were reviewed for _.the safety evaluation. 'No unreviewed safety question was identified.

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Deaver Valley Power Station Unit I-1991 Report of Facility Changes, Tests, and Experiments Page 22 of 187 CII A NGE TITI,E Annunciators A3-47 Boric Acid Batching Tank Temperature liigh-Low,

- A3-48 Boric Acid Batching Trk Level Low CII ANGE DESCRIPTION This change will disable the Control Room Annunciators by maintaining their knife switches open.

This change will incorporate the Dark Board Concept of NUREG 0700,' Guidelines for Control -

Room Design.- The alarms are normally lit (green) when the plant is operating at power. - The-level low alarm (20%) is normally lit because the batching orocedure drains the tank after batching. The high-low alarm (70 F-100*F) can be lit if ambient temperature drops.

SAFETY EVAI,UATION SUMM A RY No credible failure modes were identified.- The boric acid batching procedure is performed on an -

intermittent basis with an Operator locally present Local indication is used during the procedure.

The affected portion of the Chemical and Volume Control System is non-safety related. Disabling the alarms does not change the system performance. Neither alarm results in a control function.

The entire Updated Final Safety Analysis Report Chapter 14 safety analysis was reviewed for applicability, particularly Section 14.1.4, Uncontrolled Boron Dilution. Since the batch tank is non-safety related, it was not mentioned in the safety analysis. Disabling the alarms cannot initiate a design basis accident. No Technical Specification limits are affected by the changes. 'No.

unreviewed safety question was identihed.

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Beaver Valley Power Station Unit i 1991 Report of Focility Changes, Tests, and Experiments Page 23 of 187 CII ANGE TITI,E Operating Manual 1.6.4.D, Reactor Coolant System Loop Clearance And Drain Temporary Change Cil ANGE DESCRIPTION Use a temporary nitrogen rig to supply nitrogen gas to the C reactor coolant loop to create a nitrogen blanket to prevent oxygen from entering the loop during loop draindown. The loop will be isolated from the Reactor Coolant System (RCS) using loop stop valves before draining.

SAFETY EVAL,liATION SUMM ARY If the temporary hose would fail, safety related components would not be afTected since the hose is secured to prevent hose whip. If the pressure regulator failed, the RCS loop would not be over pressurized since a relief valve is installed on the rig to lift at 100 psi. No change is made to safety systems since the draindown will occur with nitrogen at low overpressure. Updated Final Safety Analysis Report (UFSAR) Sections .14.1.4.2, Dilution During Refueling, and UFSAR Section 14.1.6, Startup of an inactive Reactor Coolant Loop, were identified. The use of the temporary rig will not afTect the mentioned accidents since the normal existing procedures for isolating and draining a loop controls the concerns of the accidents. Nitrogen is prevented from entering the RCS by the loop stop valves. The loop stop valve discs are pressurized to 450-500 psi for disc pressurization in accordance with approved procedures. Having nitrogen in the loop will not prevent the loop stop valves from performing their function. Technical Specification 3.4.1.4 is not afTected since the loop stop valves were closed prior to the temporary modification.

Technical Specification 3.4.1.5 is not alTected since the loop stop valves will be opeud by existing procedures which address the concerne Technical Specification 3.4.1.6 is not alTected since reactor coolant pump startup will be performed by existing procedures which address Overpressuie Protection System operation. No unreviewed safety questions were identified.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 24 of I87 -

CH ANGE TITLE ITOP 91-21, Test Of feedwater Bypass Reguhning Valves For DCP 1484 CIIANGE DESCRIPTION This procedure provides for functional testing of the Low Power Steam Generator Water Level Control System (LPSGWLCS). The test will provide data for scaling the signal to the Bypass Feedwater Regulating Valves (BFRV) and for post design change dynamic testing. _ Temporary recorders will be connected to the process racks 32 7300 cards. The test begins with the reactor at 5% power. One Steam Generator Atmospheric Steam Dump Valve (SGASDV) is opened to 30%, and data is gathered. The valve is then closed. This test is performed for each BFRV. The Operator will control Reactor Coolant System (RCS) average temperature (Tavg) during the test.

SAFETY EVAttiATION SIIMM ARY Failing open of one SGASDV was tested on the plant simulator with reactor power at 6%, and did not cause P7 or P10 reactor trip permissives to alar _m. Failing (open or closed) of BFRV results in steam generator level deviation alarms which the operator responded to as instructed by procedure. Failing closed of BFRV at 5% and 10% reactor power is within the bound of Updated Final Safety Analysis Report (UFSAR) Section 14,1.8. Failing open a BFRV is within-the analysis of UFSAR Section 14.1.19. Failing open of SGASDV is within the analysis of UFSAR Sections 14.2.5,14.2.11, and 14.2.13. Radiological consequences are not credible since this involves the secondary plant. The probability of occurrence 4 an accident is not increased because steam generator level and Tavg can be controlled by operators using the procedure and alarm response procedures without causing a reactor trip as demonstrated by the simulator runs.

In the event of failure of SGASDV or BFRV, plant transients will not exceed those discussed in UFS AR -accident analysis, as~ demonstrated on the simulator; therefore, no neiv. accidents associated with this procedure are credible. Failure of SGASDV and DFRV durirg this procedure is within the bounds of the referenced accidents. Tavg is maintained above 543 F for this test which is above Technical Specification limits. The margin of safety for Technical Specification-3.1.1.5, 3.3.1, and 3.3.2 is not reduced since the temporary recorders are connected to -

LPSGWLCS 7300 series process cards, which are downstream of the reactor protection system isolator; therefore, failure or shorting of a recorder will not atTect the process protection system.

Beaver Valley Power Station Unit 1 1991 Report of Fr.cility Changes, Tests, and Experiments Page 25 of 187 CIIANGE TITLE Procedure ! TOP-90-07 (Revision 2), "BV-1 Asiatic Clam Chemical Treatment -

Program" CII ANGE DESCRIPTION The following changes have been made to the previous safety evaluation for ITOP-90-07:

1. The change eliminated use of the Bio-Box since use of the sample baskets in the cooling tower basin was an adequate representation of the clam population percentage of the river water system.

2. The procedure was revised to move the clamicide injection point from water treatment circulating pump [lWT-P-12A] to the new reactor plant and turbine plant deposit control chemical injection stations installed under Design Change Package (DCP) 1254. This change eliminates impact on plant tiltered water system., Clamicide will be injected to each of the three River Water System headers using a metering pump for each header.

The clamicide brik units will be located near the deposit control stationt

3. The procedure was revised to use an eductor system and the temporary detoxant slurry quick feed injection system at the Unit I cooling tower basin discharge flume for detoxification of the Circulating Water System aller clamicide injection is completed, Cooling tower pump house instrument air will be used to power the air diaphragm pumps - .

used in this detoxant slurry quick feed injection system. . Temporary power will be used to power the 240 VAC auger feeder and the submersible pump (s) used for the eductor system. This temporary power supply is from a non-emergency power source and has no impact on plant operation.

4. A step was added to isolate the chlorination system from the River Water System headers, so that chlorination for BV-1 is not performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to and during .

clamicide injection.

5. The procedure was revised to momentarily open the third charging pump cooler and both control room air conditioning backup cooling coils for several minutes to pass 2 to 5 volumes of clamicide treated river w..er through these coolers.

SAFETY EVALITATION SIIMM ARY This Temporary Operating Procedure (TOP) provides a means _of exterminatiag- Asiatic clams in the BV-l_ River Water and Circulating Water Systems, in-order to maintain the design heat transfer and flow path conditions of these systems. The following is a summary of changes to the safety' evaluation for 1 TOP-90-07

1. Elimination of the Bio-Box reduces the amount of clamicide treated river water flowing into the turbine building sump. This is considered a conservative change; therefore,

. Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 26 of I87

. environmental safety is en'sanced.

2. A clamicide injection unit is connected to the River Water System chemical injection pump -

discharge piping of the Unit 1 Reactor Plant and Turbine Plant Deposit Control Stations.

These feed points are to be controlled by Chemistry Department personnel and will senice the River Water System and Circulating Water System.

Temporary lighting and temporary power shall be setup at the BV-1 Reactor Plant and Turbine Plant Deposit Control Stations (near Unit I valve pit and North Yard Elevation 710fl.) for the clamicide injection equipment. Existing teraporary power sources are located in both areas and are energized from the non-emergency power supplies. Security.

is notified in accordance with the initial conditions of this procedure of the use of a -

temporary sheher onsite at storm drain CB-II.

3. An additional source of detoxant will be injected into the River Water System and Circulating Water System at the cooling tower. hiost of the river water is dumped to the-cooling tower and used as makeup. Sullicient clam cide concentrations and contact time will exterminate clams living in the lower basin. An additional quick feed detoxification system comprised of five detoxant slurry unitt, injected by one or two_ instrument air powered air diaphragm pumps will be used to quickly detoxify the cooling _ tower basin once adequate clamicide concentration has been reachedi These pumps will be powered from permanent instmment air supply lines located in the cooling tower pump house, at -

elevation 710ft, via temporary hose connected to instrument air valve [llA-71-7]. The total weight for this equipment shall be within the limitations specified in Eh! 65380 -

"Clamicide Addition / Evaluation Program for Corbicula Control"

4. Isolating chlorine injection mto the River Water System ensures that the environment d

sampling concurrently taking place prior .to and during clamicide injection at the environment trailer, located at the outfall structure, is not adversely afTected by River Water System chlorination.

5. Having the third charging pump cooler aligned to a River Water System header should not reduce flow below Design Basis Accident (DBA) River Water System flow requirements for the other two charging pump coolers. The third cooler is closed in the Normal System Arrangement (NSA), only for cooler tube cleanliness control. Opening the control room.

backup cooling coils while a control room air conditioning condenser unit is in service wi" reduce River Water System flow below DBA River Water System flow requirements.

However, temperature in the in-service condenser unit will be munitored to ensure its condenser chiller rernains in senice. Thus, control _ room temperature should not be -

increased to the 88 F Technical Specification limit. It wo_uld be less desirab!c W swap -

operation of contral room condenser units ;o facilitate flushing one control room backup '

cooling coil at a time than to just align the control room backup coc'ing coil for several minutes to pass two to three volumes of clamicide treated' river v through the coils River Water System tubes The same logic also applies to the thiri ;ervice charging pump cooler flushing with clamicide treated river water.

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lle:ver Valley Power Station Unit 1 1991 Report of Fecility Changes, Tests, and Experinnnts Page 27 of 187  ;

fil ANGE TI'UE l'"o 405, Rev. O, Replacernent ofIleat Circuits ET 2F,250,513 Cll ANGE DESCRIPTION The purpose of this design change is to replace Nelson heat trace circuits ET-28 on line 3/4" 31- e 55-1503 Q2, ET 250 on line 1"-SI 1541503-Q2, and ET-513 on line 1" LW-150152 with Chemelex heat trace circuits since replacement Nelson heat trace circuits are not availab!c on site.

SAFETY EVAI,UATION SUMM AID' This design change does not alter Updated Final Safety Analysis Report (UFSAR) Section 0.3, Emergency Core Cooling; Section 8.5, Emergency Power System; or Section 11.2.4, Liquid Water Disposal Systems, nor does it affect Technical Specification Section 3/4.5.4. This change does not constitute an unreviewed safety question.

UFSAR Section 6.3, Emergency Core Cooling, and Section 11.2.4, Liquid Waste Disposal System, are not altered, as this change merely constitutes a replacement of heat trace circuits and does not affect any of the equipment in either system. Section 8.5, Emergency Power System, is-not changed since there will still be redundant heat tracing with separate circuits.

This change entails the replacement of Nelson heat trace :ircuits with Chemelex circuits. None of.

tiie systems referenced in UFSAR Sections 6.3,8.5 or i1.2.4 are changed and this change does not create a new type of accident or malfunction. e The basis for l echnical Specification Section 3/4.5.4 does not change.

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l' lleaver Valley Power Station Unit 1 1991 Rcport of Facility Changes, Tests, and !!xperiments Page 28 of 187 CII ANGE TITI4 DCP 501, key.1, improve Response of Feedwate Flow Cor. trol Valves Cil ANGE DESCitilTQN The purpose of this design change is to improve the stroke speed and

• low flow" control iesponse of the main feedwater bypass now control valves. The response improvement should reduce plant trips due te steam generator level control problems.

The valve stroke time will be redxed from seventy (70) seconds to less than ten (10) seconds when the controlling solenoid valves are de energized. Also, the valve trim flow characteristic will be chang, from equal percentage to linear. The flow characteristic change will increase the rate of flow change at " low" flows (less than 1000gpm) when the transfer from bypass flow to main feed flow is begun.

SAFETY EVAL,UATION

SUMMARY

Implementation of this design change is considered to be safe. No change to the Updated Final Safety Analysis Report (UFSAR) or Technical Specifications is required as a result of this change.

The probability of occurrence or the consequence of an accident or malfunction of equipment imponant to safety as previously evalanted in UFSAR Section 14.1.8, Loss of Normal Feedwater, and Section 14.1.0 lixcessive lleat Removal Due to Feedwater System Malfunctions, will not be increased. This design change will improve the stroke speed and low flow control response of the main feedwater bypass flow control valves. Also, the valve trim flow characteristics . vill be changed to increase the rate of flow change at low flows when the transfer from bypass flow to main feed flow is begun. Therefore, the possibility of an accident is not increased since this design change will provide the requircJ response improvement which should reduce trips due to steam generator level control.

The modifications to be incorporated under this design change will not adversely alrect the operation of the Feedwater System as described in UFSAR Section 10.3.5 or steam generatai water level control as described in UFSAR Section 7.7..l.74 Plant trips due to steam generated level control problems associated with the feedwater bypass regulating valves should be reduced.

The margin of safety as defined in the basis for any Technical Specification will not be ::cduced since there are no Technical Specifications associated with the feedwater bypass flow control valves. The subject valves are not considered containment isolation valves as defined in Technical Specliication 3/4-6.3 (Table 3.6-1),

i.i Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 29 of I87 CilANGE TITIE  !

DCP 683, Rev. O, Renovate Unit i Warehouse CII ANGl' DESCRIPTION The purpose of this modification is to renovate a portion of the Unit I storeroom, which became available upon completion of the new ofTsite warehouse, for the Nuclear Operations Unit. The area will be renovated to provide the following:

1) Maintenance, and Instrument and Control (!&C) Department supenisoiy staff ollice space,2)

Electrical Maintenance shop, storage, and Foreman's oflice, 3) I&C shop, calibration room, storage, and Foreman's ollice, 4) Materials Management issue booth, ofrice, and controlled envirorment storage, and 5) communications room. -

Additicnally, the following renovations will also be made: 1) a new entrance to the lunch room,2)

• a Building Maintenance Foreman's oflice will be provided,3) the Radeon Foreman's oflice will be reduced in size,4) the women's locker room will be expanded, 5) the Radeon count room will be relocated, and 6) the containinated I&C shop will be relocated.

SA FETY 1;VALO ATION SUMM ARY This design change is considered to be safe and does not effect the bases of the Technical Specifications. In addition, this modification is non-safety- related and does not present an <

unreviewed safety question. A chang: to the Updated Final Safety Analysis Report (UFSAR) is not required by the installation of this design change.

This modification entails tenovating portions of the Unit i storeroom and old I&C shop, and will not affect any safety related components. Therefore, the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated ist UFS AR Section 14 is not increased.

By making these modifications to the stcreroom and senice building, the possibility for an accident or malfunctior. of a different type than an > previously evaluated in the UFSAR will not ,

be created, since the ventilatior, system in the I&C contaminated shop will still exhaust to the common monitored ventilation vent through prefilters and particulate filters.

The margin of safety as defined in Technical Specification Bases 3/4.7.14 will not be reduced, since fire protection will still be provided in the senice building safety related areas.

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Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 30 of187 CilANGE TITLE DCP 775, Rev. O, Chlorine Analyzer Replacement in Outfall Structure CilANGE DESCRIP flON The purpose of this concept is to provide continuous " free chlorine" moniti a ig at the outfall stmeture as required by NPDES pennit PA0025615. This will be done by re . icing the Wallace and Tiernan Chlorine analyzer with Xertex Model 924 " free chlorine" measurement system This new system will meet the continuous sampling criteria established by the Environmental  :

Protection Ay ack ' EPA). Additionally, the Deckman conductivity and dissolved oxygen ,

monitoring are ..:n e .juired by the EPA. These will be removed. The dissolved Oxygen recorder and conductivity recorder have already been deleted from the chemistry lab environmental panel.

The dissolved oxygen recorder and conductivity recorder will be removed from all affected drawings.

All instrumentation except the Leeds and Northrup recorder will be located in or around the '

outfall structure and building. The Leeds and Northmp recorder will be located in the Unit 1 chemistry laboratory. None of this equipment is necessary for safe operation or safe shutdown of Unit 1.

SAirETY EVAI,11ATION FifMM ARY This pioposed design change will not involve an umeviewed safety question. The Chlorine analyzer is not considered to function in any Updated Final Safety Analysis Report (UFSAR)

Chapter 14 accident analysis. UFSAR conclusions are unafrected by the proposed modifications.

The proposed design change will not require a change to the Technical Specifications.

The proposed change will require changes to the Updated Final Safety Analysis Report. The recorders will be removed from all afTected figures.

The proposed changes will not involve an unreviewed environmental question, and will not result in any releases to the environment. No changes will be required to the Environmental Protection Plan.

No design basis accidents from UFSAR Chapter 14 are affected by the proposed design change, which afrects the Circulating Water System only.

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

The probability of occurrence for an accident previously evaluated in the safety ai+13 sis report and the consequence of an accident previously evaluated in the safety analysis .eport will not be-increased because this system does not function during an accident.

The probability of a malfunction of equipment important to safety as previously evaluated in the

llcaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Enperiments Page 31 of187 safety analysis report and the consequence of a malfunction of equipment important to safety as-previously evalui.ted in the safety analysis report will not be increased because this modification does not affect any UFSAR Chapter 14 accident analyses.

Failure modes of the proposed design change which were reviewed includui failure of the analyzer unit to function. This is consistent with the UFSAR analysis, The possibi!..y for an accident of a different type than prevkasty evaluated in the safety analysis report will not be created because all changes are being performed away from safety related equipment.

The possibility for a malfunction of equipment impo; tant to safety of a different type than previously evaluated in the safety anaiysis report will not be created because circulating water system pressure boundaries and system operation are not changed.

No changes in parameters that afrect the course of any accident analysis suppo..ing Technical

! Specification bases, or tesult in exceeding the acceptance criteria for fuel cladding, Reactor-Coolant System boundary, or containment boundary are included in this design change.

The margin of safety as defined in the basis for any Technical Specification will not be reduced since the Chlorine analyzer is not included in the Technical Specifications.

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Heaver Velicy Powcr Station Unit 1' 1991 Report of Facility Changes, Tests, and Experiments Page 32 of 187 Cil ANG E TITI.lj l DCP-784, Rev. O, Automatic Wet Sprinkler Protection for Anti-contamination Clothing (Anti C) Storage Area and Potentially Contaminat:d Area (PCA) Shop l llallway .

CII ANGE I ESClQP11J,JN .

The purpose of this modification is toinstall a new sprinkler system in the Antl.C storage area and 1 the hallway outside of the PCA shop per American Nuclear Insurer's (ANI) recommendation.

ANI made this recommendation as a result of the amount and types of combustible materials that 4 are present in the area. The sprinkler header will tie into the existing PCA shop supply header.

SAFETY EVAI,IIATION SilMMAlW The proposed design change will not involve an unreviewed safety question. The proposed design change will not require a change to the Technical Specifications or the Updated Final Safety Analysis Report (UFSAR).

The probabib af occurrence for an accident previously evaluated in the safety analysis report will not be mereased because this modiner. tion does not affect any fire protection systems that protect any safety related equipment. The evaluation of Section 9.10.3 of the UFSAR is not affected by this modification.

Tae consequence of an accident previously evaluated in the safety analysis report will not be increased because this modification is providing an additional means to suppress a fire that may occur in the Anti C room and the helway outside of the PCA shop where only portable fire extinguishers and fire hoses were available previously. This modiGention does not alrect the availability of exist:ng portable fire suppression equipment.

The probability of a malftmetion of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this modification does not protect or involve any areas that contain safety related equipment. No existing fire barriers are being added, deleted, or altered by this modiGeation, The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this rnodification does not protect or involve any areas that contain safety-related equipment. No existing fire barriers are being added, deleted, or altered by this modi 0 cation.

The possibility for an accident of a different type than previously evalur.ted in the safety analysis report will not be created because failure or in-operability of this modification will require that backup fire suppression equipment be made available per UFSAR Section 9,10.5.2.

The possibility for a malftmetion of equipment important to safety of a different type than previously evahiated in the safety analysi s report will not be created because no safety related

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Heaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 33 of187 equipment is affected by this modification in any way.

Failure modes of the proposed design change which were reviewed included pipe rupture, loss of the pumps, isolation ormain supply header to PCA shop area.

The rnaspin of safety as dcEned in the basis for any Technical Specification will not be reduced because no Technical Speci0 cation bases are alrected by this modification.

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i lleaver Wiley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 34 of 187 Cil ANGE TITLE DCP 804, Rev. 1 Program to Correct the Control Room iluman Enginecting Discrepancies CII ANGE DESCRIPTION The purpose of this modification is to correct human engineering discrepancies (IIED's) which were found during the control room design review (CRDR). The modifications consist of relocating various switches, indicators, and recorders. All of these changes were evaluated and recommended by the CRDR.

SAFETY EVAL,llATION SilMM ARY This design change is considered to be safe and does not involve an unreviewed safety question or Technicel Specification change.

Correcting human engineering discrepancies on the control room control panels will not increase the probability or conwquences of accidents previously evaluated in the UFSAR. These are no changes to the system function or to automatic actuations. Changes pertain only to how information is displayed to the control room operators and correcting the llED's will climinate potential sources of operator error in diagnosing problems.

The llED resolutions will not create a new type of accident since all the original design criteria will be met including separation, fire nrote: tion, and seismic requirements for the control panels.

There is no change or in. pact on any Technical Specification.

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Beaver Valley Power Station Unit i l 1991 Report of Facility Changes, Tests, and Experiments Page 35 of 187 Cll A NG E TITI.I' DCp.814, Rev 1, Control Room Chlorine Detectors Cil ANGE DESCRIPTION This design chenge will replace the chlorine detectors installed by Design Change Package 156 since the detectot response time chn not prevent the chlorine toxicity limits of the control room from exceeding NRC Regulatory Guide 1.95 and 1.78 requirements The new chlorine detectors are duct mounted probes with local process and control units as installed in Unit 2. No changes to the 2/3 logic to sohd state protection system witi be made at this tim The chlorine probes will be installed above air intake damper VS D-401 A. ,--

SAFETY EVAIMNIION SUMM ARY lt has been determined that 'his modification will not increase the probability of an occurrence or the consequences of an uccident or malfunction of equipment important to safety as previously evaluated, nor cicate a new type of accident not previously evaluated in the Updated Final Safety Analysis Report (UFS Alt t This modification will also not adversely alrect the margin of safety as defined in the basis for any Technical Specification section and an unreviewed safety question is not raised.

This modification requires Unit 1 UFSAR Figure 9.13 2 to be revised. No changes to the Unit 2 UFSAR are required by this change.

Unit 1 Technical Specification 3.3.3.7 does not require a revision as a result ofimplementing this change.

UFSAR Sections 2.1.5,7.8.5 and 9.13 A discuss control room habitability due to an accidental chlorine release. The duct mounted chlorine probes will detect a chlorine concentration ofless than or equal to 5 ppm. The response of the chlorine probes and associated equipment will limit the chlorine concentration in the control room to less than or equal to 15 ppm by volume '

(45mg/cm3 ) within 2 minutes of detection. A high chlorine signal or loss of power will trip the associated chlorine detectors bistable in the solid state protection system and alarm in the control room. Should 2/3 detectors alarm, the control room will be automatically isolated and pres;urized by the Control Room Emergency Bottled Air supply System (CREBAPS). Revision 1 of the design concept added channel trip switches to facilitate surveillance of the detectors, provided for access to the new probes, and prevented probe failure from automatically tripping the associated bistable in solid state protection. The addition of the switches and access to the probes does not impact the results of the safety evaluation since these changes provide added n.aintainability and operability to the Chlorine Detection System. Preventing probe failur; from tripping the associated bistable in solid state protcetion does not invalidate the conclusion of this safety evaluation because the revised inputs for tripping the solid state protection system (high chlorine and loss of power signals) do not reduce the original margin of safety of the Chlorine Detection System described in the UFSAR. The local processing units will still be

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-lleaver Vtiley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 36 of 187 administratively checked by operations for probe failures as the original drip type chlorine detectors were checked.

The high probe failure rate at Limerick Power Station as documented by 40 LER's indicates the probe is extremely sensitive and can cause a false trip signal to be generated. If during testing or normal operation 2/3 probes alarm, the control room would be isolated and the CREll APS will be initiated if the bottled air supply system drops below 1825 psig, the plant will be required to be shutdown by Technical Specification 3.0.3. The present high probability of probe failure causing spurious actuation of the bottled air supply system outweighs the added conservatism of tripping the solid state bistable on a probe failure.

Accidental release of chlorine is already addressed in UFSAR sections 2.1.5,7.8.5 and 9.13.4.

This modification will not reduce the margin of safety of Technical Specification 3.3.3.7 since the duct-mounted probes and associated equipment will limit the Chlorine concentration inside the control room to a maximum 15 ppm by volume (45 mg/m 3 ) within 2 minutes of detection The setpoints of the new chlorine detectors will detect a chlorine concentration orless than or equal to 5 ppm.

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Ikaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 37 of 187 Cil ANGE TITI.E DCP 820, Rev. O, Computer Temperature Alarms for the Emergency Response Facility (ERF)

CIIANGE DESCitiPTION  ;

There is presently no way to alert cperators in the Unit I control room that cedain critical areas (the Technical Support Center, Emergency 0;: rations Facility, computer and cominunications sooms in the ERF) are experiencing high temperatures. This design concept proposed to install thermostats in the above areas to monitor the temperatures and to send signals to the operators' Plant Variable Computer System (PVS) console. Upon detection of a high temperature, a flashing light on the PVS console will be activated. This will alert the operators so that appropriate actions can be taken.

SAFETY EVAI,lTATION SUMM AltY The implemeniation of this design concept is considered to be safe. No change to the Technical Specifications or Updated Final Safety Analysis Report (UFSAR) is *cquired as a result of this <

modification.

This modification will not increase the probability or the consequences of any accident previously evaluated in the UFSAR. The addition of the thermostat circuitry will not affect the existing circuitry for the Plant Variable Computer System. The addition of the alarm will help provide warning of possible trouble in the ERF (UFSAR Sections 7.9,9.17, and 12.3) and/or aan.hted equipment; none of which is safety related.

The possibility for an occurrence of an unanalyzed accident will not be created with the implementation of this modification. This modification serves only to provide warning of possible trouble that, ifleft unattended, could lead to damage of systems that are not safety related. The modi 0 cation itself does not create any new failure modes.

No Technical Specification basis is affected by this modification.

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Heavcr Valley Power Station Unit i 1991 lleport of Facility Changes, Tests, and Experiments Page 38 of 187 CII A NGli TITI.E DCP 1074, Rev. O, Pressure Surge Piotection For CO2System Fire Damper Releases Cil ANGE DI:SCRIPTION The proposed modification will add a total of three pressure segulators to protect the elght dampers listed in the Design Change Package. The regulators are necessary to protect the 125 psig rated pneumatic McCabe Links (release mechanism) nom the potential 150 psig supply presstue. The CO2 pressure release mechanisms are non safety related; this modification will improve the operability of the dampers by minimizing the potential for an over pressure rupture of the link.

SAFETY EVAI,UATION SilMM ARY rhe proposed design change will not involw an unreviewed safety question. The aflbeted lines are not safety related.

t The proposed design change will not require a change to the Technical Specifications. The proposed change will require changes to the Updated Final Safety Analysis Report (UFSAR),

UFSAR Figure 9.10 2 must be updated.

The Fire Protection System is Quality Assurance (QA) Category F and the Switchgear Cooling System is QA Category 11. These systerns are not evaluated in UFSAR Chapter 14, and are non-safety related. The proposed change has no impact on UFSAR accidents.

The proposed change will affect three dampers which are not redundant. No malfunction of a single release mechanism or a single damper will impact more than one area. This is consistent with the original design.

Failure modes of the proposed design change which were reviewed include failure to actuate, inadvenent actuation, and a seismic event. This is a QA Category 11 system and is not required to function during an accident, and has no safety ftmetion.

The possibility for an accident of a different type than previously evaluated in the safety analysis -

report will not be created because the change is minor and affects the non-safety related CO2 system The possibility for a malftmetion of equipment impor', ant to safety of a different type than previously evaluated in the safety analysis report will not be created because the system has no active function to be evaluated.

No changes in parameters which afrect the course of any accident analysis supponing Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, reactor coolant l system boundary, or containment boundasy are included in this design change.

I3edver Valley Power Stttion Unit 1 2 1991 Report ot' Facility Changes, Tests, and Experiments 1 Page 39 ofI87: (

i' The margin of safety as defined in the basis for any Technical Specification will not be reduced because this design change does not impact Technical Specifications. j

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Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 40 of 187

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CII A NGE TITI.E DCP 1179, Rev. I, Replacement of Mission Check Valves CIIANGE DESCRIPTION This design change shall be adininistered as a Generic Replacement Design Change (GRDP). The basis for this change is that the replacement valves may have a dilTerent manufacturer and may have different materials than the installed valves. Additionally, these type valves can no longer be obtained from Mission Manufacturing Company, as C&S Valve Company has obtained rights to these valves. C&S Valves will teplace existing Mission valves as needed initially, the following ten check valves in the River Water System are to be replaced: RW-106,107,108,109,133,134, 135,136,197 and 198. Other check valves will be replaced by a Technical lhaluation Report  :

referencing thi; GRDC.

SAFETY EVAI.IIATION SilA151 ARY The proposed design change will not involve an unreviewed safety question as this change is a one-for one replacement of a valve with a replacement having nearly identical design characteristics but built by a different manufacturer.

The proposed design change will not require a change to the Technical Specifications. f The proposed change will not require a change to the Updated Final Safety Analysis Repoit (UFSAR). The valve manufacturer is not specified in the UFS AR, No design basis accidents from -

UFSAR Chapter 14 are affected by the proposed design change, No safety systems will be adversely afrected by the proposed design change.

The probability of occurrence for an accident previously evaluated in the safety analysis report will not be increased. The replacement check valve improves the maintainability of the valves.

Thero is no change to the system design or to the valve ftmetion.

The consequ::nce of an accident previously evaluated in the safety analysis report will not be increased because this valve replacement does not alrect the consequences of the accidents evaluated in the UFS AR.

The probabil_ity of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this modification only increases the maintainability cf the valves.  :

The consequence of a maltimetion of equipment important to safety as previously evaluated in the safety analysis report will not be increased becarse this modification does not affect the UFSAR Chapter 14 accident analyses.

Failure modes of the proposed design change which were reviewed included failure of the check

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13eaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 41 of187 ,

valve to shut, Failure of the reolacement valve to shut is no different than failme of the original l valve; therefore, failure of a check valve to shut does not introduce a different accident. ,

The possibility for an accident of a difTerent type than previously evaluated in the safety analysis l report will not be created because this is a one for-one replacement of a valve with a similar cast valve. There is no change to system pressure boundaries or system operation due to this change.

The possibility for a malfunction of equipment important to safety of a different type than '

prevluusly evaluated in the safety analysis report will not be created because systern pressure bou.idaries and system operation are not changed.

No changes i. parameters that could affect the course of any accident analysis supporting _ .

Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, l Reactor Coolant System boundary, or containment boundary are included in this design change.

The margin of safety as defined in the basis for any Technical Specification will not be reduced because this design change improves the valve maintainability.

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Ileaver Velley Power St: tion Unit i 1991 Report of Facility Changes, Tests, and Dxperiments Page 42 of 187 Cil ANGE TITIE DCP 1210, Rev. 0, Deletion of Evaporator Bottoms Conductivity Loop (CC.IIR 101)

CilANGE DESCRIPTION The objective of this change is to completely remove all parts of the Evaporator llottoms Conductivity Loop (CC-IIR-101).

This will consist of removing the conductivity probe (CC BR-101), modules (CY- DR-101 and CY IIR-101 A), recorder (CR Illbl01), and high and high-high alarm modules (Call DR.101 and CAllil llR-101). In addition, the annunciator points and inputs to the sequence of events recorder will be climinated SAFETY EVAI,UATION SilMM ARY The probability of occurrence for an accident previously evaluated in the safety analysis repo1 will not be increased because the operation of the Doron Recovery System evaporators as described in Section 9.2 of the Updated Final Safety Analysis Report (UFSAR)is not affected by this modification. A sample point exists inunediately upstream of conductivity probe CC DR 101 to allow fbr sampling of the evaporator bottoms. The probability of a beric acid leak in the evaporator bottoms loop is not incicased pa the description in Table 9.2 2 of the UFS AR.

The consequence of an accident previously evaluated in the safe'y analysis repat will not be increa. J because the consequences of an evaporator bottoms leak or overpressure condition as described in Table 9.2-2 of the UFSAR are not affected by this rnodification.

The probability of a malftmetion of equipment innportant 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 wi" not be increased because no safety related equipment is affected by this modification.

The design basis accidents from UFSAR Chapter 14 which were reviewed for potential impact by the proposed design change included Accidental Release of Waste Liquid and Gases (Sections 14.2.2 and 14.2 3, respectively). This modification has no elTect on these design basis accidents.

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

The possibility for an accident of a difTerent type than previously evaluated in tl.e safety analysis report will not be created because the limiting accident, a boric acid leak or spill, has been previously evaluated in ttm UFSAR.

- The possibility for a malfunction of equipment important to safety of a different' type than

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l Deavet Vallcy Power Station Unit l' ,

1991 Report of facility Changes, Tests, and Experiments i Page 43 of I87 l previously evaluated in the safety analysis report will'not be created because no safety-related j equipment is affected by this modification.

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Failure modes of the proposed design change which were reviewed included a boric acid leak from evaporator bottoms (Table 9.2-2 of the UFSAlt, Jtem 6). .

The margin of safety as defined in the basis for any Technical Specification will not be reduced because no Technical Specification bases are alTected by this modification. l No changes in parameters that could effect the course of any accident analysis supporting  ;

Technical Specification bases, or result in eveeeding the acceptance criferia for fuel cladding,  !

Ileactor Coolant System boundary, or contalament boundary are being made per this ,

modification.

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Ileaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments j Page 44 of187 ,

J fjjjNGE TITLE  :

DCP 1213, Rev. O, Fuel Transfer System Gemco Resolver Unit Cil ANGE DESCRIPTION During the last refueling shutdown, upender winch proxhnity switch failures caused a critical path schedule de:ay. A Gemco resolver and transducer were installed to replace the switches , using Maintenance Work Request 880002.

This design concept proposes to provide a formal engineering review and inspection and also provides the proper vehicle to initiate further corrections if deemed appsopriate. Specific actions include the following: an inspection of the Gemco equipment installation; a review of all documentation relative to this installation; a physical inspection and cortcetion, if necessary, of the conduit and the mounting of the Gemco equipment for Seismic Category 11 concerns; an updating of electrical drawings, foreign prints, and the Fuel Transfer System technical manuals; the

- assignment of mark numbers, where applicable; and the procurement of Gemco spare parts, SAFETY EVALitATION SilMMAlW The probability of occurrence for an accident previously evaluated in the safety analysis report will not be increased. The vendor certified that the new components meet the design specifications and equipment quality of the original components.

The consequence of an accident previously evaluated in the safety analysis report will not be ,

increased. The replacement of the proximity switches will have no adverse etTects on any plant systems, including the Fuel Transfer System.

The , o bability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. The function of the upender win:hes remain unchanged and no adverse etTects on any safety-related equipment are created.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because no new failure modes or potential hazards will be created by the i_mplementation of this modification.

The design basis accidents which were reviewed for potential impact by the proposed design change 'aciude those covered in UFSAR Section 14.2.1. These accidents include a stuck fuel assembly and a dropped fuel assembly.

No safety systems will be affected by the proposed design change since aone of the safety-related .

components of the fuel handli:m system are affected by this modification.

- The possibility for an accident of a difTerent type than previously evaluated in the safety analysis report will not be created. The new components serve the same function as the old proximity i

switches. No new failure modes or potential hazards are created

Ileaver Valley Power Station Unit 1  :

1991 lleport of Facility Changes, Tests, and Ihperiments - l Page 45 of187 .

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 ielated  ;

components will be affected by this modification.  ;

i Failure modes of the proposed design change which were reviewed included a failure of the new i components. This equipment failure would cause the winch to stop moving. A stuck fuel assembly has been previously analyzed.-

The margin of safety as defined in the basis for any Technical Speci0 cation will not be reduced.

No Technical Specification bases, including those of Sect;on 3.9, *lleflieling Operations", are l affected by this modi 0 cation.

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Ileaver Valley Power Station Unit i 1991 Repost of Facility Changes, Tests, and Experiments Page 46 of 187 Cll ANGC TITIE DCP 1227. Rev. 0, Enhancement of hf ain Sh-am isolation Valve Response Time CII A NGE IWSCRIPTION hiain Steam Trip Valves TV-hts-lot A, II, and C must shut in less than 5 seconc , as specified in Operating Surveillance Test 1.21.4,5,6 and Technical Specification 4.7.1.5b. The valves require major work in order to meet this time limit. Usual closure times are between 4.5 and 4,9 seconds aller work.

The air lines will be enlarged from the actuator to the auxiliary feed pump room. Tubing will be used with large radius bends instead of cibows and instead of tees Two additional valves with high flow coeflicients will be installed for each main steam trip valve (one per train) in order to increase air blowdown rate.

hiockup testing during the seventh refueling outage indicated that the main steam isolation valves (h1SIVs) will close in the 1-2 second time frame. Orifices may be installed to limit hiSlV closure time to appioximately 4 seconds.

This design change will be limited to the respective h1SIV actuator piping in the main stesm valve room, auxiliary feed pump room, and the quench spray pump room. Conduit and wiring will also have to be installed for the new solenoid operated type valves.

SAFETY EVAL,llATION SilMM ARY Updated Final Safety Analysis Report Section 14.1.13. " Accidental Depressurization of the hiain Steam Syste n"; Section 14.2.5, *hiajor Secondary System Pipe Rupture"; Section 14.2.11, "hlinor Secondary System Pipe lireaks"; and Section 14.3, " Loss of Coolant Accidents" were reviewed for applicability. None of these accidents are affected by this design change because thi.:

design change will ensure that the h1SIVs will operate in accordance with Technical Specification 4.7.1.5b.

The main steam system and the reactor coolant system were revicwed for applicability. Normal operation of these systems is not affected by this design change. .

The probability of occurrence for any of the above accidents is not changed by this design change.

The normal or accident operation of the main steam system is not changed by this design change.

The consequences of the above accidents are not increased. This design change will ensure that the h1SIVs will close within the required time span (less than 5 seconds) as required in the Technical Specifications.

The probability of a malftmetion of the hiSIVs is not increased. The new tubing, fittings, valves and associated support equipment will be seismically qualified for use in the area to ensure reliability. Single failure criteria will also be met by the installation of redundant air bleed off

I 13eaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 41 of187 valves powered from separate safety-related power sources.

l The consequences of a malfunction of the main steam isolation valves are not changed by this j design change. The worst case is that the MSiv will remain open and pennit steam flow after receiving a close signal. This is no worse than the present worst case scenario. 'l l

The credible failure modes of this design change include rupture of the instrument air lines, as well  ;

as improper operation (failure to open or close as directed, or spurious operation) of the air bleed off valves. Spurious opening of one of the air bleed ofTvalves or rupture of the air line can lead to inadvertent closure of the MSIV, which willlead to a plant trip. If one of the air bleed off valves fails to open ou command, single failure criteria is satisfied because the other redundant  ;

valve will open. -i This design chan; e a places existing equipment (tubing, fittings, valves, etc.) with larger diameter  !

air supply tubing 3 :

• valves that perform the same function, but allow for faster closure of the MS t V5.

The equipment being installed replaces existing equipment and does not change the existing flow - .

path of the control air to the MSIVs.

There are no changes in parameters that affect the course of any accident analyses that support  ;

Technical Specification bases.

This design enange will ensure that the margin of Technical Specification 4.7.1.5b 11 met. This I Technical Specification is not changed by this design cl.ange.

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llcaver Valley Power Stotion Unit 1 ,

1991 Report of Facility Changes, Tests, and Experiments Page 48 of 187 Cil ANGE TITI.E DCP 1254, Rev. O, Deposit Control of Reactor Plant and Turbine Plant River Water '

Systems Cil ANGE DESCRIPTION ,

The proposed design change will provide a method to inject chemicals into the River Water ,

Systems to reduce sedimentation fouling and to control macroinvertebrate infestaticn of the system piping. The Reactor Plant River Water System (RPRWS) and the Turbine Plant River Water System (TPRWS) will each have a chemical supply tank and associated injection pump.

Each will also have a second pump for the treatment of macroinvertebrate. The supply tanke and pumps are all non-nuclear safety related. -

SAlrETY EVAL,tIATION SilMM ARY Post design basis accident cooling requirements are described in Updated Final Safety Analysis Report Section 14.3. The addition of a Deposit Control System to the RPRWS will not impact the safety analysis.

The proposed modification will connect with the Quality Assurance (QA) Category i River Water System piping. The piping and connections up to and including the check valve at the class break in the Reactor Plant River Water System will be installed QA Category 1.

The deposit control system is non nuclear safety and cannot initiate a UFSAR Chapter 14 accident, Connections to the RPRWS and TPRWS will maintain the QA Category of the respective systems up to the class break.

The proposed system is not required to operate during an accident and will not result in t,n increase in dose from any accident or equipment malfunction.

The proposed change will not afTect the operation of the RPRWS. and will not cause a malftmetion of RPRWS equipment.

No malfunctim % Depan Control System will result in increased dose.

The toiloe , aphs list the credible failure modes associated with the proposed change and identify thc , . < P .,r reasons which conclude that these modes have been previously evaluated.

1. Pipe lheak - Separate Injection lines will connect to the A and B, RPRWS headers.

Header isolation valves in the pump enclosures will be controlled by procedure to allow only one header to be in service at a time. Check valves in each header will prevent a failure in either!M 9m draining both headers.

2. Loss of Powu equipment is not required to operate, thus a loss of power has no adverse afTect,

lleaver Valley Power Station Unit 1

- 1991 Report of Facility Changes, Tests, and Enperiments -

Page 49 of I87-The. types of possible accidents for the Deposit Control System, i e., leaks, breaks and flooding, are the same as those possible for the river water system. The Deposit Control System has no sah , function and is t.st required to operate in an accident. There are no RPRWS parameter changes associated with this modification.

The NPDES Permit tellects station elliuents, i.e., discharges from both Unit I and 2. The Environmental Protection Plan (EPP) Addresses changes in station design or operation. For this reason, a preliminary evaluation of the information contained in the design concept was made. As a minimum, two chemicals are proposed for use in the Deposit Control System; " Clam Trol" (CT.1), a chemical to control Asiatic Clams, and "Powerline DQU-04", a chemical to control silt and reduce sedimentation fouling. The use of CT-1 after 1990 has not been approved by the Pennsylvania Department of Environmental Resources (DER). Ilowever, the sedimentation treatment chemical is approved by the DER. No product can be teleased to the environment -

unless it is done in accordance with the NPDES Permit, and with the approval of Duquesne Light Company's EnvNnmental Affairs Unit. Design Change Package 1254 provides the mechanism to introduce these chemicals while the actual injection will be by approved procedure. The EPP addresses " additional construction" or " operational activities" which may significaatly alTect the :

environment. The constmction of the Deposit Control System will not affect the environment.

The DER previously approved tne use of CT-1 on a one time basis for the purpose of gathering data. There is reasonable assurance that the DER will approve the use of CT 1 for Unit 1; however, this chemical will not be used in the deposit control system without that approval, Operational activities which involve the use of CT-1 as stipulated by the DER do no significantly; affect the environment.

It is the conclusion of this evaluation that an Unreviewed Environmental Question as dermed by the NRC in the EPP does not exist.

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Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 50 of 187 CilANGE TITI.E DCP 1302, Rev. O, Solid State Protection System AC Power CII/MGE DESCRIPTION The present AC power distribution configuration for the Solid State Protection System (SSPS) trains resulted from a Stone und Webster (S&W) modification that was performed under Design Change Package (DCP) 180. The configuration conformed to the S&W practice of powering Train A loads only from Battery I and Vital Bus 1 and powering Train B loads only from Battery 2 and Vital Bus 2, llowever, DCP 180 did not account for the internal wiring of the SSPS logic bays. The Train A logic bay is powered from Busses 1 and 2, and the Train B logic bay is powered from Buses 3 and 4. A failure of Vital Bus 2 would affect both the output bay of SSPS Train B and one set of the logic bay power supplies of SSPS Train A. Thc[efore, the present AC distribution configuration does not take optimum advantage of system reliability.

This modification proposes to make a change in the power distribution to the SSPS logic bays.

The change will cause the Train A logic bay to be powered from Busses I and 3 and the Train B logic bay to be powered from Busses 2 and 4. This will be accomplished by internal wiring modifications in the logic bays.

Additionally, supervisory relays will be added for indication of an output bay power loss. Existing spaces for spare relays in the output bays will be used for this purpose. The power loss signal will be routed to the Computer and Annunciator Systems (windows A4-75 and A4 83) via the Multiplexing System. Isolation boards located in the Multiplexing System provide the boundary between the safety-related signal and the non-safety related Computer and Annunciator Systems.

SAFETY EVAIEATION SUSihl ARY ,.

The SSPS receives input signals from various plant parameters and initiates the required actuation of plant process equipment and/or a reactor trip, based on the input signals received. The SSPS affects many systems and components required to help mitigate various Design Basis Accidents (DBAs). This modification will actually help to enhance the reliability of the SSPS by providing the Train A bays with different power sources than the Train B bays This will in no way affect any of the SSPS ftmetions. Therefhe, there will be no adverse affects on any safety related components or systems, and none of the DBAs will be impacted by this modification.

The SSPS and the vital bus power supplies are all safety-related. The power loss signals from the SSPS output bays are also safety-related up to and including the isolation boards in the Multiplexing System. The signals from these boards to the Computer and Annunciator Systems are non safety related. No safety-related functions will be adversely affected by this modification.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report. None of the initiating factors of any of the previously analyzed accidents will be adversely afrected by this modification.

Heaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and I?xperiments  ;

Page$1of187  :

The proposed design change will not increase _ the consequences of an accident previously ]

evaluated in the safety analysis report. None of the equipment required to help midgate any of the previously analyzed accidents, will be adversely affected by this modification. l The proposed design change will not increase the probability of occurrence or consequences of a  ;

malfunction of equipment important to safety as previously evaluated in the safety analysis report.

No safety-related components or system functions will be adversely affected by this modification.

This modification will produce no new credible failure modes or potential hazards. The changes in the power supplies to the SSPS logic bays will be accomplished by_ internal wiring changes. i The newly added relays will meet all necessary electrical requirements and they will snake use of . >

existing spare relay locations. All changes in the SSPS bays and the wiring changes up to and  ;

including the isolation boards will be QA Category I. The modification will meet seismic design requirements. l r

The proposed design change will not create the possibility of an accident of a different type than j previously evaluated in the safety analysis report since no new failure modes will be introdi ced by this modification.

The proposed design changt ..I not create the possibility of a malfunction of a different type than  !

previously evaluated in the safety analysis report since no safety related functions will be  !

adversely affected by this modification.

i There are no ' anges in parameters which affect the course of any accident analysis supporting Technical Speufication bases, or that result in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment integrity, 1

The proposed design change does not reduce the margin of safety as defined in the basis for any-

• Technical Specification. No Technical Specification or its bases will be -afrected by this' modification, including that of Specification 3/4.3.1 and 3/4.8, 1

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lleaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Expcriments Page $2 of 187 Cll ANGE TITI,E JCP 1319,- Rev. O, Fuel Pool Weir Gate Air Supply Modification CilANGE DEScitiPTION Design Change Package (DCP) 1319 addresses NRC concerns as described in Information Notice .

89-92, " Potential for Spent Fuel Pool Draindo 'n". This modification proposes to add a series of vc!ves and check valves to the Instmment A System to provide additional assurance that a faihua of the instrument air supply to the fut' ool weir gate bladder will not result in a loss of seahog capability. The new valves will ist ' , # ,e fuel pool gate bladder from the Instrument Air System should the air supply fail.

SAFETY EVAL,1fATION SlfMM ARY There are no Updated Fiual Safety Analysis Report (UFSAR) Chapter 14 accidents which will be affeed by tLs DCP because this DCP does not . ssely alTect any safety or non. safety systems, does .ot change, degrade, or prevent actions described in UFSAR Chapter 14 Accident Analyses.

This DCP adds a check valve and a bypass valve to the instrument air supply to fuel pool gate bladder, but this will not edversely affect the safety function. A failure to close is not different than having no check valve, i.e no change.

The probability of occurrence for an accident previously evaluated in the UFSAR will not be increased. This change does not affect any safety related system and will have no efTect on any other equipment.

The consequences of an accident previously evaluated in the UFS AR will not be increased. Ofr- ,

site do es will not be increased. UFSAR Section 9.8.1 states that no part of any safety related equipment requires compressed air for shutdown.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. A malfunction of the compressed air system is not evaluated in UFSAR Chapter 14. 9 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 increasa the consequence

• of a malfunction. This DCP will not adversely afTect any safety system used to mitigate an accioent. Therefore, there should be no effect on the consequences of a malftmetion of equipment important to safety.

This DCP will not cause any new credible failure modes because the fundamental deaign features and functions of the Instrument Air System htsve not been altered.

The possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR will not be created The new valves only affect the Instrument Air System,' a non safety L

Ileaver Valley Power Station Unit 1 1991 Report ofitacility Changes Tests, and Experiments Page 53 ofI87 '

related system, and do not change accident parameters; therefore, the change is not significant '

enough to create the possibility for an accident of a different type than analyzed in the UFSAlt This DCP will not change any , .irameter which alrects-the course of any accident analysis supporting Technical Specification bases.

The Technical Specification index ar.. Section 3/4.9.11 were reviewed to determine if any bases m'ght be affected. It was determined that this DCP will not adversely affect the margin of safety l 3 as defined in the bases for any Technical Fpecification because the reliability of the Ibel pooi wear y pte air supply will be maintained, and no other equipment will be affected. .j

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Deaver Valley Power Stction Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 54 of 187 Cil A NGE TITif, DCP 1391, Rev. O, Meteorological Computer System (MET) /

Plant Variable Computer System (PVS) /

& Safety Parameter Display SystenujPDS) Elimination From The Alternate Technical Supprnt Center (ATSC)

CIIANGl DESCRIPTION Unit i equipment associated with the SPDS, PVS and MET computers will be removed from the Alternate Tcchnical Support Center (ATSC) since there is no lenges a licensing commitment to retain the ATSC.

This design change includes the removal of a recorder panel, a MET console, an indicating light  ;

box, a power transfer switch, a power distribution panel, a console table, and several raceways and conduits. Approximately 14 instrument cabics and 4 power cables will be pulled back through the floor and spared, associated cables will be Aisconnected and spared, and solhvare modifications will be made. Modifications will be made in both seismic and nor.. seismic portions of the service building and in the emergency response facility, which is a non seismic structure.

The MET and SPDS computer systems are QA Categmy 11, and the PVS is QA Category 111.

SAFETY EVAL,UATION SIIMh1 ARY There are no Updated Final Safety Anamsis Report (UFSAR) Chapter 14 design basis accidents whose anaiysis depends upon the SPDS, PVC, or the MET computer systems.

There are no Unit I or Unit 2 safety systems affected by this design change. Per UFSAR Section 7.5, the SPDS, PVC, and the MET provide supplemental information io support plant operation and are not important to safety.

Nonr. of the accident scenarios in Section 7.5 or Chapter 14 of the UFSAR depend upon the PVC, SPf6 or MET computer systems to prevent or mitigate these accidents.

These systems perform indication functions only and do not perform any control' functions; also these systems are not reibd upon for safety related information. Thus, removal of these components from the alternate TSC will not increase the consequences of previously evaluated accident.

This design change does not affect any systems or equipment important to safety. Because the scope of this design change is limited to the removal of equipment and termination of electrical hookups, no credible failure modes exist for this design change.

This design change removes equipment that is not required to perform any safety functions, and will not create the possibility of an accident of a different type.

h No Technical Specification bases in Section 3/4.3 are affected by this design change.  :

Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 55 of187 Cil ANGE TITIE DCP 1392, Rev. O, Instrument Air Filter Changes

.C*dd5EI.LDESCRIPTION

'li.* h .rument air dryer bypass filter elements IA-FD-1 and IA-FD-2 are no longer available, ano suitable replacement elements cannot be found IA-FD-1 and IA- FD-2 will be totally replaced with coalescent type filters, the same as instrument air dryer prefilter IA-FL-1. An advantage to this change, besides the necessity for it, is that a single stock item will be maintained for all three Coniponents.

An unrelated problem is with the afler filter (IA FL-2), for the instrement air dryer . u- 1).

This filter has been discontinued and was made of wool, cotton, and rayo . The filten design is

=ized to remove particulates of 4 micron or larger. Instrument Society of America (ISA) Standard S7.3, which we have committed to meet in response to Generic Letter 88-14, reo" ires 3 micron or larger filtration. A new filter element will be installed and a new stock item ^ eated that will ,

provide the degree of filtration required by IS A S7.3.

SAFETY EVAL,UATION

SUMMARY

Loss and/or malfunction of compressed air systems are bounded by the loss and/or malfunction of the cafety systems served by compressed air. All such systems are designed to fail safe and perform their safety functio.' independent of the condition of the compressed air system. No design basis accidents are impacted, Many miems in the plant (safety and non-safety) use compressed air for valve and instrument control. All safety system valves that incorporate the use of compressed air fail closed on loss of air. No part of any safety-related y sipment requires the supply of compressed air for shutdown.

The compressed air system is not req nu a engineered safety functions. This system has no impact on the probability of occurrence & acident.

The consequences of accidents are bounded by the response of the safety systems designed to mitipte those conset. . nces. The compressed air system does not impact the ability of safety.

systema to respond to an accident and therefore will not incr:ase the consequ+ces.

Equipment impomnt to safety that employs compressed air uses it only to ensure station -

operation. The operation of this equipment during engineered safety functions is not dependent on con. pressed air, and therefore the probaH"'y of occurrence of a  ! function is independent of the condition of compressed air.

The em aquences of a malfunction of eqa. aent inaportant to safety are not dependent'on the campassed air T,stera. The consequences i :ulting frc,m equipment malfimetion are dependent upor. W we..J sciety functions, which work independent of compressed air, mitigating those consequences

13eaver Valley Power Station Unit 1 - ,

1991 Repon of Facility Changes, Tests, and Experiments -

Per 56 of187 The filters could clog, fail to provide filtering, and block all compressed station instrument air

. supply. The filters could fail to provide filtering and bypass all flow to the supply or atmosphere.

In either case, the loss and/or malfunction of station instrument air will not afTect safety-related systems or prevent such systems from perfarming their engineered safety functions,.

No possibility for an accident'of a difTcrent type than those that have already been evaluated is O foreseen, since this design change will permit the system to operate similar to the way it operated prior to the change.

No new compenent fuu.ctions are being created, only improvements or replacements of existing -

equipment. The types of malfunctions that are possible will remais, unchanged.

Parameters that affect the course of an accident have never been affected by the state of the compressed air system and this wil' ~ntinue *o be tme following this change. ,

The margin of safety as defined in the oases for any Technical Specification is calculated i independent of the operation of the compressed air system and therefore will be unchanged by this modification.

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13 ewer Valley Power Station Unit 1 1991 Report of Facility Changes, Teets, and Experiments ,

Page 57 of187 Cil ANGE TITI.E DCP 1393, Rev,0, Replace intake Stmeture Instrument Air System -

CII ANGE DESCRIPTION Instrument air compressor 11A-C-3, located in the intake stmeture does not provide acceptable air quality, as required by the NRC Generic Letter 88-14. The existing system delivers air with a measured dew polni of approximately 48 F and it contains oil.  ;

The Instrument Society of America (ISA) has established air quality acceptable to the NRC in the standard, ISA-S73, " Quality Standard for Instrument Air" This standard addresses the following four (4) elements for quality ofinstmment air for pneumatic instmments:

1. Dew Point - In no case should it exceed 35 F.

2. Particle Size - Maximum particle size in the air stream at the instrument shall be three (3) micrometers.

3. Oil Content - Under no circumstance shall the maximum total oil hydrocarbon content exceed one (1) ppm percent by weight or volume under normal operation conditions.

4. Contaminants - Air shall be free of all corrosive contammants and-hazardous gasm, flammable or toxic, which may be drawn into the instrument air stream.

It is proposed to replace the existing instmment air compressor / receiver tank skid unit with a new oil free air compressor / receiver tank skid unit Since the existing system does not have air dryers, filters, afler coolers, or water traps, these components will be included in the new system.

This will reqyire installation of a breaker unit in a spare cubicle or Motor Control Center (MCC) 1-23, with associated hardware. Installation of conduit from the MCC to the compressor / drier unit will also ' oe required, along with the appropriately sized cables.

SAFE'IT EVALUATION SUMM ARY A review of the Unit 1 Updated Final Safety Analysis Report (UFSAR) Chapter 14 and Unit 2 UFSAR Chapter 15, determined that there are no design basis accidents that are affected by this.

design change.

The Unit 1 River Water System and the Unit 2 Service Water System have the potential to be affected by this design change because the Intake Structure Instrument Air System (ISIAS)'

provides air to pressure control valve PCV-RW-117, which regulstes filtered seal water pressure

-to the Unit 1 River Water Pumps, and pressure control valve 2SWS*PCVll8 for theT Jnit 2 Service-Water Pumps. Upon loss ofinstrument air to these valves, the valves will fait closed.-

This will cause a loss of seal water pressure and will cause the supply valve from the discharge of

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Beaver Valley Power Station Umt 1 1991. Report of Facility Changes, Tests, and Experiments Page 58 or187_

the service or river water pumps to open, thus, maintaining seal water pressure. ' Therefore, this

- design change will have no adverse impact on any safety systems.

No failures of this design change will adversely affect the Unit 1 River Water or Unit 2 Service

- Water Systems. Since the River Water and Service _ Water Systems are not affected by this design change, their ability to mitigate the consequences of any accidents are not affected and thus the consequences of accidents previously evaluated in the UFSAR are not increased.

This design change will ensure that clean, dry instrument air is available to all air-operated -

components in the intake structure. This will ensure that equipment reliability k not lessened and will not increase the probabili:y of a malfunction.

This design change will neither increase nor decrease the severity of any malfunctions since no equipment important to safety is affected by this design change.

The primary failure mode is the inability to maintain air pressure. This failure mode is no worse than the previously evaluated failure mode of the existing components. It is expected that the-newly installed equipment will be more reliable. q The proposed design change will not create the possibility of an accident of a different type than previously evaluated in the st 'ety analysis report.

Existing equipment is being replaced with newer and larger equipment to improve reliability The most severe failure will not be any worse than the existing worst-case scenario (loss of air supply b7any means) and will not create any new accidents.

This des'gn change will improve air quality to ISA-S73 levels with newer and more reliable -

equipment. The systems function will not be changed.

There are no changes m parameters which affect the course of any accident analysis supporting l . Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, ,

Reactor. Coolant System boundaiy, or containment integrity, The instrument air system is not required to perform any safety-related functions at either Unit 1 or Unit 2, thus the provisions of Technical Specification 3/4.7.4 at Unit I and 3/4.7.4 at Unit 2 are -

not affected.

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Beaver Vallcy Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments-Page 59 ofI87 CII A NGE TITI,E DCP 1401, Rev.1, General Electric AK Air Circuit Breaker Modification CIIANGE DESCRIPTION Remove, overhaul, and re install 38 AK-25 and 6 AK-50 safety-related breakers and additional non safety-related breakers. The Maintenance Department will remove and re-install the breakers, and General Electric Company (the original manufacturer) will refurbish and convert these circuit breakers at their Philadelphia service center. The circuit breakers to be refurbished in lots of five at a time using substitute rpare breakers during the overhaul. The concerus of NRC Bulletin 88-10, the intent of NUREG 1000, Generic Letter 83-28, and Information Notices 87-12 and 89-45 were reviewed in conjunction with this change. The change is being implemented to increase plant reliability.

SAFETY EVAI,lTATION StiMM ARY The proposed design change will not involve an unreviewed safety question. .

The proposed modification does not represent a design change to the facility as described in the Updated Fiaal Safety Analysis Report (UFSAR). This Design Change Package (DCP) _first documents the decision to establNh a program to overhaul and upgrade these breakers, and secondly, this DCP replaces the original power sensor in each breaker with a new " state-of-the-art" Micro Versa Trip RMS-9 programmer, This DCP does not change system logic, actuation devices, or auxiliary supporting features of the AK air circuit breakers.

The proposed design change will not require a change to the Technical Specifications since no design parameters are changed.

i-The proposed change will not require changes to UFSAR Sections 8.4.2 or u.4.3. No design Dasis l accidents from UFSAR Chapter 14 are affected by the proposed design change.

L No safety systems will be adversely affected by the proposed design change. This modification i will be accomplished prior to and during :;hutdown, and the afTected breakers will be, replaced i with spare breakers during refurbishment.

l The probability of occurrence for an accident previously evaluated in the safety analysis report

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will not be increased. This modification increases the reliability of the Geneul Elactric AK air circuit breakers and therefore reduces the probability of malfunction.

p L The consequence of an accident previously evaluated in the safety analysis report will not Le increased because this modification will be performed during plant shutdown.

L The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this modification increases reliability.

Beaver Valley Power St'ation Unit _ !!

1991 Report of Facility Changes, Tests, and_ Experiments -

- Page 60 of 187-The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not' be increased because no accident 'previously evaluated _will be -

- affected.

Two failure modes of the proposed design change were reviewed; breaker fails open, and breaker -

fails closed.

The possibility for an accident of a different type than previously evaluated in the safay analysis report will not be created because the safety related breakers are dual train; therefore, any single failure is not a different accident.

The possibility for a malfunction of equipment important to safety of a different type' be  ; .;

previously evaluated in the safety analysis report will not be created because the safety-rei..ed breakers are dual train.

No changes in parameters that afTect the course of any accident analysis supporting Technical-Specification bases, or result in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment boundary are included in this design change.

The margin of safet'; as oefined in the basis for any Technical Specification will not be reduced because d!is design change does not impact Technical Specifications.

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Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 61 of I87 CII ANGE TITI,E M DCP 1405, Rev. C, Solid State Output Relay Testing CIIANGE DESCRIPTION Westinghouse has initiated technical bulletin NSD-TB-84-01 due to a potential undetectable failure in the Solid State Protection System (SSPS) output relay test circuitry. During testing of the various relays in the circ titry, the master relays are powered from 48 VDC (which is also their normal power voltage although frorn s difTerent source), and the slave relay coils are powered from 15 VDC via a test light. During normal operation, the slave relay coils are powered from 120 VAC via a shunt that bypasses the test light. The slave relays are not picked up during testing because of the reduced voltage of 15 VDC. A problem could occur if the shunt that bypasses the test light does not reclose after testing is completed, and the system is returned to normal. If a safeguards actuation is then called for,120 VAC would be applied to the slave relay coil through the test light. Since the test light is not rated for 120 VAC, it may fail before the slave relay operates, thus, the required safeguards actuation will not occur.

This modification proposes to eliminate the possibility of this failure by making a minor wiring change to both Train A and Train B output relay test circuits. Unit 2 has already performed this wiring modi 6 cation, but Unit I has been performing additional testing to ensure proper SSPS operation. The testing now done at Unit I takes approximately I hour and 45 minutes to do.

Technical Specification Table 3.3-1 allows a maximum limit of two hours for surveillance testing.

This modification will reduce the time of surveillance testing by approximately 15 minutes.

This modification proposes to remove the master relay contacts frorn the push-button test switch.

Three different test switches must be used to test the circuits. Their general operation and the atTects of this modification on each of the switches is as follows:

1. The TEST / OPERATE switch, when placed to TEST, switches the slave relay power from 120 VAC to 15 VDC and switches the master relay power from normal 48 VDC to test 48 VDC. This switch will not be afTected by the modification.

2. The Master Relay Selector switch is used to select the master relay to be tested. The circuitry for this switch will not be changed by the modification.

3. When the Push-button switch is depressed, the shunt is removed from the slave relay circuit and the 15 VDC " test" power is verified by the illumination of the test light. This ftmetion will not be affected by the modification. The master relay contacts previously associated with the push-button switch will now be permanently closed and each master relay will now be energize as soon as it is chosen with the selector switch (Rather than afler the push-button switch is depressed.) Since the master relay contacts will not be tied into this switch, the master relay will remain energized as long as the selector switch chooses that particular master relay. When the push-button is released, the shunt will be verified to reclose as long as the :est light extinguishes. (Prior to this modification, the light would extinguish when the push-button was released, but it could not be determined

l Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments -

Page 62 of I87 -

ifit was because of the shunt reclosing or because of the master relay contact reopening.)

' SAFETY EVAL,11ATION StiMMARY Numerous previously analyzed accidents were reviewed to determine if they will be affected by this modification including: Partial Loss of Forced Reactor Coolant Flow, Loss of Normal Feedwater, Accidental Depressurization of the Main Steam System, Accidental Depressurization of the Reactor Coolant System, and Spurious Operation of the Safety injection System at Power (See UFSAR Sections 14.1.5,8,13,15 and 16 respectively). A failure of the SSPS could initiate a spurious safety injection signal; however, this modification deals only with the SSPS output relay test circuitry and will not increase the probability of this type of occurrence. - Although the SSPS is necessary to help mitigate the other accidents, it is not associated with their initiation.

The safety system that will be affected by the proposed design change is the Solid State Protection System; however, this modification will affect only the test circuitry of the output -

relays and will not affect any of the protective functions of the SSPS.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report. None of the initiating factors for any of the previously analyzed accidents will be afTected by this modification.

The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report. None of the protective functions of the SSPS used to help mitigate any of the previously analyzed accidents will be affected by this codification.

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. Tids modification will affect test circuitry only. All safety related equipment functions required during the SSPS " operate" mode will remain unchanged.

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. This modification serves only to help ensure that the SSPS can per#orm its required function.' It will have no affect on any other safety related components, systems, or structures.

No new credible failure modes will be created by the _ implementation of this' modification. Thu removal of the interlock of the master relay contacts with the push-button test switch is needed to help provide a means of ensuring that the shunt in the slave relay circuit reclosed 'afler testing.

The permanently closed master relay contacts will not create any new failure modes; the contacts are located in the test portion of the circuitry and will have no 'afTect on the operation of the ciremt when in the operate (normal) mode. The remainder of the circuitry will not be affected by

. this change.

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 failures or potential hazards will be -

created by this change.

Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments i

Page 63 of187 I

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 new credible failure modes will be created by this modification. In fact, this change will provide a means to help ensure that an existing ptential failure does not occur.

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

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification. The reduction in time required to perform testing will aid in adherence to Technical Specification 3/4.3.1.

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Beaver Valley Power Station Unit i

~ 1991 Report of Facility Changes; Tests, and Experiments 1

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Page 64 of I87 Cil ANGE TITI,E DCP 1431, Rev,1, Surge Line Thermal Stratification Cil ANGE DESCRIPTION The rupture restraints for the pressurizer surge line (14"-RC-86-250lR) require gap modifi:ations. The following corrective actions will be performed in compliance with 'NRC +

Bulletin 88-11, " Pressurizer Surge Line Thermal Stratification". The surge line, rupture restraints, and pipe hangers will be thoroughly inspected to determine any gross. discernible distress or structu al damage. The rupture restraint gaps will be modified. Spring hangers Sil-1 and SH-2 will require a change in size to accommodate additional vertical pipe expansion. The analysis to modify the gaps will utilize standard industry methods and . vill be accomplished in accordance with established industry design criteria. The original conclusions of this safety evaluation, as determined from Revision 0 of Design Change Package 1431, are unaffected by this latest -

revision.

SAFETY EVALtl ATION SUMM AIW None of the UFSAR Chapter 14 design basis accidents will be adversely affected by_ this modification. A rupture of the pressurizer surge line has been previously analyzed (Section 14.3,

" Loss of Coolant Accident"). This modification will serve to help reduce the probability of a surge line rupture occurring.

No safety systems will be adversely alTected by the proposed design change. The functions of all safety-related systems and components, including that of the Reactor Coolant System, will remain unchanged.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety c.nalysis report All corrective actions performed by this modification will be supported by acceptable evaluations. This modification- will follow all applicable industry codes and standards and will be designed, installed, and tested as a Quality -

Assurance Category I, Seismic Class I modification. The reliability of the pressurizer surge line will actually be increased; The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis report. The functions of aU systems required to help mitigate any of the previously analy7ed accidents will remain unchanged.

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 functions will be affected by this modification. The surge line itself will remain unchanged. The gaps of the rupture restraints will be adjusted. - Spring hangers SH-1 and EH-2 will be replaced with hangers that will accommodate increased vertical pipe expansion. However, the function of this total pipe support system will remain unchanged.

Beaver Valley Power Station Unit 1:

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1991 Report of Facility Changes,= Tests, and Experiments Page 65 of I87 The proposed design change will not increase the consequences of a malfunction of equipment-

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important to safety as previously evaluated in the safety analysis report.: No other systems or equipment, other than discussed above, will be affected by this modification.

Possible failure modes that were reviewed included a rupture of the pressurizer surge line. This :

failure has been previously analyzed as a ddan basis accident.- This modification.will be performed to actually help decrease the probabi;ii.; of this accident from occurring.

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 other systema or components, other that presiously discussed, will be affected by this modification.

The proposed design change will not create the possibility of a nialfunction of a ditterent type than previously evaluated in the safety analysis report. No new failure modes or pof e tial hazardsw' ill be created by the implementation of this modification.

There are no changes in parameters which affect the course of any accident analysis supporting Technical Specification bases, or result inaxceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any.

Technical Specification. No Technical Specification or its basis. will be affected by this modification. Technical Specification 3/4.4.4 deals with the limiting conditions _of operation for the pressurizer heaters, and is unrelated to this modification. _ Technical Specification's 3/4.4.9.1 and 3/4.4.9.2. give the pressure / temperature limits of the Reactor Coolant _ System and the -

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pressurizer, and Technical Specification 3/4.4.10 deals with the structural integrity of ASME Code Class 1, 2 and 3 components. The bases for these Technical Specifications will not be-affected by - this- modification,1 and compliance' with these Technical Specifications will be maintained.

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Beaver Valley Power Station Unit I -

1991 Report of Facility Changes, Tests, and Experiments Page 66 of 187 CII ANGE TITLE DCP 1433, Rev. O, Invert Reactor Vessel Level Instrumentation System 11ead Sensors Cil ANGE DESCRIPTION The Reactor Vessel Level Instrumentation System (RVLIS) high volume sensors on the reactor head connection are presently installed with the capillary connection on the top :md the vent on the bottom. Westinghouse letter DLW-89 614 states that this arrangement can result in air in leakage which could adversely affect calibration (accuracy) of the sensors. This design change will minimize this problem by inverting the sensors, IRCS-LIS1311 and IRCS-LIS1321, so that the capillary connection is on the bottom. The capillary lines will be rerouted and supported as necessary. This work should be done without any additional capillary tubing, and therefore, a refill of the capillary lines should not be necessary.

In addition, this design change will replace the existing Swagelok vent cap on the sensor with Swagelok tubing, an instrument valve and cap. This will allow for better control while venting the process piping during calibration This same design chr.nge will be instituted for Unit 2 by Design Change Package 1438.

This design change will maintain the reliability, imegrity (both pressure boundary integrity and accuracy of measurement), and operability of the RVLIS and will have no adverse efTects on any other equipment.

SAFETY EVALUATION SUT,1M ARY No design basis accidents will be afrected because this design change does not adversely afTect any safety or non-safety systems, does not exacerbate any existing accidents, and does not .

introduce any new hazard beyond that already considered in tne Updated Final Safety Analysis Report (UFSAR).

This design change will not adversely afTect the safety function of any system. The reliability, integrity, and operability of the RVLIS will be maintained and no other systems will be affected.

The probability of an occurrence of any accident previously evaluated in the UFSAR will not be increased. This design change will maintain the reliability, integrity, and operability of the RVLIS, 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 evaluated in the UFSAR will not be increased -

because the reliability integrity, and operability of the RVLIS is being maintained and the change will have no effect on any other equipment. This design change will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR. This design change will not adversely affect any safety system used to mitigate an accident.

iBeaver Valley Power Station Unit 1.

1991 Report of Facility Changes Tests, and Experiments Page 67 of 187-The probability of a malfunction of equipment important to safety as previously evaluated in the = 7 UFSAR will not be increased because this design change will not adversely anct, either directly 'l

-or indirectly, any equipment, including the RVLIS.

The consequences of a malfunction of equipment important to_ safety as previously evaluated in -

. the UFSAR will not be increased. This design change will not adversely alTect any parameter; which would increase the consequences of a malfunction. This design change will not adversely _

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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 design change will not cause any new credible failure modes because the furidamental design -

feature and functions cf the equipment have not been significantly altered. This is:a relatively ,

minor change-P The possibility for an accident of a different type than previously evaluated in thqUFSAR will not - [

be created because nothing is being added or altered in a way which creates the possibility of a -

differentf type of accident. This design change is minor, and the reliability, integrity,' and operability of the RVLIS will be maintained, and no other equipment will be affected.

The poss_ibility for a malfunctica of a difTerent type than any previously evaluated in the UFSAR-will not be created becausa the fundamental design features and functions will not be changed in a :

way that creates the possibili;y of a malfunction of a ditTerent type: 'This design change is minor-and the reliability, integrityL and operability of the RVLIS will be maintained 1

This design chanFe will not change any parameter which affects the course of any accident ~

- analysis supponing -Technical Specification bases. lThe Technical _ Specification index was reviewed to determine if any bases might be affected. It was determined that this design change E

will not- sdversely affect the margin of safety- as defined in the bases- for_ any_ Technical-Specification; because the reliability, integrity, and operability of the RVLIS will be maintained,:

and no other equipment will be affected.

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y Beaver Vallcy Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 68 of187 i Cil ANGE TITI E DCP 1453, Rev. O, Closed Position Indication on Motor Operated Valve's Cil ANGE DESCRIPTION The Institute of Nuclear Power Operations, in Significant Operating Event Report (SOER) 86-2, identified a pot:ntial for inaccurate position indicatien on Motor Operated Valves (MOVs) that utilize a 2-rotor limit switch control scheme. The cor ective action proposed by the SOER is to utilize a 4-rotor limit switch design to allow separate settings for valve position indication and torque switch bypass.

This same subject is also discussed in NRC Generic Letter 89-10 and the 1989 NRC Combined inspection Report No. 50-334/89-10 and 50-412/89-11.

A review of MOVs at Unit I revealed that although 4-rotor limit switch assemblies are installed,_

the majority of these valves only utilize t 2-rotor control scheme, with the other two rotors installed as spares. The following valves will be rewired to utilize one of the installed spare limit switch rotors for closed position indication:

MOV-Cl! ll5B MOV-CII-275B MOV-CH-373 MOV-Cll-l l5C MOV-Cll 275C MOV-SI-836 MOV-Cil-l l5D MOV-Cll-289 MOV-SI-867C MOV-CII-l 15E MOV-Cll 310 MOV-SI-867D MOV-Cil-275 A MOV-Cll-350 The required changes to the limit switch wiring will be confined to the Limitorque motor operator housing for each valve. Each of the affected valves and motor operators is QA Category I, seismic class 1.

This design change will allow separate settings for valve position indication (closed) and torque switch bypass.while not affecting valve control, ftmetion, or response time, and is therefore acceptable.

This design change will maintain the reliability, integrity, and operability of the MOV operators and will have no adverse efTects on any other equipment.

SAFETY EVAlliATION SilMM ARY No design basis accidents will be atrected because this design change 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 Updated Final Safety Analysis Report (UFSAR). This design change will not adversely affect the safety function of any system.

The reliability, integrity, and operability of the MOV operators will be maintained and no other systems will be affected.

13eaver Valley Power Smtion Unit 1:

1991 Report of Facility Changss, Tests, and Experiments

' Page 69 of 187 The probability of an occurrence of any accident previously evaluated in tl.e UFSAR will not be increased. This design change will maintain the reliability, integrity, and operability of the MOV operators, and it will have no_ e&ct on any other equipment; therefore, the _ probability of ,

occurrence for t.ny accident will not be increased.

, . . The consequences of an accident previously evaluated in the-UFSAR will not be increased because the reliability, mtegrity, and operability of the MOV operators is being maintained, and the change will have no effect on any cther equipment. This design change will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR. This design change will not adveisely affect any safety systems used to mitigate an accident.

The probability of a malfunction of equipment important to safety as previously evaluated'in the UFSAR will not be increased because this design change will not adversely affect, either directly or indirectly, any equipment, including the MOV operators._ >

The consequences of a malftmetion of equipment important to safety as previously evaluated in -

the UFSAR will not be increased. This design change will not adversely affect any parameter ,

which would increase the consequences of a malfunction. This design change 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 design change 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 because nothing is being added or altered in a way which creates the possibility of a -

different ' type of accident. This design change is m'mor, and the reliability, integrity, and operability of the MOV operators will be maintained, and no other equipment will be affected.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR' will not be created because 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 design change is minor and the reliability, integrity, and operability of the MOV operators will be maintained.

This design change will not change any parameter which affects the course of any _ accident:

analysis supporting Technical Specification bases.

The Technical Specification index was review' d to determine if any bases might be affected.' It was determined that this design change will no. adversely aEct the margin of safety as dermed in the bases for any Technical Specifications because the reliability, integrity, and operability of the_

MOV operators will be maintained, and no other equipment will be affected.

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x Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 70 of 187 CII A NGE T!TLE DCP 1464, Rey,0, Transfer of Battery 1-5 Loads to Battery 1-2 and 1-1 CII A NGls,DESCRil,' TION Battery duty cycle calculatiot:s for batteries 1-1 and 1-2, performed for Design Change Package (DCP) 673, i-dicated that there was inadequate operating capacity available to supply both the .

Class lE loads and the non-Class IE loads. This prompted the station to transfer the non-Class lE loads from busses DC-BAT-1 and DC-BAT-2 to bus DC-BAT-5. Since then, there have been sinbl e failure concerns over-having all the power for the non-Class IE control and protection systems powered from one source (DC-BAT-5). Two problems which occur with a single failure of bus DC-BAT-5 are:

1. The plant w"I trip and experience a concurrent loss of offsite power incident due to the opent,g of the relays and the bus transfer circuits failing to swap between the unit and i system station sersice transformers.

2. The station will lose all the electrical protection to the major power systems in the plant.

This is due to the dual redundant circuits for the electrical protection of the main generator, main transformer, and station bus supplies being powered all from DC-BUS-5.

Since DCP 673'was implemented the electrical calculations for DC-BUS-1 and DC. BUS-2 have been refined and suflicient margin has been identified that will' allow some non-Class IE loads to-be switched back to DC-BUS-1 and DC-BUS-2. This design change will implement the necessary administrative and breaker position changes to switch the 4KVS-1B loads back to DC-BUS-1, and the 4KVS-lC and PNL-DC-4 loads back to DC-BUS-2. By performing these changes, the single failure concerns over the failure of DC-BUS-5 will be eliminated since the power sources will be diversified.

DCP 1464 requests that the DC control power to 4KVS bus l A be supplied from DC- BUS rather than the DC control power to 4KVS bus IB. Also the nomenclature used in the DCP describing the involved loads (4KVS-1A and 4KVS-lC) was . revised to better xplain the significance of these breakers.

Emergency 4KVS busses I AE and 1DF are powered from 4KVS busses I A and ID respectively and switching the DC power supply to bus 1 A, instead of IB, back to DC-BUS-1 assures that the DC control power to both busses I A and ID can not be lost by the failure of a single DC Bus, (DC-BUS-5). Suflicient margins must be shown to exist for DC-BUS-1 and DC-BUS-2 to allo'w for the requested load distribution. Electrical calculations will be performed / revised to substantiate this revision.

S AFETY EVALUATION SUMM ARY The Updated Final Safety Analysis Report (UFSAR) Chapter 14 accidents which were reviewed

' include Section 14.1.8, Loss of Normal Feedwater; Section 14.1.11, Loss of Offsite Power to the

t Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 71 of I87 Station Auxiliaries (Station Blackout); and Section 14.2.9, Complete Loss of Forced Reactor Coolant Flow. It was determined that these design basis accidents would not be adversely affected by the implementation of the proposed design change.

The safety related DC busses DC-BUS-1 and DC-BUS-2 will have additional loads connected to o them, however, suflicient margins in the battery duty cycles exist based on revised electrical calculations. Therefore, this proposed design change will not adversely affect any safety systems.

Suflicient margins exist in the DC-BUS-1 and DC-BUS-2 duty cycles to allow for these additional loads. Additionally, degradation of the Class lE busses (DC- BUS 1 and DC-BUS-2) by abnonnalities in the non-class IE loads and power systems is prevented by two breakers in series, and positive breaker interlocks exist which make it physically impossible to parallel either Class IE bus to DC-L1US-5.

A loss of normal feedwater accident could result from the failure of either DC- BUS-1 or DC- '

BUS-2 since the main feedwater regulating valves would fait closed on the loss of DC power to their solenoid vahes. The probability of this occurring, however, is not increased since this modification will not increase the probability of failing bus DC-BUS-1 or DC-BUS-2.

The loss of DC-BUS-5 would cause all three reactor coolant pump (RCP) under voltage (UV)

. relays to open and result in a two out of three RCP UV reactor trip. This in turn would lead to a turbine / generator trip and a bus transfer to offsite power would try to occur. liowever, since the .

bus transfer circuits that permit the 4160 volt station busses to swap between the unit and system station service transformers are powered from DC-BUS-5, the transfer to ofTsite power would not occur and a loss of offsite power to the station auxiliaries would result. By performing this modification, diversified power sources to the RCP UV relays would exist, and therefore, it would take the failure of multiple DC busses to lead to this type of accident. The probability of this type of accident occurring due to the failure of DC-BUS-5 would, therefore, be decreased.

The probability of a complete loss of ferced reactor coolant flow due to a loss of the DC power supply to the RCP UV relays would be reduced. - By performing this modification, it will now take the failure of at least two DC busses to produce this reactor trip signal, as opposed to the failure ofjust DC-BUS 5.

Implementation of this proposed design change will have no effect on the design basis accidents reviewed nor the offsite doses associated with them, and it will not increase the consequences of

- these accidents.

The loads which are being switched to DC-BUS-1 and DC-BUS-2 were originally configured and designed to be powered from these busses.. Implementation of this design change simply involves repositioning the existing breakers back to their original configuration; i.e., the breaker contacts which are shown as normally open on UFSAR Figure 8.4-2 will now be normally closed and powered from DC-BUS-1 or DC-BUS-2. Since suflicient margins in the duty cycles exist to have these loads switched back to these busses, the probability of a malfunction of equipment important to safety will not be increased.

Beaver Valley P_ower Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 72 of 187

- Failure modes considered include:

1. Electrical shorks. The Class IE bus to non-Class _lE bus is supplied by two breakers in series, failure of one of these breakers to clear a fault would be backed up by the other.

2 Excessive loading of the safety-related batteries Revised battery duty _ cycles for DC-BUS-I and DC-BUS 2 show that suflicient margins exist to have the proposed loads placed back on these busses.

3. Seismic events. No new equipment is being installed, simply a repositioning of the existing breakers.  :;

This design change will not create the possibility of an accident of a different type than pieviously evaluated since the credible failure modes are not expected to result in an unanalyzed condition.

This modification will not install any new equipment and will not affect the opert. tion of any components as described in UFSAR Sections 7.2,1;1,4, 8.3, 8.4 or _ 8.5. The testing and inspection methods and programs for the 125 volt DC system will be maintained as described .in UFSAR Sections 8.5.3 and 8.6 to ensure their reliability. Therefore, the possibility of a malfunction of a different type is not created.-

There are no parameter changes associated with this design cbge which could affect the course of any accident analysis supporting Technical Specification Bases. The margin of safety as defined in Technical Specification Baces 3/4.3.1 and 3/4.8.2 will net be reduced. .By performing this modification, a higher degree of piant reliability will be attained since a single DC bus failure will no longer disable the electrical protection circuits for the major plant power equipment, not will it result in a spurious reactor trip due to all three RCP UV relays opening.

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Beaver Vallcy Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments c Page 73 of187 CII ANG E TITI,E DCP 1467. Rev. O, hicisture Separator Reheater Replacement Tube Bundles CIIANGE DESCRIPTION hioisture Separator Reheater (htSR) deterioration has resulted in large outage maintenance expenditures during the sixth and seventh refueling outages. Steam seal strips, located in the htSR tube bundles, have been breaking loose and entering the low pressure turbines causing impact damage to stationary and rotating components. htSR operating data was collected following the sixth refueling outage. A computer heat balance program was used to compare the act"al hiSR performance with the design parameters of the proposed components. The results indicated that a unit electrical output increase of 10.5 h1WE will result This calculation is on the conserutive side due to an assumed higher than actual MSR moisture removal effectiveness. .

During the seventh refueling outage, an additional seventy-nine (79) tubes were plugged in the C-hfSR which will further increase the unit heat rate. -

This design change will:

1. Replace the moisture separator reheater Copper - Nickel (Cu-Ni) tube bundles with a redesigned 429 stainless steel tube bundle. The moisture separation chevron sections will be replaced with improved moisture separation components.

2 Replace the reheat steam control valve [FCV-hts-100A and C] trim to enable more stable automatic control system performance.

3. Modify the reheat control system circuitry (RC time constant) to provide an automatic two hour RAMP function to be used during htSR startup.

4. Replace the MSR founh pass vent control stage (orifice) to the first point heaters with a Chromium - hiolybdenum (Cr-Mo) manual valve. At 100% load,- the valve will be adjusted to obtain the optimum flow. rate based on installed tube bundle thermocouple temperature data.

5. Replace the pipe elbows downstream of the htSR fourth pass vent control stage to the first point heaters with stainless steel elbows to reduce erosion susceptibility.

6. Install local readout devices at the reheat steam flow annubar locations. Design Change Package (DCP) 722, " Installation of Annubar Flow Instruments in MSR Reheat Steam Lhes" installed the annubars but did not complete the flow transmitter installation portion.

DCP-722 is presently in " Partial" turnover status. The DCP-722 design concept will be revised to show the present configuration as being complete to close out DCP-722. ,

7. Remove the reheat purge isolation valves [TV-SD-101 A, B, C, D] from service. The-optimum configuration for purging and preheating the four pass tube bundles design is through the fourth pass vent to condenser [MOV-SD-104 A, B, C, D].

' Beaver Valley Power Mation Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 74 of 187

8. hiodify the control logic for the reheat steam supply isolation valves [hiOV-hiS-100A, B]

and the excess steam vent valves [MOV-SD-104A, B, C, D] and [MOV-SD-105 A, B, C, D]. ' The turbine trip signal which presently realigns the excess' steam vent from the first point feedwater heaters to the condenser following a turbine trip will be removed. The turbine trip signal will_be included in the reheat steam supply isolation valve [h10V- hts-100A, B] control circuits to automatically isolate the MSR reheat steam following a turbine trip.

• SAFETY EVAL,UATION SUMM ARY No Updated Final Safety Analysis Report (UFSAR) Chapter 14 accidents will be affected by this DCP because ws 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 afTect the safety function of any system. This design change is not safety related. It will solve moisture separator deterioration problems.

This DCP will replace the moisture separator reheater tube bundles plus some oti.er minor modifications. The alTected components do not have a safety function, therefore, the probability of occurrence for any accident will not be increased.

This DCP will not afTect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR. This equipment does not impact the ability of safety systems to respond to an accident; therefore, it will not increase the consequences.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not increase because this modification is independent from the safety function of other equipment, It will not decrease the reliability of other equipment.

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; therefore, there will be no.effect on the consequences of a malfunction of equipment important to safety.

Identify the credible failure modes associated with the proposed change. List failure modes which were reviewed and identify the criteria or reasons which conclude that these modes have been previously evaluated.

Failure modes of the proposed design change which were reviewed included the equipment failure of the turbine genera'or system. Since it is not safety-related, it does not introduce a new accident.

No possibility for an accident of a different type than those that have already been evaluated is foreseen. An a:cident of a different type will not be created.

Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 75 of 187 No new safety-related component functions are being created. The failure of this equipment has no elTect on any other safety related equipment. The failure of the equipmet.t will not affect the safety function of other equipment.

There are no changes in parameters which afTect the course of any accident analysis supporting Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, ,

Reactor Coolant System boundary, or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification. No Technica! Specification basis will be affected, in any way, by thi>

modification.

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L Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 76 of 187

.CII ANGE TITI.E DCP 1470, Rev. O, Unit i Instmment & Control Shop Ventilation Hood-CIIANGE DESCRIPTIOf Several INPO findings and an incident during the past year demonstrate the need for a fume exhaust hood in the Unit 1 Instmment and Control (I&C) Shop. The objective of this design change is to provide a means of exhausting fumes from the Unit i I&C Shop. This will provide a safe method of handling Chlorine (during calibration procedures) and other hazardous materialt SAFETY EVAL,UATION SUMM ARY Updated Final Safety Analysis Report (UFSAR) Chapter 14 design basis accidents were reviewed to identify design basis accidents that could be affected by the proposed modification. This design change is non-safety related and will in no way afTect design basis accidents. It has no impact on any safety function of any system. The ventilation system has a negligibleimpact.

This design change will not adversely affect the safety function of any system. The fume hood will not be powered from a Class IE power source. It does not have a significant impact on the ventilation sysicm because the total exhaust is only 800 cfm.

This design change will not increase the probability of occurrence of an accident of the ventilation system or any other accident previously evaluated. This design change will provide a safe method of handling Chlorine during calibration procedures.

This design change will not affect any parametei which would increase the consequences of an accident beyond that previously considered in the UFSAR. This equipment does not affect the ability of safety systems to respond to an accident, and therefore will not increase .the consequences.

The probability of a malfunction of equipment important to safety as previously evaluated it, the:

safety anabis report will not be increased because the ventilation hood will not be powered from a Class IE powr source. The ventilation hood will not decrease the reliability of other -

equipment.

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

Fsilure modes of the proposed design change which were reviewed included the equipment failure of the ventilation system and power failure in the I&C Shop. No possibility for an accident of a different type than those that have already been evaluated is foreseen.

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- Beaver Valley Power Station Unit 1 -

1991 Report of Facility Changes, Tests, and Experiments-_-

Page 77 of I8'l -

No new safety-related component functions are being created.1 This equipment is not safety--

- related.' The fume hood will not be powered from a Class IE power source. This design change

. has a negligible affect on the ventilation system. The failure of this equipment will not affect the -

safety function ofother equipment.

rThere are no changes in parameters which affect the course of any accident analysis supporting Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment integrity. The proposed design chsnge does not reduce the rnargin of safety as dermed in the basis for any Technical Specification. No Technical Specification basis will be affected in any way by this design change.

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Beaver Vd!:y Power St tion Unit 1 1991 Report of Facility Changes, Tests, and Experiments.

Page 78 of I87 Cil ANGE TITI,E DCP 1476, Rev,0, Color Separation Deficiencies Resolution CII ANGE DESCRIPTION This design concept proposes to resolve color separation deficiencies identified with the annunciator cables for trip valves TV-MS-101 A, B, C. The affected cables are non safety-related and non-redundant. This modification will reroute and rewire the circuits such that each will now -

be associated with only one color train.

SAFETY EVAI,UATION SUMM ARY No design basis accidents from Updated Final Safety Analysis Repon (UFSAR) Chapter 14 are -

affected by the proposed design change.

The probability of an occurrence or the consequence of an accident previously evaluated in the safety analysis report will not be increased because the proposed change will not affect any safety-related systems and_ will not, therefore,- afrect the accident analysis. The annunciator is non- '

safety-related.

The prcbability or the consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because no equipment imponant to safety will be adversely affected.

This change will not cause any new credible failure modes because the fundamental design features ano failure modes of the equipment have not been altered.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not -

be created. This change involves non safety-related electrical systems and does not affect fluid system pressure boundaries or introduce a different type of component.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR will not be created. A malfunction such as loss of power, or a short circuit is not a different type malfunction than any previously considered.

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

The Technical Specification index and Specification 3/4.6.2 were reviewed to determine if any bases might be afTected. It was determined that this change will not adversely affect the margin of safety as defined in the bases for any Technical Specification because this change is minor and affects non safety-related annunciation which is not included in the Technical Specification, and no other equipment will be affected.

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Beaver Valley Power Station Unit i J 1991 Report of Facility Changes, Tests, and Experiments i Page 79 of 187 CII ANGE TITLE DCP 1482, Rev,1, Group 1 Fire Damper Replacement CIIANGE DESCRIPTION I

The design objective is to replace seventeen ('7) Group 1 non-UL rated fire dampers in areas with a greater than 1-hour fire loading. The replacement fire dampers will have a 3-hour UL fire rating label, will be seismically qualified per Unit I criteria, and will be qualified in close under air flow conditions. These fire dampers wiu be furnished with CO 2discharge release mechanisms in most cases, and must also close when exposed to heat. Existing CO 2 fire damper actuation lines will be adapted to the new release mechanisms. Replacement of these seventeen (17) fire dampers will include replacing only the fire damper assembly; however, some ductwork modifications and fire seal replacements may be necessary when installing the new fire dampers.

Fire dampers VS D-180, VS-D-281, and VS-D-291 are iocated in QA Category I ductwork, and therefore will be QA Category 1, Therefore, the reliability, integrity, and operability of the ductwork (Supplemental Leak Collection and Release System [SLCRS]) will be maintained. All of the other fire dampers will be QA Category F.

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This design change will replace eleven (11) Group 1 fire dampers in the Service Building Cable  !

Tray area and six (6) Group 1 fire dampers in the Safeguards Building East and West Cable Vault areas.

SAFETY EVALUATION SUMM ARY No Updated Final Safety Analysis Report (UFSAR) Chapter 14 accidents will be affected by this design change because this design change does not adversely affect any safety or non-safety q systems, does not exacerbate any existing accidents, and does not introduce any new hazard beyond those already considered in the UFSAR.

This DCP will not adversely affect the safety function of any system. The reliability, integrity, and operability of fire dampers and the Supplementary Leak Collection and Release System (SLCRS) 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 design change will maintain the reliability, integrity, and operability of the fire dampers and SLCRS, and it will have no effect on any o+er equipment; therefore, the probability of occurrence for design basis accidents will not be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased. This design change is minor and the change will have no efTect on any other equipment. This design change 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 fire dampers and SLCRS is being maintained.

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c gi Beaver Valley Power Station Unit 1

- 1991 Report of Facility Changes, Tests, and Experiments

-_ Page 80 of 187.. >

L The probability of a malfunction of equipment important to safety as previously evaluated in the--

UFSAR w!! not be increased. _ This design change is minor and the changes will not adversely -

? affect any equipment,-including the fire dampers and SLCRS.

The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This design chance wih not adversely affect any parameter which would increase the consequences of a malfunction. This design change will not adversely .

affect any safe'y system used to rrdtigate ra accident. Therefore, there will be no cfrect on the l consequences of a malfunction of equipment important to safety.

This design change 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 design change is minor, and the reliability, integrity, and operability of the fire dampers and SLCRS 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 design change is minor and the reliability, integrity, and operability of the fire dampers and SLCRS will be maintained. - The fundamental design features and ftmetions will not be changed in a way that creates the possibility of a malfunction of a different type; This design change will not change any parameter which afTects 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 design change-will not_ adversely affect the margin of safety as . defined in the bases for any -Technical- ,

Specifications,-including 3/4.7.8, because the reliability,- integrity, and_ operabaity:of the fire ,

dampers and SLCRS will be maintained, and no other equipment will be affected.

Beaver Valley Power Station Unit I  !

1991 Report of Facility Changes, Tests, and Experiments Page 81 of187 Cil ANGE TITLE DCP 1484, Rev. O, Feedwater Bypass Valve Control hiodification CliANGE DESCRIPTION This modification proposes to improve the performaae of the Low Power Steam Generator Water Level Control (LPSGWLC). LPSGWLC continuously compares measured steam generator level with a setpoint and reactos pow:r level (from N44) to regulate the bypass feedwater control valves, (FCV-FW-479,489, and 499). The present scheme uses proportional plus integral control (PI) and uses level as the only feedback variable.

Level as a function of fluid mass, is dependent on the relationship between the water and the entrained steam. Fluctuations in the mass of entrained steam can cause a significant fluctuation in-the level as sensed by the narrow range 1; vel instruuentation. This is experienced during the shrink and swell phenomena. When these oscillations reach either the Ili-Ili or Lo-Lo level setpoint, a reactor trip occurs.

Better level control would be available if cn additional feedback variable that provides a measure of entrained steam was used. The variable related to entrained steam mass that is readily available is steam flow. A study of the steam generator level control problem has been performed to assess the efficiency of using steam flow in a feed fonvard (FF) scheme.

Differences in the LPSGWLC were evaluated using both the PI and the FF control schemes.

Although no significant differences were evident as a result of simulated feedwater transients, the FF scheme proved superior following a simulated steam transient. The steam transient caused a

- severe undershoot of the PI control scheme but, affected the FF scheme to a much smaller degree.

Therefore, the proposed solution is to incorporate a FF control scheme using a steam flow signal initiated at flow elements (FE-hiS-474,484, and 494). All changes will take place in the process rack area of the Senice Building and will involve non-safety related electrical components.

SAFETY EVALUATION SUSINI ARY This n.sdification serves only to improve the performance of the LPSGWLC scheme. It will use a signal generated at a hiain Steam System flow element (FE-hts- 474,484,and 494) to help.

compensate for the affects that the steam mass has on the water level. No component or system functions, either safety or non-safety related, will be adversely affected by this change, nor will any of the previously analyzed accidents, including those involving feedwater or steam system transients, be affected. (

Reference:

Updated Final Safety Analysis Report Sections 14.1.8, 14.1.9, 14.2.5, and 14.2.I1).

L No safety systems will be affected by the proposed design change. The steam generator level and i setpoint signals are used for control and are isolated from protection circuitry. All portions of the j steam flow circuitry that will be involved in the . modification are also non-safety related even though their point of origin is from safety related flow elements (FE-hiS-474,484, and 494).

Adequate signalisolation occurs upstream of the proposed changes.

licaver Valley Power Station Unit.1 1991 Report of Facility Changes, Tests, and Experiments -

'Page 82 of 187 The propor.ed design ' change will not increase the probability of occurrence _of an accident previously evaluated in the safety analysis report. This modification will have no affect on any of the initiating circumstances for any of the previously analyzed accidents.

The proposed design change will not increase the consequences of an accident previously Jevaluated in the safety analysis report, This modification will have no affect on any of the systems or components required to help mitigate any of the previously analyzed accidents.

The proposed design change will not increase the probability of occurrence or consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis report.

All safety related equipment functions will be unafTected by this mo lification. All electrical work '

performed as part of this modification is non-safety related however, the work will b' icismically supported as applicable due to the proximity of these changes to safety related components.

The appropriate installation and testing will help to ensure that the new control scheme _ will be as reliable as the old one. However, in the unlikely event of failure in the steam flow signal circuitry no new unevaluated conditions would arise. No changes to the actions of the bypass flow control valves will result since this modification deals only with the generation of the control signal to these valves and not with the valves themselves. ~ Also, no safety related ftmetions or protection

- channels will be affected.

. The proposed design change will not create the possibility of an accident of a different type than a previously evaluated in the safety analysis repo,t 'No safety related components, systems or functions will be affected 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. No new failure modes or potential hazards will_

be created by this modification.

There are no changes in parameters which affect the course of any accident analysis supporting -

Technical Specification bases, or that result in exceeding the acceptance criteria for fuel cladding, .

Reactor Coolant System boundary, or containment integrity, The proposed design change does not reduce the margin of safety as defined in the basis _for any Technical Specification. The signals used for this modification are used for control purposes ad are adequately isolated from protection circuitry therefore, no Technical Specification-or it's-bases, will be affected by this change including, Technical Specifications 3/4,3,1 and 3/4.3.2.

Based on the above review, this modification is non-safety related, Quality Assurance (QA)

Category II. However, due to the close proximity of safety related components in the process rack area this nodification must meet seismic requirements. Although the feedwater bypass-valves will be considered as QA' Category I components, this modification will still affect only non-safety related portio _ns of the system. . _Only the control portion of the LPSGWLC will be affected. The ability of these valves to isolate feedwater on a feedwater isolation signal will be unaffected by this modification.

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11eaver Valley Pmver Station Unit 1 1991 Report of Fccility Changes Tens, and Expcriments Page 83 of I87 CII A NGE TITI.E DCP 1514, Rev. 1, Regulatory Guide M Limit Switch Upgrade Outside of Containment Cil ANGE DESCRIPTION This mooincation consists of a one for-one replacement of the existing hmit switches and associated SIS wire on approximately 92 valves in ten systerns with qualified NAh1CO limit swixhes and quick disconnect environmentally qualified cable. Additionally, most of the mountin3 plates for the limit switches will be replaced. The new limit switches will be qualified NAh1CO switches which meet the requirements of Regulatory Guide 1.97 and are seismically and environmentally qualified. This modification is to be implemented in response to commitments made to the NRC in a letter dated agust 1,1988.

Revision i adds three (3) limit switches (for hiain Steam isolation Valve test circuits), to the scope of this design change. Fail . , af the presently installed switches have been attributed to high temperatures in the main steam valve area. Therefore, they will be replaced by NAhiCO high temperature limit switches. The SIS wire to the new hiain Steam Isolation Valve (h1SIV) test switches will remain hard wired; the flex conduit for these limit switches will be replaced.

Revision 1 of this design change will not affect any of the other limit switches that were previously discussed in Revision 0.

SAFETY EVAI,UATION SUMM AHY No design basis accidents found in Updated Final Safety Analysis Report (UFSAR) Chapter 14 are affected by the proposed design change. Safety systems will not be adversely afrected Fy the proposed design change.

The probabilty of occurrence for accidents previously evaluated in the safety analysis reo 'rt will not be increased. Chapter 14 of the UFSAR does not discuss failure of a limit switch. Sh,cc this change willimprove component reliability, the probability of occurrence for previously evaluated accidents will not be increased.

The consequence of an accident previously eveluated in the safety analysit report will not be increased because the change improves component reliability and does not change the original design intent. Failure of a limit switch will not impair the ability of containment isolation valves to close.

The probability of a malfunction of equipment important to safety as previously evaluated in the 4- safety analysis report will not be increased because this change improves component reliability.

This is a one-for-one replacement with equipment which is equal to or better than the original equipment.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this is a one-for-one change which does not i o

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lleaver Valley Power Station Unit i 1991 Report of Fa:ility Changes, Tests, and IIxperiments Page 84 of 187 affect the original design intent.

Failure modes of the proposed design change which were reviewed include loss ofindication and cironeous indication. Iloth of these failure rnodes are consistent with the potential failure modes which existed prior to the change and do not represent different types of failures.

The possibility for an accident of a Jifferent type than previously evaluated in the safety analysis report will not be created because the proposed change is a one for one replacement and does not change the system operation The poscibility for a analfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because this modi 0 cation is increly an equipment upgrade.

The MSIV test limit switches are used only to verify MSIV panial closure during power operation. The safety related function of the MSIV's is to close fully, on demand. These switches are nc,i requirec' for this or any other safety related function and as explained above, no new failure modes will be introduced.

No changes in parameters that affect the course of any accident analysis suppor:ing Technical Speci0 cation bases, or that re* ult in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment boundaiy are included in this design change. Technical Speci0 cation 3/4 6.3.1 was reviewed and remains uaalrected. Al t hough the MSIV test limit switches are used during the perfi>rmance of the valve partial closure tests, as required by surveillance 4.7.1.5, this replacement will have no adverse affect on Technical Specification 3/4.7.15. Additionally, the margin of safety as denned in the basis for any Technical Speci0 cation will not be reduced because this design change improves component reliability.

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Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Enperiments Page 85 of 187 CII ANGE TITI.E DCP 1521. Rev. O, Unit I hiodification for lleat Exchanger Performance hionitoring Cil ANGE IESCRIPTION NRC Generic 1.etter 89-13 requires each licensee to establish an acceptable Surveillance hionito ing Program to ensure that safety related heat exchangers that are cooled by open loop cooling water systems (such as the River Water System) are not allowed to degrade excessively without being detected and corrective actions taken. This modification _will add thermowells, pressure taps, and Ultrasonic Flow Instmments (UFI's) to provide for performance monitoring of heat exchangers.

Where UFrs are being installed, short sections of carbon steel piping will be replaced with stainless steel sections (spool pieces). The UFI's are sensitive to the build-up of corrosion and crud on the inside of the pipe; therefore, the use of stainless steel spool pieces will minimize this '

problem. Ifit is believed that the spool piece has built up crud on the inside, it can be removed and cleaned. The UFI's will be strap?cd (clamped) onto the outside of the spool piece, and therefore are not pressure boundary cemponents. The stainless steel spool pieces will meet the -

design pressure, temperature, and code recuirements, and will be QA Category I and Seismic Cless I; therefore, the reliability and integrity of the piping are considened to be maintained. The stainless steel spool pieces will be properly insulated from the carbon steel pipe.

The thrmowells consist of a fitting welded into a hole cut into the piping, into which the threaded "well" piece is screweu. The "well" is a thin walled scaled end tube which projects down into the pipe and into which a temperature sensing deWee can be inserted. The "well" may also be inserted into one leg of a piping ' Tee' if the pipe diameter is too small to accommodate a normal perpendicular entry. The thermowells will meet the design pressure, temperature and code -

requirements, and be QA Category I and Seismic Class I; therefore, the reliability and integrity of the piping are considered to be maintained.

This design change will add unions (a threaded joint) to the liver water lines serving the charging pump lube oil cooler, Cll-E-7A, to facilitate maintenance. The other two lube oil coolers, Cli E.

7B, C, already have such unions. These unions will meet design pressure, temperature, and code requirements and will be QA Category I and Seismic Class I; therefore, the reliability and integrity of these river ' fater lines are considered to be maintained.

This design change will also add a pressure tap and instrument root valve to several of the lines, as described in the design concept. _ ;These valves will be nonnally closed and will serve as the

, class break. The valves will meet the design pressure, temperature and code requirements, and will be QA Category I and Seismic Class I; therefore, the reliability and integrity of the lines are considered to be maintsined.

Design Change Package (DCP) 1502 instituted the similar, corresponding modifications for Unit 2.

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Deover Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 86 of 187 These modifications are very minor and will have a completely insignificant effect on system flows and pressure drops.

This design change will affect the River Water System (RWS), Reactor Plant Component Cooling Water System (CCR), and charging pump lube oil lines which are a support system of the Chemical and Volume Control System (CVCS).

This design change will maintain the reliability, integrity, and operability of the RWS, CCR, and CVCS, and will have no adverse effects on any other equipment.

l

$_AFETY EVAL.IIATION S11MMARY 1

This design change will not require any changes to the Technical Specifications. Technical l Specification Table 3.6-1 lists containment penetrations and valves. The instrument root valves associated with the modification will be inside the containment penetration valves, MOV.RW-105A, B, C, D for penetrations 83,84,85,86. llowever, numerous other small valves (e g., RW-603-606) on lines that are capped during normal operation are not included in the Technical Specification table, and therefore the added valves will not be added to the table.

l This design change will require a change to the Updated Final Safety Analysis Report (UFSAR).

Figures 9.4-1 and 9.9-1 A will need to be updated. For the same reason as discussed above, Table 5.3 1 and Table 5.3 1 A will not be updated.

This design change will not involvt . n Unrev:ewed Environmental Question.

This design change will not require a change to the Environmental Protection Plan.

No design basis accidents will be atrected because this design change does not adversely affect. ,

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

This design change will not adversely affect the safety function of any system. Tine reliability, integrity , and operability of the RWS, CCR, and CVCS will be maintained and no other systems will be afTected.

The probability of an occurrence of an accident previously evaluated in the UFSAR will not be increased. This design change will maintain the reliability, integrity, and operability of the RWS, CCR, and CVCS and it will_have no efTect on any other equipment; therefore, the probability of occurrence for design basis accidents will not be increased.

The' consequences of an accident previously evaluated in the UFSAR will not be increased because the reliability integrity. and operability of the RWS, CCR, and CVCS is being maintained .

i and the change w.ll have no effect on any other equipm:nt. This design change will not affect any

! parameter which would increase the consequences of an accide eyond that previously l considered in the UFSAR. This design change will not adversely afTect any safety system used to ,

mitigate an accident.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 87 of187 t The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased because this design change will not adversely affect, either directly ,

or indirectly, any equipment, including the RWS, CCR, and CVCS. i The consequences of a malfunction of equipment impoitant to safety as previously evaluated in the UFSAR will not be increased. This design change will not adversely affect any parameter which would increase the consequences of a malfunction. This design change 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 design change will not cause any new credible failure modes because the fbndamental design features and functions of the equipment have su been significantly altered. This is a relatively minor change. The possibility ofleaks or ruptures in the affected lines already existed prior to these changes, and the probability of their occurrence is not considered to be increased due to these changes.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created because nothing is being added or altered in a way which creates the possibility of a -

different type of accident. This design change is minor, and the reliability, integrity, and operability of the RWS, CCR, and CVCS will be maintained, and no other equipment will be alTected.

The possibility for a malfunction of a difrerent type than any previously evaluated in the UFSAR _

will not be created because 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 design change is minor and the reliability, integrity, and operability of the RWS, CCR, and CVCS will be maintained.

This design change will not change any parameter which affects the conrse of any accident analysis supporting Technical Specification basas. The Technical Specification index -was reviewed to determine if any bases might be affected. It was determined that this design change will not adversely affect the margin of safety as defined in the bases for any Technical i Specifications because the reliability, integrity, and operability of the RWS, CCR, and CVCS will be maintained, and no other equipment will be affected, i

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Deover Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 88 of 187 CII A NGE TITI.E DCP 1522, Rev. O, Seismic Supporis for Pressure Indicator (PI.MS.501, 502, 503)

Tubing CII ANGE DESCI IPTION Main steem local pressure indicators PI hiss 501, Pl MSS 502, and PI hiss 503 are removed periodically for calibration and maintenanre. When the instruments are removed, the disconnected tubing is unsupported. This conditior. results in the tubing becoming bent.

The bent tubing causes fit-up dimculties for instrument reinstallation, difliculties which are exacerbated by the 850 psi steam in the tubes. Ad'P "lly, these tubes tie into the main steam impulse tubing for the pressure protection channt 9. TN s.9 so.NI, bent configuration could lead to an impulse tube leak, which could cause a a ersu im W To properly support the tubing (3) new anchors (TSA's) a. be added - one near each pressure-indicator. This will alleviate tube bending and improve the fit up of the calibrated instruments.

This design change will reduce the rist sf personnel injury and will enhance plant reliabi!ity by lessening the chance of a spurious reactor trip.

SAFETY EVAL,UATION

SUMMARY

The accidents, " Accidental Depressurization of the Main Steam System," analyzed in Section 14.1.13 of the Updated Final safety Analysis Report (UFSAR) and " Major Ru,%re of a Main Steam Pipe" analyzed in Section 14.2.5.1 of the UFSAR were reviewed. This design change will reduce the probability of damaga to tubing while these instmments are being calibrated. This will reduce the probability of a spurious low steamline pressure signal which leads to a safety injection and thus a reactor trip. These accidents are bounding and are not afTected by this design change.

The Main Steam pressure inputs to the reactor protection system will not be adversely affected.

They will continue to operate as originally designed and will have no impact on any safety ftmetions provided by this instrumentation.

The probability of occurrence for the above accidents is not increased. This design change will' provide additional structural and seismic support for the afTected Main Steam instrument tubing which will reduce the probability of tubing and connector failure.

This design change will decrease the probability of damage to the Main Steam Line pressure -

instrument tubing by providing additional seismic support for the pressure transmitter tubing.

The consequences of a failure of the seismic supports to prevent damage to the tubing does not represent an increase from the existing condition. The consequences of tube failure is bounded by the analysis described in UFSAR Section 14.1.13.

Another failure considered is fracture of a tubing support. This will not occur because the

11eaver Volley Power Station Unit I l 1991 Report of Facility Changes, Tests, and Experiments Page 89 of I87 support will be seismically designed and installed with appropriate testing.

This design change will not create the possibility of a new accident because the possible accidents are bounded by those analyzed in UFSAR Sections 14.1.13 and 14.2.5.1.

No new malfunctions are created by this design change. Malfunction of the steam line pressure transmitters are bounded by the analyses of Sections 14.1.13 and 14.2.5.1 No changes in parameters affect the course of any accident analysis supporting Technical: j 5,,ecification bases, or. result in exceeding the acceptance criteria for fuel cladding, Reactor

Coolant System boundary, or containment integrity. No Technical Specification bases are affected J by this design change.

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Deaver Valley Power Station Unit 1 1991 Report of Facility Change 4 Tests, and Expcriments Page 90 of 187 l Cil ANGE TITI.E DCP 1526, Rev. O, Removal of Out Of Service Equipment Cil ANGE DESCRIPTION .

This design change will remove unwanted, unused equipment which was spared during installation of Design Change Package (DCP) 615 in the vertical board and benchboard Section C.

The following equipment will be removed:

1. Turbine Controlled Start Temperature Recorder

2. Turbine auto Start Monitor

3. Turbine Speed Indicator

4. Vibration Phase Angle Indicator

5. Eccentricity Phase AngleIndicators Also to be removed is the DC power supply associated with the turbine auto-start monitor. The thermocouple inputs to the turbine controlled start temperature recorders will also be disconnected and spared at the vertical board and in the field at thejunction boxes.

't Since the temperature recorder itself was previour,1y spared, disconnecting the thermocouples will be of no affect. The thermocouples will still provide a signal to the computer The Vertical Board and Benchboard (VBil) openings will be patched per the applicable procedure. It must be assured that the VBB remain in a seismic Class 1, Quality Assurance Category I condition.

Since the equipment removed by this design change was previously spared, its' removal is of no affect. .

Therefore, this design change will maintain the reliability, integrity, and operability of the VBB and will have no adverse effects on any other equipment.

SAFETY EVAI.UATION SIIMM ARY No design basis accidents will be affected because this design change does not adversely affect any safety or non safety systems, does not exacerbate any existing accidents, and does not introduce any new hazard beycnd that already considered in the Updated Final Safety Analysis Report (UFSAR).

This design change will not adversely afrect the safety function of any system. The reliability,.

integrity, and operability of the VBB will be maintained and no other systems will be affected.

. The probability of occurrence for any accident previously evaluated in the UFSAR will not be

. increased. ' This design change will maintain the reliability, integrity, and operability of the VBB,

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Be:ver Valley Power St: tion Unit i 1991 Report of Facility Ch:nges, Tests, and IIxperiments Page 91 of 187 and it will have no effect on any other equipment; therefore, the probability of occurrence for design basis accidents will not be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased because the reliability, integrity, and operability of the VilB is being maintained, and the change will have no effect on any other equipment. This design change will not affect any parameter which would increase the consequences of an accident beyond that previously considered in the UFSAR. This design change will not adversely affect any safety systems used to mitigate an accident.

The probability of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This design change will not adversely affect, either directly or indirectly, any equipment, including the VilB.

The consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR will not be increased. This design chang; will not adversely afTect any parameter which would increase the consequences of a malfunction. This design change will not adversely affect any safety system used to mitigate an accident. Therefore, there will be no elTect on the consequences of a malfunction of egaipment important to safety.

This design change 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 relatiw:ly minor change.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created because nothing is being added or altered in a way which creates the possibility of a difTerent type of accident. This design change is minor, and the reliability, integrity, and operability of the VBB will be maintained, and no other equipment will be efTected.

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

This design change will not change any parameter which afRets the course of any accident analysis supporting Technical Specification cases. The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this design change will not adversely alTect the margin of safety as defmed in the b7 for any Technical Specificatiore because the reliability, integrity, and operability of the VBB will be maintained, and no other equipment , vill be affected.

I

A Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 92 of 187 CII ANGE TITLE DCP 1527, Rev. O, Pressurizer Surge Line Rupture Restraint Modifications CllANGE DESCRIPTION The pressurizer surge line rupture restraint shim stack to pipe gaps were reset dur:ng the seventh refueling outage to qualify the restraints. These changes are described in Design Change Package (DCP) 1431. DCP 14L resulted from NRC Bulletin 8811 mandated inspections of the surge line which uncovered the deficiency with the gap settinas. Subsequently, an operational restriction of 200 F, system delta T, between the Reactor Coolant System hot leg and the pressurizer was initiated. This restriction minimized thermal stratification induced line displacements such that the line will not impact the restraints. Maintaining this limited system delta T is diflicult and takes operations adoItional time to heat up and cool down the plant.

The surge line has been instmment d for operational thermal stratification gradient and displacement alTects. The data obtained has been deciphered and an analysis completed to include the thermal stratification conditions. Additionally, leak before break analysis of the surge line has

' mssfully demonstrated. Application to the NRC for approval of the leak before break

,a required. Upon approval, the requirement of postulating breaks on the surge line and sding mpture restraints will be deleted. It is proposed that the shim stacks on the rupture restraints be removed. This will disable the rupture restraints function and allow the line to move freely under higher system delta T's. With this design change the 200 F delta T limitation may be removed.

The basis for this solution is in accordance with USNRC Standard Review Plan Section 3.6.3, Leak-Before-Break Evaluation Procedures, NUREG-0800.

The analysis conducted to incorporate stratification effects does not take into account rupture restraint gaps closing under higher system delta T's. Therefore, instrumentation must also remain in place and be used to monitor heat up and cool down until the shims are removed. The instrumentation was installed under DCP-1430.

SAFETY EVA1,11ATION SUMM ARY Updated Final Safety Analysis Report (UFSAR) Chapter 14 design basis accidents were reviewed.

A rupture of the pressurizer surge line has been pwviously analyzed and bounded in Section 14.3 Loss of Coolant Accident. Additionally, a leak before break analysis of the surge line has been 1 evaluated in accordance with USNRC Standard Review Plan 3.6.3, Leak-Before-Break evaluation Procedures (

Reference:

WCAP-12727).

No safety systems will be adversely affected by the proposed design change. The functions of all safety related systems and components, including that of the Reactor Coolant System, will remain unchanged.

ller.ver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments

- Page 93 of 187

/Ji corrective actions performed by this modification will follow all applicable industry codes and standards and will be designed, installed, and tested as a Quality Assurance Category I, Seismic i Class 1 modification. The modification will not decrease the reliability of the pressurizer surge ,

line.  !

i in accordance with 10 CFR 50 Appendix A, GDC-4, calculations based upon fracture mechanics technology were performed as discussed in WCAP.12727. The calculations demonstrated that ,

the pressurizer surge line will leak in a predictable and detectable manner. The margin-requirements for allowable stresses in fracture mechanics analysis were met by the calculations of WCAP-12727. The functions of all systems required to help mitigate any of the previously  ;

analped accidents will emain unchanged. l

-i No safety related equipment functions will be afTected by this modification. The surge line itself .

will remain unchanged, it will not create a situation which would increase the probability of malfunction. [

The consequences of a malfunction of equipment important to safety as previously evaluated in <

the UFSAR will not be increased. Ruoture of a surge line restraint, whether it be aller the design change or 'oefore the design change, will result in the same consequences.

No new credible failure modes will be created by the implementation of this modification. There .

will be no ftmetional changes to any safety related equipment, systems or structures. The rupture j

of the pressurizer surge line has been previously analyzed as a design basis accident bounded by_

the Loss of Coolant Accident.

The function of the pressurizer remains the same. There is no configuration change such that an -

accident of a different type is created.

No new failure modes or potential hazards will be created by the implementation of this-modification. The only failures / malfunctions that will exist are those that presently exist. ,

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

Reactor Coolant System boundary, or containment integrity.

The proposed design change does not reduce the margin of safety as defined in the basis for any .

' Technical Specification. No Technical Specification basis will be afrected by this modificationc 1

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Beaver Volley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 94 of I87 Cll ** N@ TITI& 3 DCP 1530, Rev. O, Insallation of Diesel Engine Mounted Day Tank Drain Valves CII A NGE DESCRIPTIOb' The Emergency Diesel Generators (EDG's) engine mounted fuel oil day tanks, EE. TK10A, D currently have an cibow and drain plug at the low point drain connection. The objective of this Design Change Package (DCP) is to replace the elbow and drain plug with a drain valve and piping to facilitate water removal and sampling. The new piping and drain valve will meet applicable codes, be seismic class 1, and are considered to be of essentially equivalent reliability and integrity for the purpose of sealing the low point drain. Seismic supports may be added for the new piping and valve au necessary. Therefore, this DCP will maintain the reliability, integrity, and operability of the EDG Fuel Oil System (EDGFOS) and will have no adverse effects on any other equipment.

SAFETY EVAL,11ATION S11MM ARY No design basis accidents will be affected because this design change 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 Updated Final Safety Analysis Report (UFSAR).

This design change will not adversely affect the safety ftmetion of any system. The reliability, integrity, and operability of the EDGFOS will be maintained and no other systems will be affected.

The probability of an occurrence of any accident previously evaluated in the UFSAR will not be mereased. This design change will maintain the reliability, integrity, and operability of the EDGFOS, and it will have no effect on any other equipment; therefore, the probability of occurrence for any design basis accidents will riot be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased because the reliability, integrity, and operability of the EDGFOS is being maintained, and the -

change will have no efTect on any other equipment. This design change will not affect any parameter which would increase the consequences of an accident beyont that previously considered in the UFSAR. This design change will not adversely afTect any safety systems used to mitigate an accident.

The probability'of a malfunction of equipment i,aportant to safety as previously evaluated in the UFSAR will not be increased. This design change will not adversely afrect, either directly or-indirectly, any equipment, including fac EDGFOS.

The consequences of a malfunction of equipment important to safety as previously avaluated in the UFSAR will not be increased. This design change will not adversely affect any parameter which would increase the consequer,ces of a malfunction. This design change will not adversely . o

lleover VcJley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments i Page 95 ofI87 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 design change 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.

I The possibility for an accidem of a different type than previously evaluated in the UFSAR will not be created because nothing is being added or altered in a way which creates the possibility of a difTerent type of accident. This design change is minor, and the reliability, integrity, and operability of the EDGFOS will be maintained, and no other equipment will be effected.

The possibility for a_ malfunction of a different type than any previously evaluated in the UFSAR will not be created because 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 design change is minor and the reliability, integrity, and operability of the EDGFOS will be maintained.

1 This design change will not change any parameter which affects the course of any accident analysis supporting Technical Specification bases. The Technical Specifiestion index was reviewed to determine if any bases might be affected. It was determined that this design change .

will not adversely affect the margin of safety as defined in the basca for any Technical Specifications because the reliability, integrity, and operability of the EDGFOS will be maintained, and no other equipment will be affected.

Ileaver Valley Power Station Unit 1 i 1991 Report of racility Chtnges, Tests, and Experiments 1 Page 96 of187 Cil A NGE TITI.E DCP 1532, Rev. O, Permanent Vibration Proximity Probes on Safety injection Pumps SI F 1 A,111 CII ANGE DESCRIPTION Vibration monitoring of the Low licad Safety injection (LilSI) pumps is performed utilizing a  ;

portable vibration instmment. Data procured by this method has been inconsistent, panly because it is difficult to revisit the exact location of the previous readings and panly because ofinherent i difliculties with a portable instrument. Test procedures were revised to provide additional detail for the accurate positioning of the vibration shaft stick, but this did not eliminate the inconsistent readings.

The NRC resident inspector identified this concern via Unit 1 Unresolved item 88-08 02, which was subsequently updated in InspecCon Report 89-04. The concern way that the inconsistent readings could mask real pump degradation.

This design change permanently installed two proximity probes per pump. The probes will minimize or climinate operator and V stick variab!cs and associated data retrieval inconsistencies.

SAFETY EVAI,IIATION

SUMMARY

No Updated Final Safety Analysis Report (UFSAR) Chapter 14 accic:ents will be alTected by this -

design change because this change does not adversely affect any safety or non safety systems, does not exacerbate any existing accidents, and does not introduce any new hazard. ,

This design change will net adversely affect the safety function of any system. A failure of the proximity probes will not prevent the safety injection pumps from performing their safety function.

The probability of occurrence for accidents previously evaluated in the UFSAR will not be increased. This design change will slightly improve the reliability of the LilS1 pumps by providing more accurate vibration data; however, it will have no effect on any other equipment. Therefore, the probability of occurrence of any accidents will not be increased.

The consequences of an accident previously evaluated in the UFSAR will not be increased. This design change is minor and the change will have no effect on any other equipment. This design change will not afTect any parameter which would increase the consequences of an accident. This design change will not adversely alTect any safety system used to mitigate the consequences of an

-accident. This design change will slightly improve the reliability of the LilSI pump. ,

The probability of a malftmetion of equipment important to safety as previously evaluated in the UFSAR will not be increased.- This design change is minor and the changes will not adversely ,

affect any equipment. Therefore, the probability of any malfunction of equipment important to safety will not be increasedi

Beaver Valley Power Station Unit 1 1991 Report of Fecility Changes, Tests, and Experiments Page 97 of187

'The consequences of a malfunction of equipment important to safety as previously evaluated in -

the UFSAR will not be increased. This design change will not adversely affect any parameter which would increase the consequences of a malftmetion. This design change will not adversely

. affect any safety system used to mitigate an accident. Therefore, there should be no effect on the consequences of a malfunction of equipment important to safety.

This design change will not cause any new credible failure modes because the fbndamental design features and functions of the LilSI pumps have not been altered. The permanently installed proximity probes will be seismically installed to prevent potential damage by the probes to other equipment in the area during a seismic event.

The possibility for an accident of a different type than previously evaluated in the UFSAR will not be created. This design change is minor, and the existing portable instrument is merely being replaced by another permanently installed instrument; therefore, the change is' not significant enough to create the possibility for an accident of a difTerent type than analyzed in the UFSAR.

The possibility for a malfunction of a different type than any previously evaluated in the UFSAR-will not be created. Because this design change will install the permanent instrument above the pump seals, does not affect system pressure boundaries, or adversely affect the L1-ISI system, the -

possibility for a malfunction of a different type is not created.

This design change will not change _ any parameter which afTects the course of any acekjent analysis supporting Technical Specification bases. The Technical Specification index was reviewed to determine if any bases might be affected. It was determined that this design change will' not adversely 'afrect the margin of safety as defined in -the bases for any Technical Specification because the reliability of the LIISI pumps will be maintained,- and no other equipment will be affected.

" .4

lle:vcr Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and IIxperiments i Page 98 of 187 Cil ANGE TITI.E DCP 1543, Rev. O, Removal of the Unit 1 Auxiliary lloilers Cil ANGE DESCRIPTION Proper hanc' ling and storage requirements of various hazardous materials has forced the company to upgrade its current chemical handling methods. The methods to provide the upgrade are provided in Design Change Packages (DCPs) 1419 and 1269. Various locations can be used to locate the new equipment associated with these two design changes, however, the change; will be extremely expensive if they are located in areas that are not currently equipped to handle hazardous materials. The auxiliary boiler room is currently equipped to a:fdess these concerns and guidelines. In fact, parts of the existing Auxiliary lloller Chemical Addition System could be used for boric acid additions to the Condensate System with a minimal amount of change. i This modification proposes to remove both Unit I auxiliary boilers in order to facilitate these design changes. The auxiliary boilers (QA Category 3) are not required at Unit 1; auxiliary steam can be supplied from either the Unit I hiain Steam System (hiss) or from Unit 2 via the Unit 2 hiss or auxiliary boilers.

The boundaries of change for this modification include scrapping both auxiliary boilers and associated equipment within the boiler room envelope, Piping will be terminated and capped just outside the room envelope if possible. !!!ectrical supplies will be terminated in the hiotor Control Centers (htCCs) and cables from the hiCCs to the retired components will be removed. Auxiliary boiler equipment located outside of the boiler room will be electrically disconne:ted, but will be retired in place. Conduit inside the room will be scrapped, but conduit outside the room will remain for future use as required. Related control room alarms and displays will be removed or relabeled as required. To allow access by a fork lift truck, the boiler concrete pads will be removed, and the floor will be resurfaced as required.

SAFETY EVALUATION SUMM ARY None of the previously anelyzed accidents for either Unit 1 or 2 will be affected by this modification. This change will not afTect any accident initiating events or any acciMt mitigation functions including Auxiliary Steam ' stem Isolation as discussed below.

No safety systems will be alTected by the proposed design change. No services supplied by the Auxiliary Steam System perform a safety related function except for valves [IlYV.AS-101 A &

13). In the event of a high energy line break outside containment, these trip valves isolate auxiliary steam to the safeguards area and the auxiliary building. This safety function will be unafrected by :

this modification.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report. The auxiliary boilers and the entire Auxiliary Steam System are not associated with any accident initiating event; therefore, the probability of an accident will not be increased.

Beaver Valley Power Station Unit 1 1991 Repon of Facility Changes, Tests, and Experiments Page 99 ofI87 The proposed design change will not increase the consequences of an accident previously evaluated in the safety analysis repod. None of the components, systems, or structures required to help midgate a previously analyzed accident will be adversely affected by this modification; therefore, the consequences of such an accident will not be increased.

The proposed design change will not increase the probability of occurrence of a malfunction of equipment impodant to safety as previously evaluated in the safety analysis report. No safety related components will be affected by this modification The proposed design change will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the safety analysis repon. No safety related fimetions will be affected by this modification.

No new failure modes will be introduced by this modification. This change involves the retirement of non safety related, redundant components and does not install new ones.

Termination of electrical and mechanical componer.ts (pipelines) will meet acceptable standards and will be tested as required to help ensure their adequacy.

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 system functions will be changed to the extent that a different type of accident could be created. This mo:lification will retire one of four redundant sources of auxiliary steam.

The proposed design change will not create the possibility of a malfunction of a difkrent type than previously evaluated in the safety analysis report, No new failure modes or potential hazards will be created by the implementation of this modificcior..

There are ne changes in parameters which affect the course of any accident analysis supponing Technical Specification bases, or that result in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment integrity. The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification. No.

Technical Specifications or its bases will be affected by the implementation of this modification.-

11 caver Valley Power Station Unit i 1991 Report of Focility Changes, Tests, and Experiments Page 100 of 187 CII ANGI: TITI.E DCP 1546, Rev. O, Replacement of Station Air Compressors Cll ANGl; DESCRIPTION The objective of Design Change Package 1546 is to improve the reliability of the Station Air System. llistorically, system operability problems have been associated with air quality (such as particulate, hydrocarbons, or high dewpoint) and excessive maintenance. The existing reciprocating compressors have been a maintenance problem because of their age, design vintage and over use. Ilowever, engineering has been monitoring the system with particular attention given to the overell system demand and it has been determined that the existing system configuration does not provide adequate capacity to support typical system loads. To resolve the system reliability concerns of air quality and excess maintenance and to correct an identified '

design deliciency, the proposed modifications are as follows:

1. Replace the three existing 350 SCFM reciprocating wmpressors (SA- C-1 A, til, & IC) with two 570-SCFM, oil free, rotary screw compressors. This will require replacing the 1 1/2 inch diameter tmbine plant component cooling water line to the compressors with a  ;

2 inch diameter line. Ilowever, the 12511P motors provided with the proposed compressors will not require major changes to the local power source.

2. Install a dew point meter downstream of the instrument air re. iver tank (IA-TK-1),

3. Install an oil removal coalescing filter downstream of the diesel driven compressor cunently sitting at the east end of the main operating level of the tmbine building. This will provide assurance of proper air quality to support station instrumentation if this diesel driven, oil flooded compressor is placed into service. .

4. Replace the existing heat regenerative dryer (IA D 1) with a flow regenerative type similar to dryer IA D 2. This will eliminate the reliance upon filters to' adequately dry the instrument air when the air dryer is out of service. The existing dryer (IA D-1) has similar problems as those of the compressors; age, parts, and over use.

5. Install a particulate filter downstream of the station instiument air receiver (IA TK 1).

This will remove any scale or rust due to degradation of the carbon steel receiver tank and keep such contaminants out of the instmment air system.

6. Install two flow meters, one downstream of the station air receiver tanks and one downstream of the instrument air receiver tank. This will allow operations to continuously-monitor the system demand.

t llcaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 101 of 187 SAFETY EVAL.lfATION SIIMM ale Updated Final Safety Analysis Report (UFSAR) Chapter 14 design basis accidents were reviewed to identify desip basis accidents that could be affected by the proposed modification. The station air system is non safety related and not required for safe shutdown of the plant. None of the Chapter 14 safety analyses will be affected by this modification.

No part of any safety related equipment requires the supply of compressed air for shutdown. All air operated valves fail closed. Valves that must respond during shutdown or aller a shutdown to maintain that state are either hand or motor operated.

The design change will not increue the probability of occurrence of a loss of station air or any other accident previously evaluated Sullicient air capacity will be provi6ed for station air systems. The Turbine Plant Component Cooling Water System (CCT) can adequately accon modate the Ir.rger cooling line required. Neither the station air nor the CCT system is required for a safe plant shutdown.

No equipment or systems required to mitigate the consequence of any previously analyzed accident will be affected by this modification. Tte modification serves only to enhance the reliability of the station air system by replacing the existing compressors, System air quality will be improved and excess compressor ntintenance problems will be resolved.

Replacing the existing heat regenerative dryer with a flow regenerative type dryer will eliminate the reliance upon filters to adequately dry the instrument air when the air dryer is out of service.

Air dryer rupture or malfunction has been reviewed. It does not create a situation which would increase the probability of a malfunction llecause of their location the new rotary screw compressors are not a credible source of missiles that could impact safety related equipment.

The compressed air system is not required to perform any safety related system functions. There should be no effect on the consequences of a malfunction of equipment important to safety as previously evaluated in UFSAR.

Compressor malfunction, air dryer ruptures or malfunction and filter clogs were reviewed. No new credible failure modes will be created by the implementation of this modification. There will be no functional changes to any safety related equipment, system or structure.

There is no configuration change such that an accident of a different type is created. The present configuration includes three 100 percent capacity compressors. Only two of these compressors are needed to meet the original system design. The third compressor was added by DCP 140 because of a high amount of maintenance that had been required on these reciprocating compressors. The new rotary screw compressors will be of greater capacity but are expected to require less maintenance. This design change h replacing existing equipment with similar equipment and will not create a different type of accident.

No new failure modes or potential hazards will be created by the implementation of this

i Beaver Vallcy Power Station Unit 1 i 1991 Report of Facility Changes, Tests, and Experiments  !

Page 102 of187 j i

modification. The failure of this equipment has no effect on safe plant operation and shutdown of [

the plant, and the failure of this equipment will not alTect the safety function of other equipment.

There are no changes in parameters which affect the course of any accident analysis supporting i Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding,  !

Reactor Coolant System boundary, or containment integrity. l The proposed design change does not reduce the margin of safety as dermed in the basis for any a t

Technical Specification. No Technical Specification basis will be affected in any way by this.

modification.

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llcaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 103 of 187 Cil ANGE TITI.E DCP 1557, Rev. O, Suction Pressure Gauge ft, uxiliary Feedwater Pump (PI-FW-156,156A,15611)

CII ANGC DESCRIPTION Design Change l'ackage 1557 is being implemented to provide pressure indication for the auxiliary feedwater pump suction that complies with ASME XI Program IWP-4000, and falls within accuracy (range) criteria specified therein.

Auxiliary feedwater pump suction reference pressure during testing is approximately 10 psig.

The installed gauges do not meet the range requirements of ASME XI paragraph IWP-4000. The installed gauges are 160 psig gauges, and are being replaced with 60 psig gauges.

SA FETY EVAI IIATION SilMM ARY The pressure indicated by the auxiliary feedwater pump pressure indicators (PI FN-156,156A, 15611) during testing is 10 psig. The water source for testing is the primary plant demineralized water storage tank (WT-TK 10). In emergency situations the pumps and associated gauges are supplied water from the river water system. The maximum pressure expected in this situation is 55 psig.

There are no credible failure modes associated with this change. Replacement gauges will be bought under the same specification as the current gauges.

This change only affects the indication of auxiliary feedwater pump suction pressure. It will have no effect on performance of the auxiliary feedwater system. This design change will not afrect any design l' asis accidents as previously evaluated in the Updated Final Safety Analysis Report (UFSAR).

There will be no affect on the auxiliary feedwater system due to the change of 160 psig pressure gauges to 60 psig gauges. These gauges (Pl.FW-156,156A,156fl) provide indication only.

There are no design basis accidents for which failure modes of the new gauges can be an initiating event.

This design change does not increase the probability of a design basis accident as previously evaluated in the UFSAR. This change involves only a change of gauges on the auxiliary feedwater nuction lines. The gauges will provide increased accuracy. The gauges to be installed will be eurchased under the same purchase specification (IlVS 262) as those installed except that the sange specified will be decreased to 60 psig.

This design change only provides increased accuracy for the auxiliary feedwater suction line pressure indication Technical Specifications are not affected. This design chang. does not change the acceptance limits which form the basis for the Technical Specifications.

W

lleaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments j Page 104 of 187 i Cil ANGE TITI.E l DCP 1578, Rev. O, Fuel Transfer Tube Illind Flange Modification  !

l Cil ANGE DESCRIPTION During the seventh refueling outage, the fuel transfer tube blind flange was installed such that the gaskets were mispositioned. Following the installation, the blind Dange was successfully Type 11 tested The test was successful because the mispositioned gaskets effectively covered the test port. Therefore, the gasket interspace was not pressure tested.

l This rnodification proposes a solution to ensure that the gasket interspace is successfully pressurized during future Type 11 testing. A 1/4" diameter swagelok fitting will be installed at l approximately 90 F from the test port such that the fitting leads to the gasket interspace. During Type 11 testing, a pressure gauge will be attached to the fitting to ensure that the interspace is, in fact, pressurized. A swagelok plug must be installed in the fitting at any time that the plant is in any mode other than 5 or 6.

SAFETY EVAI UATION SilMMARY The fuel transfer tube blir.J flange normally provides inside containment isolation for penetration number 65. The containment structure and the Containment isolation System are designed based on the shielding requirements and the pressure and temperatures that would be generated by the design basis accident. This modification will not adversely affect the ability of the blind flange to perform its intended ftmetion.

The safety system that will be affected by the proposed design change is the Containment isolation System. Ilowever, no adverse affect to any safety related componer t or system will be created.

The proposed design change will not increase the probability of occurrence of an accident previously evaluated in the safety analysis report. The fuel transfer tube blind flange is_ not associated with any of the potentialinitiating factors for any of the previously analyzed accide as; including, those discussed in Updated Final Safety Analysis Report (UFSAR) Sections 14.2.1,

" Fuel llanc !ing Accident" and 14.3, " Loss of Coolant Accident." (The blind flange can not be 8

associated with a fuel handling accident; ;he flange is removed from the fuel transfer tube prior to-fud movements in the cantainment building).

The proposed- design change will not increase the consequences of an accidert previously evaluated in the safety analysis report. This modification will not adversely alTect the ability of the fuel transfer tube blind flange to perform its intended function. This modification will actuahy enhance the reliability of the blind flange in performing its isolation function since it will help to ensure that the sealing gaskets are adequately tested The proposed design change will not increase the probbility of occurrence of a malfimetion of

I.

lleaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and !!xperiments Page 105 of 187 equipment important to safety as previously evaluated in the safety analysis report. No safety related components or systems will be adversely afrected by this modification.

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

f There are no credible failure modes associated with this modification. During modes of operation when the blind flange must be in place (modes I through 4), the new fitting will be isolated with a plug. The fitting and associated plug will meet applicable QA Category I and seismic requirements. Since the fitting will extend to the interspace between the two gaskets, the containment atmosphere adjacent to the fitting will remain isolated by:

!. The fitting and plug assembly, and

2. The gasket nearer the fuel transfer tube (This arrangement is simiur to the existing flange test connection).

The proposed design change will not create the possibility of an accidere of a different type than previously evaluated in the safety analysis report. Components associa.2 ' with this modification provide a passive prote::tive function, and by conforming to Quality Assurance Category I and seismic requirements, the possibility of creating an unanalyzed accident is precluded.

The proposed design change will not create the possibility of a malfunction of a d;!Terent type than previously evaluated in the safety analysis report. This modification will have no adverse afrects on any components or systems, therefore, no new failure modes or potential hazards will be created.

There are no changes in parameters which affect the course of any accident analysis supporting Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, Reactor Coolant System boundary, or containment integrity. The proposed design change does -

not reduce the margin of safety as dermed in the basis for any Technical Specification. No Technical Specification or its bases will be affected by this modification. This modification will

- help ensure compliance with lechnical Specifications 3/4.6.1.1, " Containment Integrity,"

3/4.6.1.2," Containment f.eakage," and 3/4.6.3.1, " Containment Isolation Vrives"

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Ecaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 106 of 187 Cil A NGE TITI.C DCP 1598, Rev. O, CO 8 to CO ll Relay Replacement on IE12 lF12 Air Circuit Dreaker Cubicles Cll ANGE I)ESCRIPTION A coordination problem exists with the existing CO 8 relays on Air Circuit Hrcaker (ACD) cubicles 1E12 and IFl2 rnd the refurbished General Elcetric(GE) AK air circuit breakers on large -

(300 llP) 4.!0V motors on the N and P emergency bus. The GE RMS 9 replacement sensors for the 480V bus ',reakers do not duplicate the instantaneous trip taps of the original power sensors they replaced. This creates the coordination problem in the case of the large 480V motors air circuit breakets and the high side 480V bus transformer supply breaker's CO 8 relays This design t change will replace the 12 existing CO 8 relays with CO 11 type relays to get the necessary coordination on emergency bus large motors 8N3, 8N19, 8N20, 9P4, 9Pl7, and 9P18. The 4160/480V transformer overcurrent CO 8 relays to be replaced are SI.VEll2, SI-VElll2, SIVFil2, and SI-VFill2 on all three phases. The Cb-ll overcurrent relays with extremely inverse time range are needed to maintain emergency bus (N and P) coordination with the replacement RMS 9 sensor characteristics.

S Al'ETV EVAL,tfATION StIMM ARV Updated Final Safety Analysis Report (UFSAR) Section 14.1.11, " Loss of Offsite Power " was reviewed to identify the affects of the proposed modification on this design basis accident. This modification has no impact on the loss of otTsite power event. None of the other Chapter 14 Safety Analysis will be affected by this modification, 4160/480V Station Sewice Systems will be afTectM by this design change. This design change will solve the coordination problem that exists with the existing CO-8 relays on ACB cubicles and refurbished GE AK air circuit breakers for large 480V motors on the N and P emergency bus.

This CO-8 relay replacement will not change any of previously designed air circuit breaker protective functions nc . it defeat, remove or change any of the 4160/480V emergency bus design requirements for uedicated Engineered Safety Features (ESP) equipment or any other safety related systems. The functions of all safety related equipment and systems will be unaffected by this modification. ,

The consequences of an accident after the replacement will be unchanged because it will not change any of the 4160V or 480V emergency bus design requirements for dedicated ESF .

equipment. It will not defeat, remove or otherwise change out any companion actuation device within the IE12 or IFl2 cubicle.

This modification does not create a situation which- would increase the probability of a malfunction. A malfunction of these relays, whether it be the present relays or the replacement

- relays, will resuh in the same consequences.

i Beaver Valley Power Station Unit I  !

1991 Report of Facility Changes, Tests, and Experiments  !

Page 107 of187  ;

No new credible failure modes will be createJ by the implementation of this modification; There - I will be no functional changes to any safety related systems or structures. -l l

Since the function of the relays remains unchanged, no new possibilities of an accident are created. The only failures or malfunctions that will exist are those that presently exist.  ;

There are no changes in parameters which affect the course of any accident analysis supporting [

Technical Specification b=.ses, or result in exceeding the acceptance criteria for fuel cladding, j Reactor Coolant System boundaiy, or containment integrity. The proposed design change does - i t

not reduce the margin of safety as.dermed in the basis for any Technical Speci0 cation. The Technical Specification basis for Se;ction 3/4.8.2 will not be affected in this modification.

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I Ileaver Valley Power Station Unit i 1991 Report of Freility Changes, Tests, and Experiments Page 108 of 187 Cil A ngl: TITI.E DCP 1622, Rev. O, Control Room ilVAC Instrument Air Filter Change Cll A NGE IWSCRIPTION Lhe air quality of the control room lleating Ventilation and Air Conditioning (llVAC)

Compressed Air System is monitored on a weekly basis per test tilVT 11.34.3. The results of this surveillance test program have identified a design deficiency with the configuration of the air system. The filter downstream of each of the two air dryer trains, is inappropriately placed within the system. The dryer discharge line has a branch upstream of the filter (the same holds true for both air system trains). The objective of this design change is to relocate the dryer after filter for each train to a posiCon upstream of the system branch. This will provide filtration for all of the compressed air discharged from either Train A or B.

The basis for this filter relocation is to improve the air quality to a level consistent with the requirements of instrument Society of America (ISA) Standard S7.3 *Cuality Standard for Instrument Air" Beaver Valley has committed to such a level of air quality in our response to NRC Generic Letter 88-14, " Instrument Air Supply System Problems Affecting Safety Related Equipment."

SAFETY EVAL.11ATION SilM M Ally Upd ted Final Safety Analysis Report (UFSAR) Chapter 14 design basis accidents were reviewed to identify design basis accidents that could be affected by the proposed modification. UFSAR Section 14.3.5.2 and Section 9.13.4 were reviewed; this design change will pmvide better quality air to the damper and w;!! not afTect the Containment Isolation Phase B (CID) signal to close the contial room area air-conditioning system outde,or air intake. No Chapter 14 Safety Analyses will be atTected by this modification.

No safety systems will be aversely afTected by the proposed design change. The function of all safety-related systems including Air Ventilation Systems in the controi room, and related components (including filters and dampers) will not be changed.

The design change will not increase the probability of occurrence for acci6ents considered in the control room dose analysis. It will improve the air quality to a level consistent with the requirements ofISA S7.3 " Quality Gtandard for Instrument Air "

Equipment or systems required to mitigate the consequences of previousiy analyzed accidents will not be affected by this modification. The air quality problem will be improved and resolved.

No safety-related equipment function will be afTected by this modification. The air pressure on the damper and pressure relief valve will not be reduced. It will not create a situation which would increase the probability of malfunction.

The consequences of a malfunction of equipment important to safety as previously evaluated in

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Ileaver Va' ley Power Station Unit 1 1991 Report of Facility Changes, Tests, and thperiments - 1 Page 109 of187 l the UFSAP. will not be increased. The malfunction of a damper, whether it be aller the design I change or before the design change, will result in the same conseqi.ences.

Compressor malfunction, damper malfunction and filter clogs were ": viewed. No new credible I failure modes will be created by the implementation of this modification. -There will be no J functional changes to any safety related equipment, systems or structure.

There is no configuration change such that an accident of a different type is created. The.

compressed air package was seiunically qualified at the time of purchase. It will not create a different type of accident.

No new failure modes or potential hazards will be created by the implementation of this -

modification. The failure of this equipment will not affect the safety function of other equipment.

There are no changes in parameters which affect the course cf any accident analysis supporting Technical Specification bases, or that result in exceeding the acceptance criteria for fuel clac "ng, Reactor Coolant System boundary, or containment integrity. The proposed design change does not reduce the margin of safety as defined in the basis for any Technical Specification. No Technical Specification or its bases will be afTected by this modification,

l Beaver Valley Power Station Unit 1 l 1991 Report of Facility Changes, Tests, and Experiments Page 110 of 187 CII ANGE TITIE DCP 1646, Rev. O, Instrllation of Fire Protective Wrap on Conduit IFC439002 CilANGE DESCRIPTION Floor sleeve IFC439002 is a five inch diameter conduit which breaches the floor of the Unit I cable mezzanine, an Appendix R fire barrier. The qualified fire rating for the Unit I cable  !

mezzanine floor is 1 1/2 hours. It is the objective of this design change to bring conJuit IFC439002 into conformance with the NRC Generic Letter 8610. l In response to NRC Gener!c Letter 86-10, " Implementation Of Fire Protection Requirements," a formal program has been created at Beaver Valley Power Station to evaluate, install, and monitor ,

qualified fire stops and fire seals for all conduits, raceway, instrumentation tubing, and piping which breaches any Appendix R fire barrier.

As part of this program all conduits larger than four inches in diameter that breach an Appendix R fire barrier must either (1) be sealed at the fire barrier penetration, or (2) the conduit must be cncased in fue protective wrap of the same fire rating as that of the Appendix R barrier. This >

encasement must extend from the Appendix R barrier to the point where the conduit is sealed with a qualified fire stop.

SAFETY EVAISATION SUMM ARY L Updated Final Safety Analysis Report (UFSAR) Chapter 14 Design Basis Accidents were reviewed to identify design basis accidents that could be affected by the proposed modification.

No Chapter 14 safety analyses will be affected by this modification. .;

UFSAR Section 9.10 was also reviewed. This design change installs a proper, qualified fire protective wrap for conduit floor slec"e IFC439002. Seismic qualification will be evaluated and ,

a calculetion will be done to ensure thut it will not affect the power carrying capability'of safety related cables, Therefore, any single fire wit not cause an unacceptable riso to public health and safety, will not prevent the performance of a necessary safe shutdown functioit, and will not significantly increase the risk of radioactive release to the environment.

This design change will not adversely affect ' #ctr ftmetion of any system. It will not affect the capability of the fire protection sys: m mitigate fires that could affect safety related equipn.ent. A calculation will be performed to ensure the change will not affect the power.

carrying capability of e safety related cable.

L This design change will install a proper, qualified fire protective wrap for conduit floor sleeve IFC439002 and will create a more reliable, seismically qualified, fire barrier. Therefore, the ,

probability of occurrence for any accident evaluated in the safety analysis report will not be  ;

increased.

No equipment or systena required to mitigate the consequences of any previously analyzed -

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Ileaver Valley Power Station Unit 1 1991 Report of Facility Charges, Tests, and Experiments Pege 111 of 18'/

nccident will be affected by this modification. The design change will providt more reliable fire protection.

The probz ility of malfunction af equipment impoitant to safety as previously evaluated in the safety andysis report will not be increased because a calculation will be performed to ensure that the change will not affect the power carrying capability of safety related cables. Seismic qualification of the conduit will also be evaluated.

The consequ mces of a malfuncton of equipment important to safety as previously evaluated in the UFSAR will not be increased. Malfunction of the conduit before or afler the design ihange will result in the same consequences.

No new credible ailure modes will be created by the implementation of this modification. There will be no functional changes to any safety rclated equipment, system, or structure.

This design change does not alter the wiring, equipment, or configuration of any safety related

.. circuitry, piping, or instrumentation tubing. Conduit IFC439002 contains 20 safety (clated ,

i cables. To ensure that these cables are not adversely afTected by changing the heat transfer characteristws of the conduit, a formal analysis will be 1,crformed. >

No new failure modes or potential hazards wih be created by the implementation of this modification. The only failure / malfunctions that will exist are those that presently exist.

There are no changes in parameters which affe v 'ourse of any accident analysis supporting >

Technical Specification bases, or result in exceeu...g  : acceptance criteria for fuel cladding Reactor Coolant System boundary, or containment integrity. The proposed design change d not reduce the nwgin of safety as defined in the basis for any Technical Specification. No Technical Specification basis will be affected in this modification.

Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments-Page 112 of 187 CII ANGE TITIE DCP 1684 Rev. O Feedwater Modifications Cil ANGE DESCRIPTION The proposed modification will repair or replace if required, three existing monoball res i s with conventional box frame type restraints; will secure rupture restraint shim stacks as requ.::s; and will install additional teraporary instrumentation to confirm global thermal stratification profiles.

lined on analysis of feedwater line deflections, two of the three monoball restraints appear to be restricting free thermal piping movement. During the outage a physical inspection will confirm whether the monoball restraints are bound and if corrective action is warranted. Continued operation assuming locked monoballs with thermal stratification efTects has been documented using ASME UI Appendix F to quahfy pipe support acceptability. This was done on an interim basis. RepScement of these two monoballs will eliminate the interim usage of the Appendix F criteria for support qualification, and will be done in accordance with existing UFS.3 design basis. The third monoball is of similar design and will be repaired or replaced as required.

Several repture restraint shim stacks which were afTected by piping deflections under stratified condi<. ions will be secured as required, and additional instrumentation to identify piping movements to confirm the global the .I stratification profiles will be installed by this proposed change.

SAFETY EVALUATION SUMM ARY The proposed change merely replaces three piping supports with a better design to ensure the support performs its intended function.

The affected section of feedwater piping is inside containment between the steam generators and the containment isolation valve; this ponion of the piping is safety related. The proposed modification will not adversely affect the piping or the safety function.

The monoball support acceptability (with " postulated" locked monoballs and global thermal stratification)is based on an interim qualification using ASME Ill Appendix F. This modification will restore the piping such that interim measures are not required. This should reduce the probability of an accident.

This change improves Feedwater System supports and has no adverse affect on the system or its safety function. Since the Feedwater System and its pressure boundary are not affected by this change, the consequences of a malfunction are not increased.

The modification deals with the Feedwater System piping supports. Credible failure modes include seismic event, loss of fcedwater (14.1.8), major ruptures (14.2.5.2), and other mnifunctions (14.1.9). The ir . supports will be seismically qualified The other types of failures 1

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1991 Report ofFacility Changes, Tests, and Experiments -

- Page 113 of 187 will be unalrected by this modification.

This modification will restore the restraint function to its original design and will not adversely

- affect the Feedwater System.- ,

The affected Feedwater System supports will be seismically designed and will provide foi thermal movements as intended in the original design. A different type malfunction does not exist; ,

This modification does not change any Technical Specification Parameters. Tecimical -

Specification 3.6.3.1 was reviewed. This change does not affect any Technical Specification.-

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Beaver Valley Power Station Unit 1 1991' Report of Facility Changes, Tests, and Experimentsi Page 114 of 187 CII A NGE TITI,E DCP 1702, Rev. O, Permanent Installation of Output Metering Test Circuitry Cil ANGE DESCRIPTION This design change installs test switches to allow connection of a Wattmeter for accurate assessment of generator output. This is a non-safety related piece of equipment. .

SAFETY EVAI UATION

SUMMARY

The test switches are located in the main generntor Potential Transformer and Current Transformer (PT&CT) circuits of the Main Ger.erato: and Transformer System.

The failure modes are dependent on a failure of the test switches. The failure of test swiwh is deemed a low probab:lity event as the proposed test switches are the same switch that is currently installed in the generator monitoring circuits.

The Main Generator & Transformer System is not safety related. _ It affects generator output.

Addition of the test switches does not alrect operation of the generator as the switches are only used in a test mode to monitor output of the generator. There will be no change to the performance of the generator. The test switches only monitor the performance of the generator.

Updated Final Safety Analysis Report (UFSAR) Chapter 14 was reviewed and no design basis accident analyses are affected by the proposed change. Installation of the test switches will not have any radiological consequences as the switches only monitor electrical output of the gneratur.

The addition of the. test switches cannot initiate any Design Basis Event (DDE). The design change provides test equipment for a secondary output system which is assumed to trip off during -

any DBE. The test switch i3staration in the generator output circuit does not interact with any safety-related equipment to cause a new event. No equipment important to safety is affected.

The generator test switch hddition has no Technical Specification impact.

Beaver Valley Power Station Unit l_-

1991 Report of Facility Changes l Tests, and Experiments 4 Page 115 of187 CII ANGE TITI,E DCP 1712, Rev. O, Removal ot' Pilot Wire W-2 Ten Switch CII ANGE DESCRIPTION This design change provides for removal of the pilot wire test switch for relays87-104 & 87-105 (Main Generator); relays 87 Z109,87-Z110 (IB Transformer Leads); and relays 87-Z119,87- '_

Z120 (1 A T ansformer Leads).

.The switch's normally closed contacts do not completely close. This condition simulates an open-pilot wire condition. The MCB-1 relay win then trip for faults outside the area of protection.

SAFlifV EVAL.l!ATION SUMM ARY The test switches are provided to test the pilot wire pairs for main generator transformei. . A &

IB, They are part of the 4KV Station Service System, llowever, the switches are not used to test the pilot wires.

This change removes the test switches, therefore, there is no failure mode associated with the change. The presence of the test switches in the circuit increases the probability of a normally .

closed contact not being made; this satisfies a trip permissive and thereby increases the likelihood of a unit trip. Removing the test switch increases the reliability of the circuit in that the pilot _ wire isn't artificially sensed as open. The System operation will not be changed with the deletion of the test switches.

A review of Updated Final. Safety Analysis Report (UFSAR) Chapter 14 determined that no -

design basis accident analyses arc afTected by the proposed change. Test switch removal has no-radiological consequences since the pilot wire relays only affect electric generation with no direct impact on primary systems.

Pilot wire test switch removal cannot initiate any Design Basis Event. The proposed modification does not contribute to radiological consequences.

L Removal of the test switch does not cause a new type of accident. The switch is used to test au relay that has the potential to affect electric generation. The testing is no longer performed.- The absence of the test and test switch doesn't affect plant operations.

I No equipment important _to safety is affected by the change. Remaval of pilot wire test switches does not affect Technical Specifications.

Beaver Valley Power Station Unit 1 ,

'1991 Report of Facility Changes, Testa, and Experiments Page 116 of187 CII ANGE TITLE DCP 1726, Rev. O, Remove Internals of Check Valve RW-197 CIIANGE DESCRIPTION This design change will remove the trim from River Water System valve RW-197. This design change will cause a reduced pressure drop in the River Water System and allow removal of River Water System valves RW-197 and RW-198 from the valve inspection program.

This design change was initiated to:

1. Assure better river water flow through the recirculation spray heat exchangers for Operations Surveillance Test (OST) 1.30,!2, and

2. Remove two valves from the valve inspection program that are extremely diflicult to inspect in accordance with ASME requirements.

SAFETY EVALUATION

SUMMARY

Removal of the valve trim enhances River Water System performance and eliminates a potential system failure due to possible valve failure. Water flow, under normal and abnormal conditions, is siphoned to a circulating water pipe. Valve internal removal does not affect recirculation spray performance for the main steam line break accident.

Removal of valve internals from valve' RW-197 will enhance overall River Water System performance by reducing system difTerential pressure.

Updated Final Safety Analysis Report (UFSAR).Section 14.3 " Loss of. Coolant Accident" requires one of two recirculation spray heat exchanger trains to function._ Water from the heat exchangers goes through valve RW-197, The removal of valve RW-197 internals enhances availability of the system by eliminating the potential failure of the valve to open. Valve internal removal enhances water flow through the recirculation spray heat exchangers, The checking l feature of the valve is not necessary in this abnormal system operation.

l Valve trim removal enables the containment depressurization functionc The River Water System itself does not contribute to the severity of a Loss of Coolant Accident (LOCA).

. A new type of accident situation is not presented. Valve trim removal eliminates a potential

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! accident due to valve ~ internals failure. Valve trim removal enhances the containment depressurization function during the LOCA.

l There are no failure modes of equipment important to safety created by the valve trim removal.

l . Removal of the check valve trim eliminates the potential for failure of valve RW-197 internals.

l f Techdcal Specification 3/4.6.2 "Depressurization And CoCng Systems" requires that each river

Beaver Vciley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Expcriments Page 117 of 187 water subsystem and associated recirculation spray heat exchangers be capable of passing a flow rate of 8,000 gpm. Removing valve RW-197 internals makes it more likely that a flow rate of 8,000 gpm can be achieved by reducing overall system pressure drop. River water flow shall pass through the empty body of valve RW-197 and bypass entirely the remaining check valve RW-198.

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Page 118 of187 Cil ANGE TITI.E DCP 1730, Rev. O, Solenoid Valve Changeout fil ANGE DESCRIPTION Existing Solenoid Operated Valves (SOV's) have a Maximum Operating Pressure Differential (MOPD) rating which is too low for the service condition. This design change provides for replacement of these solenoid valves.

SAFETY EVAI tf ATION StiMMARY Failure modes for the replacement valves and the valves to be replaced are the sr.me. The SOV's could fail to stroke in response to the actuation signal. This failure mode is independent of the change being made.

The change will increase the stroke time of the valves whhin the Technical Specification limits.

The probability of failure will be reduced because the SOV's are better suited for their service-condition.

There will be no effect on Design Basis Accidents because mitigation systems will continue to function within Technical Specification limits, and will limit radiological releases to- those previousiy analyzed. No Design Basis Accidents have been made more probable.

The change is not modifying the plant response because the change is still within the Technical Specification limits.

No new failure modes or new unanalyzed type of malfunction is created because the SOV's are equal to or better than those presently installed and will operate within Technical Specification limits. Technical Specification 3/4.6, Table 3.6-1, indicates that the stroke time for valve TV-RC519 is 12 seconds maximum and for valve TV-SV100A is 20 seconds maximum.- This change willincrease the stroke time from actual historical records which are approximately 5 seconds for valve TV-RC519 and 10 seconds for vsive TV-SV100A. The flow coeflicient (CV) is being decreased from 0.38 to 0.31 which is a decrease of approximately 18%. Stroke times are also expected to increase by approximately 18%.

Beaver Valley Power Station Unit 1 l-1991 Report of Facility Changes, Tests, and Expertnents Page 119 of187 Cll ANGE TITI,E DCP 1731, Rev. O, Replacement of the Control Room Air Conditioning Condensing Unit Compressor VS-E-4A CII ANGE DESCRIPTION This modification will replace the existing Y-53 or "G" series single stage compressor for control room air conditioner condensing unit VS-E-4A with a new model 2-3/4" stroke "R", 8 cylinder compressor.

The new compressor will mount directly on the existing base without any changes to the base itself. Slight changes to the existing piping and condenser frame will be made to permit installation of the new compressor. Copper tubing at the compressor suction will require modification. All pumpdown and bypass piping between the compressor suction and discharge valves will be modified slightly to accommodate the new compressor. The replacement compressor a equipped with a liquid refrigerant cooled oil cooler. Some copper tubing and unistrut type tubing supports welded to the compressor frame are required by the design change.

Some electrical control changes are also required to accommodate the " oil cooling kit" The new compressor must meet the design capacity requirements. The emergency power rystem must be able to handle the power requirements for the new compressor.

SAFETY EVAL,UATION SUMMnRY Updated Final Safety Analysis Report (URSAR) Chapter 14 design bases accident were reviewed to identify design basis accidents that could be afTected by the proposed modification. The proposed modification does not affect any safety system function and, therefore, does not impact any accident analyses.

The Control Room Area Air Conditioning System will be atTected by the proposed design change.

Ilowever, this modification will have no adverse effects on the system: No safety-related functions will be changed.

The probability of occurrence for an accident previously evaluated in the safety analysis report will not be increased. No system functions will be changed, and no previously analyzed accidents will be affected by this change.

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

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis repon will not be increased. No system or safety functions will be adversely affected by this change.

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1991 Report of Facility Changes, Tests, and Experiments

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The consequence of a malfunction of equipment important to safety as previously evaluated in the'.

safety analysis report will not be increased. The new compressor is commercially dedicated for- q'

. Quality Assurance Category I use. No equipment'or system functions will be 'afTected by the change..

The design change will not cause any new credible failure modes because the original design _

features and functions of the equipment have not been altered. The possibility of failure in the affected equip _ ment already existed prior to this change, and probability of failure occurrence is; not considered to be increased due to this change. The existing Class IE electrical equipment will remain the same for the replacement compresser, This same component has been installed for the control area air conditioning refrigeration unit VS-E-48. ,

The possibility of an accident of a different type than previously evaluated in the UFSAR will not - -

be created because no equipment or system function is changed in a way which creates the .

. possibility of a ditTerent type of accident.  !

The possibility for a malftmetion of equipment important to safety oft a difrerent- type than previously evaluated in the safety analysis report will not be created. There is no configuration '

- change such that a new failure mode or potential hazard will be created by the implementation of ~

this modification. Original design features and functions will not be affected by this design change.

There a a e changes in parameters _which affect the course of any accident analysis supporting-Technical Specification bases, or result in exceeding the acceptance criteria for fuel cladding, Reactor Coctant System boundary, or containment integrity. The margin of safety. as defined in thebuk, .my Technical Specification will not be reduced because no bases will be afTected by-tfA mufu stion.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tcsts, and Experiments Page 121 of 187 CII ANGE TITLE DCP 1736, Rev. O, Splices in Terminal Boxes TB-132 and TB-119 CII ANGE DESCRIPTION Terminal block connections TB-132 and TB 119 will be spliced. These terminal blocks are located in the auxilian feedwater pump room, and are provided for limit switches and SOV's on throttle valves TV-MS105A and TV-MS105B. This will improve these limit switches to meet Regulatory Guide 1.97 equipment qua!ification.

Corrosion was found on conductor lugs for terminal block TB-132 during the steam turbine driven auxiliary feedwater pump operating surveillance test (OST 1.24.4); In order to assure equipment qualification, a Raychem splice will be used in place of the terminal block.

SAFETY EVAlliATION SUM M A RY Limit switch leads, which are lifted to prevent the steam driven auxilian feedwater pump from starting during OST 1.24.4, will have to be lifted in relay panels PNL-REL-35 and PNL-REL-36 after the splices are installed.

There are no credible failure modes associated with this change because equipment qualification will be improved. This change only ensures continuity. The new splices will reduce the probability of failure of the above system due to improved continuity in the circuit. The system will perform the same, however the probability of failure will be reduced.

This change will increase system reliability in the event of loss of normal feedwater due to -

improved continuity in the circuit to start the steam turbine-driven auxiliary feedwater pumps.

Because the splices are an upgrade, there are no failure modes associated with this change There will be no affect on the probability of occurrence of a loss of normal feedwater due to this form of electrical connection inside the terminal boxes.

No new accident is created. A qaalified splice will keep out moisture which is causing a potential continuity problem. No new type of malfunction is created. This is an upgrade of an electrical connection.

Technical Specification 3.7.1.2b states that one feedwater pump, capable of being powered from an OPERABLE steam supply system, shall be operable. This change will improve reliability in starting the pump. The acceptance limits are based on pump discharge pressure during recirculation flow which is not affected by this change.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes; Tests, and Experiments Page 122 of 187 Cil A NG E TITI,E DCP 1737, Rev. O, Pilot Valve Pressure Switch Seat Leak By, Sensing Line Tee Addition Cll ANGE DESCRIPTION Design Change Package (DCP) 1737 and its associated minor modification package will install a swagelok tee in the censing line between pressure switches PS-RC-551 Al, 551B1, 551Cl and pressurizer safety valves RV-RC551 A, 5518, 551C respectively. This change will allow for enhanced maintenance. The swagelok tees will enable personnel to more easily fill the sensing line of the pressure switches.

SAFETY EVALliATION SliMMARY The pressure switch composed of a switch, valve body, tubing and a connector shall be qualified -

to withstand 2500 psig. Seismic qualification of the components has been maintained.

A failure of the tubing tee will cause the pressure switches to fail low which will not allow the pressure switches to monitor for this leak. The electrical portion of the switches are Category II.

They are pressure boundary only.

This design change will have no effect on the function of the pressure switches and will have no afTect on the pressurizer safety valves as described in Updated Final Safety Analysis Report-(UFSAR) Section 4.2.2.7. There will be no efTect on the rating of the sensing lines as evaluated by DCP 426. This change also does not affect the ratings of the safety valves as described in Section 14.1.7.

This change will provide enhanced maintenance. This design change will allow the sensing lines to be filled more easily and ensure the lines are adequately filled.

This modification has no etTect and does not increase the probability oflifling pressurizer safety -

valves as previously evaluated in Section 14.1.15 and Section 14.1.7 of the UFSAR. This change will have no efTect on the pressurizer safety valves as described in Section 4.2.2.7 of the UFSAR l

l The possibility of an accident or malfunction which is different than that evaluated in the UFSAR will not be created. Rating of ti.e pressure . .vitch sensing line will not be degraded from that-which has already been evaluated by DCP 426. This design change'does not initiate any L electrically operated safety equipment.

There are no effects on Technical Specifications 3/4.4.2 (Safety valves-shutdown), 3/4.4.3.'

(safety valves-operating), and 3/4.4.6.2 -(operational leakage) due to the DCP (minor modification). There will be no impact made by the change on the acceptance limits set forth in Technical Specifications 3/4.4.2 (safety valves-shutdown),3/4.4.3 (safety valves-operating), and 3/4.4.6.2 (operational leakage).

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Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 123 of187 CII ANGE TITIE DCP 1738, Rev. O, Replacement of Relays 27-RP100 and 27-RP1100 CIIANGE DESCRIPTION This design change will replace emergency bus IP under voltage relays 27 RP100 and 27 RP1100 480V, The existing relays have degraded to the point that they ere required to be replaced with a vendor recommended replacement.

SAFETY EVALUATION

SUMMARY

The present relay has a 10 CFR 21 defect which will cause an inadvertent trip. The failure modes - -

of the new relay are the same as th: present relay except that it will not cause a spurious trip for a relay fai' arc or loss of control power, The system will operate the same.

This design change has no efTect on the assumptions or radiological consequences of any design basis accident. The new relay has the same failure modes as the present relay, except that it will not cause a spurious trip on a relay failure or a loss of control power. Thus, a loss of coolant accident or other design basis accident will be mitigated in the same manner and the radiological consequences will be as previously analyzed.

There are no design basis accidents for which failure modes associated with this change can be an initiating event. This relay would affect the performance of 480V mitigation equipment.

The change does not afTect the probability of occurrence of any design basis accident. This change does not cause a new type accident since it is a direct replacement with the same characteristics as the obsolete relay.

Failure modes of the new relay do not represent new unanalyzed types of malfunctions since the change is a one for one replacement of the obsolete relays. The failure modes of the new relay are the same (or less probable) than the present relay except that the new rehy will not cause a-spurious trip'on a relay failure or loss of control power.

The acceptance limits which form the basis for Techn: cal Specifications 3/4.8.1, and 3/4.8.1.2 are not afTected by this change. This change has no impact on the acceptance limits which form the basis for Technical Specifications 3/4.8.1, and 3/4.8.1.2.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Enperiments Page 124 of 187 OMNGI' TITI.E DCP 1740, Rev. O, Diesel Generator Strip IIcaters GIANGE DESCRIPTION This design change will remove the diesel generator strip heaters circuits from the auxiliary start relay #9 and #13 contacts. This switches the heaters off when starting the diesel generators. The diesel generator heaters are powered from a non safety related panel (PNL-AC-41). The breaker switch in this panel will be used to shut the heaters oft.

The safety related auxiliary relay (ESRXA) has had the contact wire burned off from oser current through it's #9 and #13 contact. Relay (ESRXA)is a MG.G rated at 12 amps continuously. The relay was installation tested in 1988 at i1.9 amps, and maintenance measured 12.2 amps after the relay wire burned off.

SAFETY EVAL,llATION SLIMM ARY The diesel generator strip heaters do not need to be shut off while the diesel generator is running for a long period. THs was verified during installation of the heaters in 1987 and 1988. The heaters will not add a significant amount of heat during a long (2 week) diesel generator run (e.g.,

during an accident condition).

If the auxiliary start relay contacts #9 and #13 wculd weld together due to excessive current in the diesel generator strip heater circuit the other contacts on that relay might not open/close as the original design intended and may interfere with the electrical control circuit logic for a demand to start the diesel generators from the auxiliary start relays ESRXA and B. This modification would correct / eliminate this potential relay failure, therefore, the only failure modes are those already possible with the relays.

Since a relay contact burned on safety related relays ESRXA which is in a non-safety related strip heater circuit, this minor modification would remove the non-safety strip heater circuit from the safety related auxiliary relays ESRXA and D. This removal of the strip heater circuit from the safety related relay contacts would decrease the chance of contacts welding together and causing a safety related relay malfunction that renders the auxiliary start relays in-operable.

The possibility of the auxiliary start relay burning or contacts welding close could inhibit the diesel generator control panels ftmetion upon demand to start or continue to operate. No direct radiological concerns are immediately identified upon failure of relays.

This removal of strip heaters (non safety related components) from safety related relays ESRXA and B will increase the reliability of the relays.

This modification will reduce the probability of a failed safety-related auxiliary start relay (ESRXA & B) resulting from an overcurrent condition caused by the diesel generator strip heaters non-safety related circuit.

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Beaver Valley Power Station Unit 1-~

1991 Report of Facility Changes, Tests, and Experiments

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Page 125 of 187 -

Emergency diesel generator alternate AC power has been analyzed for single failure criteria and-engineered with redundancy and independency. This modification will not create a new type. ._ ,

accident and will no: change original designed safety features.

Technical Specification 3.8.2.2 identifies minimum AC bus requirements. _ Removing the heater circuits from' the safety related contacts will not. reduce the. Technical Specification 3.8.2.2 minimum acceptance limit of one 4160 cmergency bus.

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' Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 126 of 187 CII A NG E TITI,E DCP 1741, Rev, 0, Replacement of Fischer and Porter Transmitter LT-WT-104 A1 I

CII ANGE DESCRIPTION This design change will replace the nisting Fishei and Porter level transmitter on the primary plant demineralized water storage tank with a Rosemount Model #1153 transmitter. The Rosemount Transmitter has been evaluated in terms of pressure range, radiological environment (mild), seismic qualification, QA Category (1), materials of construction, temperature limits, and ,

accuracy. No adverse rfrects are expected.

This replacement will make available a Fisher and Porter transmitter to replace toe refueling water storage tank level transmitter LT-QS-100A. Level transmitter LT-QS-100A is outdoors and the :

primary plant demineralized water storage tank level transmitter LT-WT-104A1 is located in a

-heated building. Since the operating range of the Rosemount is +40 to +200 F, installing it on the reflieling water storage tank would require a heated enclosure heat trace and an annunciation circuit which aren't necessary for the primary plant demineralized water storage tank.

M f:TY EVALUATION

SUMMARY

The Ros' mount transmitter has an operating temperature range of +40 to +200 F. The existing Fisher and Porter (F & P) transmitter has an operating temperature range of-40 to +212 F.

Because level transmitter LT-WT-104A1 is in a heated building, the difference in operating-temperature will have no impact on the level transmitter or the system itself. The Rosemount transmitter will be mounted difTerently than the existing transmitter. The Fischer and Porter transmitter is a flange mounted transmitter whereas the Rosemount will be mounted on a rigid-support. This will require the installation ofprocess tubing between the flange and the transmitter like the existing F & P, the Rosemount is QA Category I for pressure boundary and is seismically qualified. Both the transmitter and the tubing will be installed seismically.

A crack or break in the process tubing would be a breach of the pressure boundary. The leak could conceivably continue until the tank -(WT-TK-10) level fell low enough .to actuate

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annunciators A6-52 and A6-54. At this point (per the operations manual) the local indicator l' would be checked and the leak would be observed and isolated. Since the transmitter is safety related for pressure boundary only, and since the transmitter provides indication only, failure of the transmitter will have no impact on Auxiliary Feedwater System performance or the probability of system failure.

Level transmitter LT-WT-104Al monitors the level of the primary plant demineralized water

storage tank WT-TK 10. These components are part of the steam generator Feedwater System and are safety related.

Since the transmitter provides indication only and any tubing leak will be detected and'isoiated, this design change does not affect the assumptions or radiological consequences of any accidents.

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LIleaver Valley Power _ Station Unit i 1991 Repon of Facility Changes, Tests, and Experiments:

Page 127 of 187.-

- The transmitter provides indication only. Failure of the transmitter will not initiate other failures.

-This design change has no impact on plant response.

Failure of the transmitter has no effect on the system since the transmitter provides indication only.13 reach of the pressnre boundary, (i.e., a tubing leak), will be detected by indication in the control room and will be isolated. No new unanalyzed typ'e of malfunction will be created by this

' design change.

Primary plant demineralized water storage tank WT-TK-10 must have a minimum; contained volume of 140,000 g'allons. Since the low level annunciator will alarm at 140,995 gallons and the; tubing is only 3/8", any tubing leak will be isolated prior to violating the Technical Specification  ;

3/4.7.1.3 volume requirement.

As stated above, this change does _not impact the Technical Specification volumetric requirement, _

therefore, the ability to provide a heat sink will not be compromised.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiraents Page 128 of167 CII ANGE TITLE DCP 1743, Rev. O, Electro flydraulic Fluid Reservoir Low-Low Level Lockout Switch Cil ANGE DESCRIPTION Westinghouse Operation and hiaintenance hiemo 122 describes recommended changes to preclude an Electro Hydraulic (Ell) fluid pump trip due to low Ell fluid reservoir icvel. This would eliminate the potential for unit trips due to spurious indications.

SAFETY EVAL.U4 TION StiMM Alq The automatic controllogic for EH pumps LO-M-9A, and -98 will be modified to preclude a trip of the pumps on low fluid level. Westinghouse Operation and Maintenance Memo 122 describes a change of pump control wiring to eliminate the possibility of unit trip due to the misoperation of the low-fluid level switch.

The control logic for EH fluid pumps LO-M-9A, LO- M-9B are not safety related.

This change will reduce inadvertent unit trips attributed to low fluid indicators in the EH fluid reservoir. Operator response for the low-level alarm will still be adhered to na alarm circuitry will still exist.

The EH fluid pumps are not necessary for safe shutdown of the plant. There are no Design Basis Accidents associated with this change. This design change will eliminate a potential unit trip due to an inadvertent low EH reservoir level switch signal.

There are no failure modes associated with a design basis accident. The pump will continue to run on a low level signal.

This change will reduce inadvertent unit trips attributed to low fluid level indicators in the EH fluid reservoir by defeating the low fluid trip relay contacts of the pump. This change does not involve the creation of a new type of unanalyzed event.

Ell fluid pumps LO-M-9A and L.0-M-9B are non-safety related, therefore, the control logic changes do not represent a new unanalyzed type of event. This change will reduce the possibility of an inadvertent unit trip due to low-fluid indication in the EH fluid reservoir.

There are no Technical Specifications associated with the turbine lubricating oil system.

- Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 129 of 187 Cil A NGE TITLl' DCP 1746, Rev. O, Modification of Steam Dump Control System CIIANGC Di'SCillPTION This DCP will make temporary modification 1-90-09 and I-90-10 permanent.

1/PEX 9000 electro-pneumatic transducers were installed m place of Fisher type 546 transducers -

as temporary modification 1-90-010 to provide for better control and increased reliability.

Temporary modification 1-90-009: 1) placed an additional summator in series with TM-408N (temporary to TM-4G3N-1) to provide the driving voltage required for the new electro-pneumatic transducers. 2) replaced driver summators TM-MS-408N, P, Q, & R with summators containing .

output limiting to prevent overdrive of demand signal when out of range, tr, timit valve diaphragm air pressure to only that required, resulting in reduced closing time when demand is terminated.

Replaced 1/V modules TM-408L 1/R, TM-40831/R, and PC-464A I/R witt the proper module for the application.

SA FE'I' ' 2 VALUATION SUMM AltY The new system design is similar to the existing system. This modification will decrease the probability of an accident.

The Steam break accident, including inadvertent actuation of the steam dumps is already analyzed.

The safe condition of steam dump valves is closed after sullicient cooldown. This modification will prevent overcooling that is possible with the present system. Valves will still epen as before the modification. This modification does not change the consequences of a relfunction.

Consequences of a malfunction are bounded by the accident analyses for Loss of Electrical Load / Turbine Trip, and Accidental Depressurization of the Main Steam System.

This modification results in better control and operation of the systcm. Therefore, no new accidents are created.

The system may fail in the same ways as before, but failure is less likely for the overcooling-condition.

System operation is not assumed in Technical Specification basis.

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Beaver Valley Power Station Unit I-1991 Repon of Facility Changes, Tests, and Lxperiments _.

Page 130 of 187 CII ANGE TITI,E DCP 1747, Rev. O, Relocation of Terminal Boxes From Flow Indicating Switee, FIS-FW-151 A,151 B,152 CII ANGE DESCRIPTION Flow Indicating Switches FIS-FW-151 A,151B and 152 are currently installed with terminal boxes mounted directly to the conduit port of the switch. However, these switches were not seismically tested in such a manne. (per IEEE 344). T10 configuration invalidates the testing documentation which was supplied with these switches.

The objective of this minor design change is to relocate the terminal boxes off the auxiliary feedwater system flow ndicating switch housings, and mount them directly to the flow indicating switch supports. This will return the flow indicating switch (FIS-FW-151 A,151B and 152) configuration to an as-tested condition.

SA FETY EVAL,UATION SUMM ARY The changes will not afrect auxiliary feedwater pump operation. The design parameters and safety ftmetion of the Auxiliary Feedwater System will not be altered.

The credible failure mode associated with this design change is failure of the flow indicating switches FIS FW-151 A,151B and 152. Failure of the switches could cause the flow control valves to open. This in tum could afTect the flow of feedwater to the steam generator during an accident. Failing high will cause the recirculation flow control valves to shut which may challenge the pump protection capability.

The relocation of terminal boxca will assure the flow indicating switches are within the seismic design criteria. Therefore, the proh Q4ity of flow switch Pailure will not be increased.

Since the credible failure modes er the flow switches will not be increased by this desiga change, the safety functions of the Auxiliary Feedwater System will not be affected.

Loss of normal feedwater, loss of offsite power to the station auxiliaries, steam generator tube rupture, and major rupture of a main feedwater pipe were reviewed. The change will not afTect the design bases of the Auxiliary Feedwater (AFW) System, and accident consequences will not-be increased.

The AFW System is a standby system, failure of the flow indicating switches will not cause an accident initiating event.

The relocation of terminal boxes will assure the design bases of the AFW System is maintained during a seismic event. Thus, the probability of occurrence of the design bases accidents will not be increased, and no new type of accident will be created.

Beaver Valley Power Station U_ nit'l-1991 Repott of Facility Changes, Tests, and Experiments

-Page 131 of_187 The terminal boxes will be mounted on the existing flow switch supports. - These suppons have been analyzed and were found to be acceptable with the additional terminal box, Therefore no new type of malfunction will be created.

Technical Specification 3/4.7.6.2 was reviewed. This design change will not affect the Technical :-

Specification.. The design change will ensure AFW pump operability is maintained under-_all conditions. There is no affect on the Technical Specification acceptance limits.

Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 132 of 187 Cil ANGE TITLE DCP.1748, Rev. O, Permanent Controls for Installation of Westinghouse LIFETIME Temperature Monitors. -

CIIANGE DESCRIPTION The moniters are being installed in various locations tu provide long term temperature baseline information as pan of the Equipment Qualification Program.

SAFETY EVAI,11ATION SilMMARY No system operating or design parameters are alTecad by this design change. The monitors to be installed are passive self contained devices that do not interact with any existing plant equipment or systems. Each enclosure contains 0.35 lbs. of aluminum, thus the total addition of aluminum is well under 1% of the total aluminum within containment. The devices are resistant to caustic spra3 The only possible failure mode, tinco the monitors are not needed for safe shutdown of the unit, is a failure of structuralintegrity caus. '3 a monitor to act as a missile. The mounting brackets are designed to withstand a seismic acceleration well in excess of that experienced from a Design Basis Earthquake (i e.,0.125g according to Updated Final Safety Analysis Report Section 2.5.2).

Qualification of equipment will prevent hydrogen generation, which would be insignificant at any- ,

(- rate due to the small amount of aluminum. Thus n!! systems will function unalTected by these devices.

As the change does not afTect the accident analyses or the performance of any mitigation system, the accident consequences remain as previously analyzed.

The only potential failure modes discussed are themselves the results of an accident. The devices are totally passive and do not impact normal operating parameters.

Since, the devices are seismically mounted as stated before, and have no impact on any initiating

titilure mechanisms, they cannot increase the probability of any event initiated by those failure mechanisms.

The only failure modes result as a consequence of a previously analyzed event. The change involvej passive devices that have no effect on the station.

These devices have no inter-relationship with any Technical Specification equipment and thus do l not affect their operability.

L L No Technical Specification equipment or operating limits are afTected by this change.

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l Ileaver Valley Power Station Unit 1 i 1991 Report of Facility Changes, Tests, and Experiments Page 133 of187 Cil ANGE TITLE DCP 1750, Rev. O, Pressurizer linter Controller Modification Cil ANGE DESCillPTION a package makes several minor modifications to the pressurizer heater controller to eliminate noise spikes it has been generating. These spikes are being picked up on containment air total pressure monitor PT-LM-100C, and are being generated whenever the pressurizer heater, bank C, is turned on from the vertical board.

SAFETY EVALUATION SUMM ARY

'C' bank pressurizer heaters are non-safety related, non-lE heaters that are not used to mitigate an accident. There are no credible failure modes associated with this moditication. This change.

eliminates a potential failure mode associated with inadvertent actuation of containment pressure monitor (PT-LM-100C) components due to noise.

Addition of noise suppressing devices to the 'C' bank controller will not change the probability of

'C' heater bank failure. The operation of the controller is unafTected by Qe design change.

There are no design basis accidents associated with the 'C' bank pressurizer heaters. These heaters are only control heaters.

A failure of the controller for the 'C' bank will have no effect on the radiological consequences of accidents described in the Updated Final Safety Analysis Report (UFSnR).

This is a minor change to reduce noise generated by the controller of the 'C' bank and will not increase the probability of the 'C' bank becoming an initiating event for accidents described in the UFSAR.

The modification will have no afTect on the heater controller. A failure of this non safety related controller will not create a new type of accident or malfunction.

Technical Specification 3/4,4.4 requires operability of at least 150kw of pressurizer heaters in modes 1,2, and 3. This change affects the controller for the 'C' bank pressurizer heaters. This bank is not supplied from a Class IE safety related power source. The Technical Specification refers to pressurizer heaters supplied from a Class IE power source.

Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 134 of187

.Cil ANG E TITI,E DCP 1751, Rey, 0, Source Range Monitor Nuclear Instrementation System (NIS) Rack Grounding CilANGE DESCRIPTION This design change installs a jumper wire that connects the bus (ground) bar of the source range detector (N-31) cabinet to station ground.

Westinghouse recommended this change to eliminate noise afTecting source range detector h-31 output. The change meets requirements set forth in Westinghouse Control and Electrical Systems Standard Section 4.1, Document Section 3.16. This standard was written for field installation of NIS triaxial c41es.

SAFETY EVAI,UATION SilMM ARY This design change incorporates proper grounding techniques to increase the reliability of the source range monitor and does not have a failure mode associated with it. Failure modes are the same as those previously analyzed.

The addition of proper grounding increases the reliability of the source range monitor. This change also decreases the probability of N-31 source range failure.

There are no design bases accidents associated with the installation of proper grounding for the N-31 source range monitor cabinet. The monitors will function as before to detect uncontrolled rod withdrawals when the reactor is shutdown (Peference: Updated Final Safety Analysis Report Sections 14.1.1 and 14 ).4) -

There are no udiological consequences associated.with the addition of proper grounding for the N-3) ource range cabinet in the . mtrol room. The source range detectors will operate as before and radiological consequences of events that take credit for the detectors remain the same.

There are no failure modes associated with the addition of proper grounding. The detector is a monitoring instrument to detect the approach of criticality.

The addition of proper grounding increases the reliability of the N-31 source range monitor. It will not increase the probahility of occurrence of a design basis accident.

The installation of proper grounding will not cause an unanalyzed type of accident. The instrumentation is for monitoring and mitigating certain subcriticality events, it can not initiate them.

The addition of proper grounding, as recom,nended by the vendor, does not impact Technical Specifications.

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' Beaver Valley Powe 'tation Unit !

1991 Report of Facility Changes, Tests, and Experiments Page 135 of187 Cil ANG E TITI,E DCP 1752 Rev. O, Source Range Monaor Penetration Wiring Change CIIANGE DESCRIPTION Rewiring at the penetration for NI-31 will be performed due to a defective penetration pin.

Source range detector penetration RCP-9A Pin 5 is defective. Pin 6 of penetration RCP-94 is currently spare and will be used for the N-31 sc ;rce range monitor.

SAFETY EVAL,UATION SUMM ARY The NI-31 detector will operate as before. Credible fai:are modes are as before. ,

The NI-3 detector will perform the same function as before. This change only involves a cable relocation to the spare penetration pin location. There are no non-failure modes associated with this change.

Containment integrity will be maintained as this is merely a cable relocation.

As stated previously, the source range will function as before to provide monitoring during .

shutdown. System functions will remain as previously analyzed and radiological consequences are the same.

These monitors are for the detection of events and since they will operate as before the wiring change , no accident probabilities have changed.

No new types of accidents are being developed. The source range monitor will still perform its intended function. Other system and core parameters remain as before.

Source range monitor N-31 will function as before meeting the operability requirements.of Technical Specification 3.9.2.

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! Beaver Valley Powcr Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 136 of187 CII A NG E TITIJ' DCP 1753, Rev. O, Remosal of Check Valve Internals for Valves IRW-133,134,135, and 136 CII ANGE DESCRIPTION This design change will remove valve internals from River Water System check valves in the supply to the control room air conditioning condensers and the redundant cooling coils (valves IRW- 133,134,135,136). The System is being changed to improve river water flows.

SAFETY EVAL.UATION

SUMMARY

No operating or design parameters wi'.1 be atTected by this change. The NSA position for manual cross-connect valve IRW-139 will be changed from open to closed. Updated Final Safety Analysis Report (UFS AR) Figure 9.9-1 A must be revised.

The check valves will have fewer parts; there are no new failure modes associated with this change. The original pressure boundary and flow paths are being maintained except that valve IRW-139 will be closed.

The river water system is designed to meet single failure criteria (UFSAR Section 9.9.2). The safety ftmetion of check valves IRW-133,134,135 and 136, is to prevent cross-connecting the "A" and "B" river water headers such that a single pipe break in either header could drain both headers through normally open cross connect valve IRW-139. These check valves are located in the 3" supply lines _ downstream of the 6" X 3" reducers, to the control room air conditioning condensers and to the redundant cooling coils. Without these check valves preventing reverse flow, a break in the 6" header could potentially drain both redundant headers. The two 3" headers sersing the redundant cooling coils are separated by normally closed valves IRW-142,143,144, and 145.

The proposed modification removes the check valve function from this portion of *! - system, but the safety function (isolation of redundant headers) is accompliu.:d by shutting the cross-connect valve IRW-139 and maintaining it administratively locked closed. Valves IRW-142,143,144, 145 must also be administratively controlled to prevent crea*.ing a cross-connect flow path between redundant 3" headers. These valves (IRW-142,143,144,145) are shown in UFSAR a Figure 9.9-1 A as open; per Operating Manual Figure 30-2 they are shown closed. The UFSAR must be revised to correctly reflect the normal system arrangement position of these valves The physical change to the check valves will not increase the probability of failures. This change will reduce flow restrictions and increase Oow.

This change does not affect the assumptions and radiological consequences of accidents because no changes are prc j ased which will reduce river water flow.

These components are not accident initiators for any design basis accident evaluated in UFSAR l

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Ileaver Valley Power Station Unit 1 1991 Report of Facility Clenges, Tests,' and Experimer.u -

- Page 137 of 187-Chapter 11.

-This change creates no naw failure modes. The isolation function provided by the check Aves--

will now be provided by nn administratively controlled closed valve.

No new failure modes were identified, no new unanalyzed type of malfunction was created, Reverse flow within a header is prevented by operating pump pressure and flow, and by closed isolation valves. Ily closing valve IRW-139, reverse flow between headers is prevented.

Technical Specification 3/4.7.4 requires two Reactor Plant River Water System subsystems to be:

operable. The proposed change does not aTect the operability of the River Water System; system flow rates will not be adversely affected.

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Beaver Valley P_ower Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 138 of 187 Cil ANGE TITI,E -

DCP 1755, Rev. O, Disconnecting Two Pressurizer; aters CHANGE DESCRIPTION This design change will disconnect two pressurizer heaters (59 and 45) at the pressurizer and at the reactor containment penetration.

Pressurizer Heaters 59 and 45 have shorted to ground rec.dering two heatei banks inoperable.

Each bank of heaters is made up of three emersion hea:ers connected in a delta format, and fed -

from its own breaker in the power distribution panels located in the cable vault rooms. To isolate the damaged heaters, two circuits were de-energized causing a loss of power to the four working heaters powered from the same circuits. This design change will restore power to the four-working heaters.

SAFETY EVAL,UATION SUMM ARY None of the safety systems and systems important to safety are affected by the two heaters taken out of senice. The affected heaters are fed from a Category 11 power source. Four of the six heaters in the two heater banks are being returned to senice.

The change and/or the failure modes associated with the change will not affect the probability of failure of the systems since two of the three heaters out of each bank would be returned to senice.

The change will not affect the performance of safety systems. The afTected heate" are used for the control group only and are fed from a Category Il power source.

Design basis accidents described in the Updated Final Safety Analysis Report (UFSAR) are not alTected by the change. UFSAR Section 14.1.16, " Spurious Operation of the Safety injection System," assumes that the pressurizer heaters are non-operable. The control heaters are not fed from an emergency power source.

Technical Specification 3/4.4.4 requires operability of at least 150kw of pressurizer heaters in modes one, two and three. This change will have no effect on the' Technical Specification requirement since each heater is rated at approximately 18kw, and 76 heaters will remain in service.

Heaver Valley Power Station Unit i 1991 Feport of Facility Changes, Tests, and F.xperiments Page 139 of 187 D_1 ANGl TITIE DCP 1756, Rev. O, Removal of Check Valve Internals at IRW.158 and 159 CII ANGE Di'SCitiPTION This design change will remove the nternal parts from Itiver Water System check valves (IRW.

158 and 159)in the cooling water supply to the charging pump coolers. The River Water System is being changed to improve niver water flows SAFl:TY EVAI,11ATION StiMM ARY No ope ating or design parameters will be arTected by ihis change. The normal system alignment for valves IRW 161,162,163 and 165 will be changed from normally open to administratively locked closed Updr a J Final Safety Analysis Report (UFSAR) Figuie 9.9.l A must be revised to reflect the new valve positions.

There are no new failure modes associated with this change. The original pressuie boundary and flow paths are being maintained except that administrative controls will be utilized to maintain separation of redundant headers.

The physical change to the check valves will not increase the probability c,f failure or initiate additional failures.

The system will perfor msistent with the design. Train separation will be provided by administratively locked ch .1 valves instead of check valves.

This change does not afrect the assumptions and radiological consequences of the above accidents because redundant system separation is maint r ' and cooling water flow is not reduced.

These components are not accident initiators for any Design Basis Accident evaluated in UFSAR Cha;...r 14, This change has no elTe on the probability of occurrence of any of the previously analyzed accidents.

No new failure modes are created. Train separation is now provided by closed isolation valves.

No unanalyzed malfunctions are created by the proposed change. There is no change to, the system pressure boundary.

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Technical Specification 3/4. 4 requires two Reactor Plant idver Water System subsystems to be operable. The propcsed change does not ailbct the operability of the River Water System; each cooler can be aligned 10 eitht header and a locked closed vah provides header separation.

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11eaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experimcnts Page 140 of 187 CII ANf;l'. TITIE DCP 1757, Rev. O, Ilydrogen Recombiner Wiring Modification Cil ANGE DESCRIPTION A wiring discrepancy has been found. The thermocouple input to the recombiner heater controller is from heater output instead of reactivity chamber temperature as specified by the vendo- The temperature of the reactivity chamber is a necessary input to the controller for system operability. This wiring change for the A hydrogen recombiner will ensure proper operation of the associated heater bank.

S A FF(ly I'NAI,lTN1 ION SilM M AIW There will be no new credible failure modes associated with this change. All failure modes are as previously evaluated. This change will ensure operability of the A hydrogen recombiner. No increase in the probability of a hydrogen recombiner failure is initiated This change ensures the operability of the hydrogers recombiner and will increase the reliability and capability of the system. The System will function as intended. The assumptions and offsite dose predictions are as analyzed in Updated Final Safet> analysis Report (UFSAR) Sec. ion 143.

This design change involves accident mitigating equipment. Failure of this equipment does not initiate a Design llasis Accident.

There is no new type of accident created by this change. This change only restores the equipment to it's original intended function. Accidents are as analyzed previously.

This design change will ensure system operability and will not change Technical Specification acceptance limits. This change will ensure compliance with Technical Specifications,

l lle:.ver Valley Power St: tion Unit 1 1991 Report of Facility Changes, Tests, and Experiments i Page 141 of 187 CII ANGE TITI,13 DCP 1780, Rev. O, MoJification to Control Room N Conditioner VS-E-411 Support l

Cil A NGE I)ESCRIPTION l This design change will modify the structural support for the condenser on the Control Room air conditioner unit (VS E-411). The support will be redesigned to allow removal and reinstallation of the compressor without disassembly.

'l he current location of the condenser support interferes with removal of the compressor.

SA FETY EVAI,UATION SilMM ARY The operating and design parameters of the control room air conditioner unit will not be alrected I by this change. The supports will be redesigned to the original design criteria including seismic requirements. I Failure of the suppon could cause failure of the air conditioning unit. The modification of the support does not change the failure modes or probability of failure _ of control room ventilation systems since the support will be seisraically designed.

The modification of the support will not affect the performance of the control room ventilation system since no operating or design parameters will be changed. No parameters or systems are . ,

affected such that equipment will not perform ftmetions required to meet assumptions made in accidents identified.

Fai!uce of the control room air conditioner unit will not initiate any accidents.

Modification cf the suppon does not change the probabiuty of occurrence of any accident.

Modification of the suppon does not change the opecation of the control room air conditioner er the response of the control room emergency prer arization system.

Failure of a control room air conditioner unit is analyzed and a redundant air conditioner with backup cooling coils is available. The redundant air conditioner and backup cooling coils will meet design basis requirements. No common mode failure is introduced by this change.- No new faihue modes of hstdied equipment imponant to safety are created by this change and therefore no new unanalyzed types of malfunctions are created.

- The control room air temperature is maintained less than 88 F in accordance with Technical Specification 3/4.7.7. Since the modification of the support will not affect the operation of the air conditioner unit there 'will be no change to Technical Specification acceptance limits. j

lleaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 142 of 187 Cil ANGE TITIE DCP 1800, Rev. O, Reloca'e Reactor Vessel Overpressure Protection Relays CII ANGE DESCRIPTION The function of the Relays for Overpressure Protection System will be relocated from the auxiliary relay cabinets (RK AULREL A and B) to the auxiliary safeguards cabinets (AUX.

RPTST A and B). The auxiliary relay racks are not qualified to IEEE-279. Westinghouse ha issued a 10 CFR 21 notification that the auxiliary relay ra:ks should not contain safety related equipment.

b'AJETY EVAL liATION Si1MM ARY This design change affects the control circuit for Power Operated Rc'icf Valves (PORVs) PCV-RC-455D and 456. No operating parameters are a'rec tec since the '. clays will be the same and wircJ identically. The relays willjust be in a different cabinet.

Since the relays will be identical and will be installed in a qualified cabinet, there are no failure modes attributed to this design change.

Since the only change is to locate the relays in a qualified rack, the probability of failure will not be increased.

There are no changes to performance. The relays are of the same type and wired the same.

The alTected relays have no affect on the Updated Final Safety Analysis Report accidents.

The affected relays cannot initiate inadvertent PORV actuation since they are locked out during power operation.

The probability of Over Pressure Protection Systera (OPPS) failure is not increased because the system is identical except for cabinet qualification.

The plant response is identical with the relays in an adjacent cabinet.

There are no new failure modes created because the relays and wiring are the same.

The design change upgrades OPPS to !icensing basis requirements, i e. seismic qualification.

11eaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 143 of187 CII ANGE TIEC DCP lh: , Rev. O, Permanent Removal of Exciter llearing Oil Drain CllANGE DESCRIPTION This design change removes the generator exciter bearing oil drain temperature indicator (T1 Til-211 local). Westinghouse Turbine Generator Operation and Maintenance Memo (OMM) 049 recommends the removal of the exciter bearing local temperature indicator to prevent inadvertent giounding of the electrically insulated exciter bearing.

SAFETY EVAI,UATION SifMM ARY The local bearing oil drain temperature indication (TI TB 211) thermocouples provide the plant computer with the "same" oil drain temperature and the bearing metal temperature.

The local temperature indicator is being removed. No new credible failure modes exist. The exciter bearing parameters as still adequately monitored as is also stated by the Westinghouse OMM 049 This change, and specifically the turbine generator lubricating oil system, does not affect any systems important to safety, therefore, the probability of a safety system failure is not increased.

The removal of the exciter bearing oil drain temperature indicator does not have any effect on the performance of any safety system because the turbine generator lubricating oil system does not alrect the performance of any safety systems.

No Updated Final Safety Analysis Report Chapter 14 accidents will be affected by this change.

The turbine generator system has no cirect on the radiological consequences of any accident.

No failure modes are credible with the removal of the exciter oil drain local temperature indicator.

The turbine generator lubricating oil system does not have safety- functions, therefore, no probabilities of occurrence of any accident will be increased. The removal of the local temperature ,

indicator can not cause any accident. The temperature indicator sensing element is being removed - '

from the pipe thermowell.  !

No safety related equipment is affected by this change. No Technical Specifications apply to the  !

turbine generator lubricating oil system. I 4

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Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 144 of 187 r

CII ANGE TITI,I; TER 48, Rev. O, Pressurizer Scaled Reference Leg Level System - UFSAR Figure 7.2-4 Cll ANGE DESCRIPTION Technical Evaluation Report (TER) 48 updates drawings 08700 RK-lC, 6C, and Updated Final Safety Analysis Report (UFSAR) Figure 7.2-4 to reucct the as built configuration of the ,

pressurizer level and pressure transmitter sensing lines.

The vent lines to the condensate pots at the level transmitters upper tap connections actually run '

through the crane wall but are shown on drawing 08700 RK-lC and UFSAR Figure 7.2-4 as ending inside the pressurizer cubicle. Valves and connections to transmitters sharing common sensing lines are not accurately depicted on drawings 08700 RK-lC, -6C and UFSAR Figme "

7.2-4.

$AFETY EVAI,UATION SUMM ARY UFSAR Figure 7.2-4, Pressurizer Scaled Reference 1.eg Level System, was a typical drawing for Westinghouse plants and was not meant to be specific for Heaver Valley Unit 1.

The change involves the level and pressure cc, trol of the pressurizer in the Reactor Coolant System. There are no credible failure modes associated with the change. No failure modes were created that did not exist previously because the figure represents the actual plant conuguration.

The change does not affect the probability of failure of the pressurizer pressure and level sensing lines since the only change was adding a takeoff for a pressure transmitter. There is redundancy with 3 identicalloops.

The performance of pressurizer pressure and level indication is not affected since pressurizer pressure and level are still indicated with the same performance. The change does not alTect design basis r.ccidents in the UFSAR since the revised figure represents the original as built configuration of the plant.

c No Technical Specifications or Technical Specification Basis are alTected by the change.

The change does not involve an unreviewed safety question.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 145 of 187 CII ANGE TITI.E TER 1101, Rev. O, Resolution of SSFE Observation WRS ME-001 for the River Water System VOND/FL OW Discrepancies

[lI ANGC DESCitiPTION Technical Evaluation Report (TER) 1101 responds to Safety System Functio:d Evaluation Program (SSFE) observation WRS ME-001 (and WRS-ME-002 Items 'A" and 'E2'), which had identified inconsistencies among the River Water System flow diagrams, VONDs, and isometric drawings. The TER updated the drawings to add / revise nomenclature and equipment identification.

TER 1101 revised flow diagrams related to the River Water System in order to correct discrepancies between the flow diagrams and the VONDs. In addition. revisions were made tc the flows basci on discrepancies with the isometric drawings. The effect is to produce more accurate drawings, with no adverse elTect on the system. (The flow diagrams have corresponding Updated Final Safety Analysis Report figures.)

SAFETY EVAL.UATION SUMM ARY No changes were made to any plant equipment or system, only to drawings. The drawings changed were those flow diagrams and isometric drawings that describe the River Water System.

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No credible failure modes are associated with the drawing changes. No physical changes were made to any plant equipment. Referring to the TER, it can be seen that most of the changes involved nomenclature, only. Those changes that were made to the description of the configuration of the system have no credible failure modes.

None of the drawing changes had any effect on the probability of failure of the River Water System.

This TER and the resultant drawing changes have no affect on the performance of the River Water System.

River water temperature determines, in part, the time required to depressurize containment and maintain it depressurized afler a pump suction double ended rupture (UFSAR Section 14.3.4.1).

The Recirculation Spray System rejects heat to the Piver Water System following a Loss of Coolant Accident (LOCA)(UFSAR Section 14.3.4.2). The radiological consequences of a LOCA (UFSAR 14.3,5) are not directly affected by the River Water System.

The changes made to the drawings of the river water system, have no afTect on the river water temperature, the capability of the recirculation spray heat exchangers, or the radiological consequences of a LOCA.

UFSAR Section 6.4.3 describes how the River Water System is monitored for leakage by means

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13eaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 146 of I87 l of radiation monitors, and states that defective subsystems are shut down if leakage above allowable values is detected, j

No design basis accidents are initiated by the failure of the River Water System, and there are no l failure modes associated with the change. t The change has not modified the plant response in any way that creates a new type of accident. f The drawing changes made by the TER enhance the description of the River Water System and l the identification ofits components, but does not change the described configuration significantly. .

No failure modes are associated with the change. -

UFSAR Table _9.9-3 Identifies the design basis requirements for river water flow that are the acceptance limits for Technical Specification 3.7.4.1 and bases section 3/4.7.4. This TER does not affect these acceptance limits, since it has no affect on rimr water flow rates.

lleaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 147 of187 Cil ANGE TITI,E TER 4964, Rev. O, Resolution of SSFE Observation RSS-ME 004 and RSS ME 005 CilANGE DESCRIPTION Technical Evaluation Report (TER) 4964 was written to respond to Safety System Functional Evaluation Program (SSFE) observations RSS-ME 004 and RSS-ME 005, which had identified inconsistencies among the Recirculation Spray System flow diagram, VOND, isometric drawings, and operating manual valve list. The TER updated the drawings to add / revise nomenclature, add equipment identification, make minor additions to the piping configuration (added reducers, caps, and plugs), and change valve position indications to reflect the Normal System Arrangement (NS A) positions listed in the operating manual valve list, TER 4964 revised the flow diagram for the Containment Depressurization System (8700 RM.

35A) to correct discrepancies among the flow diagram, the VOND, the isometric drawings, and the operating manual valve list. The effect of the drawing changes is to increase the accuracy of the drawing. There is no adverse effect.

SAFETY EVAI,UATION SUMM ARY "

No changes were made to any plant equipment or system, only to drawings. The drawings changed were the flow diagram and isometric drawings that describe the Recirculation Spray System.

No credible failure modes are associated with the drawing changes. No physical changes were made to any plant equipment. Referring to the TER, it can be seen that most of the changes involved nomenclature, only. Those changes that were made to the . description of the configuration of the system have no credible failure modes. Those changes to the description of the valve position indication, also have no credible Silure modes, since they reflect the NSA position already established in the operating manual valve list. .

None of the drawing changes has any effect on the probability of a previously analyzed accident.

The drawing changes resulting from this TER have no efTect on the performance of the River Water System.

Section 14.3.4.1 of the Updated Final Safety Analysis Report (UFSAR) states that the time to depressurize the containment and the capability to maintain it depressurized below 1 ATM after a pump suction double-ended rupture depends (in part) on the design' of the Containment Depressurization System. The changes made to the flow diagram have no affect on this design. '

Also, the changes have no afrect on the radiological consequences outlined in UFSAR Section 14.3.5.

' No design basis accidents are initiated by the failure of the Recirculation Spray System'(or the Quench Spray System).

Heaver Valley Power Station Unit i 1991 Report of Fccility Changes, Tests, and Experiments Page 148 of187 The change has not modified the plant response in any way that creates a new type of accident.

The drawing changes made by the TER enhance the description of the Recirculation Spray System and the identification ofits components, but does not change the described configuration significantly, The changes made by the TER to the valve position indication on the flow diagram, were made only to reflect the NSA positions t.lready established by the Chapter 13 Operating Manual valve list.

No failure modes are associated with the changes.

UFSAR Section 6.4.2 describes the design basis requirements, for the quench and recirculation spray subsystems, that are the acceptance limits for Technical Specification 3/4.6.2 and 3/4.6.2.2 This TER does not affect these acceptance limits, since it has no affect on the capability of the Containment Depressurization System to depressurize containm:nt (to below 1 ATM)in less than-60 minutes following a LOCA, and since it has no affect on the capability of the system to maintain the depressurization of containment afler a LOCA.

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Heaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 149 of 187 Cll ANGE TITI.E TER 5293, Rev. O, Firo Protection System hfodifications (JI ANGE DESCRIPTION This Safety Evaluation applies to the following System Releases: SR. 268 7, SR 268 9, SR-268-10, SR 268-12 and SR 26814.

SAFFTY EVAI,UATipN SUMM ARY SR 268 7 invo l ves flow switch FS FF-751, additional work per NCR-238 and instrument tubing support modific, tion per Engineering Memorardum (EM) 70455. Flow switch FS FF-751 is a fire protection flow swi:ch that monitors for water flow down-stream of gate valve IFF-757 and provides a signal for annunciator A13 22 (" CABLE VAULT 110SE REEL STATIONS GATE VALVE FLOW"). It ftmetions as a water flow detector by monitoring the gate valve and thus giver remote indication of an active fire protection system in operation at the Cable Vault hose reels so that operator awareness is incret. sed, Updated Final Safety Analysis Report (UFSAR)

Section 9.10, Fire Protection System, was reviewed and it was determined that this portion of TER 5293 will not decrease the effectiveness of the fire protection of these areas. Therefore, the probability or consequences of a malfunction of safety related equipment in the Cable Vault areas will not be increased.

System retcaso SR-268-9 (Cable Vault Door Position Indication) provides electrical supervisors of fire doors used for access to areas protected by automatic total flooding gas suppression systemt This tuodification was rcquired to comply with 10CFR50 Appendix R which states ". .

areas protecud by automatic total flooding gas suppression system shall have electrically supervised self-closing fire doors. . . . Magnetic door contacts with alann features were provided on each of the following doors: Cable Vault stairwell door (CV35 3), Cable Vault double door to Vent Room (CV35-4), West Cable Vault stainvell door (CV35 2) and West Cable Vault door to Auxiliary Feed Pump Room (CV35-5). The electrical supervision is monitored inside the Security Center at the Data Gathering Panel (DGP 3) located in the control room. If a door is opened, a time delay installed in DGP-3 will provide suflicient time for normal access, llowever, if the door is not closed within the time delay period, an alarm'on the fire protection annunciator panel will alert control room personnel. This modification will not decrease the -

effectiveness of the fire protection system but it will increase operator awareness.

System release SR-26510 involves rework in the Primary Auxiliary Building. This system release involved modifications to piping and supp .ts in order to resolve clearance violations and as suc does not impact upon any safety function of the fire protection system.

System release SR26812 (Item 14 and 16) involves work done to the fire dampers in an exi: ting duct in the service bulid ng and also medifications to the oil waste lines for auxiliary feedwater pump room drains. The fire dampers were installed to prevent the spread of a fire throughout the ductwork if there was a fire in the ventilation systent The drains enter an oil interceptor and then a drain tank where a sample pump takes suction and discharges back to the drain tank via a

l lleaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 150 of 187 radiation monitor (RM DA-100). Two pumps in parallel, pump water to the turbine plant oil separator (through normally open valve T%DA.105A). Valve TV DA 10511 shuts automatically on a high radiation signal from radiatbn monitor IW DA 100 and valve TV-DA 105A opens.

This system ensures that any contamisated water is detected and then contained in the tunnel sump rather than being transferred to the river water system. Therefore, the probability or consequences of an accident previously evaluated in the Updated Final Safety Analysis Report (UFS AR) will not be increasc<i.

SR 268-14 involves relocation of smoke detectors above Train A output cabinet. This modiiication was performed in order for easier access to the detector during periods of preventative or corrective maintenance. The function of the smoke detector was not impeded by relocating it, and as such, does not impact upon any safety function of the fire protection system.

The UFS AR was reviewed and it was determined that NRC reporting in accordance with 10CFR 50.59 and updating of the UFSAR are required because the.se modifications will change the system or component as described in the UFSAR (Sections 9.10 and 14). Additional information needs to be added identifying the changes to the fire protection system. Chapter 14 of the UFSAR was reviewed and it is concluded that these modifications will not increase the probability or consequences of any accident previously evaluated in the UFSAR nor will these modifications create the possibility of an accident different than any already evaluated.

The Technical Specifications were reviewed and it was determined that a change to the Technical Specifications is involved (Section 3.3.3.6). Table 3.3-10 needs to be updated to include the new smoke detectors installed during this refueling outage. Technical Specifications (Section 3.3.3.6 and 3.7.14.3) have been reviewed and it was determined that these modifications have no impact on the existing Technical Specifications. Ilowever,it should be noted thtt NRC Amendment 18 states that upon implementation of the modifications the Technical Specifications will be modified to incorporate the limiting conditions for operation and surveillance requirements for the new equipment.

13eaver Valley Power Station Unit 1 1991 Report of Focility Changes, Tests, and Enperiments Page 151 of 137 CII ANGE TITLE TER 5580, Rev. O, Replace River Water Valve RW-580 with a 3/4" Stainless Steel Contromatic Itall Valve Cil ANGE DESCRIPTION This Technical Evaluation Report (TER) proposes to replace River Water *B" header pressure ,

switch isobtion valve RW-580, utilizing a Generic Replacement Design Change (Reference Design Change Package 484). Design Change Package (DCP) 484, entitled Contromatics Ball Valve Specification Upgrade (llVS-210)" states in Rev. 1:

The general evaluation will be performed prior to procurement; the specific evaluation will be requested by Operations Department with an Engineering Memorandum.; the specific evaluation for an identified plant location will be performed with a Technical Evaluation Report by the Nuclear Engineering Department.

Operations requested via Engineering Memorandum 65405 that valve RW-580 be replaced with a Contromatics ball valve. Technical Evaluation Report 5580 proposes to replace a 3/4" carbon steel,600 lb. ANSI class globe valve, manufactured by Pacific Valves, with a 3/4" stainless steel, 150 lb. ANSI class ball valve, manufactured by Contromatics. Based on the six specific (TER) evaluation attributes documented and approved in DCP 484, Rev, 1, which are addressed in TER 5580, the fit, form, and function of the valve remain unchanged. The Engineering 10CFR50.59

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evaluation for DCP 484 concluded that neither the ftmetion nor the configuration will change, and that functions of systems utilizing these valves will not be affected by DCP 484, Replacement of RW-580 will not change the function or system configuration and will correct a leak-through problem.

SAFE'IY EVALUATION Si1MM ARY The proposed change is considered safe and will not involve an unreviewed safety question.

The proposed change will not require a change to the Technical Specifications, Valve RW-580 is not addressed in the Technical Specifications.

The proposed change will require a change to the Upd.ited Final Safety Analysis Report

_ (UFSAR). Figure 9,9-1B shows the old valve type and must be updated.

There are no UFSAR Chapter 14 design basis accidents impaaed by .his TER because the fit, form, and function remain unchanged.

The probability of occurrence for accidents previously evaluated in the safety analysis will not be .

increased. This change does not affect any Chapter 14 analysis, and does not, therefore, affect the :

probabhity of those accidents.

i 13eaver Valley Power Station Unit 1' 1991 Report of Facility Changes, Tests, and Experiments Page 152 of 187 H

The consequences of an accident previously evaluated in the UFSAR will not be increased. This . 1 change is minor and the change to the instrument isolation valve will not affect any other safety j systems or components. l l

The probability of a malfunction of equipment important to safety as previously evaluated in the j UFSAR will not be increased. This change only replaces an isolation valve; a malfunction of the River Water System 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 change will not adversely affect any parameter which .

would increase the consequences of a malfunction. This change will not adversely afTect any l safety system used to . mitigate an accident. Therefore, there should be uo 'effect on the consequences of a malfunction of equipment important to safety, implementation of changes described in this TER will not cause any new credible failure modes because the fit, form, and functions of the equipment have not been altered.

The possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR will not be created. This change is minor, and the new ball valve functions as an isolation  ;

valve similar to the replaced globe valve; therefore, the change is not significant enough to create  !

the possibility for an accident or malfunction of a difTerent type than analyzed in the UFSAR.

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

The Technical Specification index and Specification 3/4.7.4 were reviewed to determine if any.

bases might be affected. It was determined that this change will not adversely affect the margin of - _

safety as defined in the bases for any Technical Specification because the reliability of the isolation valve will be maintained, and no other equipment will be afTected. j a

l w 6 w = nc , . + _ a y- _.

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lleaver Valley Power Station Unit 1 1991 Report of F cility Changes Tests, and Ihperiments Page 153 of 187 Cil ANGE TITI.E TliR 5795, Rev. O, A3 47 Iloric Acid llatching Tank lli Lo and A3-48 Ilonic Acid Ilatching Tank Level Low CII ANGE DESCitiPTIOfi The Boric Acid Ilatching Tank level alarm (Setpoint 20%) is normally lit becs.use the batching procedure completely drains the batching tank aller flushing. The temperature alarm (111-100, LO- 70) can be lit depending on ambient temperature in the Primary Auxiliary Building. The same alarms at Unit 2 were climinated by Design Change Package 1075.

In accordance with the dark board concept of NURl!G 0700, Guidelines For Control Room Design, the above alarms which are normally lit (green) with the unit at power, are to be disabled by maintaining the associated knife switch open.

F AFETY EVAL,UATION SUMM AltY Input to annunciators A3-47 and A3-48 from temperature indicating controller TIC Cll 100 and from level controller LC-Cil-101 respectively will be disabled by maintaining the knife switch open. Local indication will remain unaffecte i. Control of batch tank steam jacket temperature controller TCV-Cll-100 is not changed.

The lloric Acid Batching Procedure 1.7.4J does not refer to the alarms. Acid batching is used only on an intermittent basis with a local operator present. Local indication is used in the batching procedure.

The botic acid batching tank is not a safety related piece of equipment. Local indication is available and used in the batching procedure. The batching operation is not alrected.

The afrected portion of Chemical and Volume Control System (i.e., Boric Acid Batching Tank Cil-TK 5) is not safety related. Disabling the alarms does not change system performance as neither alarm results in a control function.

The entire UFSAR Chapter 14 (Safety Analysis) was reviewed for applicability, particularly Section 14.1.4, Uncontrolled Boron Dilution. Since the batch tank is not safety related, it was not mentioned in the safety analysis.

Disabling inputs to annunciators A3-47 and A3-48 can not initiate a design basis accident because the boric acid batching tank and support components are not safety related.

Ovetall ph response as well as the batching operation, is not afTected by this change. The alTected portion of the Chemical and Volume Control System is not used to mitigate the consequences of an accident.

Disabling the alarms does not affect or,eration of the boric acid batch tank. Also, it neither i

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Deaver Valley Power Station Unit 1 1991 Report of Facili:y Changes, Tests, and Experiments ~

Page 154 of 187  :

changes the failure modes of the equipmc.it involved nor introduces any new potential hazards-

' since affected equipment is not safety related.

No Technical Specification limits are affected by this change. Boric acid batching tank level and -

temperature are not Technical Specification related. Disabling the alarm will have no affect on Boron concentration as local indication is required.  ;

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 155 of187 CilANGE TITI E TER 5797, Rev. O, Loop Fill IIcader Press liigh CII ANGE DESCRIPTION In accordance with the Dark Board Concept of VUREG 0700, Guidelines for Control Room Design, the reador coolant loop fill header pressure high alarm, which is normally lit (green) with the unit at power, will be disabled by maintaining the knife switch open.

The alarm is normally lit due to leakage through reactor coolant loop fill header flow coatrol valve FCV Cil-160 and/or Reactor Coolant System leaking through the loop fill Motor Operated Valves (MOV-RC-556A,B,C)

A similar action was performed at Unit 2 by Design Change Package 1094.

SAFETY EVAL,UATION SITMMARY Input to alarm window A3-60 from the loop fill header pressure transmitter PT-CII 160 will be disabled by maintaining the associated knife switch open. Benchboard indication will remain available. Status of the loop fill header pressure will be available via the bench board indicator.

The afrected portion of Chemical and Volume Control System (CVCS) is not safety related.

System failure probability is not increased as control room indication will remain available. The loop fill header is not safety related - disabling of the alarm will not change the performance of tnis poition of the CVCS.

The entire Updated Final Safety Analysis Report (UFSAR) Chapter 14 (Safety Analyses)'was reviewed for applicability. Since the fill header is not safety related, it was not mentioned in the accident analysis.

The reactor coolant loop fill header is not safety related and the portion between the loop fill MOVs and loop fill header flow control valve FCV-CII-160 is normally isolated at power. The plant response is not changed, since the fill header is not used at power and not used to mitigate the consequences of an accident.

No new unanalyzed typer of malfunction are created as a control system is not affected.

l The change will not affect any limits in Technical Specifications. Reactor coolant loop fill header pressure is not a Technical Specification related item. Disabling of the alarm will not afrect loop fill header flo control valve FCV-Cll-160 stroke time.

Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 156 of 187 i

CII ANGE TITI,E TER 5951, Rev. O, Drawing No. 8700 lW-43A Discrepancy Cll A NGE I)ESCRIPTION ,

Neither plant or procedures are changed by this Technical Evaluation Report (TER). This TER will correct a drawing discrepancy by showing a collection cylinder and associated tubing attached to the oxygen analyzers.

PAFETY EVAI ATATION SilMMAltY The only credible failure is tubing damage and gaseous waste leaks.

This drawing change shows the tubing connection to the collection tank, and does not change system parameters. No new type of accident is postulated.

The waste gas accident considers the rupture of the gas surge tank with the release of its radioactive gas inventories to the environment. This TER does not affect the assumptions and radiological consequences of this postulated accident. ,

o Techmcal Specification 3.11.2.6 requirements will not be affected by the change. Surveillance requirement 4.11.2.6.1 will not be afrected by the change.

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lleaver Velley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 157 of187 Cil ANGE TITI,[i TER 6035, Rev. O, Pipe Section Replacement in River Water System Due to Cavitation Damages CilANGE DESCRIPTION This Technical livaluation Report (TER) replaces a cavitation alTected carbon steel pipe section P'pe class 121) with TP 304 stainks steel material (pipe class 153A). The affected pan ofline 3.0. 14" WR.136 121 will be changed to 14" WR 638153 A.

SAFETY EVAL,UATION SUMM ARY This change affects the make up water line to the Circulating Water System The affected pipe section operating pressure is 70 psi at a tempervure of 87'F, design pressure is 87 psi at a temperature oi 100'F, and the fluid carried is water. The stainless steel material has higher allowances and is resistant to cavitation damages. This change to stainless steel material will increase the reliabili,y of the system.

The Circulating Water System is a non-safety related system and it is independent of emergency core cooling requirements (Reference Updated Final Safety Analysis Report Section 10.3.4.2, Page 10.3-11). The worst possible postulated break in the Circulating Water System is rupture of the main condenser inlet expansion joint with failure of the associated condenser in!ct valve to close.

The circulation water system does not contain radioactive materials.

The change of material for this pipe section will not alter the piping configuration compared to the existing piping. The change of material for the pipe section will not create a new type of accident.

The existing safety analysis will still be bounded by the much larger condenser inlet pipe break.

The change of material for the pipe section will not create a new unanalyzed type of malfunction.

No equipment impo: tant to safety is involved.

There are no acceptance limits in the Technical Specifications which will be changed, as there are no Technical Specifications for the make-up water to the condensate system, and no safety related equipment will be changed.

Ileaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 158 of187 fil ANGE TITI.E wa<

TER 6038, Rev. O, lloron injection Surge Tank 1.evel Indication Cil4NGE Di'.SCRIPTION 13arton level indication switches for the boron injection surge tank, LIS S1964 and LIS S1965 were changed to Foxboro differential pressure level switches. The end result is the same, 2 local indicators for baron injection tank surge tank level, and 2 signals are sent to both high and low level annunciators. The 11arton level indication switches were giving no indication so the system was teplaced.

S A FETY EVAL.lf ATION SilMM Ain' No operating or design parameters were changed except component mark numbers which will be changed under Technical Evaluation Report (TER) 6038. The Safety injection System is the only system affected Failure of any of the new level equipment (transmitters, indicators, and/or switches) is possible but the possibility of failure is not increased by the equipment replacement because liaran level indication switches could and did fail. A failure would include loss of one indicator and loss of a switch or transmitter.

The probability is still the same because eight new pieces of equipment perform in the same capacity as the two old combined level indicating switches. These are Category !! components provided for mdication and would not cause failure of the Safety injection system.

The change improves the reliability of surge tank level indication. This change does not affect (change) any of the original indication parameters and does not affect the system in any adverse manner.

Failure oflevel indication on the boron injection surge tank will not initiate a steam line rupture.

The probability of a Steam L.ine Break accident has not increased.

This instmmentation is provided to monitor boron injection surge tank level, and is required to ,

show that the boron injection tank is operable for an Engineered Safety Features actuation. The new components will not cause any new accident if they fail.

This change involves Category 11 instmmentation which performs the sar e function as the original equipment. There are no new failure modes.

Technical Specification Section 3.5.4.1.1 requires the boron injecti on tank to be operaUe with a minimum contained volume of 900 gallons of borated water (which was the assumption ued in the Steam Line Break analysis). Operating Manual Chapter 11, Section 4K indicates that if the low level boron injection surge tank alarm is received and both local indicators on the surge tank

Ileaver Valley Power Station Unit I  !

1991 Report of Facility Changes Tests, and Experiments Page 159 of187 read s 0,900 gallons in the boron injection tank cannot be assured. ,

i There is no impact on Technical Specifwa*.lons because 2 local indicators have been installed and {

the low level annunciator is still working (no setpoints have been changed).  :

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k M.k 11eaver Valley Power Station Unit 1 Jk 1991 Report of Facility Changes, Tests, and lixperiments ifg Page 160 of 187 Cll ANGE TITI.lj TliR 6249, Rev. O, Reclassify component 1.G CC 100 from Quality Assurance Category I to Non Nuclear Safety Class CII ANGE 1)ESCitlPTION This Technical Evaluation Report (TliR) will reclassify the sight gage instrument (LG CC-100) on component cooling water surge tank CC-TK 1 from Quality Assurance Category 1 (safety related) to non nuclear safety related (NNS)

There is no neel to classify the instrument as safety related because its malfunction would not have an adverse effect on plant operation. The unnecessary classification of components as safety related, places an unnecessary burden on tasks associated with purchasing, servicing and maintaining equipment.

SAFETY EVAL.IIATION SilMM ARY No failures are anticipated, whcre failure is postulated to result from normal operating stress, because of the low system operating pressure. The piping attachments to the tank including the sight gage are seismically designed Gage constmetion is of a formidable design using U bolt and MICA hields around the glass for its entire length. Replacements will iequire seismic qc ficanon to the original specification regardless of the reclassification.

The Component Cooling Water System can be affected through the loss of water from a break.

The reclassification to NNS class will not change the probability of failure as component integrity will be maintained by ketping the original seismic requirement. Testing and maintenance tasks will still be controlled by procedures.

Loss of water inventory from a broken gage assembly will not utfect component cooling operation because the line is small (3/4") and there is adequate makeup water available from the boron recovery supply source (i.e. Primary Grade water). Additionally the location of the sight gage coraection to the tank is above the low level alarm connection point so adequate water is always asered.

Reactor plant component cooling water (CCR) is isolated as a result of a Containment isolation Phase 13 Signal, and other accidents that wera reviewed (including Lose Of Onsite Power) do not take credit for CCR, thus no accidents were identified that couhi be impacted by this change. The Updated Final Safety Analysis Report (UFSAR) notes in Section 9.4.1,I that component cooling water is not used for accident conditions.

There is no technical change to the instrument involved, rather it is an administrative change to simplify procurement, testing and maintenance tasks.

' No failure modes are foreseen that would have a significant affect on safety equipment because of

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! Ileaver Valley Power Station Unit 1-

/ 1991 Report of Facility Changes, Tests, and Experiments j Page 161 of187 .

the small line site. Flooding is not of concern because of drain returns and the low; level tank j alarm within the contret roem to alert operators. e Technical Speci0 cation Section 3/4,7,3.1 requires that at least two primary component cooling water subsystems shall be operable. This administrative change does not affect Technical Speci0 cation acceptance limits. ,

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Ileaver Valley Power Station Unit i 1991 Report of Facility Changes. Tests, and Experiments Page 162 of 187 Cll A NGi-: TITI.1:

TER 6403, Rev. O, Replace River Water System Valves R\ '-357 and RW 359 Cil ANGl: Ill:SCRIPTION This , , sign change will replace the existing 3/4" carbon steel globe valves RW 357 and RW-359 with 3/4" stainless steel ball valves.

The existing globe valves are cortaded and choked with silt. The proposed stainless steel ball valves will minimize corrosion and the ball cabe design wili reduce maintenance time during cleaning.

SAFl:TY l'NAI,1f ATION SilMM AlW There are no new failuie modes introduced by this change.

This change replaces carbon steel globe valves with stainless steel ball valves, and thus will decicase the probability of failure of these valves. This is due to the design characteristic of a ball valve being straight through, while a globe valve would act as a silt trap.

The River Water System valves identified (RW-359 and 357) are the 'lP heades drain valve and the 'C' header pressure indicator isolation valve respectively. The replacement of these valves will not affect the performance of the River Water System, as they are 3/4 inch " Tap. Offs" from the 20" headers.

As this change affects a drain line and a pressure indicator on the liiver Water System, and does not affect the actual performance of the River Water System to supply cooling water to the required components, no assumptions or radiological consequences are affected.

, The failure of these valves will not initiate any design basis acu :,nts.

This change involves the replacement of existing valves with a different style of valve. As such, it will not create a new type of accMt.

No new unanalyzed malftm-tions will t e created by the replacement of these valves.

Technical Sr.ecification 3/4.7.4 identifies the operability requirements for the River Water System.

This change does not impact any acceptance limits that fbrm the basis of the Technical Specification

i Heaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 163 of187 l 1

Cil ANGE TITLE UFSAR 1-9 4, Dypass Valve Around Main Feedwater Control Valve i l

Cil ANGE IESCRIPTION l The proposed change requires valves FCV-FW-479, -489, -499 to be closed above 30 percent i power. These are the bypass valves around the main feedwater flow control valves and provide steam generator level and feedwater flow control during low power operation or hot shutdown when feedwater flow is below 20 percent of design flow.

The proposed change will revise Updated Final Safety Analysis Report (UFSAR) Page 10.315 to ,

indicate that the valves must be isolated above 30 percent power. UFSAR Page 10.3 27 will be revised to reference the letter from Duquesne Light Company to the Nuclear Regulatory ,

Commission on the subject. UFSAR Figure 10.3 5 will be revised to show the valves closed.

When a reactor trip occurs feedwater control valves are closed and auxiliary feedwater flow is initiated. Flow continues into the main feedwater piping downstream through the bypass valve.

Continued flow through the bypass valve combines with relatively cold auxiliary feedwater flow to cause thermal stratification in the downstream main feedwater lines. This stratification in the ,

I main feedwates lines can cause increased stress levels in the main feedwater piping and supports.

Therefore, the bypass valve around each main feedwater control valve must be isolated above 30 percent power to prevent the possibility of increased stress levels resulting from thermal stratification in the downstream main feedwater lines.

SAFETY EVAL,UATION SUMM ARY The change will eliminate the possibility of bypass flow around the main feedwater control valves at power levels greater than 30 percent. No design parameters are afrected by the change. The main feedwater system is affected by the changec Only the portion of the main feedwater system be&xn the steam generators and the containment isolation valves outside containment is reqwred during shutdown of the unit (Reference UFSAR Sectic,n 10.3.5.1.1).

The change will ensure that the affected portion of the main feedwater system (See Above) will not be subjected to thermal stratification resulting trom concurrent main feedwater and auxiliar, feedwater flow in the event of a reactor trip. Therefore, the change will decrease the probability of failure of the piping due to thermal stratification. There are no credible failure modes associated with the change that afrect the capability to safely shutdown the plant.

The performance of the affected portion of the main feedwater system will be unchanged.

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13eaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experime.its Page 164 of 187 The following design basis accidents were reviewed for potential impact by the change:

Section Iille 14.L7 Loss of External Electrical Load and/or Turbine Trip 14.1.8 Loss of Normal Feedwater 14.1.9 Excessive lleat Removal Duc To Feedwater System Malfunctions ,

14.1.11 Loss of Offsite Power to Station Auxiliaries (Station Blackout) 14.1.13 Accirlental Depressurization of the Main Steam Systern 14.2.4 Steam Generator Tube Rupture 14.2.5.1 Major Rupture of a Main Steam Pipe 14.2.5.2 Major Rupture of a Main Feedwater Pipe 14.2.7 Single Reactor Coolant Pump Locked Rotor 14.3.1 Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates Emergency Core Cooling System 14.3.2 Major Reactor Coolant System Pipe Rupture The accidents identified above assume continued feedwater flow or result in a safety injection signal (or high steam generator water level signal) to close the main and bypass feedwater control valves.

For the case where continued main feedwater flow is assumed (UFSAR Section 14.1.7, Turbine Trip), the proposed change to isolate the bypass valve around each feedwater control valve will not prevent (or inhibit) the system from delivering the design flow. For the case where the main and bypass feedwater control valves receive a signal to close, the proposed change to isolate the bypass valve will have no effect since the analysis assumes the valves are closed. The plant response will not be changed, and there are no failure modes arsociated with the change.

Based on the above, this change will have no affect on the perfbrmance of the main feedwater system, or on the assumptions and radiological consequences of the identified accidents.

There are no failure modes associated with the change. Since there are no failure modes associated with this change, the probability of occurrence of the design basis accidents identified will be unalrected.

l l Since no equipment or operating parameters described in the Technical Specifications or accident L_ analysis are affected, no acceptance limits or Technical Specification Basis will be affected.

! e The following li:,t provides references to the location of information used for the safety.-

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a. UFSAR Section 10.3.5.1.1, Condensate and Feedwater Systems (Design Basis).

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! b. UFSAR Section 10.3.5.2.1, Condensate and Feedwater Systems (Description).

c. Other UFSAR Accident Analyses as listed above.

l Based on the infonnation discussed above it was determined that the change does not involve an unreviewed safety question.

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lleaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 165 of 187 Cil A NGE TITI.E UFSAR l 9 27, Re-calculation of Minimum River Water Flow Regoirements C_II A ngl', l> ESCRIPTIO N The UFSAR is being changed to provide infonnation consistent with the latest analyses.

Calculations were performed to clarify the minimum river elevation and determine the minimum river water flow to ensure adequate coohng of the following:

1 Charging Pump Lubricating Oil

2. Diesel Generators, and

3. Control Room it was determined that the charging pump coolers are able to perform their safety function with 20 gpm of river water flow (previously 55 gpm), a maximum of 20 tubes plugged, and a 90 F river water temperature. The minimum river water flow requirements for the diesel generators is being changed from 400 to 260 gpm. The control room river water backup cooling coil is being credited instead of the control room air conditioning unit for control room cooling during accident conditions with a minimum flow requirement of 100 gpm. The Technical Specification minimum river elevation of 654 ft is used for the minimum Design Basis Accident (DBA) flow analyses instead of 648.6 ft The maximum control room temperature using the backup coohng coils is

( changed from i13 to 120'F.

SAFETY EVALUATION SUMM AltY Minimum flow requirements for the charging pump coolers, diesel generator 3 and the control room backup cooling coils have been re-calculated using mose realistic measured heat loads. The minimum river elevation used for DBA flow analysis is changed to reflect assumed failures and initial conditions.

The charging pump coolers are able to perform their safety function with 20 gpm of 90'F river water flow and a maximum of 20 tubes plugged. Operation, operating characteristics, and failure modes for the charging pump lubricsting oil coolers (Cil-E-7A, -7B, and -7C) are not alTected by the flow requirement reduction. Failure modes of the coolers remain the same. A change in the minimum flow requirements for diesel generate and control room coolers will not increase the probability of failure for these compontnts or create new faiiure modes.

There is no efrect on the performance of safety systems since the cooling provided is adequate to meet design requirements. As stated above, the changes do not affect the performance or failure modes cf the identified components and thus do not increasc the prchability or consequences of an accident or malfunction of equipment as previously evaluated in the UFSAR.

Since the changes do not create new failure modes and the changes have no efTect on the minimum performance requirements of safety systems, the plant response has not been changed

13eaver Valley Powcr Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments l Page 166 of187 l to the point where it can be c'onsidered a new type of accident or malfunction. Thus, the change does not create the possibility for an accident or malfunction of a different type than any evaluated previously in the IJFSAR.

Technical Specification requirements and associated basis for charging pump and diesel generator operability will continue to be met. Also, Technical Specification requirements and associated basis for control room equipment environmental q talification and habitability will continue to be thet.

Based on the safety evaluation summarized above it was determined that the changes do not involve an unreviewed safety question.

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Beaver Valley Power Station Unit 1 1991 Repon of Facility _ Changes, Tests, and Experiments ,

Page 167 of 187 f_II ANGE TITLE UFSAR l-9 30, Emergency Diesel Generator Fuel Sueticn Strainer Description CII ANGE DESCRIPTION The Updated Final Safety Analysis Repon (UFSAR) text is being revised to clarify the description; of the Emergency Diesel Generator Fuel Oil Suction System. The discrepancy was identified in Safety System Functional Evaluation Ob, aation EDS-ME-037 and NRC Electrical Distribution System Audit Task Number 132.

Section 9.14.4.1.3 of the UFSAR will be changed to indicate that there is a single fuel suction strainer per emergency diesel generator, in place of the existing wording which indicates that there are two fuel suction strainers per diesel generator.

S AFETY EVALUATION SUMM ARY There are no credible failure modes associatcd with the change since the change is administrative 4

in nature and the physical configuration remains as originally designed. A single strainer per train -

is adequate to provide clean fuel to each emergency diesel generator.-

The change involves the fuel oil subsystem of the emergency diesel generators. There is no physical change to the installed system, therefore, there is no effect on the probability of failure of the system. The application of single failure redandancy criteria to the fuel oil system means that failure of a single strainer will not prevent use of the redundant train. Since there is no physical chaage involved, there will be no effect on the performance of the emergency diesel generator fuel oil system.

Loss of Offsite Power Due To Station Blackout is tha cc _ design basis a::cident reviewed for potential impact by the the change. The fuel oil strainer is in-line between the fuel oil. storage tank and the fuel transfe. pump that supplies fuel oil to the diesel. The strainer provides added reliability for diesel operation. There are no direct radiological consequences in the event or a Loss of Offsite Power accident. No design basis accidents are initiated by the failure of the emergency diesel generators.

Technical Specification Section 4.8.1.1.2 specifies ~ that each ' diesel generator shall be- 4 demonstrated operable at least once per 31 days. One of the surveilled attributes that assures operability is the sampling and testing of fuel oil from the day tank. This change does not afTect Technical Specification acceptance limits. A single strainer and filter per diesel generator train: 1 assures adequate fuel oil quality. Based on the above, the change does not reduce the margin of ,

safety as defined in the basis for any Technical Specification.

Based on the safety evaluation rummarized above it was determined that the changes do not involve an unreviewed safety question. l l

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Baa'cer Valley Power Station Unit I l')91 Report of Facility Changes, Tests, and Experiments Page 168 of 187 CilANGE TIT 12 UFSAR l-9-46, Add . :ference to Electrical Calculation Program in Place of Battery Duty Cycles and DC laad List Cil ANGE DESCRIPTION The Updated Final darety Analysis Report (UFSAR) text is being revised to reference an electrical calculation program e tablished to control DC load changes. The DC loads and duty cycles are shown in controlled calculations which are updated, as required, every refueling outage.

The battery duty cycles (UFSAR Figures 8.5-9, -10, -11, and -12) and the DC load List (UFSAR Table 8.4-3) will be deleted from the UFSAR.

SAFETY EVAL,UATION SUMM AlW The electrical calculation program and associated calculations ensure the capacity of the batteries continues to be adequate to power the prescribed loads. Operating and design parameters associated with the DC power system remain unchanged (except the battery aging factcr which is being revised to comply with IEEE Std. 485-1983).

Calculations are resised as required to assure any load changes are evaluated. The DC loads are presently listed in both the calculation program and Operating Manual. Transferring information from the UFSAR to the calculation program will have no effect on the ability of the batteries to supply the DC loads for the required two hour duty cycle. Ther will be no change in the operation or performance of the DC system since the batteries will continue to have adequate capacity to supply their two hour duty cycle.

Deleting the duty cycle, etc, frc m the UFSAR has no effect on the probability of failure in the DC system because all the information being deleted will be maintained in ' calculations. The calculations show that the batteries are capable to supply their two hour duty cycle, even when a cell is jumpered. Since no new failure modes have been identified, the probability of occurrence for an accl. dent or malfunction previously evaluated in the UFS AR is r,at increased.

Since there is no change in the performance of the system, the consequences of an accident or malfunction previously evaluated in the UFS AR are not increased.

Since the performance of the system is unchanged and no new failure modes have been created, the possicihty for an accident or malftmetion of a different type than any evaluated previously has not bec a created.

No Technical Specification acceptance limits or Technical Specification basis will be affected by this change. Calculations will shov. that the batteries have the capacity to meet the two hour duty cycle and there will be no change in the operation of the DC system.

Based on the safety evaluation summarized above it was determined that the changes do not involve an unreviewed safety question.

r Deaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 169 of 187

. CII ANGE TITLE UFSAR 1 9-47, Fuel Transfer Tube Containment Isolation Test Requirements CII ANGE DESCRIPTION This change will clarify the Updated Final Safety Analysis Report (UFSAR) description of the containment isolation anangement and tes' requirements for the fuel transfer tube. The fuel transfer tube connects the containment and fuel building. Specifically the change will eliminate test requirements for the fuel transfer mbe welds and gate valve.

The Unit 1 ncorrectly updated to include a description of construction phase testing of the fu1 n m. velds and gate valve as provided in the Duquesne Light Company -

respor o NP 6.5. Construction testing included provisions for testing the leak tightnu . and liner welds during construction and an air pressure test to ensure the stru ac containment. During the operating phase, containment leakage rate tests are .:ordance with the guidelines of Appendix J of 10 CFR 50, " Primary Reactor t akage Testing for Water Cooled Power Reactors."

The conta nment leakage testing program includes the performance of Type A tests, to measure containment overall integrated leak rate, Type B tests, to measure leakage of certain containment components, and Type C tests, to measure containment isolation valve leakage rate.- The joint between the blind flange and the fuel transfer tube in the reactor containment is Type B tested to identify any leakage. Leakage from fuel transfer tube welds throughout the containment liner, would be identified during Type A testing.

The fuel transfer tube gate valve, located in the fuel building, is not tested since it is not considered a containment isolation valve. Operating license change request No.160/20, dated April 23,1990, adds the gate valve to the Technical Specification containment isalation valve Table 3.6-1. Table 3.6-1 will indicate that the valve is' listed for information only and 'is not _ r required to be Type C leak tested. In addition, other plants were contacted to determine the testing that they perfonned on the fuel transfer tube. It was found that all the plants, of similar design, Type B tested the flange inside containment. The valve outside containment is not addressed in their UFEAR or Technical Specifications and is not tested.

SAFETY EVALUATION

SUMMARY

Fuel transfer tube weld or gate valve leakage are the c .redible failure modes identified. There are no credible failure modes that affect operations or t... apability to safely shutdown the plant,-

Information contained in the following UFSAR Sections describing the! containment isolation system, containment leakage testing, and design basis accidents were reviewed for the evaluation.

5.3, Containment Isolation System, 5.6, Containment Tests and Inspec* ions, 14.3.1, Loss of Coolant Accident - Small Break LOCA Evaluation Model, and 14.3.4, LOCA - Containment Evaluation.

o Beaver Valley Power Station Unit 1 -

- 1991 Report of Facility Changes, Tests, and Experiments Page 170 of187

'It was determined that the performance of safety systems or systems important to safety were not--

affected by the change. Only the description of the fbel transfer tube containment isolation -

arrangement and test requirements in the UFSAR was affected.

The change will have no impact on the assumptions and radiological consequences of the design basis accidents reviewed. The failure modes associated with the change cannot be an initiating event for the design basis accidents or any new type of accident.-

Therefore, the proposed change wiU not increase the~ probability of an accident or malfunction previously evaluated in the UFSAR.

The plant response will be unchanged,0 rings associated with the blank flange at the end of the transfer tube will provide the necessay containment isolation functioni Containment integrity will continue to be demonstrated by Type A, and Type B containment leakage testing.

Fuel transfer- tube weld or gate valve leakage will not create a new type of accident or malfunction. The blank flange prevents leakage through the transfer tube in the event containraent isolation is required. Thus, eliminating gate valve leakage is- not required for -containment .

isolation.

Should fuel transfer tube weld leakage develop, it would be identified by Type A containraent leakage rate testing and corrected.

The following Technical Specifications were reviewed:

- 3/4.6.1, Primary Containment Integrity, and -

3/4.9.4, Containment Building Penetrations.

Since the change will not affect equipment operability _or surveillance requirements, the basis ~ for the Technical Specifications v ill be unaffected, Based on the safety evaluation summarized above, it was determined that no unreviewed safety question is involved.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments-Page 171 of187 CII ANGE TITI,E UFSAR l-9-48, Control Room hQ Values Cll ANGE DEScitIPTION This change revises the LQ values provided in the Updated Final Safety Analysis Report (UFSAR) for use in control room habitability analyses. The values currently in _use were determined using methodology shown to be less than optimum. NUREG/CR-5055 provides an improved methodology.

The development of new, less overly-conservative LQ values was initiated as a corrective action to address problems with the Unit 2 locked rotor accident analysis performed in support of the a-Unit 2 fuelload. Analyses assuming the previous values showed control room doses in excess of 10 CFR 50 Appendix A, GDC 19 criteria. Although the immediate need was to address Unit-2 concerns, new values were determined for Unit 1. These values will be used for all control room habitability analyses performed in the future. Existing analyses will not be revised unless a specific need arises.

The new LQ values for the significant release points differ from the previous LQ values by factors of about 10 to 40.

SAFETY EVAL 4UATION

SUMMARY

No safety systems or systems important to safety are directly affected by this change to the-UFSAR. This change modifies parameters that are used in safety analyses performed pursuant to -

other structure, procedure, or test changes. The LQ values to be changed are used in performing safety analyses, and therefore, do not have an impact on the performance of existing plant systems.

All UFS AR Chapter 14 accidents resulting in a radioactivity release are potentially affected by this change. The revised LQ values would reduce the magnitude of the consequences for postulated accidents due to the change in analysis inputs. However, the revised values h:,vc no impact on the actual consequences that could occur as this change does not modify plant systems, structures, components, procedures or tests.

It could be said that an underpiediction of the postulated control room doses could result in a condition under which actual doses during an accident could adversely affect control. room habitability and, therefore, the' operators' ability to respond to an accident. However, the revised methodology was shown to have a similar tendency toward underprediction as the' previous methodology. Thus, there is no discernible increase in the postulated consequences.

There are no failure modes associated with this change other than an underprediction of the actual-

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Page 172 of 187 imeteorological conditions As noted previously, the tendency for under prediction is roughly -

equivalent in both methodologies.

The vilues affected by this change are used in _ the assessment of postulated accidents to demonstrate control rooni habitability, This change cannot affect the probability of occurrence for-an accident or create a new type of accident or malfunction.

Technical Specification 3/4.9.15 addresses control room habitability system requirements. These systems are intended to maintain post-accident control room doses to values less than 10 CFR 50 Appendix A, GDC 19 for the duration of the accident. The Basis for this Technical Specification does not specifically address X/Q values.

Lower postulated X/Q values have the elTect of increasing the margin between the postulated ~

consequences and GDC 19. Future modifications to systems, structures, components, procedures : .

or tests enabled by the increased margin could again reduce the margin. Should the new -

methodology underpredict the X/Q values, these subsequent modifications could have the affect of decreasing the margin below that which currently exists. However, as noted previously, the tendency for _underprediction is equivalent in the new and old methodologies. Thus, it is concluded that there is no discernible reduction in the margin.

Based on the safety evaluation summarized above, it was determined that no unresiewed safety question is involved, w

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l Beaver Valley Powcr Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 173 of 187 CII ANGE TITLE UFS AR l-9-54, Charging Punip Flow Requirements CII ANGE DESCillPTION The maximum runout flow for the safety injection charging pumps is being changed from 550 gpm to 560 gpm. The minimum charging /high head safety injection (HilSI) mp curve in the UFSAR is being changed to a nominal representative curve. The minimum rformance curve (MOP) for the charging pump will be maintained in the appropriate Operatir . Surveillance Test (OST).

A 5 gpm tolerance on simulated sealinjection flow and 11HSI branch line throttling could result in a maximum pump flow of 560 gpm during performance of OST 1.11.14 or under accident conditions. The MOP curve is being removed from the UFSAR to avoid duplicate reviews and changes each time the curve is revised.

SAFETY EVAL,UATION

SUMMARY

Normal pump operation is not affected by the change. Pump design parameters of maximum flow, rtmout current draw, and minimum available net positive suction head (NPSil) are affected by this change. The system affected by the change is the Chemical and Volume Control System (CVCS).

The only failure modes identified are pump failure caused by motor overload or inadequate NPSH. The increased maximum flow has been evaluated by Westinghouse and the pump manufacturer. It was determined that this change does not increase the probability of failure.

Adequate NPSH is available at the higher flow based on Duquesne Light calculations.

The change will not afTect the analyzed performance of the CVCS. Flow tolerances are included in the system analyses.

The small break Loss Of Coolant Accident (LOCA) and Main steam line break accidents were reviewed for potential impact by the change. The change does not affect the assumptions' or radiological consequences of the accidents reviewed. The increased maximum flow accounts for the tolerance in the test method, and the system performance assumptions in the analyses are not affected.

Failure of a charging pump is not an initiating event for any accident. The change does nos increase the probability of occurrence or consequences of an accident or malfunction previously evaluated in the UFSAR. Single failure of a HHS1 pump has been analyzed for all relevant accidents. This change will not increase the probability of failure or create any common mode failure condition.

Failure of a HHSI pump due to motor overload on NPSH deficiency is among the postulated pump failure conditions. However, this change will not increase the probability of these failure j l

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Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 174 of187 l modes.

The acceptance limits in the accidents identified above include Peak Clad Temperature less than 2200 F for the Small Break LOCA, and Departure from Nucleate Boiling Ratio greater than 1.3 for the Main Steam Line Break accident. This change has no effect on analyzed accidents since flow tolerances are included in the current system performance analysis. The change does not reduce the margin of safety as defined in the basis for any Technical Specification.

BasJ on the safety evaluation sununarized above, it was detennined that no unreviewed safety question is involved.

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Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments ,

Page 175 of187 I

CII ANGE TITI.E UFSAR l-9-56, Addition of Supplementary Leak Collection and Release System Heat Removal Function to the System Description CIIANGE DESCRIPTION Updated Final Safety Analysis Report (UFSAR) Section 6.6 describes the Supplementary Leak Collection and Release System (SLCRS). Currently the UFSAR states that the System'r purposes are to maintain a negative one-eighth inch water gauge pressure in containment contiguous areas (to prevent radioactive leakage), and to filter contaminated air. This proposed change will add a description of another system function: removing heat from areas in the Auxiliary Building (Charging Pump Cubicles) and Safeguards / Cable Vault following any design basis accident (resulting in a Containment Isolation Phase B - CIB) coincident with a loss of ofTsite power. In -

such an event, tSe normal air conditioning systems (VS-AC-7, 8, 9, ll A and llB) no longer operate, leaving the SLCRS as the only means ofmaintaining area temperatures.

An internal investigation following the September 1991 discovery of a blocked off West Cable Vault SLCRS duct, led to the conclusion that the SLCRS was not conigured as originally intended. An October 1991 evaluation revealed that the SLCRS ventilation flow rates for the Charging Pump Cubicles were inadequate to maintain the temperature of the pump motors below their environmental qualification limit of 123 F. As a result of these evaluations, it was concluded that the design basis of the SLCRS had to be reconstituted, especially since most of the flow rates assumed in design basis documentation (Drawing 8700 RB-2B) had not been verified by testing -

. since they were established by calculations performed in the 1977 to1983 time frame.

An NRC inspection (50-334/92-02) conducted in January,1992, concluded that Beaver Valley Power Station had been lacking sufficient design control over the SLCRS and had not been adequately verifying all design bases. The auditors agreed that incorporating the heat removal function of the SLCRS in the UFSAR was a good idea to enhance personnel knowledge of this-  ;

design basis.

SAFETY EVALUATION SUMM ARY The minimum required air flow rate in certain areas exhausted by the SLCRS is affected by the change. This change involves an administrative requirement,_ rather than an actual physical change. The performance of SLCRS, remains the same, realignn.ent to the filtered pathway on the same Engineered Safety Features (ESP) and. high radiation signals as previously, and maintaining the same 36,000 t 10 % cfm flow rate required by Technical Specification 3.7.8.1.

The ability of the SLCRS to achieve negative pressure requirements will remain the same, since the heat removal requirements envelop them. The distribution of area flows may be altered, and will be verified by testing in their new pattern.

No true physical failure modes are introduced by this change, whicl' is an administrative or technical one that represents a new understanding of the SLCRS design / licensing basis. - SLCRS

Beaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Exceriments Page 176 of 187 will continue to operate, as stated above, in its previous manner. There is now a possibility that the various area flow rates will occasionally be found below the limit established in the new suiveillance requirements of test IBVT2.16.5; however, the flows must be corrected upon discovery, and will be verified on an 18-month frequency.

The following list provides references to the location ofinformation used in the safety evaluation.

UFSAR Section 6.6,14.2.5,14.3 Unit 1 Operating Manual Chapter 16, "SLCRS" Udt 1 Design Basis Document 16, "SLCRS" Incident Report 91-74/LER 91-32 NRC Inspection Report 50-334/92-02 and Notice of Violation Engineering Memorandum 102081 The primary system explicitly affected by the change, as stated above, is the SLCRS. Systems implicitly affected are those with environmentally qualified equipment located in areas where the SLCRS provides exhaust flow, including: 480 V distribution, Hydrogen Control, Auxiliary Feedwater, Safety Injection (Low Head SI & High Head SI), Recirculation Spray, Containment Isolation, and Reactor Vessel Level Instrumentation System.

As stated above, no physical changes are occurring as r. result of this change. All currently required SLCRS performance parameters will be verified as before. This change, by reflecting the requirement that SLCRS remove heat from the safety-related systems named above, will ensure the proper post-accident operation of identified systems by ensuring that their environmental qualification (10CFR50.49) limits are met during accident situations.

As stated above, overall SLCRS performance is unchanged, but the distribution of SLCRS flows may be affected. The change documents the heat removal requirements of the SLCRS in the design basb This information will provide background for the sutveillance test program that has been established to ensure the SLCRS meets these requirements for the remainder of plant operating life As with the SLCRS, the immediate operating performance of the safety-systems employing the St.CRS for post-accident ventilation is unchanged The proposed documentation changes will assist in assuring that the systems' ability to mitigate Design Basis Accidents (DBAs) is not compromised by excessive temperatures.

The DBAs reviewed are those that would result in a CIB, in conjunction with a loss of offsite power (LOOP); namely, a loss of coolant accident (UFSAR Section 14.3) and a main steam line break (UFS AR Section 14.2.5)

No physical system performance parameters are affected by the change. Since the systems function as before,- and their post accident operability will be assured by the program improvements being implemented, the affected accidents will proceed as described, will be mitigated in the same manner, and will thus have the aame radiological consequences.

This change, admir.istrative ia nature, does not create any DBAs, since it does not deal in any 1,

z Beaver Valley Power Station Unit lj

_ 1991 Report of Facility Changes, Tests, and Experiments ?

. Page 177 of 187_'

manner with systems or components involved in the initial trensport of main steam ~or'_ reactor-coolant.- The design requirements identified involve __the performance of systems that mitigate a DBA, The proposed change does not increase the probability of occurrence or consequences of an-accident or malfunction previously evaluated in the UFSAR.

This change does not create any new failure modes, and is meant to support establishment.and verification of design basis requirements for the SLCRS. Thus, the plant response continues physically as before, but will now have controls designed to assure more effective operation..

Tids change does not physically affect any equipment, except to ensure adequate cooling will' be programatically maintained. Therefore it does not create any malfunction mechanisms, and the systems op.. ate as before.

No accident of a ditTerent type than evaluated previously in the UFSAR is created. This change reflects administrative program enhancements that relate to the performance of safety.related.

mitigation equipment, and- does not involve any system or plant function that can act as an accident precursor.

No new type of malfunction is created, since as indicated throughout this evaluation, no physical system or equipment characteristic is being altered. Assuring adequate ventilation flows in all:

situations will help prevent equipment malfunctions.

No Technical Specification limits or bases are explicitly affected by this change. Total flow rate __

of 36,000 i 10 % cfm, as identified in Technical Specification 3.7.8.1, remains the same. . Its basis -

is related to radiological concerns and _is also unchangedc No explicit operability criterion is -

changed. An ~ implicit criterion of safety-related equipment operability, namely Environmerital Qualification, is enhanced. The proposed activity does not reduce the margin of safety as defined -

in the basis of the Technical Specifications.

-Based on the _ safety evaluation summarized above, Lit was determined that no unreviewed safety question is involved.-

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l- Derver Volicy Power Station Unit 1 1991 Repost of Facility Changes, Tests, and Experiments Page 178 of187 CII ANGE TITLE UFSAR l-9 57, Remove The East and West Cable Vaults From The List Of Areas Served By The Supplementary Leak Collection and Release System To Maintain Negative Pressure Cil ANGE DESCRIPTION UFSAR Section 6.6 describes the Supplemental Leak Collection and Release System (SLCRS).

This system is, as stated in the Updated Final Safety Analysis Report (UFSAR) text, designed to maintain negative one-eighth inch water gauge pressure in most areas contiguous to the reactor containment. This proposed change will remove the east and west cable vaults from the list of contiguous areas served by SLCRS. UFSAR Section 6.6 and Figure 6.6-1 will be revised to clearly indicate the current system requirements for the SLCRS. The change can be considered an administrative one to a design basis document, and has no actual physical effect on any system or compcnent.

An analysis performed by the Radiological Engineering Section concludes that a negative one-eighth inch pressure is not required to contain radioactive effluents in the cable vault areas of Unit 1 In fact, the NRC Safety E*,aluation Report (SER) for UFSAR Section 6.6 (SER Section 6.2.3) indicates that no credit could be taken for SLCRS as regards dose evilection 5 these areas.

Therefore, the 10CFR100 offsite dose limits were still met assuming 100% unfiltered leakage from containment contiguous areas that do not contain ESF components pas.ing recirculating fluids from the reactor containment sump. The negative ditTerential pressure criterion is thus being removed because it is an unnecessary surveillance restriction.

SAFETY EVALUATION S11MM ARY The parameter affected by this change is the requirement that the SLCRS be able to establish negative one-eighth inch water gage ditTerential pressure in the east and west cable vaults. This parametec can be considered a design criterion, which is verified by a surveillance test. Actual sysiem performance is not altered, however.

Heat removal in order to meet equipment environmental (temperature) qualification limits is considered to be a design basis criteria for the SLCRS. However, flow requirements to satisfy heat removal criteria are established independently of flow requirements to satisfy negative pressure criteria; thus, elimination of the cable vault differential pressure requirement in no way impacts heat removal. The west cable vault area has been determined to have no adverse temperature effects due to a lack of SLCRS ventilation.

No new failure modes are introduced by elimination of the cable vault differential pressure requirement, which is a stated design basis. The actual perfonnance of the system is not altered directly. The potential exists, as a result of the relaxation of this requirement, for the negative pressure in the cable vaults to eventually degrade; however, as stated above, no cut is taken for this parameter in post-Loss Of Coolant Accident (LOCA) dose release analyses.

i Beaver Valley Powc St: tion Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page 179 of 187 The following list provides references to the location ofinformation used in the safety evaluation.

UFS AR Sections 6.6.1,6.6.2,6.6.3, and 14.3 BVPS Letter ND3SHP: 1482, dated 24 October,1991 West Cable Vault Ventilation BVPS Unit 1 Calculation 3700-DMC-2652, dated (1/20/92)

The safety systems or systems important to safety afTected by the change is the Supplementary Leak Collection and Release System (SLCRS).

This change, which only afTects a design basis parameter that was used as an acceptance criterion for SLCRS suiveillance testing, does not phycically affect any equipment and thus, introduces no new failure modes.

As stated before, the SLCRS will continue to perform as before, although the SLCRS is not necessary for containing radioactive leakage from the cable vaults. Should degradation occur in the negative pressure of the east and west cable vaults, no effort vill be made to monitor or restore it. Ilowever, heat removal flow requirements will be established as new, separate design basis criteria. Since some amount of air removal will still be required to meet temperature criteria, some degree of negative pressure will always exist.

Unit I does not take credit for the collection of containment penetration leakage by the SLCRS.

Releases from Emergency Core Cooling System equipment and piping circulating containment sump water, as well as fuel building release;, are considered to be collected; however, the performance of the C1.CRS in those areas is not affected by the proposed change.

Elimination of the requirement to maintain negative one-eighth inch water gauge pressure will not initiate any accident discussed in UFSAR Chapter 14. The change only affects performance requirements of accident mitigating equipment.

No design basis accidents have been afTected. The probability of_ occurrence for the LOCA, an accident the SLC sS is designed to help mitigate, is totally independent of the SLCRS's ultimate performance The proposed change does not increase the probability of occurrence or consequences of an accident or malfunction previously evaluated in the UFS AR.

No new types of accidents have been created. The plant response to a Design Basis Accident will remain the same, since the analysis for a LOCA did not take credit for the collection of penetration leakage. The lack of a negative pressura requirement for the east and west cable vaults is a performance criterion, not a physical equipment change, and so will not result in the failure of any system to perfomt its safety function .

No new types of malfunctions have been caused by thia change, which is administrative rather than physicalin nature. Any equipment qualification issues (involving equipment failure due to an implicitly reduced flow rate requirement) will be addressed in a separate safety evaluation.

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. Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments :  ;

- Page 180 of 187

-The proposed change does not create the possibility of an accident or malfunction of a different

- type than previously evaluated in the UFSAR..

Technical Specification 3.7.8.1 only deals with total flow through thelSLCRS filter banks, and--

does not address area differential pressure. Thus,- no explicit Technical Specificatic,a criteria or bases are affected. 'The only pieces of equipment that are_ im . citly affected by the change are -

those located in the cable vaults: 480 V emergency motor control centers ES, E6, El1, E12, & _

E14 (Technical Specification 3.8.2,1), and the hydrogen. analyzers (Technical Specification 3.6.4.1). Although the operability requirements for this equipment is supported by the hiat removal capability of the SLCRS, these requirements are separate from the differential pressure <

requirements. The licensing basis requirement for the SLCRS is t'ne offsite dose limitation of 10 CFR 100. The margin between the ofTsite dose calculated for design basis accidents and the limit of 10 CFR 100 is not reduced by this change, since no credit was assumed for the SLCRS in the

~offsite dose calculations..

Based on the safety evaluation summarized above, it was determined that no unreviewed safety .

question is involved.

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Beaver Va' / Power Station Unit 1

.. 1991 Report of Facility C.enges, Tests, and Experiments Page 181 of187 CII ANGE TITLE UFSAR l-9-58, WCAP 12966, 20 Percent Steam Generator Tube Plugging Analysis Program Engineering and Licensing Report Cil ANGE DESCIllPTION WCAP-12966 justifies operation at full power with as many as 20 perce it of the tubes plugged in any one steam generator as long as the thermal design flow (TDF) limit is met.

The previous justification (WCAP-11591 dated September,1987) for 10 percent tube piugging was outdated vased on the number of steam generator tubes plugged during the 8th refueling outage.

SAFETY EVAL,UATION

SUMMARY

WCAP 12966 indicates that steam generator steam temperature, pressure and steam flow will decrease as a result of the decrease in steam generator heat transfer surface area. Systems affected by this change are the Reactor Coolant System, the Main Steam System and the Feedwater System.

A leaking plug or failure of the mechanical seal are credible failure modes associated with the tube plugging operation. This change, however, is to increase the allowable number of tubes which may be plugged. There is no new failure mode introduced by this change.

An assessment of system integrity concluded that component integrity is expected to be maintained. No new performance requirements are being imposed; operation does not significantly impact the design transients and thus does not challenge the d: sign of these components.

For Nuclear Steam Supply System primary components, since the thermal design flow is maintained, and the reactor coolant temperatures associated with the proposed design condition will remain unchanged, the proposed 20 percent tube plugging will not efTect the existing design basis analysis n indicated in WCAP 12966. Section 6.3.3.3 indicates that the reactor coolant pump motor rotor windings temperature rise due to this change,_could result in an accelerated level of mechanical aging which could result in rotor windings failure. This will not afTect the safety related ftmetion of the reactor coolant pump motor.

For the Main Steam System, full power mass steam flow rate will be slightly less than for the current operating conditions and the full power volumetric steam flow rate will increase by approximately 4 percent. These will have no adverse efTects on the Main Steam System or its components It is estimated in the WCAP that piping thrust loads and the Main Steam Isolation Valve (MSIV) seat differential pressure will be increased by approximately 8 percent over those expected in the current condition if a rapid closure of the MSIVs were to occur.

For the Condensate and Feedwater System, the Auxiliary Feedwater System, and the Blowdown

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Beover Volley Power Station Unit'lL

1991 Report of Facility Changes, Tests, and Experiments Page 182 of187, and Sampling System,'the increase in the allowable steam generator tube plugging level to the 20

. percent lim t has been determined in Section 8 to have no adverse effects.

The reduction in primaiy side volume necessitated a reanalysis of the uncontrolled Boron Dilution event. A proposed Updated Final S fety Analysis Report (UFSAR) change to Section 14.1.4 will reflect this reanalysis. The new analysis confirms the conclusions presented in_the UFSAIL For the steam generator tube rupture, the 20 percent tube plugging evaluation indicates a slight'-

increase in the calculated break flow and consequently in the calculated radiation doses; however, the UFSAR results remain valid provided TDF is maintained.

The WCAP 11591 (1987) loss of coolant accident analysis concluded that the limiting case.was maximum Safety Injection (SI) and Moody discharge coeflicient Cp = 0.4 which resuked in a Peak Clad Temperature (PCT) of 1918'F, For the 20 percent tube plugging analysis, the limiting .

case is minimum Si and C 9 = 0.4 which results in a PCT of 2149 F Proposed UFSAR TT s and Figures will reflect this change.

All r. oects of the proposed steam generator tube plugging have been evaluated in the WCAP and no na r different accidents or failure modes have been identified.

Since the increased tube plugging does not challenge the integrity of the steam generators and the'.

change is not expected to indirectly affect any safety equipment, the abiSty of any safety related >

equipment to perform its function is not affected.

All LOCA and Non-LOCA' a cidents have been evaluated or re-analyzed for the efTects of the increased tube plug (ng. All acceptance ciiteria continue to be met.

Based on the safety evaluation summarized above, it was determined that no unreviewed safety question is involved.

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- Beaver Valley Power Station Unit 1

' 1991' Report of Facility Changes, Tests, and Experiments Page 183 of187

[li ANG E TITI.E Temporary Mo'dification For Blocking Open Two Power Operated Relief Valves (PORVs)

[lI ANGE DESCRIPTION This safety evaluation supported blocking open two PORVs, (PCV-RC-455C) and (PCV-RC-.

45SD), to provide over pressure protection for the Reactor Coolant System during a shutdown .

condition. This was necessary since both trains of the Over Pressure Protection System (OPPS) were out of service due to the Auxiliary Relay Cabinets being considered non-seismic. The Auxiliag Relay Cabinets housed the OPPS Relays.

SAFETY EVAI,UATION The probability or consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR) will not be increased since blocking open the PORVs will allow the reactor coolant system to be protected from an over pressure condition while the plant is shutdown.

The radiological consequences of an accident will not increase because blocking open the PORVs will prevent any over pressure challeriges to the reactor coolant system while the over pressure protection system is considered out of service.

To prevent over pressurization of the reactor coolant system at shutdown, either OPPS must be in service or there must be at least a 3.14 square inch hole in the reactor coolant system. Blocking open the PORV's creates this hole. Therefore the posibility of an accident or malfunction of a ditferent type than any previously evaluated in the UFSAR is not created.

The design basis for the low pressure setpoint is preventing over pressurization of the reactor coolant system during the inadvertent starting of a High Head Safety Injection pump and assumes no operator action for ten minutes. Blocking open the PORV's prevents the setpoint from being reached and the Technical Specification bases are not affected.

No unreviewed safety question is involved.

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Beaver Valley Power Station Unit 1 1991 Repon of Facility Changes, Tests, and Experiments-

- Page 184 of 187

' [II ANGE TITf,E Sump Pump Installation For Clam Study Program Discharge Sampling CII ANGE DESCRIPTION A temporary submersible pump was installed in the "D" intake bay. The pump provides a supply of water to a trailer for evaluation of clamicide effectiveness.

SAFETY EVAL,liATION The probability or consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR) _will not be increased since the modification only affects one safety related component, 2SWS-P-21 A, which is a redundant pump. No Unit I safety related components were afTected.

Failure of the service water pump will not increase the consequences of an accident because of redundancy. Only one pump is required for safe shutdown.

The possibility of an accident or malfunction of a different type than any previously evaluated in the UFSAR is not created because failure of a service water pump has already been addressed.

Ne Technical Specification bases are affected. The requirements of Specification 3/4.7.4.1 were still met, Based on the above information, no unreviewed safety question is involved,

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g 13eaver Valley Power Station Unit 1 1991 Report of Facility Changes, Tests, and Experiments Page IF5 of 187 CII A NGE TITIE Temporary Replacement of PCV-R\W130A With A Substitute Part DIANGE DESCRIPTION PCV-RW-130A is the pressure segulator for the back-up cooling water to river water pump WR-P-1 A seals and motor. The regulator had failed and delivery of a new regulator would take 5 to 6 weeks. This safety evaluation supported a modification which installed a similar regulator, made of different material, for a short duration. This enabled the back-up cooling system to be returned to service.

SAFETY EVAL UATION The probability or consequences of an accident described in the UFSAR will not be increased since the replacement regulator functions the same as the original and has the same pressure rating. Also the short duration of use will preclude corrosion problems for the substitute part.

The radiological consequences of an accident will not increase since this temporary modificat;on does not affect any design basis accidents.

The possibility of an accident or malfunction of a difTerent type than any previously evaluated in the UFSAR is not created since the replacement regulator functions the same as the original. The original part is made from stainless steel to prevent corrosion caused by river water. The l

replacement part is made from carbon steel, but for the short duration of installation (5 to 6 weeks), corrosion will not be a problem.

The Technical Specification bases require two operable river water pumps. This temporary modification does not afTect the acceptance limits of the Technical Specification. Two pumps are still required and this modification returns a pump to se. .ce.

No unreviewed safety question is involved.

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- Beaver Valley Power Station Unit i '

1991 Rei ort of Facility Changes, Tests, and Experiments L Page 186 of 187 CII ANG E TITI,G Temporary Modification To Install A Temporary safety Injection System Reset / Override Train l A Push-button i

Cil ANGE DESCRIPTION i l

During performance of the Train A Solid State Protection System Surveillance (MSP-1.04) it was discovered that the Safety injection System reset / override Train A push-button, located on the l bench board, had an intermittent conhet which is used to block a safety injection signal. 'A temporary modification was needed to install a temporary push button which corrected the intennittent contact. The normally inste.lled push-button cannot be safely reached while the plant  ;

is operating. This safety evaluation supports this temporary modification.

SAFETY EVAlsUATION The probability or consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR) will not be increased since the replacement push-button is identical to the original and is installed near the original push-button. The replacement push-button is seismically installed and does not violate any train separation requirements.

The radiological consequences of an accident will not increase since the repiscement switch is identical to the original. Neither the replacement switch or the original switch can prevent a -

safety injection signal from being generated. .The Safety Injection System will be available for all design basis accidents.

I The possibility of an accident or malfunction of a difTerent type than previously evaluated in the UFSAR is not created since the replacement switch will function and be installed the same as the original switch. The replacement switch is within a few feet of the original switch on the same bench board so operator action will not be alTected.

No Technical Specification bases are afrected since the replacement switch will fimetion the same as the original svcitch. A safety inject 5n signal can't be reset for 75 seconds. This will remain unchanged by this modification.

No unreviewed safety question is involved.

9 Beaver Valley Power Station Unit i 1991 Report of Facility Changes, Tests, and Experiments Page 187 of 187 CII ANG E TITI.E TV-BD 101 Al, A2, B1, B2, Cl, C2 Mechanically Blocked Open CII ANGE I)ESCRIPTION Scheduled maintenance on these blowdcwn valve operators during 8R waused the valves to fail shut. The valves were the drain path for steam generator draindown, which was required for steam generator maintenance during the outage. The blowdown valves had to be mechanically blocked open using a chain fall to permit steam generator draindown.

SAFETY EVAI,UATION SUMM ARY The probability aiid consequences of an accident described in Updated _ Final Safety Analysis Report (UFSAR) Appendix D - High Energy Line Breaks - was not increased since the blowdown lines did not meet the criteria for high energy line breaks while in mode 6.

Radiological consegaences were not increased since the defeated automatic isolation was only an environmental qualification concern for a high energy line break. In mode six the line was not a high energy line and manualisolation was still available with TV-BD-lOOA, B, C.

The possibility of an accident or malfunction of a difTerent type than any previously evaluated in the UFSAR is not created since automatic isolation is not required of this line while in mode 6, and manual isolation was available.

No Technical Specification basic were atTected by this modihcation since these valves were not referenced in the Technical Specifications.

No unreviewed safety question is involved.

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Duquesne Licfit Company

'Af BEAVER VALLEY POWER STATION ,

UNIT NO.1 DOCKET NO. 50-334 LICENSE NO. DPR-66 I

ATTACHMENT 2 l

ERRATUM IDENTIFIED IN THE 1990 REPORT OF FACILITY CHANGES, TESTS, AND EXPERIMENTS I

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Beaver Valley Power Station Unit t 1990 Report of Facility Changes, Tests, and Experiments EltRATUM Identilnd in the 1990 Report i

CII A NGE TITI,E Temporary Operating Procedure TOP l-90-9, Transfer PAB South Sump to a HIC CII ANGE DESCRIPTION A new temporary procedure was developed to transfer the contents of the Primary Auxiliary Building (PAB) South Sump to a Iligh Integrity Container (HIC) located in the Solid Waste Area using an air o, erated diaphragm pump and temporary hoses. This procedure also provides-instructions to dewater the IIIC during transfer.

SAFETY EVALUATION

SUMMARY

l The Service-Water 1q!!d Waste System (SWS) and Primary Vents and Drain System are not safety related. No safety related equipment is located in South East PAB elevation 722' or in the east -

trench or Solid Waste Building. In the event a temporary component would leak or fail, the failure would be similar to installed component failure end would be bound by Updated Final Safety Analysis Report Section _14.2.2 and 14.2.3 analysis. No Technical Specifications are affected by this procedure. No unreviewed safety questions exist.

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