ML20247H582

Annual Rept of Facility Changes,Tests & Experiments for 1988

ML20247H582
Person / Time
Site:	Beaver Valley
Issue date:	12/31/1988
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8907310073
Download: ML20247H582 (88)

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?,- r* eaver Valley Power Station Shippingport, PA 16077 0004 v Pros wn=ee Group (81216434256 n

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, July 20, 1989 U. S. Nuclear Regulatory Commission Attn: Document Control-Desk washington, DC 20555 l

Reference:

Beaver Valley Power Station, Unit No. 1

Docket No. 50-334, License No. DPR-66 l 1988 Report of Facility Changes, Tests and Experiments Gentlemen

1 This letter forwards the 1988 Annual ' Report of Facility L -Changes, Tests and Experiments, in accordance with 10 CFR 50.59.

-The report covers the period January 23, 1988 ~through January 22, 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, be f J. D. Sieber

' Vice President Nuclear Group cc: Mr. J. Beall, Sr. Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr. P. Tam, Sr. Project Manager Director,. Safety Evaluation & Control (VEPCO) 8907330073 881231 PDR R ADOCK 0500033,4 pyg l

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4 g BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 TABLE OF CONTENTS Page(s)

Facility Changes . . . . . . . . . . . . . . . . . . 1 - 45 Procedure Changes and Temporary Modifications . . . 46 - 85

.UFSAR Change . . . . . . . . . . . . . . . . . . . . 86 i

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'BVPS-1 Annual Report of Facility Changes,.

Tests and Experiments for 1988 FACILITY CHANGES ENGINEERING DESIGN CHANGE NO. 513 CHANGE TITLE-New Style Rockwell-Edwards T-58 Valves DESCRIPTION AND PURPOSE OF REVISION-Rockwell-Edwards has discontinued production of the Model #3624 Globe Valve (Westinghouse designation is T-58). Replacement parts will only be available for 2 or 3 more years. The. purpose of this modification is to evaluate if the new style Rockwell-EdwardsLModel #36124 Globe Valve is a suitable replacement for use at Beaver Valley Pover- Station Unit No. 1. The task of this Design Concept is for Engineering to evaluate the feasibility of using;the never style valve to meet the same functional requirements when' replacement is made by Power Stations on an "as needed" basis.

SUMMARY

OF SAFETY EVALUATION The Safety Analysis is performed on the basis that'the overall function of the Globe Valve vill not change and also the affected system function vill not be changed. 'The. five systems affected are identified in the " Boundaries of Change" Section of1this concept.

This modification does not necessarily fall into a category of a design change.

to the operating plant as defined in 10CFR50, Section 59(a)(1). Replacement of

-the T-58 valve as outlined in the Design Concept does not change the facility or procedures as identified in the Updated Final Safety Analysis Report (UFSAR) or change the Technical Specifications. The primary reason for the valve replacement is a maintenance function.

One of the major items to be considered for the independent Engineering evaluation is that the replacement valve (Rockwell Model #36124) meets or exceeds the design code requirements for the applicable system as identified in the Updated Final Safety Analysis Report. Seismic analysis should address placement of the new valve, whether lighter or heavier in weight, into the l

entire system as evaluated in the Stress Analysis Report.

l The independent engineering ( uation concluded. that replacement of the T-58 1

' Valve Model 3624 vith Model 36124 is safe and does not create an unreviewed safety question.

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1 BVPS-1 Annual Report of Facility Changes, i Tests and Experiments for 1988

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i ENGINEERING DESIGN CHANGE No. 523 CHANGE TITLE l Install Instrument Air Dryers b

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l DESCRIPTION AND PURPOSE OF REVISION This design change vill improve the. Station Air Compressor and Aftercooler portion of the Turbine Plant Component Cooling Vater System and the Instrutent Air portion of the Compressed Air System.

The design change vas initiated prior to issuance of Generic Letter 88-14,

" Instrument' Air Supply System Problems" and Institute of Nuclear Power Operations Significant Operating Experience Report (SOER) 88-1, " Instrument Air System Failures". An assessment of the instrument air systems at Beaver Valley Power Station Unit No. I was performed in response to Generic Letter 88-14.

The assessment results are documented in a letter from J. D. Sieber to the NRC dated February 17, 1989. The station's response to SOER 88-1 is also presented in the referenced letter.

SUMMARY

OF SAFETY EVALUATION This design change vill improve the operation of the Compressed Air System.

Also, as it is stated in Updated Final Safety Analysis Report Section 9.8.1, "No part of any safety-related equipment requires the supply of compressed air for shutdown."

Implementation of this design change is considered to be safe. No change to the Technical Specifications is required. A change to Updated Final Safety Analysis Report Section 9.8, " Figure 9.8-1, and Table 9.8-1 is required as a result of this change.

1 Page 3 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 576 CHANGE TITLE Modifications of the Air Start System and Compressors for the Diesel Generator DESCRIPTION AND PURPOSE OF REVISION The purpose of this modification is to improve the reliability of the Diesel Generators' Air Start Systems by modifying the air start compressors and by installing additional check valves, pulsation chambers, aftercoolers, air dryers, and associated piping, filters, valves, and instrumentation to the air start piping. Also, air start motor blovdown valves vill be replaced because the presently installed valves leak and no spare parts are available to repair them; they vill also be relocated for improved accessibility.

Redundant check valves vil! be installed downstream in series to allow for downgrading the compressors to 0.A. Category II as required per American National Standard N18.2a-1975/ANS-51.8, since the compressors are not required for safe shutdown of the plant.

SUMMARY

OF SAFETY EVALUATION This design change is safe to implement. No unreviewed safety question is involved nor are any Technical Specification bases affected. No changes to the Updated Final Safety Analysis Report or the Technical Specifications are required.

These modifications do not affect the ability of the air start system to ,

provide five consecutive diesel generator starts as stated in Section 8.5.2.3 of the Updated Final Safety Analysis Report, nor vill these modifications 3 increase the probability of failure of the air bottles as described in Section 8.5.2.4. Single failure criteria described in Section 8.5.2.6 are not affected. These modifications do not affect the response of the diesel generators to a station blackout (loss of offsite pover) of the emergency buses as described in Section 14.1.11 of the Updated Final Safety Analysis Report.

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Page 4 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING SECTION CHANGE NO. 589 CHANGE TITLE Fire Damper Linkage DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this modification is to: inspect existing fire dampers, assign mark numbers to each fire damper, supply drawings'for each type of fire damper, issue MVR's to repair or replace any defective fire dampers, and qualify existing fire dampers and control room HVAC motor operated smoke dampers for a 1 1/2 hour fire rating. This modification vill also insulate control room HVAC ductwork and install a Seismic Category I fire damper in the control room restroom exha m .u. Furthermore, it vill install access panels in ductwork to any inaccessible fire dampers, reinstate the C07 protected areas, and revise Appendix R Fire Hazards Analysis Report to make tne primary auxiliary building a single fire area.

SUMMARY

OF SAFETY EVALUATION The modifications made under DCP 589 are considered to be safe since they do not present an unreviewed safety question nor do they affect the Bases of the Technical Specifications. This design change should improve the reliability of the Plant's fire dampers to perform their intended function.

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Page 5 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 645 I

I CHANGE TITLE Primary Grade Deaerator Vacuum Discharge Piping Modifications DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to re-route the Primary Grade Water Deaerator Vacuum discharge piping so that it will be released via a monitored path (the process vent of the -ooling tower). This change eliminated the ,

potential for unmonitored dischartes through the deaerator vacuum discharge piping resulting from mechanical '.ailures (such as diaphragm failure) in the deaerator.

SUMMARY

OF SAFETY EVALUATION This design modification to the primary grade water deaerator vacuum discharge piping is considered safe and does not constitute an unreviewed safety question, nor does it affect the base of Technical Specification 3/4.11.2.1.

Additionally, a means of sampling the primary grade deaerator vacuum discharge vent is now provided. This modification vill not increase the possibility of a hydrogen-air explosion nor vill it allow gaseous vaste to backflow into the Auxiliary Building.

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• Page 6 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 i

ENGINEERING DESIGN CHANGE NO. 648 CHANGE TITLE Modify Unit 1 Security Computer DE'.ORIPTION AND PURPOSE OF MODIFICATION This design change is to modify the existing Unit I security system computers and multiplex units as necessary to insure compatibility with the upgrade Unit '

II security system. The two existing Unit I security computers vill be removed and replaced with a new security computer which is to be installed in the guardhouse. Unit I multiplex units (MUX) vill be modified by replacing several cards and connectors in each MUX. Changes to the software and data base vill be required to insure compatibility with the Unit II security system.

SUMMARY

OF SAFETY EVALUATION The security computer system is an essential part of the overall plant security. This modification is designed to be integrated in steps so that the Unit I portion of the system remains operational at all times. Therefore the overall plant security will not be reduced during the modification phase.

Since this equipment is not safety related, this DCP vill not adversely affect the safety of the plant as considered in the Updated Final Safety Analysis Report and Technical Specifications and is, therefore, acceptable.

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. l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 654 {

CHANGE TITLE Emergency Response Facility (ERF) Air Conditioning and Intergraph Power Supply i

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DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to provide sufficient ventilation and electrical power to maintain the availability and reliability of the Intergraph Computer in the ERF. An additional 115,000 BTU of air conditioning vas installed in the ERF Computer Room, an additional 36,000 BTU of air conditioning was installed in the ERF Communications Room, and additional power conditioning and special grounding was installed for the Intergraph computer.

SUMMARY

OF SAFETY EVALUATION The proposed design change vill not involve an unreviewed safety question because the intergraph computers are not Class lE equipment and are not connected to any safety-related equipment. Failure or malfunction of the ERF ventilation system or the Intergraph computers will not challenge any safety systems or initiate any accidents.

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l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 1

ENGINEERING DESIGN CHANGE NO. 657 CHANGE TITLE Plant Chemistry Monitoring Instrumentation I

DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to upgrade the Beaver Valley Power Station Unit No. 1 plant chemistry equipment in the area of the Vater Treatment Demineralized Water Control System and the Steam Generator Blowdown.

Demineralized Vater Control System. Also, this upgrade is intended to improve the reliability and availability of the various systems requiring quality grade water.

SUMMARY

OF SAFETY EVALUATION Implementation of this design change is considered to be safe. No change to the Updated Final Safety Analysis Report or Technical Specifications is required as a result of this change. This modification vill upgrade the Plant Chemistry Equipment associated with the subject systems and is intended to also improve the reliability and availability of the various systems requiring quality grade water.

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BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l ENGINEERING DESIGN CHANGE NO. 668 l

CHANGE TITLE Inadequate Core Cooling Instrumentation System DESCRIPTION AND PURI'0SE OF MODIFICATION This modification till replace and combine the functions of the Reactor Vessel Level Instrumentation System and sub-cooling margin monitor. It vill also upgrade portions of the Incore Thermocouple System for environmental qualification. The purpose for this modification is to achieve compliance to NUREG 0737, Item II F.2 with exceptions as noted.

SUMMARY

OF SAFETY EVALUATION This design change is considered safe and does not involve an unreviewed safety question. No new accidents or malfunctions are created and the probability of a malfunction or accident is not increased. This design change vill necessitate a change to the Updated Final Safety Analysis Report and the Technical Specifications.

The inadequate core cooling instrumentation system vill provide only monitoring at safety related systems and vill not directly perform control functions to prevent or mitigate accidents previously analyzed in Updated Final Safety Analysis Report. The system vill provide information to the operators to mitigate the consequences of accidents and malfunctions of safety systems.

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' Page 10 )I BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE No. 680 CHANGE TITLE Replace Reactor Coolant Pump #3 Seal Clamp j DESCRIPTION AND PURPOSE OF MODIFICATION The purpose is to maintain or increase the operability of the reactor coolant pump No. 3 seal clamp for improved seal life. This modification eliminates the i No. 3 seal potentially tunning dry leading to boric acid crystallization and accelerated abrasive wear.

SUMMARY

OF SAFETY EVALUATION The change does not impact existing plant systems, which support or cor.pliment the operation of the reactor coolant pumps. The effect on the health and safety of the public has not been changed, as has been previously analyzed in the Updated Final Safety Analysis Report. The change is intended to increase the wear life of the number three reactor coolant. pump seal which leads to improved reliability of reactor coolant pump operation.

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Tests and Experiments for 1988 ENGINEERING DESIGN CEANGE NO. 693 CHANGE TITLE Anticipated Transient Vithout Scram (ATVS) Mitigating System Actuation Circuitry (AMSAC)

. DESCRIPTION AND PURPOSE OF MODIFICATION 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, 2) initiate auxiliary feedvater flow, and 3) isolate steam generator blowdown flow in the event of an ATUS. This vill- involve-installation of a logic cabinet with tie-ins to the feedvater system and the -

turbine trip. Steam generator blovdown isolation is initiated by a signal' from pressure switches in the discharge lines of the auxiliary feedvater pumps. The purpose of this change is to achieve compliance with 10CFR50.62.

SUMMARY

OF SAFETY EVALUATION ATVS is not a design basis event for Beaver Valley Power Station Unit No. 1 (b7PS-1). However, 10CFR50.62 requires AMSAC protection be backfitted into the design for additional defense against a possible ATUS event. An ATVS event is not currently evaluated in the Beaver Valley Power Station Unit No. 1 Updated Final Safet*' Analysis Report.

An analysisLof ATUS events at Westinghouse plants _ has been documented in VCAP-10858P-A and accepted by an NRC SER . issued by Mr. C. E. Rossi'on July 7, 1986. The analysis shows that RCS pressure is reduced to below the ASME Boiler and Pressure Vessel Code Level C' Service Limit stress criteria of_3200 psig for the Beaver Valley Power Station Unit No. 1 limiting accident, Complete Loss of Normal Feedvater, if AMSAC is implemented. Thus, installation of AMSAC improves the consequences following an ATVS.

The probability and consequences of a malfunction of either AMSAC or the Reactor Protection System and other current design equipment important to safety is not increased due to design features which provide protection against single failures ard other deleterious system interactions.

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Page.12 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 712 i CHANGE TITLE Vater Supply to Clean Intake Bays DESCRIPTION AND PURPOSE OF MODIFICATION Once or twice a year a flange and valve had to be installed in place of a check valve's internals (either RV-6 or RV-7) to support a diver who cleaned the river water intake bays. This modification vill install a 1-1/2" veldolet and a 1-1/2" bronze globe valve in line number 6"-SVV-6-121 between valves RV-313 and RV-9 so that vater can be supplied to support the divers without removing the internals of check valve (RV-6 or RV-7) when cleaning the intake bays.

SUMMARY

OF SAFETY EVALUATION This design change is safe in that the installation of 1-1/2" veldolet and 1-1/2" bronze globe valve between RV-313 and RV-9 vill 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 to the Updated Final Safety Analysis Report have been identified.

This modification vill not affect the river water systems and can reduce the number of man-hours required to clean the intake bays.

Page' 13 - l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l

p L ' ENGINEERING DESIGN CHANGE NO. 733

. CHANGE TITLE 4

Refueling Cavity Vater Level Alarm DESCRIPTION AND PURPOSE OF MODIFICATION On August 21, 1984 the reactor cavity seal ring at the Haddam Neck Plant failed resulting in draining- of approximately 200,000 gallons of water from the reactor cavity to the lower levels of the containment building in 20 minutes.

To comply with the NRC Information Notice 93 issued December 17, 1984 the

.Duquesne Light Operations Department has implemented a plan whereby~after the occurrence of a refueling' cavity lov vater level alarm the refueling operations

. vill be suspended until the situation is corrected.

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'Both the high and low cavity water level alarms are annunciating on the same annunciator vindow at the present time. It is better to provide separate high and lov alarms to let the refueling operators know if the water level is too-high. or 'too lov. Occurrences of a high water level alarm vill not unnecessarily delay the refueling operations and vill not increase outage costs or the possibility of impacting the overall outage schedule.

SUMMARY

OF SAFETY EVALUATION The design change is safe in that the separation of high and lov alarms will not affect any equipment important to safety as previously analyzed or create any nev malfunctions or accidents. The ' Technical Specifications are not l

affected and no changes to the Updated Final' Safety Analysis Report have been identified. This design change does not present an unreviewed safety question.

During the refueling operation, approximately 27 ft. of water is on top of the refueling cavity seal (at least 23 ft of water shall be maintained). It is necessary to suspend the refueling operation when the water level reaches lov level alarm because of the potential for uncovering a spent fuel assembly if the refueling cavity water seal failed.

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A Page 14 BVFS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 740 CHANGE TITLE Class lE Inverter Bypass

. DESCRIPTION AND PURPOSE OF MODIFICATION The 120V AC Distribution System is made up of vital distribution and non-vital distribution. There are four vital instrument busses and one non-vital instrument bus provided in Beaver Valley Power Station Unit No. 1. Each of these busses are powered from their own uninterruptible power supply (UPS).

The inverter in each of the UPS converts DC battery power into AC power.

Inverters are considered to be not absolutely reliable and are normally fitted with an automatic bypass feature to protect against an inverter failure. At the present time there is only one manual bypass existing for transferring the vital instrument loads from a failing power source to the respective alternate.

This lack of an automatic transfer feature opens the station to the possibility of tripping every time an inverter fails. This problem can be solved by installing a solid-state transfer switch on each of the busses so that each bus can be transferred from a failing source to its alternate source before the equipment povered from that bus vill be affected.

SUMMARY

OF SAFETY EVALUATION This design is safe in that the modification vill not affect any equipment safety related as previously analyzed or create any new malfunctions or accidents. This changed vill be restricted to the five instrument busses, the five UPS units and annunciator system. Other minor changes to non-safety i related equipment may be required. No changes to the Updated Final Safety Analysis Report have been identified. This modification is safety related.

The solid state switches should increase availability by preventing reactor trips caused by an inverter failure. The function of the svitch is identical to the manual switch, and will not increase the probability or consequences of any accident previously analyzed. The solid state switch should reduce the number of challenges to the reactor trip system due to inverter failures.

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• Page 15 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 1

ENGINEERING DESIGN CHANGE NO. 744 CHANGE TITLE ITE-HK Power Circuit Breaker modification l

DESCRIPTION AND PURPOSE OF MODIFICATION As discussed in INPO SER '5-83, at Limerick Power Station, pre-operational testing of 36 Brovt Boveri Electric (BBE) Model 5HK1200-350 circuit breakers identified one breaker which inadvertently closed on its own following completion of spring charging. .The breaker was returned to BBE (formerly ITE) for analysis. Subsequent investigation determined the problem was intermittent and could be corrected by adding a light spring to the breaker close latch.

The purpose of this design change is to perform the BBE recommended fix of installing a latch anti-shock spring to the SHK ITE breakers to improve their reliability.

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SUMMARY

OF SAFETY EVALUATION This design change vill perform the BBC Brown Boveri, Inc. (current vendor name) recomn; ended fix of installing a latch anti-shock spring to prevent inadvertent breaker closing following completion of the breaker's spring charging cycle. This vill improve the reliability of the breaker. This vill also, therefore, improve the reliability of the 4KV station service system and is, therefore, considered a safe, and even a desirable change. Since both Category I and Category II breakers are involved in this design change and to prevent problems on retrofitting, all parts for this modification vill be procured to meet Category I, Seismic l

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• • Page 16 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 748 CHANGE TITLE Live Load Valve Packing DESCRIPTION AND PURPOSE OF MODIFICATION This design change provides a relatively constant load on a valve's stem packing by using springs under the valve's gland nuts to minimize leakage during transients and wear conditions.

SUMMARY

OF SAFETY EVALUATION The proposed design change vill not involve an unreviewed safety question. The change is limited to the following non-safety related, Category 2 valves:

MS-12, MS-11, FCV-MS-100C. These valves are not used to prevent or mitigate the consequences of a malfunction of equipment important to safety as previously evaluated in the Updated Final Safety Analysis Report.

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l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 750 CHANGE TITLE Cooling for.the Rod Control /MG Room, Cable Tray and Switchgear Rooms 7

l DESCRIPTION AND PURPOSE OF MODIFICATION The scope of this design change includes the following:

1. Incta11ation of chilled water cooling coils to cool air exhausted from the switchgear area.

2. Installation of ductvork to direct cooled air to suction side of the switchgear area supply fan.

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3. Installation of packaged chiller units on the turbine deck to supply chilled water to the cooling coils. Chilled water putaps vill also be installed in this area.

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4. Installation of two booster pumps to supply rav vater to the l chiller units.

5. A inn speed controller vill be installed to control the switchgear ventilation fans (VS-P-17, VS-F-18) in order to maintain a constant static air pressure in the contiguous areas. This pressure vill be lover than the air pressure in the control room pressure beundary.

! 6. Permanently mounted cabinet coolers vill be installed on control l rod drive cabinets.

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SUMMARY

OF SAFETY EVALUATION l

This design change is safe and does not present an unreviewed safety question.

Installation of coolers on the rod drive power and logic cabinets vill increase the overall reliability of the system. Installation of speed controllers on the ventilation system fans vill provide additional assurance that the control room pressure boundary be maintained at a positive pressure relative to the svitchgear area. The chiller units and associated equipment vill increase the reliability of equipment in the affected areas. No changes to the Technical Specifications are required. An Updated Final Safety Analysis Report revision vill be required.

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BVPS-1 Annual Report of Facility Changes,  !

Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE No. 751 CHANGE TITLE Disk Pressurization DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this modification is to install permanent tubing to supply pressurized water to any Reactor Coolant Svstem Loop Stop Valve for disk pressurization from any Safety Injection Accumulator. When the loop isolation valves are to be closed the pressurized water is injected between the discs to assure an adequate seal. The installation of permanent tubing vill replace the practice of running temrarary hoses from the accumulators to the 3omp isolation valves. The tubing vill not be permanently connected to eitner the loop isolation valve or the accumulators. When needed connections to the Loop Isolation Valves and the accumulators from the tubing vill utilize temporary hoses via quick-disconnect couplings. Blind flanges presently installed at the loop isolation valves and the accumulator drain vill be replaced with flanges that have quick-disconnect couplings permanently installed.

SUMMARY

OFSAFETYEVALUf,fN Thir design changt is considered safe. No change or addition to the Technical Specifications is required. The modifications of this design change do not affect the operation of the Reactor Coolant Loop Stop Valves or the Safety Injec+. ion Accumulators during normal or accident conditions. An Updated Final Safety Analysis Report revision is required as a result of this change.

The flanges to be installed, although they are located on safety-related equipment, do not affect safety-related functions, nor do they act as boundaries for safety-related flovpaths. Furthermore, the pressurized water being utilized from the Saftsty Injection accumulators is borated to greater than 2,000 ppm of boric acid, which is equal to or greater than the boric acid concentration at any time of the Reactor- Coolant System. Thus, it does not introduce positive radioactivity into the Reactor Coolant System. The Loop Stop isolation valves disks are designed for use at pressures of 100 to 2,500 psig. inis system vill be utilized only during outages where work is being performed on the Reactor Coolant System in modes 5 or 6 when either the Overpressure protection System is in service or the vessel head is removed.

I Therefore, no new accident situations are created.

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. 1 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE No. 756 l CHANGE TITLE Primary Grade Vater Tank Modification l

DESCRIPTION AND PURPOSE OF MODIFICATION For this design change, the floating deck in each Primary Water Storage Tank will be replaced with a floating membrane in order to maintain the oxygen content of the primary water below 0.10 ppm. In order to protect the floating membrane, additional tank modifications (such as removing internal piping or relocating instrument connections) vill be performed as necessary.

Furthermore, an overflov line vill be added on each PG Vater Tank in order to direct any overflow to the Fuel Handling Building Sump. A loop seal of 4 feet of water is to be located just outside the tank nozzle to prevent any direct contact of air with the water inventory in the tank. The tank's natural syphoning freeze protection will be replaced with a forced pump flow arrangement. Freeze protection vill also include heat tracing the loop seal.

SUMMARY

OF SAFETY EVALUATION This design change is safe in that the new floating membrane for each Primary Vater Storage Tank vill not affect any equipment important to safety as previously analyzed or create any new malfunctions o: accidents. No change to the Technical Specific ations is required. A change to Updated Final Safety Analysis Report Figure 9.7-3 is required. This design change does not present an unreviewed safety quest 1 a

• • Page 20 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 I

ENGINEERING DESIGN CHANGE NO. 768 CHANGE TITLE j Additional Emergency Lighting - Appendix R DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to provide additional emergency lighting to comply with 10CFR50 Appendix R Section III.J vhich states " Emergency

, lighting units with at least an 8-hour battery power supply shall be provided in all areas.needed for operation of shutdova equipment and in access and egress routes thereto".

The lighting units to be installed by Design Change Package (DCP) 768 vill supplement the Appendix R lighting previously installed by DCP 268 and DCP 641 and vill include the following new safe shutdown equipment and access routes thereto:

(1) Stairvell S-2 (clean shop area)

(2) The turbine buildup - northwest corner i (3) Ine main steam valve room i (4) The primary auxiliary buildup El. 722 - Blender Room (5) The Quench Spray pump room (6) The main intake structure - cubicles B & C (7) Additional Idghting may be required in other areas as determined by additional lighting tests to be performed.

All the Appendix R emergency lighting, including the previously installed lighting, will be tested to meet the latest NRC guidance.

SUMMARY

OF SAFETY EVALUATION

! This design is safe in that the modification vill not affect any safety-related l

equipment as previously analyzed or create any new malfunctions or accidents.

No change to the Technical Specification is required. No Updated Final Safety Analysis Report change is required.

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BVPS-1 Annual Report of Facility Changes,  ;

Tests and Experiments for 1988 1 ENGINEERING DESIGN CHANGE NO. 777 l l

CHANGE TITLE-Protection Upgrade of Radiation Detectors RM-219A and 219B DESCRIPTION AND PURPOSE OF MODIFICATION In the past, the weight of the flexible conduit on the connectors of Radiation Detector RM-RM219B (High Range in Containment) and the potential for damage to the electrical connections in the connectors has been considered a problem in the Nuclear Operations Department. Design Change Package No. 800 (Electrical Containment Penetration Assys.) vill eliminate the flexible metal base and run the cable bare into the bottom of the detectors. The purpose of this design change is to provide a protective steel enclosure around the detectors and protect the cable connection into the instrument.

SUMMARY

OF SAFETY EVALUATION The design change is safe in that the modification to provide a protective steel enclosure around the detectors and protect the cable connection into the instrument will not affect any equipment important to safety as previously analyzed or create any new malfunctions or accidents. The Technical Specification is not affected and no changes to the Updated Final Safety Analysis Report have been identified. This design does not present an unreviewed safety question. The final design of steel enclosures shall be submitted to the Radiological Control Department to determine that enclosure configuration vill not affect the operation or sensitivity of the detectors.

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~** Page 22 I BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 198L

' ENGINEERING DESIGN CHANGE NO. 778 1

CHANGE TITLE Auxiliary River Vater Pump VR-P-9A, 9B Seal Water Pipe Change to Stainless Steel DESCRIPTIC0i AND PURPOSE OF MODIFICATION The purpose of this modification is to replace the auxiliary river water pump seal and motor cooler carbon steel piping and valves with stainless steel tubing and valves. The pumps are mark numbers VR-P-9A and 9B. Internal corrosion of the carbon steel pipe is causing a reduction in flov to the river water pumps seal flush and motor cooler water lines, warranting the change to stainless steel piping and valves. The primary objective for the design change shall be to restore the necessary water flows for pump operation.

SUMMARY

OF SAFETY EVALUATION The design change affects Beaver Valley Power Station Unit No. 1 river water piping within the alternate intake structure only. The' intent of the design change is to make a one-for-one component replacement within the affected pipelines with functionally identical equipment. The piping and valves to be replaced are 0.A. Category II, not safety-related components. Therefore, this design change vill not adversely affect the safety of the plant as considered in the Updated Final Safety Analysis Report and Technical Specification and is therefore acceptable.

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Page 23 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE No. 798 CHANGE TITLE Upflow Conversion DESCRIPTION AND PURPOSE OF MODIFICATION Westinghouse has developed and qualified a program for field modifying the reactor vessel lover internals assembly to reduce the potential for fuel rod 1 damage resulting from baffle joint jetting. Vith this modification the coolant downflow path in the baffle / barrel region is converted to an upflow path. The current Unit #1 is "downflow" design. The objective of this conversion is to reduce the hydraulic pressure differentials, which exist across the baffle joints in the downflow configuration. Reducing these pressure differentials, via the upflow conversion, results in a substantial reduction of the coolant jetting through the baffle joints and results in improved fuel rod reliability.

The core barrel at Unit #1 has fifty-two locations where fuel ass <e.0 des could be exposed to baffle jetting.

SUMMARY

OF SAFETY EVALUATION This modification is considered safe and does not involve any unreviewed safety questions. This changes the flow path from being downflow between the core barrel and baffle plate to upflow and has the effect of increaalng the core bypass flow from 4.5% to 6.5% (including thimble plugs removal). The core average temperature is up 0.9'F and core outlet temperature is up 1.3*F. This vill cause DNB penalty 2.8% plus 2.3% of rod bow penalty and 1.5% of axial blanket penalty. The total DNB penalty is still within the generic 9.1% DNB penalty margin. The effect of these changes have been evaluated by Westinghouse, including effects on the reactor internals, the fuel assembly integrity, the core barrel plug, the thermal hydraulic design analysis, and the appropriate LOCA, non-LOCA and SGTR postulated accidents. The peak clad temperature is acceptably below 2200'F limit of 10CFR50.46.

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BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 .

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I ENGINEERING DESIGN CHANGE NO. 800 l

l CHANGE TITLE ,

Containment Penetration Assemblies RM-RM-219A & B l

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DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to install two new containment penetration assemblies in spare containment penetration holes. These new assemblies vill have full environmental qualification and vill contain two triax feedthroughs l each. Additionally, this modification '11 install new environmentally qualified triax cable from the contain- at high range radiation monitor detectors (RM-RM-219A&B) to the penetration assemblies to eliminate the problems associated with the splices on the existing coax cable. By making this change, an environmentally qualified cable with a separate ground vill be provided. This will enhance the operation of the containment high range radiation monitors by providing an outer metal jacket which will filter out stray signals and help keep the inner insulation from breaking down during post accident environment conditions. The existing coax cable vill be removed and the existing feedthroughs of the penetration assemblies in containment penetrations AB and All vill be abandoned in place.

The purpose of Revision 1 of the design concept is to use 2-two triax feedthrough containment penetration assemblies instead of 2-six coax (2 active and 4 spare) feedthrough containment penetration assemblies. By making this change, an environmentally qualified cable with a separate ground vill be provided. This vill enhance the operation of the containment high range radiation monitors by providing an outer metal jacket which vill filter out stray signals and help keep the inner insulation from breaking down during post accident environment conditions.

The purpose of Revision 2 of the design concept is to change from penetration B6 to G7 due to physical interferences in containment which prevent installation. The new penetration (G7) is currently a spare similar to the original penetration. This revision does not affect the conclusions of the safety evaluation.

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SUMMARY

OF SAFETY EVALUATION The modifications made under this Design Change are considered to be safe and do not constitute an unreviewed safety question, nor do they affect the bases of the Technical Specification. There are no required changes to either the Updated Final Safety Analysis Report or the Technical Specifications, as a result of this Design Change.

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'BVPS-l' Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 800 (Cont') .

SUMMARY

OF SAFETY EVALUATION (Cont.)

The design of the electrical containment penetration assembly (ECPA) includes Viton 0-rings as aperture seals between the header plate and the penetration nozzle slip-on flange, and Midlock stainless steel compression fittings on the feedthrough modules, as leakage limiting boundaries. The requirements of 10CFR50 Appendix J for the type. of containment penetration, which includes recilient seals and flexible' metal seal assemblies, is to perform a " Type B Test". The testing requirements- and acceptance criteria are identified asSection III.B of Appendix J. The areas between the seals, the interconnecting passages in the header plate, and the spaces in the feedthrough modules act as a pressure chamber which is used to monitor the leakage- rate of the penetration. Testing, in accordance with the vendor technical manual, vill assure that 3he ECPA in the as-installed condition vill limit leakage to less than l'x 10~ sec/sec. Further, because additional surveillance recommended by the vendor and because both- an 0-ring aperture pressure test and a Type B pressure test are conducted, a Type A' integrated leak rate test for this electrical penetration is not considered necessary for this design change installation.

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, .tf BVPS-1 Annual Report of. Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 808 CHANGE TITLE Large Bore Primary Snubber Elimination DESCRIPTION AND PURPOSE OF MODIFICATION

Recently, 10CFR50 Appendix A, General Design Criteria 4 (4-86) has been changed to allow for the elimination of Reactor Coolant (RC) Loop breaks as a design basis. This change can allow for Leak Before Break analysis technique to be

. applied to the Reactor Coolant Loop piping. This design change vill remove thirtya.of the thirty-six Large Bore Bergen-Paterson Snubbers presently installed 'on the Steam Generator (S/G) and Reactor Coolant Pump (RCP) supports to mitigate a pipe rupture event. The snubbers to be removed are RC-HC-1A through 9A, 11A, 1B through 8B, 11B, 12B, 1C through 9C and 11C. All lower snubbers on the Steam Generator and RCP frame vill be eliminated. Two of the upper.four snubbers on each Steam Generator which are RC-HC-9A, 11A, 11B, 12B, 9C and 11C will be replaced by rigid struts. And also, four (4) Grinnell snubbers on the 8" bypass' lines are replaced with 2 1/2" x 5" Grinnell snubbers. Snubbers clamps are to be replaced. Specifically snubbers RC-USS-101, 102, 105 and 106 are to replaced, including clamps. Clamps only will be changed out on RC-HSS-103 and 104. This change vill be financially beneficial due to decreased plant down time and reduced man-rem exposure.

SUMMARY

OF SAFETY EVALUATION This design change is safe. Elimination of the upper steam generator snubbers perpendicular to the Hot and Cold Legs was due to the negligible thermal movement which allowed these snubbers to be replaced with struts. The lover S/G and RCP snubbers were eliminated since the original design basis RC Loop LOCA loading was eliminated per GDC-4. An analysis of the other branch line breaks (Accumulator, Pressurizer Surge, Residual Heat Removal, Main Steam, and Feedvater) vhen combined with Operational Basis Earthquake (OBE) and Design Basis Earthquake (DBE) loading per the original plant design basis concluded that' sufficient margin exists which permits the elimination of all the lover snubbers.

The upper struts maintain the same design capacity as the original snubbers replaced which are used to mitigate the effects of a postulated Main Steam Line Break only.

Code Case N-411 has been approved by the NRC for use at Beaver Valley Power Station Unit No. 1. The 2% damping to be used for all frequencies represents a conservative use of N-411. (Code Case N-411 which was accepted by the NRC and documented by letter dated April 8, 1987 under Docket No. 50-334, allovs for the use of increased dauping values of up to 5%).

Page 27 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 816 CHANGE TITLE Loose Parts Monitoring System Upgrade DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this change is to upgrade the existing Loose Parts Monitoring System in order to provide better operability and reliability. Revision 1 of this design concept discusses the need to relocate the existing steam generator sensors and to provide new sensor mounting plates for these, and the additional sensors, as per the vendor's recommendation. Existing cabling and conduit vill be extended to the new mounting areas. The existing mountings on the upper and lover vessel may be replaced for compatibility with the new system. This revision does not affect the original safety evaluation.

SUMMARY

OF SAFETY EVAtt1 TION Implementation of this design change is considered to be safe. No change to the Technical Specification is required; however, the Updated Final Safety Analysis Report description of the Vibration and Loose Parts Monitoring System must be reviewed and revised after design change implementation. By performing this modification, the reliability of a previously installed monitoring system, designed to help preclude these types of accidents, vill be enhanced.

The loose parts monitoring system vill not affect primary or secondary systems design basis. The monitors vill be attached to primary system components (Steam Generators and Reactor Vessel) in a manner which vill not degrade their l pressure boundary. The loose parts monitoring system does not have any i protection or control functions which would affect primary or secondary system performance.

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BVPS-1 Annual Report of Facility Changes, L Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 817 CHANGE TITLE Replacement'of Reactor Coolant System (RCS) Narrow Range Protection Resistance

' Temperature Detectors (RTDs)

DESCRIPTION AND PURPOSE OF MODIFICATION The. RCS narrow range Resistance Temperature Detectors ' supply temperature signals to the . reactor protection- and control , systems. Some RTDs supply signals to the reactor protection' system, separate RTDs supply signals to the reactor control system, and some RTDs are installed spares and can be used for either protection or control purposes if they meet the requirements.

The installed RCS narrow range Protection RTDs (Rosemount-176KF) are qualified to the criteria of IEEE 323-1971. . NRC Rulemaking, 10CFR50.49 and Reg. Guide 1.89. requires that replacement for these RTDs purchased after May 23, 1980 meet the criteria of IEEE 323-1974. Since the Rosemount 176KF does not meet IEEE 323-1974 criteria and DLC has no Rosemount spares purchased' prior to this-date, a qualified replacement must be identified and procured.

The installed RCS narrow range spare RTDs are a mixture of Rosemount 176KF and Sostman RTDs. The Sostman RTDs cou2d be used in place of a failed RTD in a control channel but would not be suitable for a protection channel because their time response differs from the Rosemounts. The installed spare Rosemount RTDs can be used in Protection Channels if Environmentally Qualified (EO) splices are made to the field cable at the RTD and at the containment penetrations.

All of the RCS narrow range RTDs vill eventually be replaced by DCP 698, RTD Bypass Manifold Elimination; however in the interim, means must be provided for replacing failed RTDs with qualified spares.

The objective of this design change vill be:

1. Identify RTD that can be used generically as a replacement for the Rosemount 176KF in the RCS narrow range protection channels.

The replacement RTDs shall be qualified to IEEE-323-1974 and be similar to the Rosemount RTDs including physical dimensions, calibration data, accuracy and time response.

L 2. Replace currently inoperable RTDs (TRB-RC-412B and 421D) with qualified replacements.

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.t BVPS-1 Annual Report of' Facility Changes.

Tests.and Experiments for 1988 '{

' ENGINEERING DESIGN CHANGE NO. 817 (Cont.)

DESCRIPTION AND PURPOSE OF MODIFICATION (Cont.)

3. In the event of additional. RTD failure (s) occurring at the Sixth Refueling, steps vill be taken as necessary to obtain a  !

complement of protection and ' control RTDs for operation. These steps may include buying additional replacements or E0 splices of installed spare Rosemount RTDs.

4.. In-plant ' verification testing, i.e. calibration and time response, vill.be performed to verify that existing plant design criteria tre met.

SUMMARY

OF SAFETY EVALUATION The safety parameters that may be affected by replacing the protection RIDS are the reactor trip setpoints and time delay for the Overtemperature Delta-T (OTDT) trip, and the Overpower Delta-T (OPDT) trip.

The replacement RTDs vill be calibrated to the current criteria and no setpoint changes vill be required.

The time delay assumed in the accident analysis for the OTDT trip is given in Updated Final Safety Analysis Report Table 14.D-3 as 6.0 seconds. This is the {

total time delay, from the time the temperature difference in the coolant

' loops exceeds the trip setpoint until the rods are free to fall. Two seconds are allowed for RTD bypass loop fluid transport delay and thermal capacity effects. A maximum of 4 seconds is allowed for the sum of the RTD time constant; ' the lag compensation (AT and TAVG summator filters) time constant; the trip circuit channel electronics time constant; the reactor trip breaker delay time; and the gripper release time, i i

The trip response time constant vill require verification by performance of j BVT 1.3-1.1.8. The acceptance criteria is a maximum of 4.0 seconds. The I I

reactor trip breaker delay time and the gripper release time are absolute response times, whereas the RTD response, the lag compensation and the trip circuit channel electronics response should be measured as response time constants. Adequate performance of the time response testing assures that the OTDT time response assumed in the Updated Final Safety Analysis Report is valid. It also should be noted that th's time constant acceptance criteria L

(4.0 seconds). contains no uncertainty allowance, and therefore, the test measurement uncertainty should be included in the measured time.

The proposed replacement may also be used for a control system RTD. The current control RTDs are Sostman brand. Therefore, as long as the maximum time constant of the replacement RTD is less than the time constant of the Sostman RTD, it can be used as a control RTD.

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Page 30 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l.

ENGINEERING DESIGN CHANGE NO. 817 (Cont.)

SUMMARY

OF SAFETY EVALUATION (Cont.)

The calibration accuracy requirements in the purchase specification for the replacement RTD is the same as that which was used for the Rosemount RTDs.

Furthermore, the RTD cross-calibration procedure requirements vill be met, and therefore the uncertainty is within the current design basis, and no technical specification changes are required.

The E0 splices of the RTD cables vill bring the cables into compliance with current requirements. If the E0 splice is performed after the cross-calibration, caution should be used to assure that the splice does not affect the RTD reading.

The proposed changes vill be within the existing safety analysis, and are, therefore, considered safe and acceptable. These changes were evaluated in the NRC Safety Evaluation related to Amendment No. 118 of the Technical Specifications.

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Page 31 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE No. 819 CHANGE TITLE Type "C" Leak Test Connections Modifications DESCRIPTION AND PURPOSE OF MODIFICATION This design change vill make the following modifications:

1) Replace blank flanges currently installed on some of the Type C leak test connections with permanently installed, modified flange assemblies, which have Svaglok fittings with removable plugs to allow for Type C leak test equipment connections. By performing this modification, time would be saved by eliminating the process of removing the existing blank flanges and installing a temporary Type C test flange, then removing the test flange ar.d installing the blank flant.e along with a new flexatallic gasket.

2) Replace the reactor plant sample system containment penetration Type C test connection plugs with a Whitey valve and removable plug. This modification vill eliminate the radiological hazard of having contaminated water being sprayed on test personnel when removing test plugs whenever there is pressure in the system due to leaking isolation valves.

3) Replace some of the blank flanges downstream of the vent and drain valves on various systems, with modified flange connections which have Svaglok fittings and removable plugs. This vill allow for a quicker connection of vent and drain hoses used during nitrogen purges, and vent and drain operations of contaminated systems.

SUMMARY

OF SAFETY EVALUATION The modifications made under this design change are considered to be safe and do not involve an unreviewed safety question, nor do they reduce the margin of safety as' defined in any Technical Specification Basis. The original intent of retaining system fluid pressure boundaries vill be maintained by replacing the existing blank flange connections with modified plugged connections. ,

Revisions to Updated Final Safety Analysis Report Figures are required to j reflect the modifications performed under this design change. j J

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BVPS-1 Annual Report of Facility Changes, l Tests and Experiments for 1988 i

l ENGINEERING DESIGN CHANGE No. 823 CHANGE TITLE Replacement of FV-204, 205, 292 and 293 l

l DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of this design change is to replace four 3/4" Pacific Globe Valves (FV-204, 205, 292 and 293) cith four 3/4" Kerotest packless diaphragm valves.

The reason for the replacement is that the Globe Valves are leaking and replacement parts are not available to repair them.

SUMMARY

OF SAFETY EVALUATION This design is considered to be safe and does not constitute an unreviewed safety question, nor does it affect the basis of any Technical Specification.

Updated Final Safety Analysis Report Figure 10.3-5 needs to be revised to show the new type of valves.

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'** Page 33 BVPS-1 Annual Report of Facility: Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 829 CHANGE TITLE Feedvater System Upgrade D$SCRIPTION AND PURPOSE OF MODIFICATION A feedvater system task force was formed to investigate problems associated with this system and recommend solutions. One recommendation is to reduce the pressure drop across ~ the main feedvaterL regulating valves. This can be accomplished.by reducing the discharge head of the main feedvater pumps and re-sizing the main feedvater regulating valve trim to match the system performance. This change should result in a reduction in the velve internal fluid velocity.- This reduction should help in meeting the desfred objective of decreasing the' frequency of vibration ralated failures. A secondary objective associated with this change is to reoace the amount of power used by the main feedvater pumps.

The main feedvater pump impeller diameter vill be reduced to 21.09". This is the same impeller size as that. in the Unit #2 main feed pumps. Further reductions in.the impeller diameter are possible; however, reductions beyond.

21.09" have not currently been tested and, therefore, uncertainties exist.

The main feedvater regulating valve trim (cage and plug) vill be re-sized to match the pump and system characteristics. Hydraulic analysis indicates'a CV-requirement of 530-620 at full power. The full open CV of the new trim vill be specified.at 735 to allow margin and fix an acceptable operating position.

This design results in a reduction of the valve pressure drop from about 300 psi. currently to about 220 psi with the new trim.

These changes have been judged to have a negligible affect on the steam generator level control system response characteristics.

These changes are not expected to have a.significant effect on the Technical Specification requirement of 10 seconds closing time for the main feedvater regulating valve.

SUMMARY

OF SAFETY EVALUATION The changes of this design change vill not adversely affect the safety of the plant.

Updated Final Safety Analysis Report Section 14.1.9 was reviewed to determine if it would be affected by this design change. Section 14.1.9.2 lists assumptions which were made for the basis of the feedvater system malfunction transient.

Page 34 l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 1

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l l ENGINEERING DESIGN CHANGE NO. 829 (Cont.) {

SUMMARY

OF SAFETY EVALUATION (Cont.)

The assumptions will be revised to reflect a reanalysis for Case A. This case assumes accidental opening of one feedvater control valve at zero load I

i conditions. This case has been reanalyzed by Westinghouse for a feedvater flow rate of 210% of nominal feedvater flow to one steam generator and a feedvater temperature of 32*F. Case B is an accidental openi.g of one feedvater control valve with the reactor at full power. The flow value assumed for this case is presently 170% of nominal feedvater flow to one steam generator.

A design analysis was performed in support of the design cc.icept. The design analysis calculated the maximum flows that would occur ':ith the proposed new valve trim and impeller. These values are 150% for Case A and 133% for Case B. Even after applying a 10% conservatism factor, the max flows vould only be 165% and 146% for Case A and B, respectively. These maximum calculated flows are still less than the safety analysis limits of 210% and 170% of nominal flow assumed in Section 14.1.9.2, for Case A and B, respectively. Therefore, even after the design change, the existing safety analysis vill be valid.

This change is also not expected to change the closing time for the main feedvater regulating valve. Therefore, the 10-second stroke time of Technical Specification 3.6.3.1, as listed in Table 3.6-1 for penetrations 76, 77 and 78, vill still be satisfied.

This design change should result in a reduction in the valve internal fluid velocity, which is expected to help decrease the frequency of vibration related failures.

The changes proposed are within the existing safety analysis basis and vill not adversely affect any other equipment. Therefore, this design change vill not adversely affect the safety of the plant as considered in the Updated Final Safety Analysis Report and is, therefore, acceptable. l 1

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Page 35 BVPS-1 Annual Report of facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 832 4

CHANGE TITLE Charging Pump Cubicle Flooding Protection DESCRIPTION AND PURPOSE OF MODIFICATION The purpose of the modification is to ensure that the charging pump cubicles are protected against a postulated component cooling water system pipe rupture and subsequent flooding condition. Updated Final' Safety Analysis Report Section 9.4.3.1, Component Cooling Water System, states that a pipe rupture can cause a flood condition to a height of approximately 12 inches over the floor (El. 735'-6") of the Auxiliary Building.

The modification vill ensure that the charging pump cubicles are protected against such flooding by the use of water tight seals at all concrete slab joints over the cubicle roof area and by designing the cubicle hatch covers to have sufficient solid height in excess of 12 inches above the floor level.

SUMMARY

OF SAFETY EVALUATION Implementation of this design change is considered to be safe. No changes to the Technical Specification is required. The Updated Final Safety Analysi, Report description of the charging pump cubicle flooding protection against a failure of the component cooling water system piping must be reviewed and revised after implementation of this modification. .By performing this modification, the operability of the charging pumps vill not be adversely affected.

Page 36 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 833 CHANGE TITLE TV-FP-105, 106, 107 Trim Changes DESCRIPTION AND PURPOSE OF MODIFICATION This modification vill change the seat design of three fire protection isolation valves from metal seating to a soft seat configuration. The valves are containment isolation valves in the fire protection water supply lines for containment. As such, the valves are subjected to an air leakage test (Type C test program). The purpose of this modification is to improve the capability of these valves to function as a containment isolation valve while maintaining the capability to function as a fire protection isolation valve.

SUMMARY

OF SAFETY EVALUATION This design change vill not adversely affect plant safety. The ability of the valves to function as a containment isolation valve vill be improved, thereby reducing the possibility of containment leakage during an accident. The fire protection function of the valves is unaffected. No changes to the Updated Final Safety Analysis Report or Technical Specifications have been identified.

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. 1 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l ENGINEERING DESIGN CHANGE NO. 834 CHANGE TITLE {

J Jib Cranes in Reactor Containment ]

L I DESCRIPTION AND PURPOSE OF H0DIFICATION {

The purpose of this modification is to install two jib cranes on the operating level of the Reactor Containment in the area adjacent to Steam Generator RC-E-IC and the Equipment Hatch. These cranes will be able to handle small lifts so that the polar crane, which is currently the only means of lifting items in the containment, is available for larger lifts on a more timely basis to avoid delays in outage work.

SUMMARY

OF SAFETY EVALUATION This modification is safe to implement. No unreviewed safety questions exist and no Technical Specifications are affected. The design and load sizing of the jib cranes does not compromise the structural integrity and seismic performance of the Containment Building Structural members, nor does operation of the cranes affect safety because of in-plant procedures. Since the crane hoists contain aluminum and PVC materials, the hoists vill be removed from the containment when not in use during Modes 1 through 4 to prevent the formation of. Hydrogen gas during a Design Basis Accident situation. No changes to the Technical Specifications are required. Updated Final Safety Analysis Report Figure 5.1-1 vill be revised.

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Page 38 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l

ENGINEERING DESIGN CHANGE No. 835 CHANGE TITLE Access Panels for Group I Fire Dampers DESCRIPTIOil AND PURPOSE OF MODIFICATION The purpose of this modification is to provide access panels through ductvork adjacent to fire dampers that are inaccessible for inspection. This will allow for inspection of the fire dampers to ensure operability and freedom from blockage by foreign objects during fire damper trip tests.

SUMMARY

OF SAFETY EVALUATION This modification is safe to implement. No unreviewed safety questions are created nor are changes required to be made to the Updated Final Safety Analysis Report or the Technical Specifications. This modification vill also reduce radiation exposure te Test personnel by reducing the amount of time required to gain access to the dampers.

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.- d BVPS-1 Annual Report of Facility Changes, )

Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 840 CHANGE TITLE Security Protection to Equipment Hatch Barrier and Vital Area Penetrations DESCRIPTION AND PURPOSE OF MODIFICATION This modification is to install a physical barrier which consists of either 2" '

deep grating or 3/4" diameter round bars velded to provide no more than a 6" square opening at any spot on the following areas:

1. Air intake louver near primary grade water pump room docr.

2. Station diesel generator roof exhaust vents.

3. Security diesel generator intake and exhaust louvers.

4. Equipment hatch missile barrier area.

The purpose of this modification is to ensure vital equipment areas are properly secured in order to prevent intruder attacks and to ensure compliance with NRC 10CFR73.50 and 10CFR73.55, " Requirements for Physical Protection of I

Licenced Activities in Nuclear Power Reactors Against Radiological Sabotage".

SUMMARY

OF SAFETY EVALUATION This design change modification to the vital equipment area penetrations is considered to be safe and does not constitute an unreviewed safety question.

In fact, installing a barrier on a vital area vill ensure the vital equipment's operability to perform its safety function in the event of any accident.

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"- Page 40 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 848 CHANGE TITLE Thimble Plug Removal DESCRIPTION AND PURPOSE OF MODIFICATION Thimble plugs are placed in fuel assemblies in locations where there are no Rod Cluster Control Assembly (RCCA) source rods or burnable poisons rods to minimize bypass flow. .During a refueling outage, the thimble plugs usually have to be moved to different fuel assemblies due to the new loading pattern.

This movement can become critical path time. Therefore, the primary advantage for thimble plug elimination or removal is the time-saving realized during a refueling outage as a result of not having to handle or move the thimble plugs.

SUMMARY

OF SAFETY EVALUATION Thimble plugs are used to limit the flow through guide thimble tubes. The removal of these plugs results in a 2% increase in core bypass flow and reduce the amount of core flov available for core heat removal. As part of the upflow modification (DCP 798), Westinghouse was asked to perform the analysis required to determine the increase in core bypass flow and evaluate its impact on plant margins and the corresponding impact on Updated Final Safety Analysis Report LOCA and non-LOCA transients. Westinghouse VCAP-11639 (Upflow Conversion Safety Evaluation Report, Beaver Valley Unit 1) documented and concluded that the removal of the thimble plugs vill not adversely affect the safety of the plant as considered in the Updated Final Safety Analysis Report and Technical Specification, and is acceptable.

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BVPS-1 Annual Report of Facility Changes,

-Tests and Experiments for 1988 ENGINEERING DESIGN CHANGE NO. 858 CHANGE TITLE Boron Recovery' Test Tank Transfer to Liquid Vaste  !

DESCRIPTION AND PURPOSE OF MODIFICATION Presently the boron recovery test tank pump (BR-P-3B) cannot transfer water

- from the boron recovery test' tank (BR-TK-2B) to the high-level liquid vaste

. drain tanks (LV-TK-7A -6. B) without recirculating vater back to BR-TK-2A.

Since this alignment would mix the contents of BR-TK-2A and BR-TK-2B, a new flow path must.be installed to allow BR-P-5B to transfer water from either j B,. while providing a recirculation path

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BR-TK-2A or BR-TK-2B to LV-TK-7A &

back to the respective tank being transferred without affecting the other tank's contents.

This design change vill install 1 1/2" stainless steel pipe with a 1 1/2"~

stainless steel isolation ball valve from boron recovery lines 1 1/2"-BR 152 to 1 1/2"-BR-156-152, to provide the above mentioned flow path.

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SUMMARY

OF SAFETY EVALUATION This design change is considered to be safe and does not present an unreviewed safety question, nor does it affect the basis of any Technical Specification.

Updated Final Safety Analysis Report Figure 9.2-3 vill require a revision to reflect the addition of the new line and valve.

This dtsign change vill only provide a flow path from the boron recovery test tank (BR-TK-2B) to the high-level liquid vaste draiu tanks (LV-TK-7A & B) via the BR-P-5B test tank pump. This modification vill not degrade the boron i recovery system's design basis capability as stated in Updated Final Safety Analysis Report Section 9.2.1, nor will it increase the consequences of an accident / failure previously analyzed in Updated Final Safety Analysis Report Table 9.2-2.

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[L BVPS-1 Annual Report of Facility Changesi R

Tests and Experiments for-1988 lt . ENGINEERING DESIGN CHANGE NO.'867 1

CHANGE TITLE Fast Bus Transfer Latching Relays and Breaker Switches

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,: DESCRIPTION AND PURPOSE OF MODIFICATION i

.The purposes of this modification are as follows:

1) ~ Replace Turbine / Generator primary (1AST) and secondary (2AST) trip circuits MG-6 relays (62-ASTX-1 and 2 and.162-ASTX-1 and 2) within.high speed latching relays. These are to maintain a

turbine generator trip once executed.

2) Install knife switches into fast bus transfer breaker closing l

coil circuits for normal 4KV bus supply breakers. Proper l alignment ~of these knife switches will prevent a reverse fast bus transfer.while allowing a fast bus transfer in the desired direction; however, a reverse fast-bus transfer can be aligned by re-positioning the knife switches under administrative controls.

The first modification is to prevent equipment damage caused by a reset of the turbineitrip relays after a turbine trip signal has been produced. These latching relays vill ensure that a turbine trip. signal remains until manually reset by an operator. The second is to prevent a~ realignment of the 4KV bus supply breakers back to the original power source, which may have been lost, until the cause of the first transfer has been determined. Normal manual operations of the breakers are not affected by this design change.

SUMMARY

OF SAFETY EVALUATION This modification is safe to implement. No unreviewed safety questions exist and no change to the Updated Final Safety Analysis Report or the Technical Specifications are required. This modification does not affect any equipment

required to place or maintain the reactor in a safe condition.

This modification vill reduce the probability that a turbine trip signal vill reset itself by use of latching relays. Since the latching relays directly replace the existing turbine trip relays, no other portions of the turbine trip circuits are affected. This modification vill also prevent the L'

l- possibility of a reverse fast bus transfer once a fast bus transfer has taken place. The alignment of the knife switches shall be administratively controlled.

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Page:43 1

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1 BVPS-l' Annual. Report of Facility Changes, Tests and Experimental for.1988 ENGINEERING DESIGN CHANGE NO. 871 1

CHANGE TITLE I

' Reduction of' Heater Dra' ins and Condensate Systems Design Pressure j DESCRIPTION AND PURPOSE OF MODIFICATION

" 1The scope'of this design change includes the following:

1. Installation of a relief valve on the discharge of each heater. j drain pump. '

l

2. Installation of a common discharge line from the relief valves to 1 the heater' drain tank emergency drain line near the condenser l penetration.

J Documenting a reduction in the design pressure of'the heater drain. j 3.-

pump discharge piping based on installation of the relief valves. .j i

4. Documenting a reduction in the condensate system design pressure l based on operating data and analysis in accordance- with code l

. requirements.

The purpose of these changes is to resolve discrepancies which exist between design pressures and allovable pressures of installed components.

I

SUMMARY

OF SAFETY EVALUATION

.This change is safe and does not present an unreviewed' safety question. The l relief system to be installed on the heater drain' pump discharge piping vill -l not affect the normal operating characteristics 'of the system and therefore ~j will not affect the feedvater transients analyzed. The relief system vill l L normally operate only during a reactor trip with a feedvater isolation. The

! changes in system design. pressures are administrative and required to ensure I code compliance for installed components. No changes to the Updated Final Safety Analysis Report or Technical Specifications are required.

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Page 44 l BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 i

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ENGINEERING DESIGN CHANGE NO. 878 CHANGE TITLE Thermal shield Bolt Replacement DESCRIPTION AND PURPOSE OF MODIFICATION One of twelve bolts that hold the top of the thermal shield to the core barrel was found to be severed. The bolt was removed for examination by the Westinghouse Materials Technology Department. The preliminary report reveals that the available evidence suggests fatigue crack initiation and growth as l the cause of fracture. A heavy oxide present on the fracture supports the possibility that the bolt spent several years in the reactor after its final failure. Also, because of the lack of a significant overload area, it appears that the final loads placed on the bolt were relatively lov as opposed to the fatigue limit and loads necessary to initiate failure. Therefore, m.

I l initiation vould be facilitated by the high mean load present from the b.it torque, by a mechanical ' defect' on the as-received bolt, and by the stress l concentration in the radius where the shank meets the head of the bolt". All other bolts were Ultrasonic Tested and exhibited no degradation.

This modification proposes to remove the remainder of the severed bolt and to replace it with one of the same material and similar design except for the locking mechanism. The replacement bolt vill have a stellite locking bar, mechanically restrained in place, rather than a velded locking key. The replacement bolt vill meet all original design criteria. This installation vill be performed underwater by a diver using long-handled tools.

SUMMARY

OF SAFETY EVALUATION Implementation of this modification is considered to be safe. No change to the Technical Specification or the Updated Final Safety Analysis Report is required as a result of this modification. The replacement bolt vill meet all original design criteria and vill be installed to the original torque value.

Therefore, the Reactor Vessel Internals Description and Safety Analysis (Updated Final Safety Analysis Report Sections 3.2.2 and 14.3.3 respectively) remain valid. .

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i Page 45 i

BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 l

ENGINEERING DESIGN CHANGE NO.1044 l CHANGE TITLE l

Bypass Status Indication in the Main Control Room for the Containment Hi-Hf Pressure and RVST Luv Level Transfer to Recire Bistables DESCRIPTION AND PURPOSE OF H0DIFICATION The containment hi-hi pressure and the Refueling Vater Storage Tank (RVST) lov level transfer to recirculation bistables when placed into the " bypass" position do not provide status indication in the main control room.

Currently, when the bistables are put into the " bypass" position, the control room operator may be unaware that the respective channel is disabled and unable to perform its intended safety function.

This modification vill be accomplished by providing " bypass" inputs to existing status lights "HHCP Press. Ch. Trip" and "RVST Low Level Ch. Trip" in the main control at benchboard section "A-13" and revising the status light vindows to read "HHCP Press. Ch. Trip / Bypass" and "RVST Lov Level Ch.

Trip / Bypass".

SUMMARY

OF SAFETY EVALUATION This modification vill enhance plant operations as the status of the bistables

" trip / bypass" is now available to the operator. This modification in no vay impacts the operation of any plant equipment as described in the Beaver Valley Power Station Unit No. 1 Updated Final Safety Analysis Report or impact any Beaver Valley Power Station Unit No. 1 Technical Specification. Also, the addition of the " Bypass" status indication conforms to IEEE-279 Para. 4.13 (1971) and the Beaver Valley Pover Station Unit .No. 1 Updated Pinh1 Safety Analysis Report, Section 7.3.2.1.5 for " bypasses". The new cabic t' quired for this modification vill have a non-Category 1E function and vill be pulled and routed in neutral conduit and trays. This vill maintain and conform to n11 existing color separation and design requirements at Beaver Valley Power Station Unit No. 1.

1

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Page 46 BVPS-1~ Annual Report of Facility Changes, Tests and Experiments for 1988 i_

Procedure Changes and Temporary Modifications CHANGE TITM:

, Jumper and Lifted Lead for Air Ejector Stub Tubes DESCRIPTION OF CllANGE:

This temporary ' modification adds additional flow measuring instrumentation to the discharge of the air. ejectors for flow verification. Normally, this instrumentation'is not in the discharge flowpath but is only aligned momentarily fox flow reading as desired. 1 i

SUMMARY

OF SAFETY EVALUATION:

This temporary modification will not affect the function or operation of the system as described in Updated Final Safety Analysis Report Sections 10.3.6 and 10.3.8. This modification does not affect any equipment important to safety as the operation of the radiation tnouitor and associated trip valves are not affected. No unreviewed safety question exists.

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~ BVPS-1 Annual Report of. Facility Changes, Tests and Experiments for 1988

-Procedure Channes and Temporary Modifications CllANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-13, " Temporary Bypass of the L

Filtered Water Supply IIcader to the Unit 1 and Unit 2 River Water / Service Water Pump Seal Injection".

l DESCRIPTION OF CHANGE.

l L This TOP provides instructions fcc the installation, use and removal of a temporary fire hose'in the Intake Structure to supply filtered water to the Unit 1 Reactor Plant River Water Pumps and Turbine Plant River Water Pumps, and the Unit 2 Service Water Pumps while the normal filtered water supply piping is repaired.

SUMMARY

OF SAFETY EVALUATION:

This temporary bypass performs the: same function as the permanent line. The temporary hose used is rated to 250 PSIG which exceeds the system pressure. Adequate flow capu ity is ensured by requiring a 2 1/2 inch hose which exceeds the 2 inch header the temporary hose is replacing. The Unit 1 Reactor Plant River Water Pumps and the Unit 2 Service Water Pumps ~do not rely on the filtered water supply for the safe operation and shutdown of Unit 1 or Unit 2. Both units pumps have safety related seal water backups which provide the pumps in the event of a loss of the filtered water supply (Reference Unit 'I Updated Final Safety Analysis Report Section 9.9, and Unit 2 UFSAR Section 9.2). The temporary hose is routed into the pump cubicle through the ventilation duct tnd no missle shic] ding is afrected. In .the event of a hose rupture, operations would receive low pressure alarms'in the Control Room and tho' backup supplies would provide seal water until the condition was. corrected. Since the ' filtered water supply is not required for pump operability, the margin of safety is not reduced. No unreviewed safety question exists.

Page 48

... BVPS-I Annual Report of Facility Changes, ,

Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CllANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-10, " Reactor Coolant System Leak' Test and 10 Year Inservice Inspection (ISI) Hydro".

DESCRIPTION'0F CHANGE This TOP provides instructions to perform a hydro of the Reactor Coolant Sys+cm as required by ASME Section XI, Article IWB-5000. The performance of this hydro also fulfills the Mode 3 leakage test required for;the refueling outage. The Hydro Test consists of raising

'RCS pressure from a normal' operating pressure of 2235 PSIG to a test pressure range of 2280-2310 PSIG with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time.

SUMMARY

OF SAFETY EVALUATION:

The test pressure range is 2280-2310 PSIG which does not exceed the system design limit of 2485 PSIG. This procedure does not remove or alter any pressure control or overpressure protection equipment that is normally in service in Mode 3. The probability and consequences of an accident during the performance of this procedure is enveloped by the accident analysis of Updated Final Safety Analysis Report Section 14.

Normal system configurations are utilized to perform this procedure.

The Technical Specification Safety Limit of 2735 PSIG will not be exceeded (Technical Specification 2.1.2). No unreviewed safety question exists.

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Page 49

.; BVPS-1. Annual Report of Facility Changes, Tests and Experiments'for 1988 Procedure Changes and Temporary Modifications

. CHANGE TITLE:

Operating Manual Procedure 1.44D.4.F, " Returning the Auxiliary Building Ventilation to Exhaust to the Ventilation Vent with an Auxiliary Building Ventilation Exhaust Fan on Clearance".

DESCRIPTION OF CHANGE:

This new' procedure provides necessary guidance for returning the Auxiliary Building Ventilation to the Normal System Arrangement. (NSA)

.with an . Auxiliary Building Ventilation Exhaust fan on clearance and dampers [1VS-D-7-1, 7-3, 2A and 4A] realigned to the filter banks due to an automatic actuation signal from [RM-IVS-102A or B]. A momentary jumper is used to realign the dampers since the fan control switch is unavailable due to the fan being on clearance.

SUMMARY

OF SAFETY EVALUATION:

The jumper used in this procedure is only momentarily applied to the control circuit and its removal is doubic-verified and documented in the Nuclear Shift Supervisors Log or Reactor Operators Log. The momentary jumper used will not alter the systems configuration, function or performance as defined in Updated Final Safety Analysis Report Section 9.13.2. No unreviewed safety question exists.

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.Page 50

.- BVPS-1 Annual Report of. Facility Changes, j l' ' Tests and Experiments forL1988 l

Procedure Changes and Temporary Modifications CIIANGE ' TITLE:

k 1emporary Operating Procedure (TOP) 1-88-01, Rev. 1 " Moisture Separator Reheater Tube Leak Test"'

' DESCRIPTION OF CHANGE:

This.. TOP is 'a' revision- to the original TOP (Rev. 0) to extend the applicability of the test to all the' rcheaters.

l SUMMAh, OF SAFETY EVALUATION:

This test does not stress the reheaters beyond their normal operating or design limits or change the installed design of the reheaters. The reheaters are- not important to safety and are not required to respond to any accident evaluated .in Updated Final Safety Analysis' Report Section 14. There are no Technical Specifications that apply to this l equipment. No unreviewed safety question exists.

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Page 51 L W BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 I

Procedure' Changes and Temporary Modifications CHANGE TITLE:

l~ Maintenance Work Request (MWR) 880100 (Jumpered/ Lifted Lead),  !

l. " Substitution of Secruity Door as Boundary for Control Room'~ Pressure l ' Boundary Instead of Normal Innet Door".

DESCRIPTION OF CHANGE Substitution of inner door by security door as pressure boundary allows the inner door to be replaced by MWR 880100. Security door seal..is enhanced by taping off the seal area and locking out the security card reader to maintain this seal by preventing access.

SUMMARY

OF SAFETY EVALUATION:

An equivalent ~ boundary door will be used to prevent leakage during the replacement of the inner door. Ti - security door being used as a control room boundary has a greater fire rating (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> vs 1.5 heurs)-

than the normal boundary door. Taking the security door out of service and sealing the door removes methods of failure of the boundary. The additional area added to the control room area does not contain any additional equipment which could affect the safety analysis. No unreviewed safety question exists.

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Page 52

.: 'BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-14, " CIA Testing of CIA Valves

-[TV-1FP-105, 106 and 107]".

DESCRIPTION OF CHANGE:

This TOP performs a partial CIA test of CIA valves [TV-1FP-105, 106 and 107] in response to Incident Report 1-88-39. This TOP will satisfy Technical Specification Surveillance 4.6.3.1.2.a, whose requirements

.will.be met in future performances of Operating Surveillance Test 1.1.4 as recommended by Incident. Report 1-88-39. This . TOP 'provides instructions for activating a CIA signal from the Train B Solid State Protection System K613B output relay to these valves and verifying that the valves close. The other componerts activated by. the K613B relay will be in a condition that will not affect plant operation.

SUMMARY

OF SAFETY EVALUATION:

This TOP verifies the ESF function described in Updated Final Safety Analysis Report Sections 5.3 and 14.3 to ensure containment integrity.

The Supplementary Leak. Co11cetion and Release System is manually aligned to the main filter banks.so that the availability of automatic transfer to the filters will not be required during the test. The fire protection headers are isolated upstream of the CIA valves so that no l l water is released onto safety-related equipment during this test. When testing is complete, the . deluge valves and hose reel headers are drained to prevent water hammering if the system is required to operate while the plant is in a cold shutdown condition. The TOP ensures that

~

no fire exists in containment during the test which would require use of the deluge valves or hose recis in containment. The procedure also ,

! ensures that the Igron Recovery Syecm as not in service during this I test so that damage to the system will not occur when component cooling water is isolated to the evaporator and heat exchangers. No other tests will be in progress on the Solid State Protection System. No unreviewed safety question exists.

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. .. BVPS-1 Annual Report of Facility Changes, u

' Tests'and Experiments'for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

p Temporary Water Supply for Diver Cleaning of Intake' Boys DESCRIPTION OF CHANGE:

This temporary modification utilizes an intake structure screen wash-pump to supply water for intake bay cleaning. A redundant pump remains available'for screen wash use.

SUMMARY

OF SAFETY EVALUATION:

The screen wash system is not reviewed as important to safety in Updated Final Safety Analysis Report Section 9.9. The screen wash pump and the temporary hose are. located outside of the river water pump cubicles so that a failure of this modification will not affect any equipment important-to safety. No unreviewed safety question exists.

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.Page 54,.

p -( BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 I, ~ Procedure' Changes and Temporary Modifications-

" CHANGE TITLE:

Temporary Air Regulator on "A" Main Feed Regulating Valve _

DESCRIPTION OF CHANGE:

This temporary modification provides for local manual control of the "A" Main Feedwater Regulating Valve 'should the pneumatic controller further degrade.

SUMMARY

OF SAFETY EVALUATION:

The feedwater isolation function.of this valve is unaffected due to the modifications location. as upstream of the feedwater isolation solenoids. ESF response operation and response times are unchanged.

Any malfunction created by misoperation of this . modification will. be bounded by existing protective actions that will trip the feedwater isolation solenoids and terminate the malfunction.

4 No unreviewed

. safety question exists.

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l Page 55

. BVPS-1 Annual Report of Facility Changes, l'

Tests and Experiments for 1988.

Procedure Changes and Temporary Modifications

'CllANGE TITLE:

Temporary Modification to River Water "Y" Strainer Blowdown Line DESCRIPTION OF CilANGE:

r.

This modification routes.the blowdown from the river water strainers to the discharge portion'of the River Water System to minimize liquid l waste inventory.

SUMMARY

OF SAFETY EVALUATION:

'The strainer is isolated when blowdown line is in use. The drain hose is connected to the River Water System downstream of all system components. System function remains as described in Updated Final Safety Analysis' Report Section 9.3. A failure of the drain hose would release' the -water contained in the hose and strainer to the building sumps. -The resulting opening into the River Water System discharge piping is relatively very small and could admit air to the system at l

this point with no effect on systen. operability No unroviewed safety question exists.

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Page 56 BVPS-1 Annual Report of Facility Changes, Tests and' Experiments for~1988 Procedure Changes and Temporary Modifications CliANCE TITLE:

Temporary Operating Procedure (TOP) 1-88-19, " Maintenance of Steam Generator Chemistry in Mode 3~by Processing Blowdown via the Standby Blowdown System".

DESCRIPTION OF CliANGE This TOP utilizes the Standby Blowdown System to maintain steam generator chemistry while in Mode 3 with the main condenser out of

. service. Feedwater is supplied to the Steam Generators from the turbino plant demineralized water storage tank. using the Dedicated Auxiliary Feedwater pump.

SUMMARY

OF SAFETY EVALUATION:

The ' probability and consequences 'of an accident or malfunction of equipment evaluated in the Updated Final Safety Analysis Report are not increased because an accident in Mode 3 is bounded by_the loss of feedwater safety analysis which assumes plant power operation. As required by Technical Specification 3.7.1.2., the Auxiliary Feedwater System will be available to supply water to_the steam generators. _The blowdown flowpath' utilizes existing blowdown trip valves to isolate the flowpath and existing radiation monitor to monitor activity levels as described in Updated Final Safety Analysis Report Section 10.3.8.2.

This TOP requires sampling to verify that limits- given in Technical Specification 3.7.1.4 are not exceeded. No unreviewed safety question exists.

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'49 Page 57 s BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications L CHANGE TITLE:

L

' Operating Manual Procedure. 1.24.4.W, " Emergency- Use of '[1FW-P-2]

Turbine-Driven Auxiliary Feedwater Pump".

DESCRIPTION OF CHANGE:

E This ' procedure provides instructions to place the Turbine-Driven Auxiliary Feedwater pump in service in the event that tho' pump is not operating when required for auxiliary feedwater flow.

SUMMARY

OF SAFETY EVAL.UATION:

This procedure does not render the auxiliary feedwater pumps or-

. associated flowpaths unavailabic. In the event of a malfunction of the:

Turbine-Driven Auxiliary. Feedwater pump, two motor-driven auxiliary feedwater pumps are available to provide feedwater to the steam generators. This procedure operates the Auxiliary Feedwater System as described in Updated Final Safety Analysis Report Section- 10.3.5.1.2.

This procedure, in -attempting to return the turbine-driven pump to service, is consistent with the operability requirements of Te:hnical Specification 3.7.1.2, No unreviewed safety question exists.

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.Page 58 y BVPS-1 Annual Report of Facility Changes, a Tests and Experiments for 1988

l. .

l . Procedure Changes and Temporary Modifications l;

h ~ CHANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-21, " Post-Maintenance Pressure Test for [TV-1CV-150C)".

DESCRIPTION OF CHANGE:

L This.= TOP will test' for'the valve being fully seated and if the repair work was successful. The test method is the same method used during Type C testing.

SUMMARY

OF SAFETY EVALUATION:

The containment vacuum pumps do not prevent or mitigate any accidents analyzed in. Updated Final Safety Analysis Report Section 14. During the performance .) f - this test, -[TV-1CV-150C] will be available to perform.its containment isolation function upon receiving an isolation signal, The valve and piping under test are designed to operate at the test pressure used in this TOP. -I f. [TV-1CV-150C] malfunctions during this test, the other series isolation valve [TV-1CV-150D] is.available

- for isolation. Technical Specifications allow deviations from Action Statements:to prove operability. No unreviewed safety question exists.

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.X6 Page 59

@g. . BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988

+ 7

' Procedure Changes and Temporary Modifications

- CHANGE TITLE:

Temporary.. Operating Procedure. (TOP) 1-88-20, " Stroke Tim 4ng RCS Cold Leg Sample Containment Isolation Valve.[TV-1SS-102A1]".

DESCRIPTION'0F CHANGE; This . TOP. obtains a' closing. stroke time for [TV-ISS-102A1] with the remote valve position indication unavailable by isolating sample flow with [TV-1SS-102A1] and observing downstream pressure decrease.

SUMMARY

OF SAFETY EVALUATION:

This TOP' utilizes a n'ormal sampling flowpath and the evaluation in Updated Final Safety Analysis Report Sectioti 9.6.3. applies to this procedure. Accidents involving the sample system are not evaluated in-Updated. Final Safety Analysis Report Section - 14.1. Containment penetration is designed in a manner such that special ' operational' test procedures can be used to' test the containment isolation valves for operability..(Updated Final Safety Analysis Report Section 5.3.1). This test will verify the operability of the containment isolation valves as-L required by Technical Specification 3.6.3.1, No unreviewed safety question. exists.

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Page'60L

- BVPS-1 Annual Report of Facility Changes, Tests 1and Experiments for 1988 Procedure Changes and Temporary Modifications

' CHANGE TITLE:

Temporary Operating Procedure.(TOP) 1-88-25, " Steam Generator Blowdown Mixed. Bed Resin Regeneration".

' DESCRIPTION OF CHANGE:

1

'This TOP. regenerates potentially radioactive resin using a vendor-supplied' vessel and the Stations Water Treating System' .

l

SUMMARY

OF' SAFETY EVAI$UATION:

Updated Final . Safety Analysis Report Section 14.2 states-that the

~

Primary Auxiliary Building (PAB) Sumps will collect any waste liquids that spill. Although this regeneration will not be performed in the.

PAB, the Chemical Waste Sump in the water treating area or the specially constructed temporary enclosure for the vessel will collect

-any spills. Updated Final Safety Analysis ' Report Section 14.2 also takes credit .for administrative controls and batch processing. This

' TOP and the vendors vessel procedures are administrative controls and the. regeneration -will be a batch process. The water treating systems

.affected by this TOP are not required for safe shutdown of' the Reactor

.(Updated' Final Safety Analysis Report Section 9.11). This TOP requires sampling of the Chemical Waste Sump and if necessary, preparation of a Liquid Radioactive- Waste Discharge Authorization prior to discharging the' sump.if the-sump contains radioactive liquid. In the event the vendor vessel fails and resin spills from the vessel, a temporary.

enclosure has been constructed to contain the volume of the vessel.

Radiological Control personnel have performed an evaluation to ensure that if a' failure of the vessel occurs and the enclosure also fails

such' that- all contents of the vessel are released to the river, no

. Technical Specifications or Offsite Dose Calculation Manual requirements would be violated. This evaluation has determined that

-Quarterly Technical Specification organ doses would be exceeded if all of the resin beads were discharged to the river. To' address this concern, a precaution has been added to this TOP stating that this TOP should not be performed if it is raining or rain is imminent. No unreviewed safety question exists.

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Page 61

.; BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications l

CIIANGE TITLE:

Temporary Operating Procedure-(TOP) 1-88-22, " Pressurizing the Control-

~

Room Bottled Air ~ Pressurization System '(CREBAPS). Tanks Using. a Temporary Air Compressor".

l- -DESCRIPTION OF CllANGE:

This TOP provides instructions for .usb g a temporary breathing air compressor.in conjunction with an existing compressor in order to.

. rapidly restore' CREBAPS to an' operable condition as required by Technical Specifications.

SUMMARY

OF SAFETY EVALUATION:

This- TOP restores CREBAPS pressure to a normal operating range so that' the system will function as described in Updated Final Safety Analysis Report Section'9.13.4. The CREBAPS tanks are pressurized in accordance with approved Operating Manual-procedure 1.44A,4.!! and the tank header relief valves are utilized to prevent overpressurization of CREBAPS.

This TOP is used only for restoring .CREBAPS to a normal operating condition .and does not manipulate any other functioning portion of CREBAPS used to pressurize the control room. The temporary air supply hose' from the temporary air compressor is. secured ~in place to prevent hose whipping if the hose fails, so that no safety-related equipment in the area should be affected. The Technical Specification margin of safety 13 maintained since this TOP restores CREBAPS. tank pressure to an operable range as' described by Technical Specification 4.7.7.2. The temporary compressor fills CREBAPS with only breathable Grade D air (filtered and monitored for. carbon monoxide). No unreviewed safety question exists. j l

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Page 62-q, ' < BVPS-1 Annual Report of Facility' Changes,

. Tests and Experiments for.1988 Procedure Changes and Temporary tiodifications CIIANGE TITLE:

Jumpered and Lifted Lead to Disable Trip Function of [LT-1QS-101A and B]

DESCRIPTION OF CilANGE:

This ' modification disables the automatic shutdown of the Quench Spray Chemical Addition Pumps [1QS-P-4A, 4B, 40 and 4D] and the automatic.

closure of the Chemical Addition Pump Discharge Valves [tl0V-1QS-104A and B]. This modification is necessary because of the discovery of-inadequate train separation on [LT-1QS-101A and B] Chemical Addition Tank' Level transmitter power 1 cads. Trip is disabled by pulling the output fuses.

SUtt?1ARY OF SAFETY EVALUATION:

The trips disabled are not involved in any safety function discussed in the Updated Final Safety Analysis Report, nor do they prevent any- j safety function from occurring, therefore, disabling them does not affect the probability or consequences of any analyzed accident (Updated Final Safety Analysis Report Section 6.4.2). The function of these level transmitters is to provide equipment protection. Since the safety function of the Chemical Addition Pumps would be complete when a low-low level occurred, their malfunction at that point would have no safety implication. By disabling these level transmitter signals, this modification prevents a single failure from affecting -both trains of chemical injection due to the lack of adequate train separation.

Disabling these decreases.the consequence of an accident. Technical Specification 3.6.2.3 basis margin of safety.is unaffected by disabling this level signal since the safety fur ction of the system remains unchanged. No unreviewed safety question exists.

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Page 63 Jg- BVPS-1 Annual' Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

Temporary Operating Procedure (TOP) J-88-24, " Transferring.Potentially Contaminated Na0ll Solution from Drums to the Chemical Waste Sump for Discharge".

DESCRIPTION OF CHANGE:

This TOP provides instructions for pumping potentially radioactive NaOH solution from the Chemical Addition area sump into 55 gallon drums,

.. pumping the contents of the drums into the Chemical Waste Sump and then neutralizing and discharging the solution in accordance with the-applicable water treating and radiological control procedures.

SUMMARY

OF SAFETY EVALUATION:

This TOP requires radiological control personnel to monitor the transport of the drums. Leakage inside the buildings will be collected in the Chemical Addition area sump or the Chemical Waste Sump in accordance with the accidental release analysis of Updated Final Safety Analysis . Report Section 14.2.2. Administrative controls and batch handling ensures control of process as described in Updated Tinal Safety Analysis Report- Section 14.2.2.3. This procedure will not reduce the margin of safety for release of . radioactive fluids as

-defined .in Technical Specification 3/4.11.1. The Water Treatment System is not considered to be important to safety. No 'unreviewed safety question exists.

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Page 64 BVPS-1' Annual Report of Facility Changes,.

' Tests and Experiments.for 1988 Procedure Changes and' Temporary Modifications

, , CHANGE TITLE:

Temporating -Operating Procedure'. (TOP) 1-88-05(06)(07),.

" Pressurization /HAFA Test of.IB(IA)(IC) Steam Generator".

DESCRIPTION 0F CHANGE:

Required Inservice Inspection testing to be used in conjunction with HAFA International Test Procedure for. Acoustic Emission Leak Detection on'[1RC-E-1B(1A)(IC)] Steam Generators.

SUMMARY

OF SAFETY EVALUATION:

Inservice inspection is in accordance with Updated Final Safety-Analysis Report.Section 4.5.1.4. The testing being performed utilizes

-normal- system arrangements and is within the design pressure and temperature limitations of the system. . The equipment being utilized is not- required to be available in Mode 5 or 6. Administrative controls are in place.in this procedure to preclude ~ overpressurization of the steam generator, to alert the operators of specific indications to monitor, to detect a steam generator . leak and to ensure the steam generator -is at'the proper temperature prior to pressurization. This test will not violate any Technical Specification limiting conditions for operation. No unreviewed safety question exists.

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Page 65

, BVPS-1_ Annual Report of Facility Changes, J JTests and' Experiments.for 1988

. Procedure' Changes and Temporary Modifications l CHANGE-TITLE:

' Operating Surveillance Test (OST) 1.33.30,'" Containment Hose Station Air Test".

DESCRIPTION OF CHANGE:

L ,This new ,0ST, formerly Temporary Operating Procedure 1-87-24, is a 3-l- year test.that requires that the hose stations in containment are l . verified operable by testing the' isolation valves and verifying that no-

~.fjow blockage exists.

SUMMARY

OF SAFETY EVALUATION:

l;

' Updated Final Safety' Analysis Report,Section 14.7.1 does not postulate a fire in containment concurrent with a dropped fuel assembly. 'This test' is performed while the plant is shutdown in order to avoid violating any Appendix R concerns for attaining plant shutdown. The primary means of fire suppression capability for containment are the deluge systems, which are unaffected by this test. The fire hose-stations in containment specified in Technical Specification 3/4.7.14.4 are an improvement to the fire protection program, not assumed in accident- analyses -(SER to Technical Specification Amendment 89). The fire, hose header will remain dry, precluding possible flooding or inadvertent dilution. .For any postulated fire in containment, manual No unreviewed

~

operator action is necessary for supression function.

safety question exists.

1 .

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Page 66

, .BVPS-1 Annual Report of Facility Changes, Tests and Experiments:for-1988 , .

Procedure Changes and Temporary Modifications CHANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-09, " Discharging the Gaseous-Waste' Surge Tank'(1GW-TK-2] to Atmosphere".

DESCRIPTION OF CHANGE:

This ' procedure discharges the contents of the Gaseous Waste l Surge Tank to the Decay Gas Injector utilizing the nitrogen . flushing decay. tank.

bypass line.

SUMMARY

OF-SAFETY EVALUATION:

The . Gaseous Waste. System will not be operated.outside of its design parameters and no equipment important to safety will be stressed beyond

-its design operating limits. The required gaseous. waste discharge permit is M adequate administrative control to ensure that the consequences of ' b ' . malfunction will not exceed those which were previously. evaluated for in the Updated Final Safety Analysis Report.

Upon receiving an alarm from the Gaseous Waste Effluent Monitor, the operator will terminate the discharge using existing ' station operating procedures. No unreviewed safety question exists.

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Page 67

.. BVPS-1. Atuiual Report of Facility' Changes, Tests and Experiments for 1988

' Procedure Clianges and Temporary t! modifications

'CllANGE TITLE:

Operating tianual Change Notice to Daily'lleat Balance Log DESCRIPTION OF CllANGE:

This- change corrects an error in calculation to make the Daily licat ,

Balance more accurate. This change adjusts the calculation. method to

' account for maximum expected feedwater injection flow to the blowdown system.

SUFIF1ARY OF SAFETY _ EVALUATION:

Updated Final Safety Analysis Report Section 14,' Table 14.3-2'a'ssumes reactor power is '102% of core design rating of 2652 f1Ve. This change

-will cause a - 0 1*. . change in calculated reactor power which is conservative. Normal, system configurations are not changed. No unreviewed safety question exists.

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Page 68 L

c <

BVPS-1 Annual Report of. Facility Changes, LTests and Experiments'for 1988  :

Procedure Changes and Temporary Modifications

• CHANGE TITLE:

Temporary Operating Procedure (TOP) l' 12 , . " Control Room Air Compressor Performance Test".

DESCRIPTION OF CHANGE:

This TOP checks the' performance of the control room air compressors and establishes baseline' data for a new balance :of plant Operating Surveillance Test for the Control Room Air Compressors.

l

SUMMARY

OF SAFETY EVALUATION:

r. There is no change being made to any system component or to the operation of the system components as described in Updated Final Safety

Annylsis Report Section .9.13.4 or Technical Specification 3/4.7.7.1.

This TOP is. checking equipment which could' impact on safety if inoperable. Monitoring this equipment for.a potential malfunction will enchance the margin of safety. No unreviewed safety question exists.

i ..

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)

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=Pa'ge 69

, ~=BVPS-1LAnnual Report of Facility Changes,.

Tests and Experiments for'1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

' Operating' Surveillance Test (OST) 1.43.7A,- " Alternate Noble Gas

-Monitors'[RM-IVS-109, 110]~ Function'a1. Test".

DESCRIPTION OF CIIANGE:

'This OST provides surveillance testing in accordance with Amendment-113' to Technical Specification 3.3.10, which allows use of these monitors as. alternates; provided surveillance requirements. applicable to the associated-inoperabic channel are met'.

SUMMARY

OF SAFETY EVALUATION:

'This etest involves equipment not included in.the Updated Final Safety Analysis Report. This test and the equipment tested does not affect-equipment- described in .the Updated Final Safety Analysis Report, and does not cause system or component actuation. This test ~ verifies radiation monitor- operability .in accordance with Technical L

Specification'3.3.10, Table- 3.3-13. No unreviewed safety question-exists.

L

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.i Page 70'

. BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

Operating Manual Procedure 1.51.4.D, " Station Shutdown - Cooldown from

' Hot. Shutdown (Mode 4) to Cold Shutdown (Mode 5)".

1 DESCRIPTION OF CHANGE:

This change ' eliminates the need to pressurize and subsequent !y drain the containment fire protection hose rack piping during outages.

During -an out age. , .a manual isolation valve will be left open and the

, remotely-operated trip valve will be left closed.

SUMMARY

OF SAFETY EVALUATION:

None of the Updated Final Safety Analysis Report Section 14 accident analyses postulate a fire _ concurrent with .the accident. All containment areas required to be prot ected by fire suppression still have a separate automatic deluge system. Level of protection by the water suppression system remains the same as described in Updated Final Safety Analysis Report Section 9.10 and the Appendix R review. As manual action is already required for a fire requiring hose rack operation, opening the remotely-operated valve will not hinder response capabilities. No unreviewed safety question exists.

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Page 71

    .                                                             .BVPS-1 Annual Report of Jacility Changes, u

Tests and Experiments for 1988: Procedure Changes and Temporary Modifications CHANGE TITLE: Operating Manual Procedure 1.18.4.AH,- "Uso and llandling of Sca/ Land Vessels"' DESCRIPTION OF CHANGE: This procedure provides guidelines for use of vendor-supplied radiation waste vessels that.will be used to transport waste to be compacted off-site.

SUMMARY

OF SAFETY EVALUATION: The hand 11_ng of solid waste is not evaluated in the Updated Final Safety Analysis Report. The handling of solid waste and packaging of radioactive materials is controlled by Radiological Control Manual Procedures. There are no applicable' Technical Specifications fcr non-solidified solid waste. No unreviewed safety question exists. l l

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                                                                                                                                                               'Page 72

m. .BVPS-1 Annual Report of Facility Changes,

                                                               , Tests and Experiments for.1988 L                                    Procedure Changes and Temporary Modifications l
             ' CllANGE TITLE:
                        -Temporary Operating Procedure (TOP) 1-88-016, "liydro Test of.[3WR-P-1AJ Vacuum Breaker Valve [1RW-486]".

l1 y DESCRIPTION OF CllANGE: L. This TOP provides instructions to support.a hydro test after welding has been completed on the valve..

SUMMARY

OF SAFETY EVALUATION: The river . water pump being tested will be' out of service and will not be relied upon to perform its safety function. . A rupture of the vacuum -

                        '. breaker piping due- to the applied test pressure will not affect the operability of the in-service river water system or other components in-the A cubicle due to the small volume and energy of the released water.

This test requires that the hydro test be completed satisfactory. prior to returning the pump to service. No unreviewed safety question exists. L______=_--_--__-____

       .      (

Page 73

    ..                                                                                                                BVPS-1 Annual. Report of Facility Changes, Tests and Experiments for 1988

Procedure Changes and Temporary Modifications CHANGE TITLE:

Temporary. Operating-Procedure (TOP) 1/2-88-01,." Shutdown of the Gaseous Waste System While Unit 1 and Unit 2 are at. Power". DESCRIPTION OF CHANGE:

                                                                    -This TOP.provides instructions to shutdown the Gaseous Waste System so that piping can be installed in the disposal header by DCP-645.                                  The Gaseous Waste System is returned to service after the modification is complete.

SUMMARY

OF SAFETY EVALUATION:

                                                                   'This TOP will not cause the inventory of radioactive gases in the Unit i volume control tank or surge tanks, or in the Unit 2 Gaseous Waste
                                                                     . System to increase . greater than the amount assumed in the Unit 1 Updated Final Safety Analysis Report Section 14.2.3 or Unit 2 Section 15.7.1.                                If a malfunction occurs during the performance of this TOP,  ,

the consequences have been analyzed and are enveloped by the Unit 1 Updated' Final Safety Analysis Report Section 14.2.3 and the Unit 2 UFSAR Section 15.7.1. The Unit I and Unit 2 surge tanks, decay tanks and storage tanks pressure will be decreased as low as possible per existing procedures before shutdown of the Gaseous Waste System to minimize the possibility of an overpressure condition and to provide gas storage inventory if needed. Equipment with over pressure relief to the Gaseous Waste System is shutdown to preclude the possibility of additional inventory to be stored. The Unit 1 Main Condenser Steam Jet Air- Ejectors are aligned to'the Turbine Building atmosphere upsteam of the system radiation monitor. Radiological Control Personnel are required to sample the discharge and if high, notify the Control Room to isolate the discharge. The Unit 2 air ejectors discharge to the Turbine Building atmosphere via the radiation monitor. When the modification js complete, the system will be returned to service in accordance with Technical Specification 3.3.3.1 and 3.3.3.10. No unreviewed safety question exists, t 1

j , .i '- A; [< . , . Page 74. i BVPS-1-Annual Report'of. Facility Changes, ., Tests:and Experiments for-1988 ( Procedure Changes and Tempor3ry Modifications

                      ' CHANGE TITLE                                                                                                    ,

1 Opekating .Surva111ance . Test (OST) 1/2.49.9, " Liquid and Gaseous Effluent' Monitoring Instrumentation Channel Functional Test". DESCRIPTION CF CHANGE: This change deletes the check of remote meters end alarms.

SUMMARY

OF SAFETY EVALUATION: [his test: does not change the method or frequency of testing.of equipment listed in Updated Final Safety A1.alysis Report Section 11.3.3.3. The- remote indications to be deleted are provided for

                           . maintenance and calibration use (Updated Final Safety Analysis ' Report
                            -Section 11.3.3.2).              The radiation monitors listed in Updated Final Safety Analysis Report Section 11.3.3.3 will continue to function 'as required by -Technical Specification 3/4.3.3.9 and 3/4.3.3.10. No unreviewed safety question exists.

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Page 75 l

   .                                              BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988                          i i

Procedure Changes and Temporary Modifications- j CHANGE TITLE: Temporary Operating Procedure (TOP) 1-88-18, "[MOV-ISI-869B] q Operability Test". { ' i l' DESCRIPTION OF CHANGE: j This TOP provides a stroke test of [MOV-ISI-869B) to verify the valves' operability after cable replacement. { j

SUMMARY

OF SAFETY EVALUATION: l One charging pump will be available for the boron injection tank i flowpath. Updated Final Safety Analysis Report Section 14 assumes one train of ECCS available. In the event of a Train A failure, an operato- will be stationed at the valves necessary to place Train B in l servict ithin approximately one minute. Technical Specification bases state diat either train is capable of limiting peak cladding temperatures within acceptable limits. No unreviewed safety question exists. l l l l l l L-__.__ _ - _ _ - - - _

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Page 76

    ....                              BVPS-l' Annual Report of Facility Changes,
                                            . Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

Operating Manual Procedure 1.24.4.T, " Draining and Refilling Steam Generators". DESCRIPTION OF CHANGE: This change adds the option of opening the reactor trip breakers, placing contro1' switches for [1FW-P-3A & 3B] in PULL-TO-LOCK and isolating steam to the Terry Turbine instead of putting dummy level signals in the steam generator narrow range level indication.

SUMMARY

OF SAFETY EVALUATION: l :. All. equipment is being'used as designed and as stated in Updated Final Safety Analysis Report Sections 7.4.1, 10.3.5 and 7.1.2. Updated Final Safety ' Analysis Report. Section 14.2.5 assumes no-load conditions because of greater mass in the steam generat. ors. This change limits i the fill to .less than the high level trip setting. The control rods l are not required 1.o be operable in Modes 4, 5 and 6 (Technical Specification 3/4.1.3.1). The auxiliary feedwater pumps are not required to be operable in Modes 4, 5, and 6 (Technical Specification 3/4.7.1.2), No unreviewed safety question exists. l l 1 l-l l-l l l L l

4 Page 77

   ,                                                           -BVPS-1 Annual Report of Tacility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CllANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-23, " Leak-Testing [1BR-E-13A(13B)], Primary Water Storage Tank Heater" DESCRIPTION OF CllANGE: This TOP provides instructions to Icak test the primary water storage tank heater using primary grade water from the discharge of the-Primary Water Supply Pumps'.

SUMMARY

OF SAFETY EVALUATION: l' The procedure requires the operator to check that no leaks exists on the test rig. This will reduce the probability of a release of liquid waste as described in Updated Final Safety Analysis Report Section 14.2.2. In the event. of a Icak or pipe break, the water would be collected in the Primary Grade (PG) Pump Room Sump as described in Updated Final Safety Analysis Report Section 14.2.2. The primary water supply system is non-safety related and no safety related equipment is located in the PG Pump Room. The test pressure will be less than the design pressure and relief protection is available. No unreviewed safety question exists. 1 I l 1 1 1 l l l l 1

   .i Page 78
 .                                   BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications Cl!ANGE TITLE:

Operating Manual Procedure 1/2.56A.4.J, " Personnel Injury". DESCRIPTION OF CHANGE: This change adds instructions for the landing of an emergency medical helicopter on-site for the transportation of injured personnel to an offsite facility.

SUMMARY

OF SAFETY EVALUATION: Chapter 2 of the Updated Final Safety Analysis Report describes the rationale for acceptability of the site chosen to construct Beaver

                       'Vallny Power Station (BVPS) and establishes the portion of the design basis which resolves from " hazards" arising from the site location.

Therefore, air traffic studies described in the Updated Final Safety , Analysis Report are not intended to include flights associated with I plant activities. If the specific hazards associated with these particular flights have been bounded by the current design basis, other sections of the Updated Final Safety Analysis Report will be unaffected also. These hazards would be limited to external fires, explosions and externally generated missiles. Since BVPS is designed to accomodate external fires and since explosions or missiles would obviously have consequences far less severe than a design basis tornado, no other Updated Final Safety Analysis Report chapters will be affected. No unreviewed safety question exists.

        ?, e Page 79
      ,                             BVPS-1 Annual Report of Facility Changes.

Tests and Experiments for 1988 Procedure Changes and Temporary Flodifications CilANGE TITLE: Temporary Operating Procedure (TOP) 1-88-28, " Venting Low IIcad Safety Injection Pump Supply to Charging Pump Piping". DESCRIPTION OF CilANGE: This. temporary procedure provides instructions to remove a gas bubble discovered in the safety injection piping located between the low head safety injection pumps and the high head safety injaction pumps. This l -- procedure utilizes volume control tank pressure to vent the gases out I existing vent valves. The venting method requires de-energizing a motor-operated valve in the flowpath. SUrld!ARY OF SAFETY EVALUATION: l The motor-operated valves (fl0V) to be manipulated are cycled open on a quarterly frequency for ASitE requirements. The venting procedure is performed on one safety injection train at one time. Adequate time exists for operator action to restore normal system arrangement in an i accident condition due to function of flewpath not being required on ! safety injection initiation. The procedure requires an operator stationed at the motor control center when the r10V is de-energized so that the response may be as quick as possible. The Low llead Safety Injection (SI) pump is in PULL-TO-LOCK to prevent an automatic SI start i from over pressurizing the venting rigs. Starting of SI pump can be timely initiated on receipt of automatic SI signal from the Control Room. The procedure requires that the t!OV be electrically closed after manual operation to ensure that valve wedging will not occur. Action statement of Technical Specification 3.5.2 allows 72 hours to restore safety injection system. No unreviewed safety question exists. a

      .e
                                                                                                                                                       'Page 80 BVPS-1 Annual Report of. Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CHANGE TITLE:

Operating Manual Procedure 1.6.4.AM, " Returning an RCS Loop to Service with Fuel in.the Vessel". DESCRIPTION OF CHANGE: This change incorporates a previously-approved Temporary Operating Procedure 1-8b-?4 into a permanent operating manual procedure.

SUMMARY

OF SAFETY EVALUATION: This procedure will not cause any equipment important to safety to be operated outside of its design parameters. This procedure is within the envelope of the Updated Final Safety Analysis Report description of the startup of an isolated Reactor Coolant System loop (Updated Final Safety Analysis Report Section 14,1.6.1.2) and ensures all interlock i conditions are met, This procedure meccs the intent and wording in all applicable Technical Specification (3.4.1.4, 3.4.1.5, 3.7.2.1). No unreviewed safety question exists. l l I l 1 1

Page 81

                                      ,                                                      BVPS-1 Annual Report;of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications
                                               ' CHANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-29, '" Venting the Low Head Safety' Injection Pump Supply to Charging Pump Piping". j DESCRIPTION OF CHANGE:

                                                    'This temporary procedure provides instructions to flush a gas bubble discovered in the safety injection piping using a low head safety injection pump to provide forced flow through a shutdown charging pump.

lIle venting method requires that a motor-operated valve be cycled in the safety injection vent flowpath.

SUMMARY

OF SAFETY EVALUATION: The motor-operated valve (MOV) to be manipulated is cycled open at a quarterly frequency in .accordance with ASME requirements. This procedure requires that the MOV be electrically closed after manual operation to ensure that wedging will not occur. The . performance of this -procedure does not affect the availability of equipment required for accident response. This procedure is to be performed in Mode 5 to preclude removing plant equipment from service that is required to be operabic in other modes. The possibility of an unmonitored release at the Refueling Water Storage Tank (RWST) vent has been considered and a Technical Evaluation performed to show that a positive pressure in the RWST airspace will not occur due to the release of gases in the RWST. No unreviewed safety quesion exists. 1 i l: i

. _ _ _ _ _ _ _ _ . - _ _ . _ . _ _ _                  _ _ . _ _ _ _ . _ _ _ _ ...-m_  _ _ _   .u   __.-.mA                           _ - _ . _ _ . _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ . _ . _ _ _ - _ - _ . _ - _ _ _ _ _ _ _ _ _ . _ - _ _ . _ _ _ _ _ _ _ . _ . _ _ _ . _ _ . _ . _ _ _ _ _ . _ - _ _ _

n $ Page 82

                    ,                                    BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1986 Procedire 2   Changes and Temporary Modifications tilANGE TITLE:

Temporary Operating Procedure (TOP) 1-88-30, " Venting the Low Head Pump Supply to Charging Pump Piping" l DESCRIPTION OF CHANGE: This temporary procedure facilitates an engineering evaluation of the extent oi gas buildup in the safety injection piping and provides a means of removing the gas. A temporary hose will be utilized to vent the gas bubble from the low head safety injection pump discharge to a  ! primary drains transfer tank.

SUMMARY

OF SAFETY EVALUATION: One train of low head safety injection pump to charging pump Emergency Core Cooling System flow is isolated during this procedure, however continuous communication is maintained between the Control Room and an operator stationed at the charging pump suction isolation valves so that the valves will be opened and this procedure terminat ed if a safety injection occurs anytime during the performance of inis TOP. The temporary hose utilized is rated in excess of the shutof f head that can be developed by the low head safety injection pumps. The primary drains tank has adequate relief protection for the protection of equipment. A temporary pressure gauge monitored during procedure { performance ensures that pressure to the primary drains tank 411 not challenge the installed over pressure protection. Technical Specification 3.5.2 Action Statement allows 72 hours to restore two trains to service. No unreviewed safety questions exists. I l l l

Page 83

q. ?BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 3

x Procedure' Changes and Temporary Modifications

                                       . CHANGE TITLE:
                   ,                                            Temporary Operating Procedure ' - (TOP). 1-88-32, " Determination of Hold Times for Stroking Air-Operated Yalve".

DESCRIPTION OF CHANGE: This procedure will provide. data for establishing repeated valve. stroke requirements by determining the necessary air. operator recharge time lto

                                                              -obtain consistent valve stroke times.

SUMMARY

0F SAFETY EVALUATION:

                                                            .The= valves- to be -stroked by this test are already stroked at a l quarterly frequency in accordance'with ASME requirements, and this test
                                                              . utilizes the same test method. Most.of the valves will only be stroked twice.so that tlie possibility of a malfunction will not be increased.

Failure o f these. valves during tuis test would leave them in their safety position. .No unreviewed safety question exists. 1 I

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_ _ _ _ . _ . - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ . . - - - _ _ - _ _ _ - _ _ _ - _ _ . _ _ - - - - - - - . - - - - - _ - - _ - - - ~ ..-. - - - - - _

 -e Page 84

, BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 Procedure Changes and Temporary Modifications CllANGE TITLE: Temporary Operating Procedure (TOP) 1-89-01, " Post-Maintenance Test for [1Cll-P-1A] Benchboard Control Switch Replacement". DESCRIPTION OF CHANGE: This procedure is a post-maintenance test to verify that the vontrol switch installed provides t'ie proper control logic to the remainder of the control circuit. By performing continuity checks from the terminal

                                  . board, through the control switch and back to the terminal board for each set of control swizch contacts and for each control sicitch position, the proper co.. trol functions will be verified.

SUBL >!ARY OF SAFETY EVALUATION: This test ensures that the control of the pump is available to mitigate the accidents evaluated in Updated Final Safety Analysis Report Section 14 The operability of the pump will be verified by satisfactory performance of its surveillance test. As a continuity test, this procedure does not involve the potential for damage to any unit equipment (Updated Final Safety Analysis Report Section 7.3.2.1.5). The station equipment that could possibly be affected by this test is limited to the 1A Charging Pump, whose loss (one of two operable pumps) is analyzed for in Updated Final Safety Analysis Report Section 6.3.1.2. No unreviewed safety question exists.

g ' l . 9 ,W l Page 85 BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 b u l> Procedure Changes and Temporary Modifications

 ,                       CHA_N,GE TITLE:
                        ' Te:9orary Operating Procedure (TOP)             1-88-33, " Leak Test of [1RC-68]

r;i ack Valve" (, : DEf.CRl"rION OF 1 ANGE: L .

h. toccoure verifies the function of [1RC-68], Primary Relief Tank Nitrogen Supply Check Valve by performing a reverse flow leak check on

                       .the valve.

SUMMARY

OF SAFETY EVALUATION: Any leakage past the check valve will be collected by the Gaseous Waste System and stored in the gaseous waste decay tanks. This TOP conveys gas' through normal system flowpaths analyzed by the Updated Final Safety' Analysis Report. In t.he event of a leak in the temporary jumper being utilized, the jumper is readily isolabic by closing manual valves. The Updated Final Safety Analysis Report has analyzed gaseous releases with far more potential for activity than this procedure could release. No unreviewed safety question exists.

             --                                                                                          Page 86 k

J BVPS-1 Annual Report of Facility Changes, Tests and Experiments for 1988 UFSAR CHANGE CHANGE TITLE Updated Final Safety Analysis Report Table 5.2.5, Reactor Coolant Water Chemistry Specification DESCRIPTION OF CHANGE The RCS Lithium concentration specified in the Updated Final Safety Analysis Report will be revised to be variable from 0.7 to 3.75 ppm as a function of Boron concentration. This change is based on Westinghouse recommendations as discussed below.

SUMMARY

OF SAFETY EVALUATION The probability of an accident previously evaluated in the Safety Analysis Report will not be increased because: , 1 a) below 1200 ppm Boron the Lithium is maintained within original i Westinghouse design. b) in the short time the plant is operated above 1200 ppm Boron (Unit 1 cycle 7 beginning of cycle boron concentration is 1283 ppm), the Lithium will be maintained as recommended by Westinghouse in the attachment to Standard Information Package SIP 5-1 Rev. 4 for operation above 1200 ppm. Westinghouse has determined that no appreciable Zircaloy cladding corrosion will occur with this regime. Also, since Beaver Valley Unit 1 has seen little evidence of pressurized water stress corrosion cracking (PWSCC) thus far in operation and since Unit 1 previously operated during cycle 1 and 2 with similar high pH chemistry, the probability of increased corrosion of pre-initiated PWSCC is low. In addition, 100% eddy current is planned for the steam generators in the future and PWSCC levels will be specifically evaluated. Based on those findings, the chemistry regime in use will be reevaluated. Reference Westinghouse Technical Bulletin NSID-TB-88-03. The consequences of an accident previously evaluated will not be increased because the plant was designed and evaluated for the use of Lithium in the RCS. The probability of a malfunction of equipment important to safety will not be increased because the plant was designed to operate using Lithium and the Lithium concentrations will be kept within recommended guidelines (NSID-TB-88-03). The consequences of a malfunction of equipment important to safety will not be increased and should be decreased due to reduced ex-core radiation fields. The possibility of an accident of a different type than any previously analyzed will not be created because the Lithium will be kept within Westinghouse guidelines (NSID-TB-88-03). The possibility of a malfunction of a different type than any previously i evaluated will not be created. L______._________.____ . _ . . _ _}}