ML20043C701

1989 Rept of Facility Changes,Tests & Experiments. W/900523 Ltr

ML20043C701
Person / Time
Site:	Beaver Valley
Issue date:	12/31/1989
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9006060077
Download: ML20043C701 (94)

Text

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I 3 Drsmalmie @ Coppiny Rg;'y!'*v"a**'Staa -

. Shippingport, PA 15077-0004 ,

@p,$."$nU" uni , o, .

1 May 23, 1990

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.U. S.- Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Reference:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 1989 Report of Facility Changes, Tests and Experiments Gentlemen:

.This letter forwards the 1989 Annual Report of Facility Changes, Tests and Experiments, in accordance with 10 CFR 50.59.

The, report covers the period November 1, 1988 through October 31, 1989 .to coincide with the annual FSAR update. A brief description of each facility and procedure change is provided with a brief summary of the safety evaluation for each change.

Very truly yours, LL w

/J. D. Sieber UVice President Nuclear Group cc: Mr. J. Beall, Sr. Resident Inspector Mr. T. T. Martin, NRC Region I Administrator Mr. A. W. DeAgazio. Project Manager Mr. R. Saunders.(VEPCC)

/ Y 9006060077 891231 / l 4

.PDR ADOCK 05000412 I F R PDC

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. l DUQUESNE LIGHT COMPANY-BEAVER VALLEY POWER STATION UNIT NO. 2 DOCKET N0. 50-412 LICENSE NO. NPF-73 l

1989 REPORT OF FACILITY CHANGES TESTS AND EXPERIMENTS

, BEAVER VALLEY POWER STATION, Unit 2 1989 Report of Facility Changes, Tests & E::periments Table of Contents Testina Procedures Eggg Cycle 1 Extension 1 Operation of Gland Steam Exhaust Ventilation Without HEPA and Charcoal Filters 3 Special Test Equipment for Baseline Heat Rate Test 4 Temporary Instrument Air Dryer 5 litalth Physics ProcedMrt'd Change in Unit 2 Radiation Monitor 2 CHS-RQ101 Databanes 6 Change in Unit 2 Digital Radiation Monitor System Sample Flow Setpoint control (Effluent Monitors) 7 coeratina Procedures Temporary Modification - Modification to Borate l the Unit 2 Spent Fuel Pool 8 j l

Temporary Modification - Defeating 2FWE*P23A, B l

" Motor Driven Auxiliary Fced Pump", Auto Start from 2FWE*P22 Turbine Driven Auxiliary Feed Pump 9 Operating Manual Chapter 2.18.4A through F, l Flushing Ion Exchanger Resin to Holding Tank /

Shipping Container Procedures (Rev. 3) 10 l

Operating Manual Change Notice 2-88-525/2-89-077, Change Normal System Alignment to Open for 2GWS-530 11

! Temporary Modification - Block of High Vibration l Trip on 2HVR-FN-201C, "C" Containment l

Recirculation Fan 12 Temporary Modification - Block of the Main Turbine Electrical Overspeed Trip 13 Temporary Modification " Block of High Vibration Trip on 2HVR-FN-201A, "A" Containment Recirculation Fan 14 1

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• i s BEAVER VALLEY POWER STATION, Unit 2 1989 Report of Facility Changes, Tosts & Experiments ,

4 i Table of Contents Ooeratina Procedureu (Continued) Eggs Temporary Modification - Charging Pump Vents 15 l Temporary Modification - Installation of Blank Flanges on Charging Header Relief Line and Relief Discharge to Pressurizer Relisi' Tank, with Relief Valve (2CHS*RV8144) Removed for Repairs 16 i

Temporary Operatir.g Procedure 2-89-01, Post Maintenance Hydro Test of 2CHS*HCV-186 17 Temporary Operating Procedure 2-89-02, Temporary Fuel Pool Purification 18 l Operating Manual Chapter 2.21.3, Main Steam System (2 MSS) Valve List 19 Operating Manual Procedure 2.18.4X, Transferring Waste from the Decant Tank to a f, hipping Container (HIC) 20 .

Operating Manual Procedure 2.18.4Y, Dewatering Shipping Containers 21 Operating Manual Chapter 2.18.4Z, Transferring Spent Resin Using a Portable Resin Transfer Pump 22 Operating Manual Procedures 2.6.4C, D, N & W, Draining an Isolated Reactor Coolant Loop / Filling and Venting a drained Coolant Loop / Reactor Coolant System Isolated Loop Recovery / Isolating a Reactor Coolant Loop 23 Temporary Modification - Suction Pressure Gauge Installation on (2WSS-P21), " Resin Transfer Pump" 24 Operating Manual Change Notice 2-89-123, Operating Surveillance Test (OST) 2.16.4, " Fuel Building Ventilation System Verification Fuel Movement" 25 Operating Manual Change Notice 2-89-150, Procedure 2.6.4N, Reactor Coolant System Isolated Loop Recovery and Venting 26 Modification - Resize Flow and Restricting Orifices (2SWS*FE105A,B,C,D) and (2SWS*RO104A,B) 27 Temporary Operating Procedure 2-89-23, Service Water Flush 28 Temporary Operating Procedure 2-89-07, Temporary Supply of Power for MCC*2-E-05 and MCC*2-E-06 29

- BEAVER VALLEY POWER STATION, Unit 2 1989 Report of Facility Changes, Tests & Experiments Table of Contents operatina Procedures (Continued) Pagg ,

Operating Manual Change Notice 2-89-60, Procedure 2.6.4I, Draining the RCS for Refueling 30 Temporary Operating Procedure 2-89-9, Dewatering Shipping Containers (HIC) 31 Temporary Operating Procedure 2-89-10, Transferring Spent Resin Hold Tank Contents to a Shipping j Container (HIC) 32 l Operating Manual Change Notice 2-89-87, Procedure 2.18.4Z, Transferring Spent Resin Using Portable Resin Transfer Pump (Rev. 4) 33 l )

Operating Manual Change Notice 2-89-86, Procedure l 2.18.3, Solid Waste Disposal System (2WSS) Valve List (Rev. 3) 34 ,

l Operating Manual Change Notice 2-89-74, Procedure '

2.7.4Z, Removing the Reactor Coolant Filter From i Service 35 L Temporary Operating Procedure 2-89-29, Throttle the "A" cold Leg RTD Manifold Flow 36 (2SWS*MOV105A) "RSS Heat Exchanger 21A Outlet" l De-Energized Open 37 j Temporary Modification Block Open (2HVD-MOD 22B) 38 Temporary Operating Procedure 2-89-28, Station Air Compressor Load Cycle Test 39 Temporary Operating Procedure 2-89-26, Draining the l Recirculation Spray System Heat Exchangers' Service l Water System Supply Header 40 l Operating Surveillance Test 2.48.10, Lead / Plastic Seal Quarterly Review 41 Temporary Modification - Allowing Addition of a  ;

Vacuum Gauge With Isolation valve to  !

(2GSS-FLTA256A(B)) Compartment Housing 42 l l Temporary Modification - Styrofoam Ball in Local Reactor Vessel Level Indicator 43 Temporary Modification - Temporary Pressure Indicator Installed at (2SWS-1101) 44 l

BEAVER VALLEY POWER STATI!N, Unit 2 1989 Report of Facility Changes, Tests & Experiments Table of Contents Operatina Procedures (Continued) Page Temporary Modification - Temporary Flow Indicator Installed Downstream of (2SWS-1101) 45 Temporary Modification - Temporary Pressure Indicator Installed at (2SWS-1015) 46 ,

Temporary Modification - Temporary Flow Indicator Instelled Downstream of (2SWS-1015) 47 Temporary Operating Procedure 2-89-35, Install and Remove Jumper From the outlet of Gaseous Wasto (GW)

Oxygen Analyzer to GW Prefilter 48 Modification - Turbine Trip on Turbina Thrust Bearing Failure 49 Encility Changes Desian Chance No. 836, Rev. 1 - Unit 2 Chlorine Detection Modification 50 No. 857, Rev. 0 - ATWS Mitigating System -

Actuating Circuitry (AMSAC) 51 No. 879, Rev. 1 - Modify Unit 2 Main Feedwater Regulating Valves 52 No. 941, Rev. 0 - Addition of Piping to ,

Transfer Spent Resin to an HIC 54 No. 997, Rev. 0 - Add Vent to 2nd Point Heater, drain Receiver 55 No. 1050, Rev. 1 - Steam Generator L'nwdown j Domineralizers 56 No. 1056, Rev. 0 - Modification of Reactor Plant Sample System 58 No. 1071, Rev. 0 - Revised Annunciator Inputs to Window Al-3H 60 No. 1075, Rev. 0 - Boric Acid Batch Tank Annunciator Inputs for A2-2F 62 No. 1093, Rev. 0 - New Stud Racks for Reactor Vessel Head Studs 64

BEAVER VALLEY POWER STATION, Unit 2 1989 Report of Facility Changes, Tests & Experiments Table of Contents Facility Chanaes (Continued) Page Desian Ch 4DER No. 1094, Rev. 0 - Deletion of Computer Point PO148D and Associated Annunciator Input for 2CHS-PT160 65 No. 1123, Rev. 0 - Spent Fuel Holst Re-Rate 2MHF*CRN227 67 ,

No. 1126, Rev. 0 - EV-2 Large Bore Primary component Snubber Elimination 68 No. 1136, Rev. C - BV-? Small Bore Snubber Optimization 71 No. 1151, Rev. 0 - Turbine Lube Oil Reservoir and #6 Bearing Pedestal Vacuum Instrumentation 73 No. 1152, Rev, 0 - Turbine Trip System l

Modification 75 No. 1158, Rev. 0 - Loose Parts Monitoring System Alarm Inhibit from Rod Control 77 No. 1194, Rev. 0 - Emergency Air Lock Insulation 78 No. 1218, Rev. 0 - Fuel Transfer System Upender Shock Absorbers Deletion (2FNT-TFT22) 79

No. 1270, Rev. 0 - Permanent Utility Tie-Ins for the Outage Trailer Complex 81 No. 1352, Rev. 0 - Containment Air Recirculation Cooling coils Piping Modifications 82 No. 1380, Rev. 0 - Modification to the Containment Air Recirculation (CAR) Fans 83 Technical Evaluation Reoort (TER)

No. 862, Rev. 0 - Acceptance of Temporary Fire Loop 84 No. 978, Rev. 0 - Add Block Valves to the Containment Instrument Air System to Isolate 21AC-SOV235A and B 86 No. 987, Rev. 0 - Addition of Clean out Parts to the Turbine Lube Oil Demister 2TML-DMST21 Drain Line 87

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, Benv0r Vallcy Pow 3r StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 1 of 87 GHANGE TITLE Cycle 1 Extension CHANGE DESCRIPTION ,

An extension of Cycle 1 was desired to avoid shutdown during a period of peak system load. Westinghouse conducted an evaluation of BVPS Unit 2 Cycle 1 to extend the cycle burnup beyond end-of-life conditions. The findings showed that a core life extension posed no unreviewed safety questions. Extended operation was approved with the following restrictions.

1. Coastdown is achieved by power level reduction (vs.

temperature reduction).

2. The Tavg program for rod control, pressurizer level, and steam dump systems is not changed. (The plant should be operated within the normal Tavg/ Tref band.)

SAFETY EVA1,UATION

SUMMARY

i 1). Per the Westinghouse evaluation, "The effects of extended Cycle 1 operation on the design bases and postulated accidents... therefore the conclusions presented in the FSAR remain valid." No restrictions limiting the EOL Burnup were found during a review of FSAR Sections 4.2, 4.3, and 4.4 and during a review of NUREG - 1057, "SER related to the operation of Beaver Valley Power Station Unit 2", supplements 1 through

6. Technical Specification Base 3/4.9.14, Amendment 12 allows a maximum burnup of 60,000 MWD /MTU. The highest burned assembly with an overall burnup of 17,000 MWD /MTU will be approximately 19,400 MWD /MTU (34,700 MWD /MTU for cycle 2 maximum burnup), well below this limit. Therefore, the probability of an accident will not increase.

2). Per the Westinghouse evaluation, " Fuel performance and FSAR design basis accidents have been shown to be acceptable by demonstrating that the cycle extension results satisfy the design and safety limits for Cycle 1." Also, Technical Specification bases 3/4.9.14 gives an upper burnup limit of 60,000 MWD /MTU, this number comes from NUREG/CR 5009. The NUREG examines the changes that could result in the DBA assumptions from using extended burnup fuel (to 60,000 l MWD /MTU). The only DBA consequences affected by use of extended burnup fuel is the potential thyroid dose from a fuel handling accident. In the SER for Technical Specification Amendment 12, the NRC approved a 20% increase in the potential l thyroid dose from a fuel handling accident.

Benver Volicy P vor Statien Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 2 of 87 3). Plant equipment will be operated in accordance with BVPS Unit 2 Technical Specifications and operating Manual. The plant operational limits given in the Westinghouse evaluation do not differ from operation during the cycle except for being at reduced power. This continuation of operation under normal system arrangement will not increase the probability of a malfunction of equipment important to safety.

4). Per the Westinghouse evaluation, "The plant power capability is evaluated considering the consequences of those accidents examined in the Bsaver Valley Unit 2 FSAR using the accepted ,

design basis. It is concluded that the cycle 1 extended  !

operation will not adversely affect the ability to safely operate the core up to a measured burnup of 17,000 MWD /MTU "

Also the plant will be operated under present Technical Specifications and in normal system arrangement.

5). The possibility of an accident will not increase because the plant is run in normal system arrangement, except that the fuel will be burned longer. Therefore, only the fuel changco significantly. Technical Specifications allow a burnup of 60,000 MWD /MTU for an assembly. This will not be exceeded. .

The reactivity coefficients will change because of the extra burnup and these factors 5:ere found to be within the FSAR constants. Therefore, the reactivity coefficients at EOL and at 17,000 MWD /MTU will not increase the chances of an accident.

6). The possibility of an equipment malfunction will not increase because the plant is running identically to the way it was running at EOL, except it is at reduced power.

7). The increased burnup itself does not decrease the margin of safety since the operational parameters that do determine l safety (power distribution and reactivity control) are not changed. Technical Specification bases 3/4.9.14 sets a peak limit of 60,000 MWD /MTU burnup for the fuel pool. This maximum burnup is much higher than the burnup that will be achieved during cycle 1 (19,400 MWD /MTU). (Maximum burnup for cycle 2 will be approximately 34,700 MWD /MTU.)

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Be;vor Vallcy Pow 0r LtCtion Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 3 of 87 CHANGE TITII Operation of Gland Seal Steam Exhaust Ventilation Without HEPA and Charcoal Filters CHANGE DESCRIPTION 2BVT2.26.2 " Gland Seal System (GSS) Filtration System Carbon Canister Removal," and 2BVT 2.26.3 "GSS Filtration System Inplace Leak Test" could not be performed due to accumulation of water in the filtration system which destroyed the filter media. The Testing Group recommended a temporary modification removing the filtration media from both trains; Ref. DCP-881. Engineering is evaluating the filtration assembly drain system for a workable solution. The temporary modification posed no unreviewed safety question.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of a steam generator tube rupture (the most limiting accident) as described in Section 15.6.3 of the UFSAR is not increased. The radiological consequences of a steam generator tube rupture are based upon taking no credit for the Gland Steam Exhaust Filters, and are therefore, not increaued by removal of the HEPA and Charcoal Filters (Reference UFSAR Section 15.6.3.3)

The probability and conseq0ences of a malfunction for equipment important to safety is not increased since no safety related equipment is affected by this change. An unmonitored release is not possible because the discharge path through the leak collection normal exhaust ventilation vent is monitored by (2HVS-RQIl01), whose output is monitored by the DRMS. (

Reference:

Sections 9.4.15.3, and Figures 6.5-2 and 9.4-16)

Removal of the filters does not affect the operation of any other components of the system.

The Bases of Technical Specification Sections 3/4.7.1.4 and 3/4.11 concerning secondary activity and effluent releases are not affected by this change.

4 Be ver Vallcy Pow 0r StOticn Unit 2 j 1989 Report of Facility Changes, Tests, and Experiments 1 Page 4 of 87 1

l CHANGE TITLE Special Test Equipment for Baseline Heat Rate Test CHANGE. DESCRIPTION New test 2BVT 11.26.6 " Baseline Heat Rate Test," required a temporary modification. Special test equipment was placed into service to obtain test data and then was removed from service making restoration complete. This temporary modification posed no unreviewed safety question.

SAFETY EVALUATION

SUMMARY

The probability and consequences for a loss of onsite power and loss of feedwater flow indication as described in UFSAR Secticns 8.2, 10.4.7.3 and 15.2.7 will not be increased. No increase in '

doses will be created.

The probability and consequences of a malfunction of the equipmert will not be increased since it is analyzed for failure in UFSAR Sections 8.2, 10.4.7.3 and 15.2.7. Malfunction of the equipment involved will not result in increased doses.

No unanalyzed accidents or malfunctions will be involved in the test. During the test, operation will continue to be within the core safety limits defined in Technical Specification 2.1.1.

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Benv0r V311cy Pow 0r StOtien Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 5 of 87 CHANGE TITL2 Temporary Instrument Air Dryer CHANGE DESCRIPTION Upon performance of 2BVT 11.34.1 " Instrument Air Surveillance Test," an instrument air dewpoint of -30*F could not be obtained l as required by Generic Letter 88-14 " Instrument Air Supply System Problems Affecting Safety Related Equipment" and Instrument Society of America Standard S-7.3 " Quality Standard for Instrument Air."

The Testing Group recommended installation of a temporary instrument air dryer in parallel with existing instrument air dryer 2IAS-DRY 21 to provide the required dewpoint of -30*F per l UFSAR Section 9.3.1.1.2. engineering generated DCP-1354 to remove i 2IAS-DRY 21 and move temporary installed dryer into its place l permanently. This temporary modification posed no unreviewed l safety question. i SAFETY EVALUATION

SUMMARY

The probability of an accident previously evaluated in the Safety l Analysis Report will not be increased since air operated valves 1 fail to a safe position on loss of air, i The consequences of an accident previously evaluated in the Safety Analysia Report will not be increased, since no credit is taken

for the Station Instrument Air System in the accident analyses.

The probability and consequences of a malfunction of equipment important to safety will not be increased, since the system is not required for safe shutdown (as stated in UFSAR Section 9.3.1); the power supply is from a QA Category II non-safety related source; and only air operated valves and dampers would be affected.

The possibility of an accident or malfunction of a different type will not be created since only the instrument air system will be affected. There are no Technical Specifications that apply to the Station Instrument Air System.

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, BenvCr Vallcy P; wor StStiCn Unit 2 l 1989 Report of Facility Changes, Tests, and Experiments l Page 6 of 87 '

CHANGE TITLE Change in Unit 2 Radiation Monitor 2 CHS-RQ101 Databases CHANGE DESCRIPTION The engineering unit conversions (i.e., monitor cpm to uCi/cc) for ,

both channels of monitor 2CHS-RQ101, the Reactor Coolant System I Letdown Monitor, were revised as a corrective action to an operational problem involving increasing background count rate.

This count rate had increased to a level exceeding established alarm setpoints by a large margin, rendering the monitor I essentially inoperable. The effect of the background (deemed to be due to build-up of corrosion products on letdown line internal surfaces 2CHS-RQ101 is an adjacent-to-line monitor), was I --

mitigated by raising the discriminator setting on the monitor to 600 kev, rather than the previous value of about 67 kev. The change in discriminator level would change the monitor response,

, necessitating the change in engineering unit conversion. It was

! this change in performance of an instrument described in the UFSAR j that identified the need for a safety evaluation.

SAFETY _EVALUATICN

SUMMARY

l The safety evaluation concluded that there would be no change in the probabilit/ or consequences of an evaluated accident or malfunction as this monitor does not have associated automatic actions that could affect plant systems. Additionally, although the increase in discriminator level lowers the sensitivity of the monitor, many of the radionuclides associated with fuel failures have energies higher than 600 kev. Thus, the monitor is still capable of supporting emergency operating procedures and emergency action levels. Data base changes do not affect the physical arrangement or physical performance of the monitor. Similarly, no l

unanalyzed accidents or malfunctions are created as a change in the data base does not affect the physical arrangement or physical performance of the monitor. This monitor is not addressed in the basis of any Technical Specification.

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l Benvor Valloy P;w0r Staticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 7 of 87 CHANGE TITLE ,

Change in Unit 2 Digital Radiation Monitor System Sample Flow Setpoint Control (Effluent Monitors)

CHANGE DESCRIPTION The effluent radiation monitors in the Unit 2 Digital Radiation Monitoring System (DRMS) were described in the UFSAR as operating in automatic sample flow control. In automatic sample flow control, the sample flow is adjusted whenever the process stream flow rate changes in order to maintain isokinetic sample flow.

These monitors can also operate under setpoint control, in which the desired sample flow is maintained at a fixed value, regardless of process stream flow rate. The automatic sample flow control has caused operational problems with many of the effluent monitors. An ~ evaluation by plant personnel determined that it )

would be preferable to operate in setpoint control. AS the proposed mode of operation differed from that described in UFSAR section 11.5.2.3.5, a safety evaluation was performed.

l' SAFETY EVALUATION

SUMMARY

The proposed mode of sample flow control is provided by the vendor

! as an alternate to automatic flow control. The modes are changed

! via a change in monitor software. There is no increase in the

! probability or consequences of analyzed accidents or malfunctions in that the physical operation, i.e., interface to other plant components, and the physical arrangement of the monitor, is not affected by this change. These monitors are not credited in any accident analysis. With regard to malfunctions, the monitors will continue to provide status and warning alarms when sample flow is out of the control band. There are no unanalyzed accidents or l malfunctions created by this change as no new failure mechanism is cretted by this change in software.

With regard to reductions in the margin of safety in any technical specification, non-isokinetic sample flow could have an adverse affect on radiological measurement made to show compliance with radiological effluent Technical Specifications. The setpoint for the sample flow is set to maintain isokinetic conditions at normal process flow rate. If the process flow rate is higher than normal, the sample collection will be conservative. If the process flow rate should be less, alarms will notify plant operators that the sample flow is not isokinetic, allowing for implementation of corrective measures. Thus, the margin of safety will not be reduced.

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Benvor Valloy Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments I '

Page 8 of 87 CHANGE TITLE Temporary Modification - Modification to Borate the Unit 2 Spent Fuel Pool CHANGE DESCRIPTION A temporary hose was installed on valve (2FNC-33), " Cask Washdown Hose Connection Supply Line Isolation", and was routed to the Spent Fuel Pool to allow Spent Fuel Pool boration.

SAFETY EVALUATION

SUMMARY

An initial condition to the procedure controlling Spent Fuel Pool boration (Temporary Operating Procedure 2-88-23, " Initial Boration of the Unit 2 Spent Fuel Pool") states that no fuel is being stored in the Spent Fuel Pool. Updated Final Safety Analysis Report Section 15.7 analyzes accidents resulting in a radioactive  :

release from a system or component. Specifically, the fuel handling accident assumes radioactive material is stored in the Spent Fuel Pool. With no fuel in the Spent Fuel Pool no unreviewed safety questions exist. In addition, the Technical Specifications related to the spent fuel pool are not applicable until fuel is stored in the Spent Fuel Pool.

Beaver VO11cy Pow r Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 9 of 87 CHANGE TITLE Temporary Modification - Defeating 2FWE*P23A, B "Notor Driven Auxiliary Feed Pump", Auto start from 2FWE*P22 Turbine Driven Auxiliary Feed Pump CHANGE DESCRIPTION A temporary jumper was installed to prevent the motor driven auxiliary feedwater pumps fro- auto-starting on a turbine driven auxiliary feedwater pump discharge low pressure signal. Technical Specification Amendment 6 allowed the removal of the auto-start signal.

SAFETY EVALUATION

SUMMARY

The auto-start feature was an enhancement feature for starting 2FWE*P23A, 23B. There is no effect on the Updated Final Safety Analysis Report failure probability. Auxiliary feed flow is designed and analyzed for one motor driven pump to supply heat removal based on 2/3 Steam Generator Lo-Lo level or main feed pump trip. This jumper does not affect this design. Since the auto start has been removed from the Technical Specifications, no Technical Specification safety concerns are involved. No unreviewed safety questions exist.

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Be2VOr Vallcy Pow 0r StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 10 of 87 CHANGE TITLE Operating Manual Chapter 2.18.4A through F, Flushing Ion Exchanger Resin to Holding Tank / Shipping Container Procedures (Rev. 3)

CHANGE DESCRIPTION The solid waste procedures were revised to reflect the as-built conditions of the plant to allow proper Ion Exchanger resin flushing.

SAFETY EVALUATION

SUMMARY

The procedure changes did not change the site facility. The changes reflect operation of the plant as built, and improve reliability and safety. The changes do not affect the Technical Specifications and no unreviewed safety questions exist.

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. Beaver Volicy P: wor Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments l Page 11 of 87 l

CHANGE TITLE Operating Manual Change Notice 2-88-525/2-89-077, Change Normal System Alignment To Open For 2GWS-530 CHANGE DESCRIPTION The normal system alignment for valve 2GWS-530, "(2GWS-TRP25)

Downstream Isolation on Discharge to Plant Sample System" was changed from closed to open for ALARA concerns.

SAFETY EVALUATION

SUMMARY

The Reactor Plant and Process Sampling System is non-safety related and is not required to function during an accident. The margin of safety will be increased because non-operation of this manual valve reduces radiation exposure. No unreviewed safety questions exist.

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Be;ver Vollcy Pow r StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 12 of 87

.QHANGE TITLE Temporary Modification -

Block of High Vibration Trip on 2HVR-FN-2010, "C" Containment Recirculation Fan CHANGE DESCRIPTION To allow for long term trending on 2HVR-FN-201C, " Containment Recirculation Fan", Maintenance requested that the high vibration trip be blocked.

SAFETY EVALUATION

SUMMARY

No credit is taken for running the Containment Recirculation Fans in the Updated Final Safety Analysis Report (UFSAR). Additionally, the Fans are within the UFSAR Section 3.5.1.2.1 Missile Safety Analysis, should the fan fail and become a missile. The remaining two (2) fans will maintain the containment within the Technical Specification temperature requirements. No unreviewed safety questions exist.

Be;v r Vallcy Powcr St3ticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 13 of 87 CHANGE TITLE Temporary Modification -

Block of the Main Turbine Electrical Overspeed Trip CHANGE DESCRIPTION To decrease the potential for spurious turbine trips, the turbine electric overspeed trip was blocked. This block was made permanent during the first refueling outage under DCP 1152.

SAFETY EVALUATION

SUMMARY

The electric overspeed trip is a secondary trip. The primary means of overspeed trip is by the Electro-Hydraulic Control System. Updated Final Safety Analysis Report Section 15.2 does not take credit for the electric overspeed (secondary) trip in its accident analysis. Also, the secondary electric turbine trip is not covered by the Technical Specifications. No unreviewed safety questions exist.

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. . 1 Beaver vollcy Pow r Statien Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 14 of 87 l

pHANGE TITLE Temporary Modification -

" Block of High Vibration Trip on 2HVR-FN-201A, "A" Containment Recirculation Fan l

CHANGE DESCRIPTION )

To allow for long term trending on 2HVR-FN-201A, " Containment Recirculation Fan", Maintenance requests that the high vibration trip be blocked.

SAFETY EVALUATION

SUMMARY

Nc credit is taken for running the Containment Recirculation Fans in the Updated Final Safety Analysis Report (UFSAR). The fan is within the UFSAR Section 3.5.1.2.1 Missile Safety Analysis, should the fan fail and become a misrile. No Technical Specification safety concerns exist as the Containment temperature parameters will remain in the required range. No unreviewed safety questions exist.

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. Be;v0r Volicy Pow 0r Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 15 of 87 CHANGE TITLE Temporary Modification - Charging Pump Vents CHANGE DESCRIPTION A new valve was added to the charging pumps vent lines to facilitate charging pump venting. This change will be made permanent through the DCP process.

SAFETY EVALUATION

SUMMARY

The addition of the new vent valve does not change the function of the charging pump system. Site Engineering has concluded that seismic consequences from valve addition does not affect the previous Updated Final Safety Analysis Report evaluation. The safety margin has been increased and the probability of charging pump malfunction has been decreased since venting of pumps reduces chances of binding due to hydrogen buildup. No unreviewed safety questions exist.

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. B3cvor Valloy Powor Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 16 of 87 CRANGE TITLE Temporary Modification - Installation of Blank Flanges on Charging Header Relief Line and Relief Discharge to Pressurizer Relief Tank, with relief valve (2CHS*kV8144) removed for repairs CHANGE DESCRIPTION (2CHS*RV8144), " Regenerative Heat Exchanger Relief Valve" was removed for repairs and blank flanges were installed in its place so that the normal charging flow path could be maintained. The flanges were removed and the relief valve reinstalled prior to leaving Mode 5.

SAFETY EVALUATION

SUMMARY

The blank flanges are to be installed only in Modes 5 and 6, with the reactor coolant system temperature less than 200*F. The potential for overpressurization of the regenerative heat exchanger by volumetric expansion from hot letdown with charging isolated does not exist, so the Updated Final Safety Analysis Report Section 9.3.4.2.4 analysis is not violated. While the blank flanges were installed alternate boration flow paths remained available. No unreviewed safety questions exist.

Benvcr Vallcy Powcr Stotien Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 17 of 87 CHANfdL TITLE Temporary Operating Procedure 2-89-01, Post Maintenance Hydro Test of 2CHS*HCV-186 CHANGE DESCRIPTION A temporary procedure was developed to check the integrity of welds after (2CHS*HCV-186), " Reactor Coolant Pump Seal Header Flow Control" was rewelded into place to correct its orientation.

SAFETY EVALUATION

SUMMARY

During testing hydro test will not induce any water into the reactor coolant pump seals due to valve alignment and established clearances. The small amount of demineralized water induced when the valve is recurned to service (approx. 1 gallon) will not affect the boron concentration of the Reactor Coolant System (Core at EOL) . The Safety Analysis of Updated Final Safety Analysis Report Section 15.4.6 will not be violated. No unreviewed safety questions exist.

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BeSvOr Valloy P wcr StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 18 of 87 CRANGE TITLE Temporary Operating Procedure 2-89-02, Temporary Fuel Pool Purification CHANGE DESCRIPTION A temporary procedure was developed to provide temporary spent fuel pool filtering. Spent Puel Pool Purification Filters

[2FNC-FLT21A, 21B:l were plugged with silt and replacement filter media was not readtly available.

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l SAFETY EVALUATION

SUMMARY

Technical Specification and Updated Final Safety Analysis Report

, (Section 9.1.3.3) design requirements are written for the Spent '

Fuel Pool while irradiated fuel assemblies are stored. Since no l irradiated fuel is stored in the Spent Fuel Pool, the change is l within safety analysis. Any leakage or temporary line spillage I will be collected in the fuel Building Sump and temporary lines are not in an area which can jeopardize safety related equipment.

No unreviewed safety questions exist.  !

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. Be:v r Vallcy Pow r Staticn Unit 2 l 1989 Report of Facility Changes, Tests, and Experiments '

Page 19 of 87 CHANGE TITLE Operating Manual Chapter 2.21.3, Main Steam System (2 MSS) Valve List CHANGE DESCRIPTION The Westinghouse turbine manual recommends that prior to cold start of the turbine -

generator unit all valves in the lines supplying steam from ahead of the throttle valves to the moisture separator reheaters be closed and remain closed until the high pressure turbine exhaust steam is less than 200'F. This change corrects the Normal System Arrangement (NSA) position of valves 2 MSS-192, 193 to closed, due to the recommendation.

SAFETY EVALUATION

SUMMARY

Updated Final Safety Analysis Report (UFSAR) Section 10.3.3 does not address these valves as they are located downstream of the main steamline trip valves. Also with 2 MSS-192, 193 closed, the cooldown rate of the reactor will be reduced thereby reducing the consequences of an accident described in UFSAR Section 5.2.

Valves (2 MSS-192, 193) are not addressed in the Technical Specifications. No unreviewed safety questions exist.

Be;vor Val 1Gy Pow;r StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 20 of 87 CHANGE TITLE Operating Manual Procedure 2.18.4X, Transferring Waste From the Decant Tank to a Shipping Container (HIC)

CHANGE DESCRIPTION A new procedure was created to transfer waste from the Decant Tank to a shipping container. This procedure was written to allow the use of equipment added to the solid waste system by DCP 941.

SAFETY EVALUATION

SUMMARY

This procedure implements the operation of design changes to the Rad Waste System installed under DCP 941. Updated Final Safety Analysis Report (UFSAR) Section 15.7 is not affected by this new procedure. A malfunction of equipment or operator error could produce a spill, but according to UFSAR Section 11.4.2.7, the condensate Polishing or Waste Handling Building would contain any spills. The new procedure does not impact Technical Specification 3/4.11.1 or 3/4.11.3. No unreviewed safety questions exist.

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- Be ver V311Cy Pow:r StCticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 21 of 87 CHANGE TITLE Operating Manual Procedure 2.18.4Y, Dewatering Shipping Containers  ;

i CHANGE DESCRIPTION A new procedure was created to provide initial conditions, instructions and checklist to augment vendor High Integrity Container (HIC) dewatering procedure.

EAFETY EVALUATION

SUMMARY

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l This procedure implements the operation of design changes to the l Rad Haste System installed under DCP 941. Updated Final Safety I Analysis Report (UFSAR) Section 15.7 is not affected by this new l procedure. A malfunction of equipment or operator error could I produce a spill, however, according to UFSAR Section 11.4.2.7, the condensate Polishing or Waste Handling Buildings would contain any spills. The new procedure does not impact Technical  !

Specifications 3/4.11.1 or 3/4.11.3. No unreviewed safety questions exist.

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. Be;vOr Vallcy Pow r StCticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments j Page 22 of 87 CHANGE TITLE l Operating Manual Chapter 2.18.4Z, Transferring Spent Resin Using a Portablo Resin Transfer Pump CHANGE DESCRIPTION A new procedure was created to transfer spent resin from the Spent Resin Hold Tank (2WSS-TK22] to a shipping container using a portable resin transfer pump.

SAFETY EVALUATION

SUMMARY

This procedure implements the operation of equipment installed in the Rad Waste System under DCP 941. Updated Final Safety Analysis Report (UFSAR) Section 15.7 is not affected by this new procedure. A malfunction of equipment or operator error could produce a spill, however, according to UFSAR Section 11.4.2.7, the condensate Polishing or Waste Handling Buildings would contain l any spills. The new procedure does not impact Technical l Specification 3/4.11.1 or 3/4.11.3. No unreviewed safety 1 questions exist.

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. Be;vor Vallcy Pow 0r StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 23 of 87 1 i CHANGE TITLE operating Manual Procedures 2.6.4C, D, N, & W, Draining An Isolated Reactor Coolant Loop / Filling and Venting a Drained Coolant Loop / Reactor Coolant System Isolated Loop Recovery / Isolating a Reactor Coolant Loop.

CHANGE DESCRIPTION operating Manual Procedures 2.6.4C, D, N, W were all revised to allow up to 2 reactor coolant loops isolated, while in operating modes 5 and 6. This will allow for more efficient refueling outages. The requirements for isolation of two loops were outlined by the Nuclear Safety Department.

SAFETY EVALUATION

SUMMARY

The Nuclear Safety Department performed an evaluation to determine i I

conditions and requirements necessary to isolate 2 reactor coolant loops. With these requirements incorporated into the procedures, the interlocks and administrative controls required by Updated Final Safety Analysis Report Section 15.4.4 are not violated.

( Licensing and Compliance also pet.'ormed a safety analysis which concluded that NUREG 1057 Section 15.4.6 criteria is met and l

sufficient chutdown margin still exists. In the unlikely event of an unplanned dilution, the shutdown margin will be greater than 1% K/K 15 minutes after the event and would be within the requirements of Technical Specification 3.1.1.2. Technical l Specifications 3.1.2.9, 3.4.1.3, 3.7.2.1, 3.9.1 and 3.9.2 are l still met by these procedures. No unreviewed safety questions l cxist.

B2cvsr Valley Powor Station Unit 2 l

1989 Report of Facility Changes,-Tests, and Experiments Page 24 of 87 CRANGE TITLE Temporary Modification -

Suction Pressure Gauge Installation on (2WSS-P21), " Resin Transfer Pump" CHANGE DESCRIPTION A stainless steel tube, with a pressure gauge, was installed to the suction of (2WSS-P21], Resin Transfer pump, so pump suction pressure may be monitored to enhance pump operation. A gauge is to be installed permanently by the DCP process.

SAFETY EVALUATION

SUMMARY

No accidents evaluated in Updated Final Safety Analysis Report (UFSAR) Section 15.7 are affected by this modification. The pump is not safety related and. is classified as non-nuclear safety class per UFSAR Section 11.4.1. An accidental spill would be handled by the building drains- in the Condensate' Polishing or Waste Handling Building. Technical Specifications 3/4.11.1 and 3/4.11.3 would not be violated. No unreviewed safety questions exist.

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B2avar Valley Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 25 of 87 CHANGE' TITLE Operating Manual Change Notice 2-89-123, Operating Surveillance Test (OST) 2.16.4, " Fuel Building Ventilation System Verification Fuel Movement" CHANGE DESCRIPTION Specific negative pressure values and air exhaust flowrate values from the Fuel Building were removed from the OST and were r(placed with steps to verify just a negative pressure exists and exhaust flow exists.

SAFETY EVALUATION

SUMMARY

Updated Final Safety Analysis Report Sections 6.5.3.2 and 15.7.4 assumes a negative pressure in the Fuel Building with no specific value set. This change will ensure that the negative pressure is still maintained. Equipment alignment or operation is unchanged and is within the assumptions of the accident analyses. The Technical Specifications do not state any flow or pressure, just the fact that the SLCRS is in service through the filters. No unreviewed safety questions exist.

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B3 aver Valley Pow 3r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 26 of 67 CHANGE TITLE Operating Manual Change Notice 2-89-150, Procedure 2.6.4N, Reactor Coolant System Isolated Loop Recovery and Venting CHANGE DESCRIPTION The requirement to borate at a flowrate of 125 gpm for 90 minutes, to allow the restoration of an isolated loop, was replaced with an equivalent mass changeover based on 125 gpm for 135 minutes.(which corresponds to 16,875 gallons accumulated flow through the bypass relief line).

SAFETY EVALUATION

SUMMARY

The Updated Final Safety Analysis Report assumes a flow through the relief line for a specific time which correlates to a set.

mass. Since the mass requirement is satisfied, the safety,-

analysis is still valid and no other malfunctions which have not been evaluated are possible. The Technical Specifications do not address loop recovery or isolation. No unreviewed safety questions exist.

B3sv0r Vcllcy Powcr St0 tion Unit 2 1989 Report of Facility Changes, Tests, and Experiments-Page 27 of 87 s

CHANGE TITLE Modification ~ -

Resize Flow and Restricting Orifices (2SWS*FE105A,B,C,D) and (2SWS*Rol04A,B)

CHANGE DESCRIPTION i

Due to a recommendation from site Engineering, (2SWS*FE105A,B,C,D), flow element on the Recirculation Spray System heat exchanger (service water side) and (2SWS*RO104A,B) restricting orifices on the diesel generator heat exchangers (service water side) were resized to increase flow through them.

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SAFETY EVALUATION

SUMMARY

Resizing the flow and restricting orifices to a larger bore, will  :'

ensure that the flow requirement defined in Updated Final Safety Analysis Report Table -9.2.2 will be met. With flows above the minimums of the table, the analysis is unchanged. Resizing the mifices could have the potential of exceeding the flow capability l t the service water pumps, but the system has throttle valves

. ch- will be used to balance the system and will be tinistratively controlled. The Technical Specification required ituw will be maintained and the margin of safety will not be reduced. No unreviewed safety questions exist.

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. Beav0r Vallcy P; war Staticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 28 of 87 CHANGE TITLE Temporary Operating Procedure 2-89-23, Service Water Flush CHANGE DESCRIPTION A temporary procedure was developed to flush the service water.

piping downstream of the Recirculation Spray heat exchangers by routing the flow of two service water pumps through one train of discharge piping. The procedure is to be performed in operational mode 5 only.

SAFETY EVALUATION SUMIO2X Updated Final Safety Analysis Report (UFSAR) Section 9.2 accident analysis indicates the system is designed for failure of one j' train. Only one-Recirculation Spray System (RSS) Train is tested (flushed) at a time, therefore the other train is available.- i Additionally the RSS is not required for mode 5 and other portions of the Service Water. System will be available to perform their safety functions. UFSAR Section 3.6 analyzes for pipe failure in moderate energy lines and covers this temporary procedure.

o Technical Specifications are not affected as they do not address maximum flow through the affected heat exchangers. No unreviewed safety questions exist.

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B3avor Valloy Powor Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 29 of 87 CHANGE TITLE Temporary Operating Procedure 2-89-07, Temporary Supply of Power for MCC*2-E-05 and MCC*2-E-06 CHANGE DESCRIPTION A temporary procedure was developed to temporarily supply 480 volt power to the Battery Chargers #1 and #2 and to Uninterruptable Power Supply (UPS) #1, 2, 3 and 4 by cross-tying Motor Control Centers MCC*2-E-05 and MCC*2-E-06. The cross-tie is required to allow preventive maintenance on transformers TR-2*8N and 9P to occur.

SAFETY EVALUATION

SUMMARY

Breakers (as required by a site engineering evaluation) are installed at each end of the cross-tie to satisfy an Updated Final Safety Analysis Report (UFSAR) Section 8.3.1.4.1 requirement that a single component failure will not affect the redundant train.

The Engineering evaluation also satisfied UFSAR Section 9.5.1 and 8.3.3.3 fire protection concerns and UFSAR Section 8.3.1.1.18 physical separation concerns. Technical Specifications 3.8.1.2 and 3.8.2.2., which require one train in service while in modes 5 or 6, will be satisfied. No unreviewed safety questions exist.

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B3sv0r Vallcy P0 war Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 30 of 87 CHANGE TITL2 Operating Manual Change Notice 2-89-60, Procedure 2.6.4I, Draining the RCS for Refueling CHANGE DESCRIPTION l Guidance and administrative controls to install a temporarv l connection from (2RCS*LT102), " Loop C Hot Leg Level Transmitter",

to (2RCS*PT402), " Reactor Side of Loop C Hot Leg Pressure Transmitter", were added to the Reactor Coolant System (RCS) drain down procedure. These instructions are to be used to provide RCS level indication when the C Loop is isolated for maintenance.

SAFETY EVALUATION

SUMMARY

The installation of the temporary line- will not effect the interlocks. and administrative controls described in Updated Final Safety Analysis Report (UFSAR) Section 15.4.4, nor will a mechanism that could cause an RCS dilution rate greater than that assumed in UFSAR Section 15.4.6 be created. If the temporary line L malfunctioned, any RCS coolant leakage would be contained within L the containment. The temporary line will not affect the shutdown i

margin required by Technical Specification 3.1.1.2 and 3.9.1 nor

! will it affect the ability to add borated water to the RCS as specified by Technical Specification 3.1.2.1, 3.1.2.3, 3.1.2.5, and 3.1.2.7. No unreviewed safety questions exist.

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, Beavor Valley PowGr Station Unit 2 1989 Report of Facility Changes, Teuts, and Experiments i

. Page 31 of 87 CHANGE TITLE Temporary Operating Procedure 2-89-9 Dewatering Shipping Containers (HIC)

CHANGE DESCRIPTION A new temporary procedure was created to dewater shipping containers. The normal procedure could not be used as the solid waste system was under going a design change.

SAFETY EVALUATION

SUMMARY

The temporary procedure follows the same format as the previously approved procedure 2.18.4Y. with the exception of the collection tank which will not be available until the design change is complete. A radcon sample bottle or equivalent will be used to collect the water. No accidents evaluated by Updated Final Safety Analysis Report (UFSAR) Section 15.7 will be affected by the performance of this procedure. No safety related equipment will be impacted by the performance of this procedure. In accordance with UFSAR Section 11.4.2.7 a radioactive material spill will be collected- in the Condensate Polishing / Waste Handling Buildings.

The new procedure does not impact Technical Specifications 3/4.11.1 or 3/4.11.3. No unreviewed safety questions exist.

. Beavor Vallsy Pow 0r Station Unit 2 i 1989 Report of Facility Changes, Tests, and Experiments l Page 32 of 87 CHANGE TITLE Temporary Operating Procedure 2-89-10, Transferring Spent Resin Hold Tank Contents to a Shipping container (HIC)

CHANGE DESCRIPTION A new temporary procedure was created to transfer waste resin from the Spent Resin Hold Tank to a shipping container via the decant tank. Further modification to the decant tank controls are pending under a design change. As the system is presently designed, waste must be transferred through the sample system, with the possibility of plugging the piping. This procedure will I allow the transfer of resin directly to a shipping container.

SAFETY EVALUATION

SUMMARY

This temporary procedure uses two previously approved procedures (2.18.4I and 2.18.4X) to allow the transfer of resin directly to a shipping container. No accidents evaluated by Updated Final Safety Analysis Report (UFSAR) Section 15.7 will be affected by this procedure. In accordance with UFSAR Section 11.4.2.7 a radioactive material spill will be collected in the Condensate Polishing / Waste Handling Buildings. No safety related equipment is impacted by the performance of this procedure. The new I procedure does not impact Technical Specifications 3/4.11.1 or 3/4.11.3. No unreviewed safety questions exist.

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'. I B2Evor VallGy Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 33 of 87 CHANGE TITLE Operating Manual Change Notice 2-89-87, Procedure 2.18.4Z, Transferring Spent Resin Using Portable Resin Transfer Pump (Rev. 4)

CHANGE DESCRIPTION The procedure was revised to include valve (2WSS-343] which was not listed in the initial revision. The valve was installed under the design change process.

SAFETY EVALUATION _EUMMARY No accidents evaluated by Updated Final Safetl' Analysis Report (UFSAR) Section 15.7 will be affected by this procedure.

According to UFSAR Section 11.4.1 this system is not safety related. Additionally, Operation of this system will not impact safety related equipment. In accordance with UFSAR Section 11.4.2.7, a radioactive material spill will be contained by the building drainage system. This change will not impact Technical Specification 3/4.11.1 or 3/4.11.3. No unreviewed safety questions exist 1

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.BOsvor Vallsy Pow 3r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 34 of 87 l

l CHANGE TITLE operating Manual change Notice 2-89-86, Procedure 2.18.3, Solid i Waste Disposal System (2WSS) Valve-List (Rev. 3) l l

l CHANGE DESCRIPTION l

l Valve (2WSS-343), which was added by the design change process, l p was added to the valve list. l o '

SAFETY EVALUATION

SUMMARY

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The valve was added and evaluated under the design chance l process. Updated Final Safety Analysis Report Sections 15.7 and '

11.4.1 along with Technical Specifications 3/4.11.1 and 3/4.11..

(; are not affected by the addition of the valve to the valve list.

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i B0cvGr Valloy Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 35 of 87 CHANGE TITLE t

l Operating Manual Change Notice 2-89-74, Procedure 2.7.4Z, Removing the Reactor Coolant Filter From Service CHANGE DESCRIPTION In order to continue Reactor Coolant Syster (RCS) cleanup using the ion exchanger with one reactor coolant filter out of service, steps have been added to have Chemistry periodically sample to monitor for degradation and potential resin break through. This was a "one time only" procedure change while a new filter was on order.

SAFETY EVALUATION

SUMMARY

i Updated Final Safety Analysis Report (UFSAR) Section 9.3 states that the filter is provided to collect particulate and resin fines greater than 25 microns. Chemistry sampling will ensure sizes less than 25 microns. At first indication of degradation of the demineralizers, the letdown flowpath will be isolated and the charging pump suction will be swapped to the Refueling Water Storage Tank. An operable boration flowpath for the RCS will be maintained as required by Technical Specification 3.1.2.1. No unreviewed safety questions exist.

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Beavar Valley Powar Station Unit 2 L 1989 Report of Facility Changes, Tests, and Experiments I

Page 36 of 87 CHANGE TITLI Temporary Operating Procedure 2-89-29, Throttle the "A" Cold Leg RTD Manifold Flow CHANGE DESCRIPTION A temporary procedure was developed to determined the throttle position of either (2RCS-32), " Loop A Cold Leg Manifold A Inlet Isolation" or (2RCS-35), " Loop A Cold Leg Manifold A Outlet Isolation", to achieve the same RTD manifold indicated flow as the other two loops while orifice 2RCS-RO100A is removed.

SAFETY EVALUATION

SUMMARY

The Updated Final Safety Analysis Report (UFSAR) already assumes the failure of a single loop, hence the failure of this loop along with the control and protection RTD's by any method are covered by the UFSAR accident analysis. Additionally no Technical Specifications will be affected by this change. No unreviewed safety questions exist.

- Basvor Valloy Pow 3r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Pege 37 of 87 CHANGE TITLE (2SWS*MOV105A) "RSS Heat Exchanger 21A Outlet" De-Energized Open CHANGE DESCRIPTION During the performance of Operating Surveillance Test 2.47.3A, Motor Operated Valve (2SWS*MOV105A] was stroked, but did not fully close before tripping. Valve (2SWS*MOV105A) was re-opened and repairs were initiated. Since Technical Specifications require an operable Recirculation Spray Heat Exchanger, the valve was de-energized in the Normal System Arrangement (NSA) open position and caution tagged to assure the availability of service water flow to the heat exchangers. The valve was returned to service after maintenance was completed during the first refueling outage.

SAFETY EVALUATION

SUMMARY

Valve (2SWS*MOV105A) is NSA open and its safety function requires it to be open to permit service water flow through recirculation spray heat exchanger 21A during long term recirculation / containment spray. Updated Final Safety Analysis Report (UFSAR) Section 9.2.1.1.3 describes isolation of the recirculation spray heat exchangers should radiation be detected.

The service water system (SWS) can be isolated by closing valve (2SWS*MOV104A] to satisfy this requirement. Technical Specification 3/4.6.2.2 and 3/4.6.2.1 describes Quench and Recirculation Spray systems to ensure containment depressurization and subsequent return to subatmospheric pressure in the event of a LOCA. De-energizing valve (2SWS*MOV105A) still provides for an operable Recirculation Spray System. No unreviewed safety questions exist.

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B32Vor Valloy Powsr Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 38 of 87 CHANGE TITLE Temporary Modification Block Open (2HVD-MOD 22B)

GH&UGE DESCRIPTION Number 2 Diesel Generator Building air supply damper (2HVD-MOD 22B]

was blocked open, to allow adequate building ventilation while the damper's. actuator was being repaired. CO2 fire protection must also be disabled while the actuator is repaired.

SAFETY EVALUATION

SUMMARY

Updated Final Safety Analysis Report (UFSAR) Section 9.4.6 requires adequate ventilation in the diesel building. By placing the damper in the accident position (open), the UFSAR concern is satisfied. While the CO2 fire protection is disabled, a fire watch in accordance with Site Administrative Procedure 9D will be initiated. The ventilation system is not covered by the Technical Specifications. No unreviewed safety questions exist.

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Benver Valloy-Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments

Page 39 of 87 CHANGE TITL2 Temporary Operating Procedure 2-89-28, Station Air Compressor 'nad Cycle Test CHANGE DESCRIPTION A temporary procedure was developed to increase the load time on the station air compressor, thus providing a more even flow of compressed air to the Instrument Air dryer in an attempt to depress the Instrument Air dew point and to monitor the Station Air compressors for overheating. This procedure is performed in conjunction with Plant Testing and Performance procedures.

SAFETY EVALUATION

SUMMARY

Updated Final Safety Analysis Report (UFSAR) Section 9.3.2 lists air system failures. This test will not increase the possibility of the listed failures. The redundancy emphasized in UFSAR Section 9.3.2 will not be compromised and per UFSAR Section 9.3.1 the , system is not required for safe shutdown. The anticipated worst case accident would be a drop in service air header pressure. If this would occur,the backup compressor would start.

If this does not satisfy system demand [2SAS-AOV105)," SAS Main Header to Service Air Header" valve will close. These failures are analyzed by the UFSAR. No Technical Specifications are involved. 'No unreviewed safety questions exist.

- Beavor VallGy Paw 0r Stttion Unit'2 1989 Report of Facility Changes, Tests, and Experiments Page 40 of 87 CHANGF TITLE Temporary Operating Procedure 2-89-26, Draining the Recirculation Spray System Heat Exchanger Service Water System Supply Headers CHANGE DESCRIPTION A temporary procedure was developed to drain the Recirculation Spray Heat Exchangers supply from the Service Water System headers. This will reduce the potential of clam buildup in the headers.

SAFETY EVALUATION

SUMMARY

The Temporary Operating Procedure provides instructions for draining the Service Water System (SWS) headers from the valve-pit. These drained headers are analyzed per Stone and Webster calculation. The loss of SWS due to water hammer was also Stone and Webster analyzed. Consequences of an accident will not be increased since operators are posted to ensure normal system arrangement of the SWS is immediately restored as describcd in Updated Final Safety Analysis Report (UFSAR) Section 9.2, Table 9.2-3 and UFSAR Section 6.2. Concerns for potential entry into Technical Specification 3.0.3 (if SWS header ruptures with cross connect valve open) are administrative 1y addressed. No unreviewed safety questions exict.

- Beavor Valloy Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 41 of 87 CHANGE TITLE operating Surveillance Test 2.48.10, Lead / Plastic Seal Quarterly Review CHANGE DESCRIPTIQH The sealed component list has been updated with 45 additional valves, which are now sealed in the Normal System Arrangement (NSA) position.

SAEETY EVALUATION

SUMMARY

No valve position alignment or equipment status is affected by this change. Updated Final Safety Analysis Report Figures 9.2-2,

-3, -4, and -5 do not show these valves as "ADC" (administratively controlled). The procedure change just increases administrative l

controls on selected valves. Administrative control of valves is not Technical Specification related. No unreviewed safety questions exist.

B3svar Valloy P; wor Station Unit 2 1989 Report of Facility Changes, Tests,_and Experiments Page 42 of 87 i

CHANGE TITLE l Temporary Modification - Allowing Addition of a Vacuum Gauge With f Isolation Valve to (2GSS-FLTA256A(B)] Compartment Housing j CHANGE DESCRIPTION A design . change added loop seals to the Turbine Gland Steam (GSS). _j system filter compartment drain piping to eliminate moisture '

buildup in the compartment. This moisture buildup causes the  ;

filter media to become destroyed. The temporary vacuum gauges are needed to monitor GSS filter operation during design change .

testing. The isolation valves are needed so the gauges can be I changed or removed without shutting down the GSS filter train.

SAFETY EVALUATION

SUMMARY

l The temporary modification does not increase the probability of a i Steam Generator tube rupture described in Updated Final Safety Analysis Report (UFSAR) Section 15.6.3. The radiological  ;

consequences of a Steam Generator tube rupture do not take credit for Gland System Exhaust filters. As described in UFSAR Section 9~4.15.3,

. Figures 6.5-2, & 9.4-16, an unmonitored radiological release is not possible because the discharge path through the leak collection normal exhaust is monitored for radiation. 'No safety related equipment is affected and the addition of the temporary equipment does not have any effect on other components  !

of the system. No unreviewed safety questions exist.

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L Beavor Valloy Powcr Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments i Page 43 of 87 l l

CHANGE TITLE' Temporary Modification -

Styrofoam Ball in Local Reactor Vessel Level Indicator l

l CHANGE DESCRIPTION l A styrofoam ball and a fitting (plastic hose) was added to the local reactor vessel level indicator to enhance ease of viewing.

The local level indicator is used in modes 5 and 6.

l SAFETY EVALUATION

SUMMARY

The length of plastic hose (1 1/2") is no more likely to fail than the installed local plastic hose used for level indications, which was previously evaluated. While in modes 5 or 6, Reactor Coolant System (RCS) temperature is less than 200'F, therefore styrofoam melting is not a problem. Per the Chemistry department, RCS chemistry would not be effected even if the styrofoam ball would melt. Technical Specifications are not affected.

. Beaver Valley Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 44 of 87 CHANGE TITLE Temporary Modification - Temporary Pressure Indicator Installed at (2SWS-1101)

CHANGE DESCRIPTION A temporary pressure indicator was installed at (2SWS-1101),

" Primary Component Cooling Heat Exchanger Discharge to Emergency Outfall Structure Sample Connection", per Site Engineering request, to gather pressure data.

SAFETY EVALUATION

SUMMARY

The temporary modification is installed downstream of all equipment which discharge to the Emergency Outfall Structure (EOS). Per Updated Final Safety Analysis Report (UFSAR) Section 9.2.1.1.2, loads which discharge to the EOS will not be affected by the addition of the pressure indicator. Per UFSAR Section 9.2.1.1.1, leak tightness will be checked on installation and minimum flows described in Table 9.2-2 will be met. Two Service Water Systems (SWS) will remain operable as required by Technical Specification 3.7.4.1 and the Standby SWS described by Technical Specification 3.7.13.1 will be unaffected. No unreviewed safety questions exist.

, 3env r Vallcy Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 45 of 87 CHANGE TITLE Temporary Modification- -

Temporary Flow Indicator Installed Downstream of (2SWS-1101)

CHANGE DESCRIPTION A temporary sonic flow indicator was installed downstream of (2SWS-1101), " Primary- Component Cooling Heat Exchanger Dischargo to Emergency Outfall Structure Sample Connection", per Site Engineering request to gather flow data.

SAFETY EVALUATION

SUMMARY

The temporary modification is installed downstream of all equipment which discharge into the Emergency Outfall Structure (EOS). Per Updated Final Safety Analysis Report (UFSAR) Section 9.2.1.1.2, loads which discharge to the EOS will not be affected by the addition of a flow indicator. The flow indicator is installed outside of piping, no equipment is affected and all minimum flows per UFSAR Table 9.2-2 are met. . Two Service Water Systems (SWS) will remain operable as required by Technical Specification 3.7.4.1 and the Standby SWS described by Technical Specification 3.7.13.1 will be unaffected. No unreviewed safety L questions exist.

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.._.,_.a _.% # m Beavor Vallcy P:wcr Staticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments

, Page 46 of 87 CHANGE TITLE Temporary Modification - Temporary Pressure Indicator Installed at (2SWS-1015)

CHANGE DESCRIPTION l

A temporary pressure indicator was installed at (2SWS-1015), l

" Primary Component Cooling Heat Exchanger Discharge to Emergency Outfall Structure Sample Connection", per Site Engineering request, to gather pressure data.

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SUMMARY

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The temporary modification is installed downstream of all I equipment which discharge to the Emergency Outfall Structure l (EOS). Per Updated Final Safety Analysis Report (UFSAR) Section i 9.2.1.1.2, loads which discharge to the EOS will not be affected by the addition of the pressure indicator. Per UFSAR Section 9.2.1.1.1, leak tightness will be checked on installation and minimum flows described in Table 9.2-2 will be met. Two Service '

Water Systems (SWS) will remain operable as required by Technical Specification 3.7.4.1 and the Standby SWS described by Technical Specification 3.7.13.1 will be unaffected. No unreviewed safety questions exist.

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BaovGr Valloy Pcwsr Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 47 of 87 CHANGE TITLE Temporary Modification -

Temporary Flow Indicator Installed Downstream of (2SWS-1015)

CHANGE DESCRIPTION A temporary sonic flow indicator was installed downstream of (2SWS-1015]", Primary Component Cooling Heat Exchanger Discharge to Emergency Outfall Structure Sample Connection", per Site Engineering request, to gather flow data.

SAFETY EVALUATION

SUMMARY

The. temporary modification is installed downstream of all equipment which discharge into the Emergency Outfall Structure (EOS). Per Updated Final Safety Analysis Report (UFSAR) Section-9.2.1.1.2,. loads which discharge to the EOS will not be affected i by the addition of a flow indicator. The flow indicator is installed outside of piping, no equipment is affected and all minimum flows per UFSAR Table 9.2-2 are met. Two Service Water-Systems (SWS) will remain operable as required by Technical Specification 3.7.4.1 and the Standby SWS described by Technical Specification 3.7.13.1 will be unaffected. No unreviewed safety questions exist.

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1989 Report of Facility Changes, Tests, and Experiments Page 48 of 87 i

CHANGE TITLE Temporary Operating Procedure 2-89-35, Install and Remove Jumper From the Outlet of GW Oxygen Analyzer to GW Prefilter CHANGE DESCRIPTION Develop a new procedure to install and remove a hose jumper from the outlet of the Gaseous Waste (GW) Oxygen Analyzer to the inlet or outlet of the GW compressor prefilter.

SAFETY EVALUATION

SUMMARY

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Site Engineering performed a safety evaluation under the temporary modification program (to install the temporary hose). The engineering analysis determined that the function of the Gaseous Waste System (GWS)- as described in Updated Final Safety Analysis  ;

Report (UFSAR) Section 11.3.2.1 is unaltered. The GWS failure i

described in UFSAR Section 15.7.1.1 assumes failure at the inlet to the charcoal delay beds; since the jumper is down stream of the charcoal delay beds, failure of these lines would have less consequences. No UFSAR Figure 3.8-31 safety related equipment is ,

affected. Technical Specification 3.3.3.10 addresses radioactive I gaseous effluent monitoring instrumentation; this jumper would i enable continuous oxygen monitoring and if there is a failure of l the jumper, the Technical Specification would require periodic sampling. No unreviewed safety questions exist.

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4 o Betysr Valloy Pow 3r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 49 of 87 CHANGE TITLE Modification - Turbine Trip on Turbine Thrust Bearing Failure CHANGE DESCRIPTION The turbine rotor position supervisory instrumentation is unreliable and has produced electrical spikes of an undetermined nature. To prevent inadvertent turbine trips, both channels have been disabled and will remain defeated until the channel noise is eliminated or when upgraded turbine supervisory instrumentation is completely installed by an upcoming design change.

7 SAFETY EVALUATION SUM M A turbine trip accident is analyzed in Updated Final Safety Analysis Report (UFSAR) Section 15.2.3. Failure of a turbine thrust bearing would not cause a turbine trip directly, but indirectly from other inputs such as generator electrical protection trip or low electro-hydraulic pressure. The turbine can also be tripped manually from the control room if required, based on rotor position alarms and position indication. The turbine trip on turbine thrust bearing failure is not required by the Technical Specifications.- No unreviewed safety questions exist.

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B3avor Valloy Powar St0 tion Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 50 of 87 CHANGE TITLE Unit 2 Chlorine Detection Modification CHANGE DESCRIPTION Unit 1 and Unit 2 have a common control room. However, instrumentation relied on by each unit to isolate the control room '

in the event of a chlorine release are not identical. This modification to the Unit 2 chlorine detectors will make the Unit 2 signal to the solid state protection racks identical (loss of power, high chlorine) to the Unit 1 signal. Additionally, test switches will be provided to facilitate testing of equipment.

Probe setpoints and response times were revised por the Design knalysis.

SAFETY EVALUATION

SUMMARY

Chlorine detectors are seismically mounted, Category I instrumentation required to isolate the control room upon detection of chlorine (UFSAR Section 9.4.1). Chlorine detection

-is required during plant operation as specified in Technical Specification 3/4.3.3.3.7. The proposed modification does not affect these requirements (detectors operate and sense high chlorine levels). This change is not an unreviewed safety question, l-

. Baaver Valley Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 51 of 87 CHANGE TITLE

! ATWS Mitigating System Actuating Circuitry (AMSAC)

CHANGE DESCRIPTIOH The objective of this change is to provide a backup system diverse and independent from the existing Reactor Protection System to 1) initiate a turbine trip, and 2) initiate auxiliary feedwater flow. This will involve installation of a logic cabinet with tie-ins to the feedwater system and the turbine trip. The purpose of this change is to achieve compliance with 10 CFR 50.62.

SAFETY EVALUATION

SUMMARY

This design change is considered safe and does not involve an unreviewed safety question. The change does not create any new accidents or malfunctions and does not affect the operation of existing protection systems. This change will necessitate a change to the FSAR. No change is required to the Technical Specification at this time.

- Benvor Vallcy P:wcr Stcticn Unit 2 1989 Report of Facility changes, Tests, and Experiments Page 52 of 87 CHANGE TITLE Modify Unit 2 Main Feedwater Regulating Valves CHANGE DESCRIPTION At the present time, a number of operating problems are associated with the Unit 2 main feedwater regulating valves (2FWS-FCV-478/

488/498), including hydraulically induced plug oscillation and increased packing wear. This design change will solve these problems by increasing reliability, operability, and maintainability of Unit 2's main feedwater regulating valves. The following items have been included in this DCP

1) Qualify and install hydraulic / pneumatic actuator design and associated modified bonnets, similar to the one successfully used at Unit 1.

2) Replace the existing valve stem with one of approved material that has been chrome plated. This will increase the valve stem's resistance to packing wear and increase stem packing life.

3) Increase the valves trip closure time from 5 seconds to 7 seconde, in order to allow for more flexibility in setting control system response to transients.

4) Modify the instrument air arrangement to support operation of these valves by moving the pneumatic booster within the circuit and increasing tubing size. These changes will improve valve controllability.

SAFETY EVALUATION

SUMMARY

This design change is safe in that the modifications of Unit 2 main feedwater regulating valves will not affect any equipment important to safety as previously analyzed or create any new malfunctions or accidents. The Technical Specifications are not affected and there are only minor changes of the Updhted Final Safety Analysis Report (UFSAR) Chapter 10. Increasing the trip stroke time has been analyzed by Westinghouse Electric Corporation and Stone and Webster Engineering Corporation to ensure there is no impact on present accidents already described and reviewed in UFSAR Table 15.0-6. With the proposed modifications, the steam generator feedwater system will be maintained as it is now. The modification will not affect the systems function. The system will function to supply heated feedwater to the steam generators under all load conditions, maintaining level within the programmed band as before the modification.

Benv0r Vcllcy Pow 0r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 53 of 87 The proposed modification does not increase the possibility of an occurrence or the consequence of an accident or malfunction of equipment important to safety as referenced in Updated Final Safety Analysis Report, Chapter 15. Current analysis indicates that DNBR has reached a minimum within 5 seconds (i.e. before feedwater isolation valves are expected to close) during feedwater system malfunctions. The additional feedwater, due to the extended closing time of the feedwater regulating valves, would not present a safety concern because the minimum DNBR would still occur prior to feedwater isolation valve closure. Results of steamline break analyses also show that although peak containment temperature is not affected, the increased water inventory increases the dryout time, thereby causing the long-term temperature to maintain magnitude for a slightly longer period of time. The containment temperature Equipment Qualification envelope is not exceeded because of this scenario. In both the above analyses, there will be no increase in the consequences of an accident as described in the FSAR.

This modification will not create an accident or malfunction of a different type than previously evaluated in the Updated Final Safety Analysis Report. Since the Technical Specification for feedwater isolation response times continues to be met, increasing the Flow control valve (FCV) stroke time from 5 to 7 seconds has no impact on the non-LOCA accident analysis as described in the UFSAR. The failure of this FCV has no effect on the safe shutdown of the plant; therefore, an accident of a different type than previously evaluated in the UFSAR will not be created.

This modification has no effect on Technical Specification Table 3.3-3, Table 3.3-4, Table 3.3-5, Table 4.3-2, and Table 3.6-1.

Technical Specification safety margins will be maintained without change.

This design change does not involve a unreviewed safety question.

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4 Becvor Volicy P:wsr Stction Unit 2 1989 Report of racility Changes, Tests, and Experiments Page 54 of 87 CHANGE TITLE Addition of piping to transfer spent resin to an HIC.

CHANGE DESCRIPTION This modification will add the capability of transferring radioactive spent resin from the Spent Resin Holding Tank (2WSS-TK22) to a High Integrity Container (HIC) via the metering pump (2WSS-P23). The HIC's will be located in the existing laydown cubicle at El. 46'-6" near the truck bay of the condensate Polishing Building. This modification includes the installation of a 3-way plug valve (remotely operated) in the decant tank recirculation line, piping through the drum storage area to the laydown area, a new wall and cover to shield the HIC, a bypass line around the metering pump to transfer resin in +,he event of pump failure, and a tie-in to an existing dewatering line in the truck bay.

This modification will reduce the number of containers shipped for offsite disposal containing spent resin. The HIC can hold approximagely 60ft.3 of resin, whereas the 55 gal. drums can only hold 4 ft max., excluding the cement required, he modification will allow sealing or disconnecting of the floor urain in the laydown area to prevent any radioactive resin from entering the building drains system. All valves will be operable from low radiation areas via a reach rod.

SAFETY EVALUATION

SUMMARY

The safety aspects of this modification deal with personnel radiation protection and minimizing radioactive releases. The solid waste system is designated as non-safety related and is not located near any safety-related components.

The personnel exposure rates outlined in Section 12.4 of the UFSAR will not be altered by this modification. The piping is routed within existing shielding walls and is designed to the requirements outlined in the design basis of the Solid Waste System (Section 11.4.1). Thus the integrity of the new piping is equivalent to that which is already installed. All new valvos located in high radiation zones will be provided with remote operators to minimize personnel exposure.

The portable bypass pump connections are located outside of the decant tank cubicle. This bypass will eliminate the need to store resin in the transfer linen and decant tank should the metering pump fail.

The new shield wall and cover installed in the laydown area will minimize operator exposure while preparing for and terminating a resin transfer. A curb in this cubicle will localize any resin slurry spill. The existing drain in the cubicle will be isolated from the building drains system (which flows to the storm sewers) to eliminate the possibility of radioactive releases.

This design change does not involve an unreviewed safety question.

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- Be v0r Volicy P wcr Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 55 of 87 CHANGE TITLE Add Vent to 2nd Point Heater, Drain Receiver CHANGE DESCRIPTION I Add 6" vent piping between the 2nd Point Heater and the Drain .)

Receiver Tanks. Close the valves in the existing 16" vent lines between the Drain Receiver Tanks and the 18" Extraction Steam System lines. These vents are required to reduce the back l pressure in the drain Receiver Tank and allow the heaters to drain correctly.

SAFETY EVALUATION

SUMMARY

J This system is not safety-related. This change is required to l allow the 2nd Point Heaters to drain correctly. This vent design has been analyzed for normal and transient operating conditions.

The existing 16" vent is not required and will be isolated.

This design change does not involve an unreviewed safety question.

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Be0Vor Volicy Pow 0r St0ticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 56 of 87 CHANGE TITLI Steam Generator Blowdown Demineralizers CHANGE DESCRIPTION l

l The existing permanent design of the Steam Generator Blowdown (BDG) system normally directs the BDG to the feedwater fourth point heaters where it becomes mixed with condensate flow. The Condensate Polishing System is unable to remove the BDG impurities after they are diluted with condensater therefore, the required secondary side chemistry cannot be achieved. l

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l A temporary BDG deminuralizer system, located between the blowdown l tank (2BDG-TK21) and the main condenser, is currently in operation. l It uses rented demineralizer skids to condition the BDG prior to its l discharge to the condenser hotwell. This modification proposes to l permanently install a BDG demineralizer system by performing the following: .

1 replacing all temporary hoses (both metal and rubber) with piping,

- rotating the two heat exchangers presently used 180' in order to

, facilitate tube removal and minimize pipe runs, l

adding new cooling lines for these heat exchangers from condensate (CNM) and secondary plant component cooling (CCS),

removing the temporary circulating water cooling pump, piping, and hoses that are presently used as cooling for these heat exchangers,

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adding temperature control valves to the heat exchangers' cooling water outlet lines, and adding a temperature indicating switch upstream of the demineralizers that will terminate flow to the demineralizers and divert it to the fourth point heaters at temperatures above 120*F.

The rented demineralizer skids will be used as the permanent demineralizers with these modifications:

relocating the skids to facilitate resin removal for regeneration, piping in all three skids in parallel with isolation capabilities so that two can be on-line with one in standby, and adding an in-line flowmeter downstream of the skids to read total BDG flow.

A sample panel will include instrumentation to monitor total conductivity, cation conductivity, pH, sodium, and silica levels. A temperature control unit will maintain samples at a desired temperature for uniform analysis. All data will be trended on a A

, Benv0r Volicy P:wcr Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 57 of 87 multi-point strip chart recorder and will be continuously transmitted to a computer located in the Unit 2 Chemistry Lab. This computer will be a commercial grade personal computer (PC) and will be equipped with software especially designated to monitor and trend water chemistry.

If present levels of conductivity, pH, silica, or sodium, are l exceeded, a trouble signal will annunciate locally and at the PC.

Since the BDG demineralizer system could potentially contain radio- l active liquid and solid waste, it is designed to meet the federal l l regulations of 10CFR20, " Standards for Protection Against Radiation."

However, any radiation contributions from the blowdown system to the secondary plant side will be minimized by the following: Steam i

Generator Blowdown Sampla Monitor (2SSR-RQIl00) will remain in operation and will automatically isolate the entire BDG system by closing containment isolation valves (2BDG-AOV100A1, B1 & C1) at a

! radiation reading below the maximum secondary side liquid limit. If primary to secondary system leakage is detected but is below the signal required for automatic isolation, system operation will be

restricted to a maximum demineralizer changeout period of 90 days or a maximum dcmineralizer contact dose rate of 31 mrem /hr. This restriction will maintain operation within the analyzed limits per 2DLS-31297, which establishes the basis for meeting 10CFR20. In order to minimize dose rates from the demineralizers and resins even l

further, a maximum demineralizer changeout period of 30 days is l recommended.

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j SAFETY EVALUATION

SUMMARY

I The implementation of this modification is considered to be safe. No I I revision to the Technical Specifications is required; however, the l UFSAR must be revised to incorporate these changes.

The probability or the consequences of any accident previously

evaluated in the UFSAR will not be increased. The purpose of this

! modification is to provide a method of demineralization to the steam l generator blowdown water so that the required secondary chemistry limits can be adhered to. All components that are affected by this modification are non-safety-related. The safety evaluations and system analyses of UFSAR Sections 9.3.2, 10.4.8, and 15.6.3 remain valid.

The probability or consequences of a malfunction of equipment important to safety will not be increased. No safety-related equipment will be affected by the implementation of this modification.

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

The implementation of this modification will not affect the bases of any Technical Specifications, including 3.4.6.2, " operational Leakage" and 3.11, " Radioactive Effluents."

l This design change does not involve an unreviewed safety question.

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Benvcr Vallcy PcwOr Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 58 of 87 l

CHANGE TITLE Modification of Reactor Plant Sample System CHANGE DESQRIPTION Various problems exist in the Reactor Plant Sample System that create personnel safety nazards and/or create difficulties in obtaining accurate samples. This modification proposes to do the following:

1) Replace leaking flow indicators with units that have a higher design pressure and relocate some of the flow indicators to the inside of the sampling hoods in order to better protect the operator in the case of a flow indicator failure.

2) Replace the rigid connections between the sample panel and the sample cylinders with flex connections and redesign the cylinder holding brackets to tacilitate safer installation and removal of the sample cylinders.

3) Install a pressure indicator, for use in obtaining cylinder samples, so that it can be verified that the samples are obtained at the proper pressure.

4) Install a check valve between each steam generator blowdown grab sample point and the common drain header to prevent cross contamination from the pH analyzers.

5) Replace the old rod-in-tube pressure reducers with units that provide the same pressure reduction capability but can be flushed by the operator without taking the system out of service.

SAFETY EVALUATION

SUMMARY

The implementation of this modification is considered to be safe.

No change to the Technical Specifications is required; only minor changes to the applicable UFSAR figures are required.

The probability of occurrence or the consequences of any accident previously evaluated in the UFSAR will not be increased. The Reactor Plant and Process Sampling (RPPS) system is not required to function during an emergency nor is it required to prevent an emergency condition. However, since reactor coolant samples are taken during normal reactor operation, the RPPS system is designed to limit potential reactor coolant losses. The sample tubing containing reactor coolant is 3/8 inch in diameter, thereby minimizing the magnitude of any postulated leak. This modification deals only with non-safety-related portions of the RPPS system. The implementation of this design change will provide a safer and more reliable method of sample collection.

The probability of any accident, including that discussed in UFSAR Section 15.6.2, will not be increased.

Benv0r Vallcy Pow r StOtirn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 59 of 87 The probability or consequences of a malfunction of equippunt important to safety will not be increased. No safety-related parts of the RPPS system that are affected by this modification interface with safety-related components of any other system.

The possibility for an accident or malfunction of a different type than previously evaluated in the UFSAR will not be created. This modification will only affect non-safety-related equipment and no new failure modes or potential hazards will be created by its implementation.

No Technical Specification bases are affected by the implexantation of this modification.

This design change does not involve an unreviewed safety question.

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Be;vor Volicy P:wcr St0ticn Unit 2 -

1989 Report of Facility Changes, Tests, and Experiments Page 60 of 87 CHANGE TITLI Revised Annunciator Inputs to Window Al-3H ,

CHANGE DESCRIPTION The recirculation pump seal pots each have a high and a low level alarm on them. The low level alarm is used to indicate that one of the two mechanical seals, used to prevent out leakage of radioactive fluid along the pump shaft, has failed. The high level alarm is used only as a reference level for filling the seal pots and to assure that an adequate water level is present to maintain the pressure between the seals slightly higher than the pressure on the outside of them.

operations desires to use the high level switch as both the reference level to fill the 'Je al pots to and also as an early warning of a j

decreasing or inadequate seal pot level. The current configuration consists of a normal or unalarmed condition below the level switch setpoint and an alarm condition at the setpoint and above. In order to keep with the " Dark Board concept," this modification proposes to reverse the action of annunciator Al-3H and the computer points for the high level switches so that an alarm condition will exist at or below the level switch setpoint. In this way, operations can fill the seal pot to the point where the computer point just clears and if an alarm comes in at a later time, it could show early detection of a decreasing seal pot level. The seal pot high level computer points (LO700D), (LO702D), (LO704D] and (LO706D] will be reworded to read

" normal /off normal."

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased. This modification will not affect the operation of the Recirculation Spray Pumps or the seal system as described in UFSAR Section 6.2.2.2.2. The ability of the Recirculation Spray System to help mitigate the effects of a LOCA or a steamline break inside of containment will remain unchanged (UFSAR Sections 15.6.5 and 15.1.5 respectively). <

i The consequence of an accident previously evaluated in the safety analysis report will not be increased. This modificatjon will have no adverse effects on any systems or equipment. The functions of the i Recirculation Spray Pumps and the Pump Seal System remain unchanged.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. The operation of all safety-related components remains unchanged. The level switches will actuate at the same setpoint and no wiring changes will be made. The respective multiple input (MI) switches for the four seal pot high level computer points will be moved to the open to alarm position. Therefore, the annunciator system will alarm at or below the setpoint rather than at or above it. This change has no more probability of failure than that of the previous setup.

Beavor Volloy Pswor St0tien Unit 2 1989 Report of Facility Changes, Tests, and Experiments l

Page 61 of 87 The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. The " Dark Board Concept" will actually enhance the operator's ability to identify legitimate alarm conditions. This modification will also supply an early warning of decreasing level in a seal pot. This is a worthy improvement since there are no level indicators on the seal pots. .

The ~ design basis accidents which were reviewed for potential impact by the proposed design change included a loss of coolant accident (LOCA) and a steamline break inside of Containment.

The safety systems which will be affected by the proposed design change include the recirculation spray system.

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created. The seal pots will be filled to a level just above the level switch setpoint.

Since the seal pots do not have any level indicators, this will be used to give an early warning if the level drops to the level switch setpoint. Overfilling of the seal system, and thus overpressurizing of the seals, will be prevented in the same way as previously done.

The caution statements in procedure 2.13.4F about overpressurizing the seal system will remain in effect.

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. The functions of all safety-related equipment vill remain unchanged.

Failure modes of the proposed design change which were reviewed included failure of the annunciator to properly alarm. No physical changes will be made to either the seal pot level switches or their associated wiring. The only change will be the movement of the MI switches from the close to alarm to the open to alarm position.

Therefore, there is no greater likelihood of annunciator failure to occur now than before.

The margin of safety as defined in the basis for any Technical Specification will not be reduced. The basis of no Technical Specification, including that of Technical Specification 3/4.6.2.2, is affected by this modification.

This design change does not involve an unreviewed safety question.

Be;vOr VallOy P wnr Stcticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 62 of 87 CRANGE TITLE Boric Acid Batch Tank Annunciator Inputs for A2-2F CHANGE DESCRIPTION The design objective is to eliminate unneeded computer and annunciator inputs associated with the Boric Acid Batch Tank (2CHS-TK25). This ,

tank is operated only on an intermittent basis, and always with a l locs1 operator present. When the tank is not being operated, alarm inputs hold annunciator window A2-2F in alarm, causing operator confusion and violating the NRC " dark board concept."

This design change will involve the elimination of the following digital inputs from the PCS System Point Database and annunciators: 1 LO140D - Boric Acid Batch Tank Low T0145D - Boric Acid Batch Tank Temp Hi Hi T-146D - Boric Acid Batch Tank Temp Lo Lo DLC Operations has temporarily deleted these points from the annunciator window by opening the associated knife switches (S1649, S1650, S1651). However, a permanent solution requires that the input leads to the above knife switches be lifted and taped back. These knife switches are located in Shared Input Termination Cabinet A (2IHA-ANN-A).

This design change will havo no effect on the control of the auxiliary steam supply (2 ASS-TSV100) for the steam jacket of the batch tank.

SAFETY EVALUATION

SUMMARY

Deletion of the above digital input points will have no impact upon any of the programs used by the Plant Computer System (PCS). The DLC Computer Operations Group has reviewed the usage of these points

, within the PCS and concluded that they can be removed from the database without affecting any other PCS function.

Deletion of these digital input points from annunciator window A2-2F will also have no adverse impact because the boric acid batch tank is always operated with a local operator present, and there is local indication for level and temperature.

Benvcr Vclicy P:wcr Statien Unit 2 1989 Report of Facility Changes, Tests, and Experiments

! Page 63 of 87 l

! The " Boric Acid Batching" procedure of Operating Manutl Chapter SB, Section 2.7.4 requires communication to be established between the Control Room and the batch tank operator. When the batch tank is drained into the Boric Acid Tank, the Boric Acid Transfer Pumps are to be stopped (by the control room operator) based on local indication of level as communicated from the batch tank operator. Therefore, eliminating the annunciator point for ' low level' will have no adverse impact on the Boric Acid Transfer Pumps (less prevention of cavitation) because the annunciator signal is not used to terminate pumping for low level. (A gravity drain of the tank is also optional.)

The batch tank is always operated with a local operator present who has local indication of level and temperature. Therefore, the signals of low level and temperature can be deleted from the PCS and the annunciators because they are not used or needed. This will have no adverse affects on safety and is indirectly a slight safety benefit because it supports the " dark board concept."

This design change does not involve an unreviewed safety question.  ;

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Beevor Vallcy Pow 0r StCticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments l Page 64 of 87  :

CHANGE TITLI New Stud Racks for Reactor Vessel Head Studs CHANGE DESCRIPTION The existing BVPS-2 reactor vessel head stud racks, being rectangular in arrangement, are difficult to load / unload, and they tip to an extreme degree when lifting and/or rotating them. The use of this current rack design is time consuming, increases exposure, and is a safety hazard.

This modification proposes to replace the existing rectangular racks with circular racks similar to those at BVPS-1. The circular racks will include a rotating base or lazy susan arrangen.ent, which will help to provide simplified alignment adjustments to improve the stud handling operations.

SAFETY EVALUATION

SUMMARY

The implementation of this modification is considered to be safe. No change to the Technical Specifications is required; only minor changes to UFSAR Section 9.1.5.2.3.3 and Figure 9.1-17 are required.

The probability of occurrence or the consequences of an accident as previously evaluated in the UFSAR will not be increased. The function of the reactor vessel stud racks remains unchanged.

The probability of an occurrence or the consequences of a malfunction of equipment important to safety will not be increased. The interactions of the racks with safety-related components and structures will not be changed. The racks will be moved, with the polar crane, to t'e inside of the refueling cavity prior to removing the studs from the vessel for refueling, and prior to returning the studs to the vessel after refueling. During the actual fuel movements and at all other times, the racks will be stored on the operating floor of the containment building.

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

This design change does not involve an unreviewed safety question.

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Benver Vo11cy Pow r Stotien Unit 2 1989 Report of Facility changes, Tests, and Experiments Page 65 of 87 CHANGE TITLE Deletion of Computer Point P0148D and Associated Annunciator Input for 2CHS-PT160 CRANGE DESCRIPTION 2CHS-PT160, which is the source for alarm point P0148D, is located between check valve 2CHS*1?O and the associated Reactor Coolant System (RCS) loop fill Motor Opercted Valves (MOVs) 2RCS*MOV556A, B and C.

Normally, this fill header is on)v utilized during filling of the RCS after an outage. During normal operation, Flow Control Valve (FCV) 2CHS*FCV160 is stroke tested by Operating Surveillance Test (OST) 2.47.3A (quarterly). This line is at that time pressurized by charging system pressure. The charging pressure exceeds the setpoint. There is no facility to bleed this pressure, so when this OST is performed, the alarm is initiated and remains in alarm. This design change will delete the alarm point P0148D from the shared input to annunciator window A2-3E and the plant computer.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased because this portion of the chemical and volume control system does not have a safety function and its failure will not affect the safety function of other equipment. There is no impact on the function of the safety-related portion of chemical and volume control system in UFSAR Section 9.3.4.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because this modification to the chemical and volume control system has no effect on consequences of an accident previously evaluated in the UFSAR Section 15.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the deletion of computer point P01480 and associated annunciator input for 2CHS-PT160 does not create a situation which would increase the probability of a malfunction. 2CHS-PT160 feeds a control room indicator 2CHS-PIl60 on the vertical board Section A.

This should provide sufficient operator indication.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this portion of the chemical and volume control system is not used to prevent or mitigate the consequences of an accident. There should be no effect on the consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR.

The design basis accidents were reviewed for potential impact by the proposed design change. There are no Section 15 accidents affected by this portion of the chemical and volume control system.

- B3avor Valloy Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 66 of 87 The safety systems which will be affected by the proposed design change include the chemical and volume control system and annunciator system.

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

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the failure of this equipment has no effect on safe plant operation and shutdown of the plant, and the failure of this equipment will not affect the safety function of other equipment.

This design change does not involve an unreviewed safety question.

. Besv0r VallCy P wcr St0ticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments

. Page 67 of P7 l-CHANGE TITLE Spent Fuel Holst Re-Rate 2MHF*CRN227 '

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l Presently, the existing 10-ton capacity hoist configuration of the spent fuel hoist does not provide adequate travel distance to provide the desired clearances between raised fuel assemblies and the fuel transfer system upender. The purpose of this design modification is to re-rate the spent fuel hoist to a dual rated (10-ton /2-ton i capacity) hoist. This will be provided by purchasing a 2-ton capacity l lifting clevis which can fit into the existing load block. The larger  !

10-ton capacity cranehook can be replaced with the smaller 2-ton i capacity lifting clevis for use during fuel handling operations, l thereby increasing the hoist travel distance enough to provide the

desired clearances. The mounting arrangement of the new and existing  !

hoist capacity nameplates will be designed for easy ,

interchangeability, allowing the hoist unit to be swapped back to 10 l tons when higher capacity is needed. I SAFETY EVALUATION

SUMMARY

l This design change is considered to be safe and does not present an unreviewed safety question, nor does it require a change to the Technical Specifications. The UFSAR will require revisions. By .

implementing this modification, the desired clearances between the l raised fuel assemblies and the upender will be provided without )

l degrading the spent fuel hoist's ability to handle light loads. l l l l

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- Benvor Volicy Pow;r St0tiCn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 68 of 87 CHANGE TITLE BV-2 Large Bore Primary Component Snubber Elimination l CHANGE DESCRIPTION 1

l Changes to 10CFR50 Appendix A, GDC 4 allows for the exclusion of reactor coolant loop breaks as a design basis. The purpose of this design change is as follows:

, 1) Remove thirty (30) of the thirty-six (36) large bore i Bergen-Paterson snubbers presently installed on the steam I

generator and reactor coolant pump supports to mitigate a pipe rupture event. The snubbers to be removed include all of the lower snubbers on the steam generator and reactor coolant pump frames, and two of the upper four snubbers on each steam generator l ring. Rigid struts will be installed in place of the six (6) j upper steam generator ring snubbers removed.

l Remove six (6) snubbers on the 8" bypass lines (two per loop).

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3) Document the elimination of surge line snubbers 2RCS-PSSP-001A and 2RCS-PSSP-001B, which were removed during pre-service inspection.

l 4) optimize / eliminate approximately 80% of the snubbers on the t following lines:

a) Three (3) 6" safety injection lines b) Two (2) RHR lines c) Three (3) 2" RTD manifold lines d) Three (3) safety injection accumulator lines

5) Replace remaining PSA-1/4 and PSA-1/2 snubbers, which cannot be eliminated by analyses with like replacements due to their high failure rate.

6) Deconing snubbers identified for disposal as needed.

By performing these modifications, financial benefits of decreased l plant downtime, reduced man-REM exposure, and reduced maintenance efforts will be provided.

- Benv0r Vallcy Pow;r StGticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 69 of 87 SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased. Dynamic analyses will be performed to show that, with the removal of snubbers and the ,

installation of struts in place of some of these removed snubbers, the Reactor Coolant System (RCS), Safety Injection System (SIS) and Residual Heat Removal System (RHR) piping, the Resistance Temperature Datector (RTD) and Safety Injection (SI) accumulator lines, and these systems components and supports will remain within their design allowable stress linits as shown in UFSAR Table 5.4-21 and 5.4-26.

Additionally, Leak-Bofore- Break (LBB) analyses will be reviewed to ensure that the flu.d leakage from a postulated defect at the highest stress location con:urrent with minimum material properties can still be detected well before the rupture of the pipe by using the Reg.

Guide 1.45 qualifie d, Leak Detection System. A seismic evaluation of the reactor coolant. and branch piping using ASME Code Case N-411 will also be performed to evaluate the acceptability of increased seismic displacements due tc the elimination of the snubbers.

The consequence of an accident previously evaluated in the safety analysis report will not be increased. This modification is made in accordance with 10CFR50, Appendix A, GDC 4 to eliminate dynamic effects associated with primary coolant loop postulated pipe 4

ruptures. The LBB analyses will be reviewed to ensure that the conclusions of the WHIPJET Program are not adversely affected, thereby not increasing the consequences of pipe whip or jet impingement accidents already analyzed.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. Analyses will be performed and reviewed to determine the margin of safety. The margin of safety for all affected components will be compared with the design bases to ensure that the proper i safety factors exist. The operability of safety-related equipment I

will not be affected by this design change.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. The replacement struts will be designed for stiffness and acceptability under load conditions. Spring hangers will be analyzed j and may have their existing cold load settings changed, and will be increased in size or replaced with new struts in order to meet the design objective. In addition, the PSA 1/4 and PSA 1/2 snubbers, l which cannot be eliminated by analyses, will be replaced with like pin-to-pin snubbers, due to the high failure rates experienced with the PSAs.

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Be vOr Volicy P:wcr Stoticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 70 of 87 The possibility for an accident of a different type than previously l evaluated in the safety analysis report will not be created. By making use of the Leak Detection System, along with the review of the WHIPJET Program, the LBB assumptions will still be valid, and a i controlled plant shutdown can be made before any pipe analyzed by this  ;

design change, could rupture. In addition, surge line stratification I effects, which is not currently evaluated in the UFSAR, will be addressed and incorporated into the design analysis.

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. 10CFR50, Appendix A, GDC 4 states that dynamic effects associated with postulated pipe ruptures of primary coolant loop piping may be excluded from the design basis when '

analyses demonstrate the probability of rupturing such piping is '

extremely low under design basis conditions. The analyses reviewed i and performed under this design change will ensure that the analyzed piping will leak before it ruptures even in the event of an earthquake. The possibility of a malfunction of equipment due to pipe I whip or jet impingement will not be created.

This design change does not involve an unreviewed safety question.

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Be0v0r Vollcy pow 0r StGticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 71 of 87 '

CHANGE TITLE '

BV-2 Small Bore Snubber Optimization l

CHANGE DESCRIPTION The purpose of this design change is to eliminate snubbers on small bore piping due to their poor reliability, and failures which lead to increased testing, radiation exposure and analyses to qualify failures. Due to analytical and economic considerations, as well as operational considerations, this design change will target the optimization of the subject piping listed in the Design Input Index i Section 1.0. By making use of the ASME Code Case N-411 damping '

values, it is expected that a minimum of 60 percent of the 400 or so snubbers associated with the target piping can be eliminated. This will result in approximately 250 snubbers being removed. The modifications will be specifically defined once the analyses are complete, then Engineering will issue those blocks of snubbers or j modifications to a line which must be performed in its entirety in I order to ensure code compliance. By performing these modifications, financial benefits of decreased down time, reduced man-REM exposure, and reduced maintenance efforts will be provided. i SAFETY EVALUATION

SUMMARY

I The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased because seismic analyses will be performed on the piping using ASME Code Case N-411 damping values to ensure that the seismic design of the piping is maintained. In addition, the existing lines analyzed by WHIPJET, as l discussed in UFSAR Section 3.6B.2.1, will be reviewed to ensure that the Leak-Before-Break conclusions are still valid. This will demonstrate that the fluid leakage from a postulated defect at the highest stress location concurrent with minimum material properties can still be detected well before the rupture of the pipe by using the l BV-2 Leak Detection System.

l The consequence of an accident previo aly evaluated in the safety analysis report will not be increased since the minimum requirements of essential safety-related systems and structures are not I

compromised, the plant can be safely shut down, and offsite does in excess of applicable guidelines will not occur. This design change will ensure that adequate clearance still exists between piping, components, and adjacent structures due to increased pipe movement which could occur. Protection against the dynamic effects of postulated pipe ruptures, where applicable, will be maintained, and the criteria of UFSAR Sections 3.6N.2.3.2 and 3.6N.2.3.2.2 will be l

met. This design change does not affect the design of the Engineered Safety Features (ESP) systems.

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• Benvor Valloy Power Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 72 of 87 The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because any increases in valve accelerations will be evaluated to ensure operability and compliance with existing equipment qualification standards. Nozzle load evaluations will also be performed to justify the increased nozzle loads, where applicable.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased. Pin-to-pin replacement struts will be installed, where needed, in place of the eliminated snubbers, and some of the spring hangers may either require upgrading or be replaced with struts in order to meet the design objectives. The PSA 1/4 and PSA 1/2 snubbers which cannot be eliminated per the snubber reduction techniques will be replaced due to their poor reliability and high failure rates. In addition, those snubbers which remain on an optimized system may be required to be replaced with larger snubbers in order to ensure compliance with the analysest however, this is not expected.

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

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

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ensure compliance with the code.

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

This design change does not involve an unreviewed safety question.

l B00VOr VallCy P wor StCticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 73 of 87 i

CHANGE TITLE Turbine Lube Oil Reservoir and $6 Bearing Pedestal Vacuum Instrumentation

! CHANGE DESCRIELI,QH l

The Turbine Lube Oil Reservoir Vapor Extractor (2TML-FN202), located on top of the oil reservoir, creates a slightly negative pressure above the reservoir oil level and in the oil drain piping connected from the oil reservoir to the turbine pedestals. The negative pressure prevents the discharge of oil or oil vapor from the turbine rotor oil seals.

The amnunt of vacuum in the oil reservoir and turbine bearing pedestals is controlled by a manual butterfly valve. As stated in the vendor instruction manual, the butterfly valve should be adjusted to obtain a 1 to 3 in. HO 2 vacuum in the last low pressure turbine bearing pedestal (#6 bearing pedestal).

presently there is no instrumentation installed to indicate the aucunt of vacuum in the turbine lube oil drain system.

The objective of this design change is to permanently install a 0-10 in. HO 2 vacuum gauge to indicate the pressure in the turbine #6 bearing pedestal and a 0-30 in. HO 2 vacuum gauge to indicate the pressure in the turbinc lube oil reservoir.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not he increased because this modification does not affect the operation of the vapor extractor or the lube oil system, but will provide local indication of turbine lube oil reservoir and #6 bearing vacuum.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because the gauges installed by this modification will provide no control functions. These guages will provide local indication only. The consequences of a failure of the vapor extractor are not changed by this modification.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because this modification does not affect any safety related 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 modification does not affect any safety-related equipment, including the turbine trip system.

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c Be3vor Valicy P,w0r Staticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 74 of 87

.; The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because no new accident scenarios are created by this modification. Loss of indication due to internals failure of the gauge will cause no system failures. Rupture of tubing will cause a small vacuum, but will not cause a loss of vapor extractor operation.

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

The design change does not involve an unreviewed safety question.

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- Benvor Valloy Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 75 of 87 CHANGE TITLE Turbine Trip System Modification CHANGE DESCRIPTION This modification is to remove the automatic trip on low electro-hydraulic fluid pressure function and the redundant electronic overspeed protection trip function from the turbine emergency trip cabinet. Both of these trip functions were shown to be unreliable during the pre-operational and start-up testing phases, and have been

" temporarily" disabled since, and have remained disabled during the first full cycle.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased because the low electro-hydraulic fluid pressure and the electrical overspeed circuitry, to be disabled, is entirely contained in a chassis of the Emergency Trip Cabinet. This trip circuitry was provided as a

" backup" electrical overcpeed trip. The primary electrical overspeed trip circuitry is located in the electro-hydraulic control cabinet.

The design basis for the redundant electrical overspeed is to provide a turbine trip signal in the event of a " stub shaft" failure, in which the electrical-hydraulic control system loses its speed signals and the mechanical overspeed trip device fails to operate. The complete failure of the turbine stub shaft is not considered a credible event.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because the removal of the low electro-hydraulic fluid pressure and redundant overspeed protection turbine trip function will not adversely affect the operation of the Turbine Generation System.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the removal of the low electro-hydraulic fluid pressure and the redundant overspeed trip functions decrease the possibility of a spurious turbine trip.

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

The possibility for an accident of a different type than previously evaluated in the safety analysis report will not be created because the redundant emergency trip system electrical overspeed is not considered in the basis of the turbine overspeed protection and is not credited in the UFSAR " turbine missile" generation probability analysis.

i Benvor Volicy Powor station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 76 of 87 The possibility for a malfunction of equipment important to safety of a different type than previously eva4uated in the safety analysis report will not be created because no safety-related equipment is affected by this modification.

The margin of safety as defined in the basis for any Technical Specification will not be reduced because the removal of the electro-hydraulic fluid prossure and redundant overspeed protection turbine trip function does not impact the technical specification surveillance requirements.

This design change does not involve an unreviewed safety question.

- Be0VOr VollGy P:wcr StOticn Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 77 of 87 CHANGE TITLE LPMS Alarm Inhibit from Rod Control CHANGE DESCRIPTION currently, alarms are received on the Loose Parts Monitoring System (LPMS) Reactor Vessel Head Channel whenever the control rods are moved.

This is to be expected during and for a brief moment following rod motion unless an input is used to inhibit the Digital Loose Parts '

Locater (DLPL) of the LPMS from alarming. This inhibitor was not part of the original design, and consequently numerous false alarms have resulted.

The purpose of this design change is to install a control rod drive

! inhibit box, purchased from Babcock and Wilcox, in the LPMS Cabinet (PNL-2 LPM). This will eliminate false LPMS alarms due to control rod movements, and thereby increase the reliability of the LPMS.

The inhibit box will be connected to an output of the rod control cabinet via the existing cable connecting the rod control cabinet, in the rod control area, to computer point C0098D, in the Control  !

Building. An additional relay will also be installed either in the l Relay Rack, in the computer cabinet, or in a separate terminal box.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased. The LPMS is a l

non-safety-related system which monitors for the presence of loose ,

I metallic parts within the RCS. The inhibiting of the LPMS alarms during l l control rod movement will not increase the probability of an accident  ;

evaluated in UFSAR Section 15.4.

Moreover, loose part monitoring will still be provided prior to and shortly after rod movement to detect any potential loose parts.

I The consequence of an accident previously evaluated in the safety )

analysis report will not be increased. This design change will not l affect any engineered safety features or Category I equipment, or their 1 ability to limit the consequences of previously evaluated accidents, l

The probability of a malfunction of equipment important to safety as

! previously evaluated in the safety analysis report will not be increased. The LPMS is a non-safety-related system. This modification will not affect Class IE equipment and will be seismically mounted such that its failure will not affect equipment which is important to safety.

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

because this design change will not affect any equipment which is important to safety.

This design change does not involve an unreviewed safety question. l l

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- Besvor Valloy Powor Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 78 of 87 CHANGE TITLE

( EAL Insulation CHANGE DESCRIPTION This modification proposes to insulate the Emergency Air Lock (EAL) barrel and outer door. This will help to conserve the heat that is introduced to the EAL's reactor end and will prevent its metal surface from being chilled to below its ASME allowed lowest service metal i temperature. The insulation will be multi-layered and sectionalized

' for ease of removal and handling. Test connections and the locking flange and sight glass at the outer door will remain uncovered.

SAFETY EVALUATION

SUMMARY

i The probability of an occurrence or the consequences of an accident as J previously evaluated in the UFSAR will not be increased. The addition of the EAL insulation will have no effect on any of the previously analyzed accidents. The addition of the insulation to the EAL will be analyzed to assure compliance with seismic requirements.

_ The probability of an occurrence or the consequences of a malfunction of equipment important to safety will not be increased. The insulation will not affect the function of any safety-related component or structure including: the concrete containment building, the steel liner, the equipment hatch, and the EAL.

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

This design change does not involve an unreviewed safety question.

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Benv0r Volloy P; wor Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 79 of 87 CHAl!GE TITLI FTS Upender Shock Absorbers Deletion (2FNT-TFT22)

CHANGE DESCRIPTION The purpose of this design change is to remove the shock absorbers currently installed on the Fuel Transfer System (PTS) upender. This will be accomplished by cutting the tie-wires and unbolting four bolts on each shock, then removing each shock absorber assembly from its mounting pedestals. The reason for this modification is that the shock absorbers have a tendency to stick in the extended and fully compressed positions, thereby either prohibiting operation of the FTS or preventing the shocks from functioning. Shock absorbers were added to the FTS design as a result of a cable break accident at another operating plant and are not required for normal PTS upender operation. A cable inspection will be performv1 during dry checkout according to refueling procedures. Additionally, DCP 1223 will install a GEMCO resolver which will stop the upender hoist prior to overtensioning the upender cable, thereby reducing the possibility of a cable break accident.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident previously evaluated in the safety analysis report will not be increased. The FTS upender shock absorbers were included in the design to prevent damage to a fuel assembly while in the upender, should a break occur in the upender cable. The shock absorbers are not required for normal operation of the FTS and their removal will not increase the probability of a fuel handling accident as evaluated in UFSAR Section 15.7.4.

The consequence of an accident previously evaluated in the safety analysis report will not be increased. The removal of the FTS upender shock absorbers will not have an effect on the fuel handling accident consequences as previously evaluated in UFSAR Sections 15.7.4.2 and 15.7.4.3.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because the removal of the shock absorbers will not increase the probability of the upender cable failing. Moreover, this modification should help to enhance the operability of the FTS by eliminating faulty components which are not required for normal operation.

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- Benvar Valloy Pcwcr Stotien Unit 2 L '

1989 R6 port of Facility Changes, Tests, and Experiments l page 80 of 87 l

The consequence of a malfunction of equipment important to safety as i previously evaluated in the safety analysis report will not be l l increased. Failure of the shock absorbers to perform their function  ;

l is not addressed in the UFSAR; however, the shocks are not required for normal operation of the FTS and would only help to prevent damage to a fuel assembly should the upender cable happen to break. The consequences of a dropped fuel assembly is already addressed in UFSAR Section 15.7.4, and would not be increased due to the implementation of this design change.

The possibility for an accident of a different type than previously evaluated- in the safety analysis report will not be created because the removal of the shock absorbers will only result in a fuel assembly being dropped and damaged if the upender cable breaks. This type of accident is bounded by the fuel handling accident evaluated in UFSAR Section 15.7.4.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the removal of the FTS upender i

shock absorbers will not affect the operability of any safety-related equipment nor cause any to malfunction.

This design change does not involve an unreviewed safety question.

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Beavor Valloy Pow 0r Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 81 of 87 g

QUANGE TITLE Permanent Utility Tie-Ins for the Outage Trailer Complex CHANGE DESCRIPTION Presently, a temporary outage trailer complex is situated east of the paved roadway adjoining the Guard House and the Unit 2 Control Room.

The purpose of this design change is to provide a permanent utility tie-in to the outage trailer complex. Domestic water supply, sewage connection, fire protection (sprinkler), water and telecommunication connection will be provided on a permanent basis. The installation will eliminate all the temporary utility connections that have laid on the walkway which cause a personnel safety hazard.

SAFETY EVALUATION

SUMMARY

The probability of occurrence of an accident prev 30usly evaluated in the safety analysis report will not be increased because the proposed modification is not safety-related, and in no way would its inoperability affect plant operation.

The consequence of an accident previously evaluated in the safety analysis report will not be increased. Because the design change is located outside, the failure of this trailer complex and its associated utilities do not impact safety-related equipment and therefore affect no accident consequences.

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

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

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

evaluated in the safety analysis report will not be created because these changes do not impact the operability of any equipment required for the plant safety.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because the proposed modification is not safety-related, and it is located outside of safety areas. Therefore, no impact to the safety-related equipment will occur.

This design change does not involve an unreviewed safety question.

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9 Bnavar Valley Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 82 of 87 CHANGE TITLE Containment Air Recirculation Cooling coils Piping Modifications

. CHANGE DESCRIPTION Containment Air Recirculation Fan (2HVR*FN201B) cannot be removed for maintenance because some piping and pipe supports are in the lifting and removal paths. The purpose of this design change is to modify those sections of piping and supports to allow for removal of the Containment Air Recirculation Fan (2HVC*FN201B), Cooler inlet lines 2SWS-004-335-3 and 2SWS-004-331-3 *will be cut and flanged to facilitate future fan removal. Pipe supports 2SWS-PSA913X, 2SWS-PSR006X, 2SWS-PSR007X, 2SWS-PSR923X, and 2SWS-PSR924X will be redesigned to allow for removal of (2HVR-FN201B).

SAFETY EVALUATION

SUMMARY

The probability of an occurrence of an accident previously evaluated in the safety analysis report will not be increased because no changes to any piping flow paths are being made by this modification. Pipe support modifications will be analyzed to meet the original design criteria per QA Category I and Seismic Category I requirements.

The consequence of an accident previously evaluated in the safety analysis report will not be increased because no changes are being made that affect the consequences of a Loss of offsite Power Accident (LOOP) as described in Section 8.2 of the BVPS-2 UFSAR.

The probability of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because no changes are being made that would affect the operation of any safety-related equipment during accident conditions.

The consequence of a malfunction of equipment important to safety as previously evaluated in the safety analysis report will not be increased because 50% redundancy exists during normal operation (1 extra fan and cooler) and 200% redundancy exists (2 extra fans and coolers) during a Loss of Offsite Power Accident to maintain the containment air temperature as described J.. Section 9.4.7.1 of the UFSAR.

The possibility for an accident of a different type than previously evaluated in tha safety analysis report will not be created because no changes are being made to any fluid flow paths, and the pipe supports being modified will be designed to the existing seismic requirements.

The possibility for a malfunction of equipment important to safety of a different type than previously evaluated in the safety analysis report will not be created because all credible failure modes are diucussed in Section 9.4.7 of the UFSAR concerning the operation of the containment Air Recirculation fans.

This design change does not involve an unreviewed safety question.

e B2 aver Valley Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 83 of 87 CHANGE TITLE Modification to the containment Air Recirculation (CAR) Fans CHANGE DESCRIPTION The stress analysis for the Containment Air Recirculation System (CARS) fan housing concluded that the housing would retain blade fragments subsequent to a blade failure. The 1/4" carbon steel housing for fan 2HVR-FN201B failed; therefore, a missile shield will be designed and added to CARS fans 2HVR-FN-201A, B, and C.

Additionally, the CARS fan supports will be modified to rectify a natural frequency vibration problem.

SAFETY EVALUATION

SUMMARY

The proposed design change will not involve an unreviewed safety question.

The proposed change installs missile shields to the Non-Nuclear Safety (NNS) CARS fans. This will provide missile protection for other components in the area. Elimination of vibration will enhance fan operation. The CARS fans are Seismic Category I, and the missile shields will be installed to seismic requirements.

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

Containment Ventilation Systems are included in Technical Specifications 3/4.6.3 (isolation valves) and 3/4.9.9 (Containment Purge); however, neither of these is affected by this change.

Containment Air Recirculation System is not otherwise included in the technical specifications.

The proposed change will require changes to the Updated Final Safety Analysis Report.

FSAR Table 3.5-12 and Section 3.5.1.2 must be revised to include reference to these missile shields.

Implementation of this DCP will not involve an Unreviewed Environmental Question and will not change the Environmental Protection Plan because this change does not concern effluent pathways and does not affect the environment.

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1989 Report of Facility Changes, Tests, and Experiments Page 84 of 87 CHANGE TITLE Acceptance of Temporary Fire Loop CHANGE DESCRIPTION The original temporary fire loop was installed in the early seventies j to serve temporary structures during the construction phase of Beaver Valley Power Station. Due to the evolution of the site over the years, l

it is now desirable to dedicate this temporary fire loop as an i acceptable permanent installation. Appropriate flow diagrams need updated to indicate the merger of Unit 1 and Unit 2 fire loops.

Not within the scope of this TER is the fire pipin~g internal to the protected buildings. This would include wet sprinkler systems and hose reels within the buildings. Internal fire protection systems were designed and installed under a separate contract. Turnover of internal systems will be handled on a case-by-case basis in conjunction with the i turnover of the building. '

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1) Yard Fire Line Piping (underground) l 2) Post Indicating Valves (sectional control valves)

3) Curb Box Valves (sectional control valves) i 4) Hydrants l 5) Hydrants Valves During the construction phase of Beaver Valley Power Station Units 1 and 2, an underground yard fire line was installed to provide fire -

protection service to various temporary structures and permanent structures during construction.

Due to the evolution of the Beaver Valley site, it is now desirable to dedicate the temporary portion of the yard fire line as an acceptable installation. This. need has arisen since the temporary structures which were supplied by the temporary fire loop are now being dedicated as permanent structures.

The scope of this TER is to perform an evaluation and provide justification for dedicating the temporary fire line as an acceptable permanent installation at Beaver Valley Power Station. This TER will:

1) Provide a technical review of the temporary fire loop installation and documentation to determine compliance with applicable codes and regulations.

2) Revise plant drawings to identify aporopriate pipelines, valves and hydrants.

3) Assign equipment identification numbers to appropriate components.

4) Recommend pre-turnover tests to be performed.

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1989 Report of Facility Changes, Tests, and Experiments Page 85 of 87 SAFETY EVALUATION

SUMMARY

Fire lines being accepted were installed and tested to NFPA requirements per 9.10.1 and inspected to meet requirements of 9.10.4 for water systems (BVPS-1). The fire lines meet requirements of Section 9.5 of the BVPS-2 UFSAR. The overall system's ability to suppress a fire is not affected. Since overall performance of the fire suppression system is not pffected, safe shutdown as described in both Units' Appendix R is stall valid. Offsite doses are not increased.

The newly accepted lines are QA Category 2 and are separated from the plant's Category F system by four PIV's identified in Site Administrative Procedure (SAP 9D), " Fire Protection", Figure 4.

Accepting the lines as Category 2 vs. Category F does not affect safe shutdown of either plant per Appendix R.

A break in a Category 2 portion of this system can be isolated, and thereby protects the Category F portion which protects the plant and ensures safe shutdown capability.

All- buildings being protected fall under the combined Unit /

Non-category F portion of Figure 5 of SAP 9D. The Category F portion of the fire loop can still be isolated if there is a line break in the Category 2 portion.

Continued testing of the Category 2 portion of the fire loop will maintain its reliability. Testing of the Category 2 portion will meet the same requirements as the category F portion.

Per UFSAR Section 9.10, Technical Specifications were placed in SAP 9D. This SAP already addresses the addition of these lines in Figures 4 and 5.

This change does not involve an unreviewed safety question, l

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Baavar Valley Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 86 of 87 CHANGE TITLE Add Block Valves to the IAC System to Isolate 21AC-SOV235A and B CHANGE DESCRIPTION Engineering Memorandum (EM) 63242 requested the addition of isolation valves into the Instrument Air Containment (IAC) System to enhance Maintenance's ability to work on Solenoid Operated Valve (SOV) 235A and B. Temporary tags were attached to each valve in the field.

This TER documents the engineering review and processes the drawings supporting this modification which adds these valves to the IAC System permanently.

SAFETY EVALUATION

SUMMARY

Per UFSAR Section 9.3.1.3.3, the IAC System is non-safety related and is not required for safe shutdown.

The consequence of an accident previously evaluated in the UFSAR does not involve the IAC System since it is not required for safe shutdown.-

Instrumentation and controls served by the system are designed such that the equipment will fail in the safe mode upon loss of air. There is no increase in probability.

The consequences of a malfunction remain constant. This system has been evaluated to allow components served by this system to fail in the safe mode upon loss of air.

A hazard design review has been completed. An accident of a different type than any previously analyzed was not identified.

A hazard design review has been completed. A malfunction of a different type than any previously evaluated was not identified.

This system is not addressed in the Technical Specification.

therefore, the margin of safety is not addressed.

This change does not involve an unreviewed safety question.

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• B*avar Valley Powar Station Unit 2 1989 Report of Facility Changes, Tests, and Experiments Page 87 of 87 CHANGE TITLE Addition of Clean Out Parts to the Turbine Lube Oil Demister 2TML-DMST21 Drain Line CHANGE DESCRIPTION Mechanical Maintenance has installed flanges and clean outs in the Unit 2 turbine lube oil demister drain line. This TER documents this modification.

SAFETY EVALUATION

SUMMARY

The demister drain line is not a nuclear safety class system. No accident previously evaluated involved this open-ended drain line.

The consequences of an accident previously evaluated in the UFSAR does not involve the turbine lube oil demister drain line since it is non-nuclear safety class.

The demister drain line is not associated with any equipment important to safet). Therefore the probability of a malfunction of equipment important to safety is not affected.

The consequences of a malfunction of equipment important to safety is not affected by the turbine lube oil demister drain line.

Per UFSAR Section 15.2.3, turbine trip has been analyzed. This is the only accident possibly caused by this system. No additional malfunction is associated with this drain line.

This syster is not addressed in the Technical Specification.

Therefore the margin of safety is not addressed.

This change does not involve an unreviewed safety question.

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