ML20203B278

10CFR50.59(b) Rept of Changes to Oyster Creek Systems & Procedures

ML20203B278
Person / Time
Site:	Oyster Creek
Issue date:	02/16/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:
References
6L40-98-20074, NUDOCS 9802240247
Download: ML20203B278 (59)

Text

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GPU Nuclear. inc.

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U.S. Route H South NUCLEAR la,*,',*ll*,**n'lu**m an Tel 609-9714000 February 16, 1998 6L40 98 20074 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 4

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Dear Sir:

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Subject:

Oyster Creek Nuclear Generating Station Docket No. 50 219 10 CFR 50.59(b) Report

. In accordance with 10 CFR 50.59(b), enclosed are the summaries of the changes to the Oyster Creek systems and procedures described in the Safety Analysis Report (S AR) for the period April 1993 to March 1995.

. If any additional information is required, please contact George Busch at (609) 971-4643.

Very truly yours,

\\% % a kg MichaelB.Roche d

Vice President & Director Oyster Creek MBR/GWB c'

Attachment

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Administrator, Region 1 NRC Project Manager NRC Resident Inspector 9802240247 980216

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PDR ADOCK 05000219 P

PDR l.l!l.l 1.11.11.11!!!.Ill.1.1

ATTACilM ENT I Procedure 312.1, Rev. 0 SE No. 000212-022 Rev. O A

"EOPS Modil'ication ta Prevent Core Spray Injection During ATWS Events" Dsscription oLChange:

This change directs manual operator actions to terminate and prevent injection from Core Spray as directed in the Level / Power Control and RPV Flooding Emergency Operating Procedures (EOP's). These EOP's are used to control reactor water level and power under the condition in which a reactor scram signal was initiated and all of the control rods did not fully insert. The EOPs direct the operator to control water level using reactor injection sources which inject outside the core shroud (i.e., Feedwater and CRD) in order to limit the reactivity effects on the unrodded core due to the injection which may be at reduced temperature. Should these systems g

be unable to maintain reactor water level above a specified level, the Operator is instructed to O

rapidly depressurize the reactor in order to use low pressure injection systems (Condensate, Fire Water and Core Spray). Duiing the depressurization of the reactor, the EOP directs that injection from all of these low pressure injection systems be " terminated and prevented" in order to ensure that as the reactor is depressurized that there is not a sudden injection of coid unborated water into the con e. An uncontrolled injection into the -

shich is still not shutdown may cause a significant reactivity addition to the core and can result in damage to the fuel cladding. The core parallel isolation valves, which are interlocked with reactor pressure, will automatically open as pressure decreases below approximately 300 psig. Thus, additional detailed instruction for the operator is needed to override the automatic opening of the core spray parallel isolation valves under the conditions of a scram failure when low pressure injection systems are required to control reactor water level.

Safety Evaluation Summary; The implementation of this edditional guidance does not adversely afTect nuclear safety or plant operation. This guidance pei nits the Core Spray System to be manually operated and is only used for events beyond the plant design basis.

There is no increase in the probability of occurrence or the consequence of an accident previously evaluated in the SAR. The proposed actions will be taken for events which are not within the bounds of any of the transients analyzed in the SAR; there is no increase in the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.

The actions are being taken for events beyond the plant design and licensing basis; this change does not create a possibility for an accident or malfunction of a different type than previously evaluated in the S AR. The ability to control the Core Spray System is being maintained; this change will not decrease the margin of safety as defined for any Te, hnical Specification. The proposed change applies only to events that are covered by the Technical Specifications; there is no impact on plant Technical Specifications as a result of this change. The proposed operator instructions do not constitute an unreviewed safety question.

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Attachmert i CFR 50.59 Report April,1993 - March,1995 Page 2 of 58 Modification OC-MM-402939-007, Rev. O SE No. 402939-008, Rev. 2 "I-8 Sump Equipment Upgrade" Description of Change:

The purpose of this modification is to simplify maintenance and reduce radiation exposure to peisonnelin the 1-8 Sump /CRD Room. The modification will replace the sump pumps, pump discharge piping and sump cover plates; and add quick-disconnect connectors to the pump power cables to facilitate pump removal and replacement for ALARA considerations. Equipment drainage and other liquid waste in the Drywell is either piped or flows by gravity to the 1-8 sump located within the Reactor Pedestal on the Drywell floor. The sump ncrmally collects 1400 gallons per day (gpd),2250 gpd maximum. Two sump pumps transfer the liquid waste to the Radwaste Building. Level switches are provided to control the pumps and alarm at a high liquid level in the sump.

i Safety Evaluation Summary:

This modification is a replacement in kind and is upstream of the containment isolation valves V-22-28 and V-22-29. This modification will simplify maintenance and reduce radiation exposure to personnelin the 1-8 Sump /CRD Room. The mooification does not effect nuclear safety or safc plant operations; does not involve an unreviewed safety question; and does not require a change to the Technical Specifications.

1 Modification OC-MM402925-002, Rev. O SE #402925-002,- Rev. O

" Main Turbine Generator Protection Relay" Description of Change:

This modification upgrades the turbine gener tor protection system to meet the requirements of General Electric (generator manufacturer) and the PJM relay subcommittee recommendations on generator protection. The modification replaced the existing electromechanical type relays for the generator protection in the Control Room with two (2) new independent microprocessor based Digital Protection Relay Systems (DPRS) "A" and "B" Two protection systems (DPRS "A" as primary, and DPRS "B" as backup) are selected for redundancy, and willincrease the system availability by providing adequate uninterrupted generator protection, even in the event of failure in one of the protection systems.

CFR 50.59 Report April,1993 - March',1995 Page 3 of 58 The modification added a 20 amp fuse in panel llF to provide IE isolation between the safety related DC power supply from battery "C" and the non-safety related generator protection relay system. The DPRS provides a new sequential tripping system which initiates a generator trip on -

signals for reverse power and all steam valves closed. The new relay scheme will maintain the existing emergency diesel generators start time sequence and meet the required sequential tripping criteria by giving the trip directly to the 4160V switch gear breakers l A and IB, upon detection of a steam valve closure.

Safety Evaluation Summuy; This modification provides additional protective features for the turbine generator. These features effect systems associated with plant transients; e.g. loss of external load, turbine trip, and loss of auxiliary power. The DPRS does not adversely effect nuclear safety or safe plant operation; does not increase the probability or consequences of an accident; or reduce the margin of safety as deiined in the Technical Specifications. For these reasons, the modification does not constitute an unreviewed safety

.. c s Modification OC-MM-402910 005, Rey, O SE #402910 005, Rev. O

" Control Room Recorders Upgrade" Resninlion of Change; GPUN performed a human factors review of the Control Room to address the requirements of NUREG 0737, item 1.D.1, regarding the Detailed Control Room Design Review (DCRDR). The DCRDR identified existing huran factors deficiencies involving sixty-six recorders. GPUN i6nt fied corrective actions to remedy the deficiencies.

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I The scope of this modification is to remove or replace the recorders with the following tag numbers: RIO5 A, B, C, D (Panel 4F); SRM-RG05 (Panel 4F); RN0035 (Panel 10F) and LPRM-RJ13 A, B (Panels 3R and SR). Remove radiation monitors RN008A and RN008D from Panel IR and install blank plates. Replace selector switches 10S I through 10S10 on Panel 4F.

Safety Evaluation Summuy; The modification does not change any nuclear safety related systems nor does it change the function of any plant system. For these reasons, it is concluded de subject modification does not have any adverse effect on nuclear safety or safe plant operations or the environment. The modification does not constitute an unreviewed safety question as determined by 10 CFR 50.59.

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Anachment I CFR 50.5i Report April,1993 - March,1995 Page 4 ef 58 G.E. T.R. NEDE 81758P-A, Rev. 0 SE #402902-002, Rev. 2 "New Generation Control Rod to be Used for Testing in the Core" Dncription of Changt GPUN installed a GE New Generation Control Rod in Oyster Creek core during the 12R refueling outage. The control rod is used as a " test rod" or " demonstration rod" to develop inservice experience.

Safety Evaluation Summary; The New Generation Control Rod was renamed the hiarathon Control Rod. The control rod has three cycles in-reactor service experience at Oyster Creek. This safety evaluation coruludes that the initial h1arathon control rod should continue to be used as a test rod for future cycles. The proposed activity will not impact the safety of the plant nor will it require a change to the Oyster Creek Technical Specification. However, the risk of unforeseen problems exist, therefore, the lead htarathon control rod will continue to be included in the GE surveillance program as outlined in this evaluation ifit meets the exposure requirements of the NRC safety evaluation.

Since the operational history of the hiarathon control rod has been very successful at Oyster Creek and other sites and the three (3) poolside inspections at Oyster Creek plus other reactor

-sites did not discover abnormalities, it is further concluded that additional hiarathon control rods can be installed in the Oyster Creek reactor core without restricting the quantity and core locations in future cycles.

Modification: MDD-OC-625B, Rev. I SE #402901-003, Rev. 2 "Feedwater and Recirculation Flow Control Systems Upgrade" Descrip. tion of Change:

A new Digital Control System (DCS) will replace both the Feedwater Control System (FCS) ard the Recirculation Flow Control System (RFCS) and will provide improved control capability and reliability and resolve the problem ofimpending hardware obsolescence.

Safety Evaluation Summary:

This modification does not pose a safety concern or unreviewed safety que tion. This modification interfaces with safety related components through devices that uo not cont digital electronics such as fuses and circuit breakers. There is no software interface between this modification and existing plant safety systems. Additionally, no new safety related digital a

Attachment !

CFR 50.59 Report April,1993 - March,1995 Page 5 of 58 components are installed. Although some existing safety related components are being relocated, functionally they are unchanged, There is no impact on safe plant operation due to the relocation of the safety-related equipment not related to DRFCS. Nw failure modes are identified and addressed, as is thejustification for reduced potential plant transients attributable to feedwater system failures. The proposed modification does not have any adverse effect on safety or environmental impact. This modification does not constitute an unreviewed safety question as determined by 10 CFR 50.59.

Modification: OC-IS-402898-003, Rev. O SE #402898-002, Rev. O

" Turbine Building Roof Overlay Over Turbine Operating Floor" Description of Changt This modification removed the roofing material from the emergency repaired areas in the northwest corner and southwest corner of the Turbine Building high roof and removed about a 12 foot wide strip of roofing material along the east, west, north and south parapets to expose the metal roof deck and existing concrete crickets. The purpose of this removal was to determine which deck areas required either "use as found", wire brushing / priming, or metal desk reinforcement. Loose gravel was removed from the whole roof for reuse, insulation board was screwed down through the existing roof covering into the metal deck, and a build-up roof overlay system was installed. The final composite roof system has a Class B fire classification in accordance with the BOCA National Building Code 1993. It was fastened to the metal roof deck to meet Factory Mutual's 1-90 requirements. Since the reroofing added additional dead load to the roof, the roof live load was reduced from 30 psf to 25 psf.

Safety Evaluation Summary; The modification did not have any adverse effect on nuclear safety, safe plant operations; did not change the probability or consequences of an accident previously analyzed; and did not reduce the margin of safety in any Technical Specifications.

Modification: OC-MM-402890, Rev. O SE #402890-001, Rev. 2

" Permanent Drywell Scaffolding" Description of_C_hangt Scaffolds are required in the drywell for servicing of piping, valves, and el:ctrical components during plant outages. The present practice is to erect the scaffolds during each outage and remove an of them prior to re-start. The purpose of this modification is to design and install a permanent drywell scaffolding basic framework that can be left in the drywell during plant

CFR 50.59 Report April,1993 - March,1995 Page 6 of 58 operation. This modification will eliminate the repeated work of erecting and dismantling scaffolding frames during each outage. The scope of this modification includes the design and installation of the basic steel frames which are assa.; led using steel tubing, joint couplers, base plates, and concrete anchors. This modification provides the basic framework (skeleton) of scaffolds only. Other temporary items which are required to form a service platform such as planks, additional posts, bracing, handrails ladders, etc., will be added onto the permanent scaffolding basic framework during each outage in accordance with 0.C. Procedure 105.2 and is not covered under the scope of this modification. Those temporary items shall be removed prior to plant re-start. The original installation package provided for the installation of permanent scaffolding in the OC drywell only. A revision to the package expanded the scope of the installation to include the Reactor Building. The generic scaffold designs for elevation 23'-6" of the OC drywell shown on B&R drawing B&R 0238 can now be installed in the Reactor Building at floor elevation 23'-6" and 51'-3" Revision 2 of this safety evaluation constitutes the safety review for those efTotts.

1 Safety Evaluation 1uminary; The purpose of this modification is to install the drywell and Reactor Building scaffolding supporting frames during an outage. These frames are designed in such a way that they can i

remain in place permanently during plant operation. This will reduce overall cost and save time by eliminating the need to assemble / disassemble scaffolding supports during each plant outage.

The structural design of this modification is in compliance with all safety requirement codes and regulations. Therefore, it is concluded that the proposed modification will not have any adverse efTect on nuclear safety or the environment.

Modification: OC-MM-402880, Rev. O SE #402880-005, Rev. O

" Removal of Railroad Rails in Reactor Building" Q.jeription of Change:

e This safety evaluation assesses the final plant configuration afler removing the railroad rails from the Reactor Building on the 23'-6" elevation. The physical arrangement of the railroad rails and proposed removal procedures are described in OC-MM-402880-002. The railroad rails serve no purpose (since railroad cars are not used to nrve equipment into or out of the Reactor Building).

The rails are currently mounted approximately I to 1 1/2 inches above the floor elevation. This produces a trip hazard and makes it diflicult to roll heavy equipment across the floor. The rails are to be removed using an abrasive concrete 6oor saw. This safety evaluation addresses the possible safety aspects associated with the rail removal project.

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CFR 50.59 Report April,1993 - March,1995 Page 7 of 58 Safety Evaluation Summarv; This modification covers the removal of railroad rails in the Reactor Building. An analysis of safety systems and functions affected by the modification has not revealed any new safety concerns introduced by this work: Specifically: The unit's ability to mitigate the effects of all Design Basis Events (DDE's) is not changed during or subsequent to the modification. There is no change to the Plant Operating License or Technical Specifications. A3 a m:,4, this modification does not involve an unreviewed safety question or environmental impact.

Modification SE #328354-002, Rev. I

" Removal of RPV Internals Vibration Brackets" Description of Change; The purpose of this Safety Evaluation is to describe the effect on safety, system performance and

- plant design resulting from the removal of reactor internal vibration program brackets and/or instrumentation conduits currently attached external to the core shroud. This activity is necessary to support installation of repair tie-rods on the core shroud and/or support inspections of shroud welds.

These brackets and conduits were installed for pre-operational testing for the purpose of determining the vibration response characteristics of key reactor internal components and recirculation loops due to dynamic forces generated by the coolant flow and pump rotation.

The results of the initial testing are documented in NEDE-13109, Class 11, June 1970, entitled

" Oyster Creek Start-up Test Results". There is no need for the vibration system presently or in the future. Only the interfering brackets and conduits will be removed. Conduits left in place will be cut at the nearest remaining bracket. The instrumentation brackets and associated conduits are non-safety related and do not suppon, directly or indirectly, any safety or operational plant function.

The removal of these brackets and conduits will be accomplished by employment of a cutting process known as "EDM" (Electro Discharge Machining). This process has been qualified and used in similar applications at other BWR's. (The EDM for this application is done remotely under water for which the process is qualified.)

Safety Evaluation Summary:

The removal of the OC reactor internal vibration brackets and associated conduits externally attached to the core shroud has been evaluated. The process (EDM) to be used and associated tooling as well as personnel performing the task will be qualified as to pres ent loose pieces or damage to the shroud structure or the recirculation system or the reactor imernals. This modification does not: li adversely effect nuclear safety and/or safe plant operations; 2) increase

Attachment I CFR 50.59 Report April,1993 - March,1995 Page 8 of $8 the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR; 3) create the possibility for an accident or malfunction of a different type than any previously identified in the SAR; or 4) decrease the margin of safety as defmed in the basis of any Technical Specification.

Modification: CCD-328333-004, Rev. O SE #328333-009, Rev. O

" Installation of 10" Tees on ESW at Intake" Qc5miption of Changc; This modification replaced a 10" elbow with 10" tee and a blind flange on each Emergency l

Service Water supply line at the intake. This new arrangement allows for inspections and clean-1 out of these lines. This would be accomplished by removing the blind flanges. This change also installed a rigging point from which the blind flanges can be lifted during inspection.

Safety Evaluq1ionSummaIy.;

The safety evaluation determined that the modification did not change the design basis of the system; did not adversely effect nuclear safety or safe plant operations; did not increase the probability of occurrence or consequences of any accident previously evaluated; and did not reduce the margin of safety in the Technical Specifications. This configuration change in no way effects the safety function of the system and does not result in an unreviewed safety question.

Modification: OC-CCD-000225-001, Rev. O OC-CCD-328333-004, Rev. O SE #328333-008, Rev. O

" Replacement of Control Rod Drive Pump Stop Check Valve (s)"

Description of Change:

This configuration change replaces the Control Rod Drive (CRD) valve V-15-0007 and/or V-i5-0010 each with a separate check valve, a globe valve for throttling, and a globe valve for shutoff.

The existing valves are stop check valves.

Safety Evaluation Summary:

The activity has no effect on the margm of safety as defined in the SAR. Nuclear safety and safe plant operations are not involved. The activity will not increase the probability or consequences of any accident or malfunction. No license requirements are compromised and no special radiological safety concern is generated. This modification is safe and may be implemented.

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CFR 50.59 Report l

April,1993 - March,1995 l

Pr.gc 9 of $8 Modification: OC-CCD-328333-003, Rey, O SE #328333-007 Rev. O

" Condensate Transfer Sump Suction Expansion Joints" Description of Chg This safety evaluation evaluates the installation of an expansion joint in the suction to each of the Condensate Transfer Pumps (P-l l-001 and P-11-002). The Condensate Transfer Pumps (either one or the other) are rebuilt, on average, once per year. Review of the machinery history indicates that since 1987 these pumps have either been rebuilt, replaced, or had their seals replaced eight times. The probable cause is due to the rigidity of the suction header. The installation of expansion joints will reduce the rigidity of the suction pipe.

Safety. Evaluation Summary; Since the function and operation of the Condensate Transfer system is not changed by the modification, the safety evaluation concluded that the change does not adversely effect nuclear safety, safe plant operations, or the environment. Since change does not result in an unreviewed safety question, this modification can be implemented under 10 CFR 50.59.

Modification: OC-CCD 328333-002, Rev. O SE #328333-006 Rev. O

" Pumps SL-P-3A/3B Motor Replacement" D.tSttiption of Change:

Two holdup tanks are provided for the storage of filter sludge from the radwaste system filters, spent fuel pool filters and reactor water cleanup system filters. Pumps SL-P-3A/3B are used to transfer the filter sludge from the tanks to the cement solidification / dewatering systems. The

_ pumps are driven by 3hp variable speed DC motors designed to control flow rate. These pumps / motors have a history of maintenance problems and failures to start. Because of the location of the pumps (Locked High Radiation Aiea), ALARA requirements are not met due to the high maintenance activity. This modification shall replace the existing motors with 5hp, AC motors. These new motors shall provide a higher torque to aid in starting and operating the pumps. The need to control flow rate is no longer required because the polymer process had been abandoned.

Saffty Evaluation Summary:

The function of the solid radwaste system will not be effected by replacing Pump SL-P-3 A/3B motors with a constant speed AC motor. This will enhance motor and pump reliability and decrease maintenance time. The safety evaluation concluded that the modification does not adversely effect nuclear safety or safe plant operations; does not increase the probability or i

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Attachment I CFR 50.59 Report April,1993 - March,1995 Page 10 of 58 consequences of an accident; does not create a new accident; and does not reduce the margin of safety in the Technical Specifications. Based on these facts, the modification does not pose an unreviewed safety question per 10 CFR 50.59.

Modificatlon: OC-CCD-328333-00l, Rev. O SE #328333-005 Rev. O

" Refuel Bridge lloist Safety Brake Upgrade" Description of Changt This activity modifies the refuel bridge safety brake assemblies by reinforcing the solenoid mount and reducing the shock load experienced when the solenoid is energized. The scope of this modification is confmed to the safety brakes for the main, aux, and monorail hoists on the refuel bridge.

Safety Evaluation Summary:

This modification opgrades the safety brakes for the refuel bridge hoists (main, monorail, and auxiliary.) The upgrade strengthens the solenoid mount as well as reducing the shock load experienced when the solenoid is energized. Other changes reduce the potential for safety brake binding and allow safe viewing of brake actuation. The Safety Evaluation concluded that the function of the safety brakes was unchanged. Based on these facts, this change does not involve an unreviewed safety question as defined in 10 CFR 50.59 and does not adversely effect nuclear safety or safe plant operations.

Modification: OC-MM-328333-004, Rev. O SE #328333-004 Rev. O "ECPMS Inlet Isolation Valve (V-551-0174) Replacement" Descrip_ tion of Changt This modification replaces the existing ECPMS inlet isolation valve which has caused the system to isolate due to its failure during operation. The current condition causes, increased exposure to maintenance personnel, since solenoid repairs / replacements are accomplished while the plant is on line, as well as increased system down time which limits the amount ofinformation being captured by the Chemistry Department with respect to the reactor water chemical condition. The new solenoid will be normally de-energized which will reduce the failure rate and thereby reduce the amount of maintenance that the valve will require. The system availability will also be increased as a result of the reduction in the ntimber of failures.

CFR 50,59 Report Apni,1993 - March,1995 Page 1I of 58 Safety Evaluation Summary 1

This modification will not change the function of the system and as such will not change the system classification. The ECPMS will maintain the capability to be manually isolated via a control switch at the data acquisition panel as well as being automatically isolated on a high RWCU system pressure.

Based on this evaluation, this modification does not involve an Unreviewed Safety Question and does not have an adverse effect on nuclear safety or safe plant operations.

Modification: OC-MM-328333-003, Rev. O SE #328333 003 Rev. O

" Emergency Service Water Orifice RelGcation" DescripJion of Change; This safety evaluation justifies the removal of orifices R021 A through D and replaces them with a single orifice in the common line downstream of V-3-88 in System 1, and V-3-87 in System 2.

The replacement orifice is sized to permit the valve to be opened wider and therefore reduce or eliminate this failure mechanism. The pipe directly downstream of the new orifices (the recovery zone) will be replaced with stainless steel pipe. This is to obviate the need for internal coating in an area where high velocity flow is expected.

Safety Evaluation Summary:

This modification replaces the arrangement of one orifice for each Containment Spray heat exchanger to a single orifice for each pair of heat exchangers. The new orifices are sized so that the system required flow with the required minimum back pressure is attainable without the need to significantly throttle the butterfly valve. The seismic capability of the system is unaffected.

The modification does not adversely effect nuclear safety or the environment. No unreviewed safety question is generated by this modification.

Modification - OC-MM-328333-001, Rev. I SE #328333-001, Rev. I "Rx Fuel Zone Level Temperature Instrument Loop Modification" pescription of Change:

The purpose of this modificatien is to streamline and upgrade the circuits and electronics for the fuel zone temperature monitoring loops. The modified loops are part of the Remote Shutdown Facility and provide reactor water level indication on the remote shutdown panel. This will improve the system maintainability and reliability, and reduce the system out of service time.

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Attachment I CFR 50 59 Report April,1993 - March,1995 Page 12 of 58 Safety Evaluation Summary; The modification does not change the function or design basis of any plant systems, and therefore does not adverrely efTect safe plant operations. It is concluded the proposed modification will not have any adverse effect on safety or the environra:nt, and does not censtitute an unreviewed safety question as specified in 10 CFR 50.59.

Modification - OC-CCD-328312-0 01, Rev.1 SE #328312-003, Rev.1 "GL 89-10 Motor Operated Valve Modification" Description of Change; This modification will improve MOV capability by changing the MOV gear ratios and control scheme for RWCU and IC valves to an "on-demand" type control. This will provide more assurance that the MOV will be capable of performing its safety function while not compromising nuclear safety or valve reliability. The "lC Valve OtTNormal" alarm will be rewired to better indicate that when the Condensate Return MOVs are not in their normal (standby) position. The IC time delay relay tolerance is reduced as a result of replacing the time delay relays with qualified more accurate timers.

Safety Evaluation Summary; The increased RWCU valve stroke times have been factored into an analysis which shows that the core will remain covered, the equipment is qualified for the resulting environment and the otTsite dose is bounded by the MSLB. Similarly, for the increased IC MOV stroke times, with a maximum high flow isolation time delay of 29 seconds, the analysis shows that the core will remain covered, the required equipment is qualified for the resulting environment and the offsite dose is bounded by the MSLB. The minimum time delay has been evaluated to be acceptable.

The IC AC condensate valve will have a reduced stroke to ensure the MOV stroke times bound the analysis. The etTect of this reduced valve open position has been evaluated for system impact and valve performance to be acceptable. This modification results in no new single failure mechanism or new accident, does not effect normal operation of RWCU or IC nor impact safe shutdown of the reactor. Therefore, it is concluded that this modification has no adverse effects on nuclear safety or the environment.

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Attachment I CFR 50.59 Report April,1993 - March,1995 Page 13 of $8 Modification - Cl302-730-5350-005, Rev. O SE #328312 002, Rev. O

" Change Thermal Overload Heaters Required by Calculation" Dnq1iption of Change:

This safety evaluation addresses the changing of thermal overload heaters (TOL's) recouunended by calculation Cl302-730-5350 005, Rev. O, "O/L licater Sizing for NSR OCNGS Generic Letter 89-10 MOVs.

Safety Evaluation Summary:

Changing of the thermal overload heaters for MOVs will not degrade the abilities of the MOVs to perform their safety functions or affect safe plant operations. No modes of operation will be alTected as a result of these changes. TOL sizes were calculated per an approved engineering standard and the results design verified. Therefore, changing of the TOLs does not involve an unreviewed safety question.

Modification - OC-MM-328226-001, Rev. O SE #328226-001, Rev. O

" Minor Modifications Resolving SQUG Outliers" Dncriplion of ChaageJ The purpose of this safety evaluation is to review the modifications which reso'.ve seismic related discrepancies and deficiencies identified during the Seismic Qualification Utilities Group (SQUG) walkdowns. These walkdowns are part of the resolution of Unresolved Safety issue (USI) A-46 as described in Mini Mod Document OC-MM-328226-001. When a component does not meet the requirement of the Generic Implementation Procedure (GIP), this item is classified as an "outlief'.

Equipment may be classified as an outlier due to its own physical attributes as well as due to interaction concerns with adjacent items. Since Oyster Creek has committed to the GIP, each outlier must be addressed and may result in a modification. This safety evaluation addresses modifications installed per OC-MM-328226-001 resulting from the resolutions of outliers. The modifications, although individually designed for each outlier for specific OC equipment, are general enough to be covered by one generic Safety Evaluation.

Safety Evaluation Summarv:

The modifications covered in this Safety Evaluation are limited to those detailed in specific change oocuments or MNCRs posted against and within the scope of OC-MM-328226-001. All work is performed with applicable codes, standards, and materials and in accordance with GPUN

CFR 50.59 Report April,1993 - March.1995 Page 14 of 58 Operational Quality Assurance Plan for NSR and RR systems, subsystems, and components. The SQUG related modifications do not increase the probability of occurrence or consequence of an accident previously evaluated in the FSAR. These modifications do not involve an unreviewed safety question.

Modification SE# 326328-001, Rev. O

" Recirculation Nonle Insulation Replacement" D_cscription of Change-ISI inspections will be performed on the Reactor Vessel Recirculation suction nozzles in 15R. To do the inspections, reflective metallic insulation will be removed. The existing insulation is supported in such a way that to remove the insulation around the Recirculation nozzles require that all the insulation in the area of the vessel skirt below the nozzles must be removed. Due to economic and ALARA concerns it is prudent to replace all the removed insulation with removable fiberglass blankets.

l Safety Evaluation Summary:

This change consists of replacing reflective metallic insulation located on the Recirculation Nozzles and Reactor bottom head area with fiberglass blankets. A transport analysis was performed in accordance with Regulatory Guide 1.82, Rev.1. This analysis concluded the new insulation inside the bioshield would not clog the suction strainer. The new insulation is similar to the existing fiberglass insulation in form, fit and function, and will react equal to or better in terms of the strainer clogging issue. The safety evaluation concluded there is no unreviewed safety question, and the change is acceptable under 10 CFR 50.59.

. Modification SE #323730-001, Rev. O

" Removal of Service Water Valves V-3-709 and C-3-710".

Description of Change:

The purpose of this configuration change is to replace a 20' tee and 2' of piping on Service Water system piping upstream of valves V-3-709 and V-3-710 and to remove valves V-3-709 and V 710. These valves and the tee are located underground on the Northwest side of the Turbine Building at elevation 15'2" BOP. The tee and valves were installed on the service water system in 1986. The welds to this tee are suspected of being corroded similarly to another tee which was inspected and being replaced on the Northeast side of the Reactor Building. That tee was also installed in 1986. Improper internal coating is suspected of being the cause of the weld corrosion.

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Attahment I CfR 50.59 Report April,1993 - March,1995 Page 15 of $8 Safety Evaluation Summarv; This configuration change simply removes two service water valves which are not required and in no way adversely effects the safety function of the system and does not result in unresolved safety concerns.

Modification - OC-MM-343713-001, Rev. 2 SE #323713-001, Rev. 2 "EDG Control Changes and Upgrades" Description of Changer This modification is being implemented to upgrade and increase reliability of the station Emergency Diesel Generators based upon the following observed trends of maintenance data and deviation reports. Some controls were becoming or were already obsolete; some relays and contacts sufTer from premature failure; overuse of terminal points from the original design was creating age induced failures; control design not closely matched to OCNGS's application; most deviations / failures of controls not required for emergency mode; many controls unnecessary and create unwarranted operator confusion. The Emergency Diesel Generators' controls will be restructured / upgraded to focus on and be consistent with OCNGS specific applications.

Safety Evalualjon Summary:

This modification enhances Nuclear Safety and Safe Plant Operations without introducing an unreviewed safety question. This modification is not a test or experiment and does not reduce the margin of safety as defined in the Technical Specifications. This modification does not increase the probability of occurrence or the consequences of any previously evaluated accident or malfunction of equipment previously evaluated.

Modification - OC-CCD-323704-001, Rev. O SE #323704-00l, Rev. O

" Core Spray Parallel Injection Valves V-20-15,21,40 and 41" Description of Change:

Valve Bonnet Pressure Locking has the potentia' to adversely effect Core Spray System Parallel Injection Valves Nos. V-20-15, V-20-21, V-20-40 and V-20-41. The purpose of this safety evaluation is to provide a justification for establishing an internal pressure equahation flowpath from the valve bonnet cavity to the Reactor Pressure Vessel (RPV) side of the valve disc. The scope of this configuration change consists of drilling a small diameter hole (~l/8") in the RPV side of the valve disc.

Attachment i CFR $0.59 Report April.1993 - March,1995 Page 16 of $8 Safety Evaloal101LS11tDIllat%

This configuration change reduces the potet:Ja! for Valve Bonnet Pressurization and does not degrade the performar.se of the Core Spray System. This safety evaluation has determined ti.at this configuration change does not: 1) adversely efTect safety and safe plant operations,2) increase the probability of occurrence or the consequences of an accident or malfu;.ction of equipment important to efety previously evaluated in the SAR,3) create the possibility for an accident or malfunction of a different type than any previously identified in the SAR, or 4) reduce the margin of safety as defined in the basis of any Technical Specification. Furthermore, this safety evaluation has determined that no new unreviewed safety question has been created and that no environmentalimpacts are involved.

Modification - WR #764981, #765275, Rev. O SE #323692 001, hev,1

" Valves V 21-5 and 11 - Replace Motors" Dmniotion of Changg; Motor-operated valves, V 21 5 and 11, are fitted with a 10 fl lb. motor while drawing 2713 Z4 indicates a 15 fl lb. inotor sh~.sid be installed. Thrust and Operator Sizing for V-21-5 and V 11 have determined that a 10 fl lb. motor does not provide suflicient margin and a 15 ft lb. motor is needed. The 10 ft lb. motors w 11 be replaced with 15 ft lb. motort i

Safety EvaluatigILS ammary:

The replacement motors will provide the operators with more thmst margin to operate the MOVs against system pressure. liowever, changing the motor size will not change the function of the -

motor operators in any other way. The replacement of these motors will not effect nuclear safety or safe plant operation because the MOVs' function remains unchanged, the electrical requirements and EDG loading remain unchanged, and the seismic design remains unchanged.

The only change is to return the thrust margin of the motor operators to their original design.

Llodification - OC-MM 323616, Rev. O SE #323616-003, Rev. O "MOV Limit Switch Modification (Phase 4)"

Description of change; The purpose of this modification is to improve the position indication of various plant MOV's.

This is to implement phase 4 of the response to SOER 86-02. There are 47 BOP type MOVs identified and 40 of these will be modified by this modification.

l

Attachment I CFR $0.59 Report April,1993 - March,1995 Page 17 of $8 l

l l

Safety Evaluaticalummary:

1 This modification will relocate the "OpEN" limit switch for various plant valves. This will l

provide more accurate valve position indication. This will not change component or system i

function nor will it change system operating characteristics. The modification does not involve an unreviewed safety question, Techniul Specification change or have an adverse impact on nuclear safety or safe plant :perations.

Modification SE #315403-032, Rev. 0

" Justification for IC AC Condensate Valves Revised Open Area" Rcicliplign of Change:

This safety evaluation justifies reducing the normal open position of the isolation Condenser (IC)

AC condensate valves (V 14 36,37) from 100% open to a minimt.m of 70% open from a system performance perspective only.

Safety Evaluation Summary This safety evaluation concludes that there will be no impact on JC performance resulting from a change in the position of the AC condensate valves from 100% open to 70% open. The actual IC heat removal capacity is higher than design and more than adequate to remove decay heat and cool down the RCS, Since there is no unreviewed safety question, the modification meets the requirements of 10 CFR 50.59.

Modification SE #315403-031. Rey,0

" Justification for Revised IC Hi Flow Trip Time Delay" Dncrintion of Change This safety evaluation jt:stifies a decreased timer delay for the isolation Condenser high flow isolation signal. The current setpoint for this TD is 32 7 seconds. A reduction to 27i seconds is proposed with a lower process limit of 25 seconds.

Safety Evaluation Summary; it is reasonable to expect that an IC high flow isolation time delay of 25 seconds will prevent an isolation of either IC under normal initiation conditions assuming a recirculation pump t:ip occurs concurrently with the initiation (automatically or manually in IC loops) and the condensate return A

-~

Attachment I CFR $0.$9 Report April,1993 - March,1995 Page 18 of $b valve open stroke times are as indicated herein. The setpoint should be 27 seconds based on the 25-second lower process limit and a 29 second upper process limit. The safety evaluation concluded that the proposed setpoint change will not have any adverse effect on nuclear safety, safe plant operations or the emironment. This modification does not create an unreviewed safety question as described in 10 CFR 50.59.

Modification - OC-CCD 315302 001, E iv. O SE #315302 073, Rev. I

" Degraded.U/V Relay Setpoint Re-evaluation" PEcliplion of Change; Several class lE MCC's were evaluated from the point of view of voltages well below the minimum pickup voltage of their starters during a worst case LOCA and degraded grid conditions. As a result, an evaluation of the MCC branch circuits was performed which concluded that their starters will experience pickup, dropout and overvoltage outside of the design limits. Their electrical protective devices (relay, fuse, thermal overload heater) may be sensing higher current than previously evaluated. The recommended changes will be implemented which will then ensure circuit protection ud operation within the design limits in various plant scenarios for their intended safety functions.

Sakty Evaluation Summary:

These changes support TSCR No. 219 (submitted for NRC review and approval) which increases the degraded grid voltage setpoints frcm 3671V to 3840V. The modification w' ill ensure that.he NSR components have adequate voltages within their design limits and improve the ability of Engineered Safeguards components to provide their Safety function No unreviewed Safety concern or emironmental concern is involved.

Modification - OC-CC23194, Rev. O SE #000822-017, Rev. O I

Adding Instrument Line Snubbers DnCIjplion of Changr The Reactor Building ventilation system overall airflow is 20% lower than design. This is a direct result of how the building Differential Pressure (DP) is maintained. The buildings' negative pressure is maintained manually. The supply and exhaust fans are nm at full speed and the buildings' negative DifTerential Pressure is increased by closing off the dampers on the supply side.

There is a Reactor Building low DP alarm which receives its signal from four (4) differential pressure indicator transmitter, one mounted on each wall of the Reactor Building, on elevation 119', Building DP is affected drastically by sudden wind changes which cause a spike in the pressure transmitter and thereby, indicating an increase iri building pressure and bringing in the 1

Attachment l' l

CFR $0.59 Report April,1993 - March,1995 Page 19 of 58 I

low DP alarm. To avoid this nuisance alarm, the buildings' negative DP is raised so that fluctuations caused by the wind will not set off the alarm. To increase the negative DP would require closing off more supply air and thereby reducing overall building airflow. This.

configuration change will install, on the static sensing line, to the outside, a snubber (muffler, silencer) which will attenuate the transient affects from the wind. Thereby, preventing the pressure spikes. The building DP can then be decreased and ovarall airflow willincrease. The performance of the system is enhanced as described above.

Safety Evaluation Summary:

The addition of the snubber to the Differential Pressure sensing system does not present an unreviewed safety question or environmental concern because the systen is not safety related, serves no safety function and no other safety related and non safety related systems are affected.

Modification JO #30143/WR#758846, Rev. O SE #000822-006, Rev. 2 Upgrade Valves V-27-1,2,3, and 4 pescription of Changt The modification te V-271,2,3, and V-27-4 upgrades the material of the shaft key and the disc taper pins, and increases the size of the disc taper pins. An analysis has indicated that the valves, as originally designed, are capable of performing heir safety function of closing against a 38.5 psig drywell pressure, but they are not capable of opening and closing at 55 psig for beyond design bais events. The modification will enable the valves to be functional for beyond design basis events.

' Safety Evaluation Summarv:

The purpose of this modification is to upgrade two (2) valve parts so that V-271, 2,-3 and V-27 4 can open and close at 55 psig for beyond design basis events. The valves are presently capable of performing their design basis functions but must be upgraded for beyond design basis events. An analysis indicates that the modification will enable the valves to function for beyond design basis events. The safety evaluation determined that no unreviewed safety question exists for this modification.

e

Attachment i CFR $0.$9 Repon April,1993 - March,1995 i

Page 20 of $8 Modification -

SE #000622-015. Rev. O "Setpoint Change-Annunciator H 3-Ft Reactor Pressure High" Descriplion of Change:

During the analysis of OC-lOSRG Report 92-10, " Assessment of the Control Rod Drive Diagnostic Test Pressure Transient, it was noted that lowering the Reactor Pressure High setpoint could improve operator response. This safety evaluation is written to change the alarm setpoint to 10 psig above the normal control band.

Safety Evaluation Summary:

The or.!y effect of this setpoint change should be an improved capability to respond to a reactor over pressure transient. Lowering the reactor pressure high alarm setpoint will not effect any safety function nor will there be any adverse effect on safe plant operations, Therefore, there is no adverse impact on nuclear safety and no unreviewed safety question exists.

Modification - #551-94, Rev.1 SE# 000561-004, Rev. I

" Installation of PVC Liner in Tank T-33-001" Dncriptio1LoLChange This safety evaluation has been completed for the installation of a 60 mil PVC liner into Sodium Hypochlorite tank T43 001. This modification will structurally repair the leaking 3" pump suction nozzle, install PVC liners through the 3" overfiow nozzle, pump suction riozzle and drain nozzle and the PVC liner in tank T 33-0)l. It has been shown that PVC is inert to 1215%

sodium hypoch'orite and the PVC liner should last to end of plant life. The PVC bonding will be given a four day cure time before being exposed to sodium hypochlorite which will provide a 97%

or greater cure time which is acceptable to polled PVC manufacturers.

Safetyfvaluation Summary:

This modification does not change the function of any plant systems or violate any Technical Specifications. Normal plai; speration will not be changed and equipment necessary for safe shutdown will not be adversely affected. Based on thejustification described herein, it is concluded that the proposed modification will not have any adverse effect on nuclear safety, safe plant operation or the environment. This modification does not create an unreviewed safety question as described in 10 CFR 50.59.

l CFR 50.$9 Report l

April,1993 - hlarch,1995 Page 2I of $8 Test Special Procedures93-007,008, Rev. O SE #000231 006, Rev. 2 l

"lest of Placing AOG into Service with IIWC in Service" Dngdpfon of Changt At present, the Augmented OfTgas (AOG) system operating procedure (OCNGS Procedure 350.1) requires that the Feedwater liydrogen injection system be isolated for two (2) hours prior to plachig the AOG into service with offgas.

The purpose of this test is to demonstrate that placing the AOG system into senice under Ilydrogen Water Chemistry (HWC) ofTgas co,nditions can be done as safely.

Sa&tyfrahtation Sunimaty; This test has been evaluated and determined not to present an adverse impact on nuclear safety or ufe plant operations. This test has also been evaluated and determined not to represent any unreviewed safety question as dermed in 10 CFR 50.50. This test will eliminate haQg to secure the liWC System for two hours when the AOG System trips, prior to placing it in service. From a safety standpoint, the worst result from this test would be a buildup of hydrogen (hydrogen enriched environment in the piping). This would cause an AOG trip, which is proceduralized.

Therefore, this test does not involve any safety concerns.

Modification - OC-CCD 000225-001, Rev. O SE #000225-010, Rev 0

" Replacement of Control Rod Drive Pump Stop Check Valve" Duttiption of Changt This modification will replace the existing stop-check valve with a separate check valve and a globe valve. The existing spring can support will be reset to account for the new load.

Safety Evaluation Summary:

The activity has no effect on the margin of safety as defined in the Safety Analysis Report.

Nuclear safety and safe plant o.perations are not involved. The activity will not increase the probability of consequences of any accident or malfunction. No license requirements are compromised and no special radiological safety concern is generated. This modification is safe and may be implemented.

I

Attachment I CFR 50.59 Report April,1993 - March,1995 Page 21 of $8 Set Point Change SE #000212-026, Rev. O

" Core Spray Relief Valve Set Point Change"

- Ducijption of Change; The Core Spray System (CSS) provides cooling water to the reactor core in the form oflow pressure spray to remove decay heat and prevent fuel clad degradation following a Loss of.

Coolant Accident (LOCA). During such an accident, CSS starts injecting cooling water aner the reactor vessel pressure is reduced to approximately 285 psig. During normal plant operation, the low pressure portion of the system is isolated by two normally closed valves (i.e. testable check valves and parallel injection valves) from the high pressure portion attached to the reactor vessel.

A relief valve is provided in the system for overpressure protection for the extremely low l

probability event of significant back leakage through the system isolation valves. The relief valve (s) had a history oflining during operability and inservice testing of the Core Spray pumps.

This is believed to be the result of drining of the valve set point (i.e.,350 psig) close to the shutoff head of the pumps (approximately 320 psig). In order to avoid valve lining, it was recommended that the set point be increased to 400 psig.

Safety Evaluation Summary; it has been concluded that the availability of the Core Spray System will remain unchanged and the safety function of the system will not be effected by the proposed set point change. There is no adverse effect on nuclear safety or safe plant operation, and no unreviewed safety questions exist.

Modification - OC-MM 402990 012, Rev. O SE #402990 015, Rev,0

" Lifting Devices in Torus Vacuum Breaker Areas" Qcieription of Change; The purpose of this modification is to design and install manually operated haing devices for each of the tcrus-to-drywell vacuum breaker valve assemblies. The installation of these lining devices will case the manual efforts involved in the removal of valve covers and internals for maintenance during each outage. This will also reduce the probability ofload drop since the valve items are presently removed manually. The maximum load these miscellaneous cranes will lin is 250 lbs.

and hence are excluded from conformance to NUREG 0612.

1 nr. -

CFR $0.59 Report April,1993 - Msch,1995 Page 23 of $8 SAfCty EvaluatiRILSummaQC The purpose of this modification is to design and install lifling devices in the Torus Vacuum 13eaker platform areas to facilitate maintenance of the vacuum breaker valves. The structural design of these installations comply with all safety functions. The proposed modification will not i

have any adverse effect on nuclear safety oc the environment.

Procedure Change SE #000661-008, Rev. 2

" RN 12A/B Setpoint Change" bittiption of Changc The air ejector off gas radiation monitoring sub system monitors exhaust from the steamjet air ejectors for gross radioactivity before release to the environment via the plant stack. The sub-system alens the operator when activity levels exceed alann setpoints, and automatically isolates the air ejector off gas system before high levels of radioactivity are released to the atmosphere.

Safety Evaluation Summary:

These revised setpoints enhance protection of public health and safety and in no way impact nuclear safety or safe plant operations. An unreviewed safety question is not involved.

Modification - M DD-OC-622B, Rev. O SE #000622 014 Rev. I Reactor Fuel Zone Level Indication System Modification" DnqIiption of Changt The Fuel Zone Level Indication System (FZLIS) provides compensated reactor vessel water level indications. The FZLIS is strictly a monitoring system and does not have any control functions.

The system monitors level from 144 inches below the Top of Active Fuel (TAF) to 180 inches above the TAF. This modification changes the programmable controller logic to use the wide range level (L2) transmitter input for level computation in all conditions.

Safety Evaluation Summary:

This modification will eliminate a source of a non-conservative bias in FZLIS ir.dication. The FZLIS is a monitoring system only, has no interface with plant control functions, and as such the change has no impact on any accident or equipment malfunction. For these sg rons, this modification does not adversely impact nuclear safety or safe plant operatio;s, involves no environmental impact and involves no Unreviewed Safety Question.

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CFR 50.59 Repon April,1993 - March,1995 Page 24 of $8 i

Procedure SE #000567 003, Rev. 2

" Evaluation of Raising the Max 112 Injection Flow Rate" Description of Change:

The purpose of this change is to determine an absolute maximum injection rate and to operate the system with the high flow trip point at any value in the range of the flow control loop, as determined by Operations, as long as MSLR levels do not exceed 480 MR/IIR and AOG throughput does not exceed 40 SCFH.

Safety Evaluation Summarv:

This change has been evaluated and determined not to represent an unreviewed safety question as defmed in 10 CFR 50.59. This change will allow a maximum hydrogen injection rate set anywhere in the range of the flow control loop, as determined by Plant Operations, as long as the restrictions of 1) 480 mr/hr max on highest main steam line rad monitor and 2) maintaining len than 40 SCFM throughout the AOG system are not exceeded. This change will allow the setting of the high hydrogen trip (FS-567-0026 and FS-567 0027) to any value within the range of the instrument as determined by Plant Operations.

Modification - OC-CCD-409757-001 Rev. O SE #409758-001, Rev. I

" Replacement of MOB South End HVAC System" Descriptiop of Changt 4

The purpose of this configuration change is to install a new replacement south end HVAC system, isolate the demineralized water transfer system and turbine building closed cooling water system from the chilled water system and abandon in-place (to the maximum extent possible) the chilled water system and the balance of the existing south end HVAC system. The scope of this configuration change consists ofinstalling a new commercial grade,30 ton rated cooling capacity, single zone, rooflop mounted package air conditioning unit with variable air volume (VAV) controls. The demineralized water transfer system supply line will be isolated from the chilled water system. Similarly, the turbine building closed cooling water system supply and return lines will be isolated from the chilled water system. The chilled water system supply and return lines to the existing south end HVAC system will be drained and abardoned.

Attachment I CFR 50.59 Report April,1993 - March,1995 Page 25 of $8 SA[CE.EYaluation Summam This safety evaluation has determined that this configuration changes does not: 1) adversely effect nuclear safety and safe plant operations,2) increase the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR,3) create the possibility for an acciaent or malfunction of a different type than any i

previously identified in the SAR,4) decrease the margin of safety as defined in the basis of any Technical Specification. Furthermore, this safety evaluation had determined that no new unreviewed nuclear safety questions have been created and that no environmentalimpacts are i

l involved. This configuration change does not violate any licensing requirements, or cause any radiological safety concerns and will not affect the plants permit condition.

Modification - OC-CCD-403037-002, Rev. O SE #403037-003, Rev. O

" Shroud IIcad Holt Reduction" WiqIiption of Changc.;

This safety evaluation evaluates the acceptability of plant operation with twelve (12) equally spaced and latched shroud head bolts (SHB) and twenty four (24) removed shroud head bolts for one fuel cycle of operation. The twelve (12) SHD to be installed during the ISR outage for fuel cycle 15 service are the new type (lGSCC resistant) bolts that were designed and constructed for the new overhead core spray sparger (not installed) and modified for use with the original core spray sparger. The 1-1/2" thick spacer that was installed on the new SHB's requires removal for the purpose of accommodating a shroud stabilizer system to be installed during the ISR outage.

The shroud stabilizer system provides features to permit installation of the overhead core spray sparger if needed in the future. As a result, the shroud head will sit 1-1/2" higher than originally designed. This condition was analyzed in the design for the overhead core spray sparger modification. A review for steam separator / steam dryer interferences shows no interference problems.

Safety Evahiation Summary:

An evaluation of the impact on nuclear safety based on plant operation with a reduced number of SHB's (12 vs 36) has been performed. It has been demonstrated that Oyster Creek can be safely operated with only twelve (12) equally spaced and latched HB's. The conclusions are still valid since such reports have not changed since origina!iy issued. This safety evaluation is only valid for one cycle of operation This safety evaluation has determined that this modification does not:

3

1) adversely effect nuclear safety and safe plant operations,2) increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the SAR, ?) create the possibility for an accident or malfunction of a difTerent type than any previously identified in the SAR,4) decrease the margin of safety as

i a

CFR $0.59 Report April,1993 - Ma:ch,1995 Page 26 of $8 defined in the basis of any Technical Specification. Since this safety evaluation has determined that no unreviewed nuc' ear safety question has been created and that no environmenN impact is involved, this modification is acceptable.

1 Modification OC PLT Proc.105.2 SE #000214-005, Rev. O

" Shutdown Cooling Room Sca%Iding" Description of Change:

This safety evaluation documents instification for scaffolding in the Shutdown Cooling Room remaining in place (including fibetglass grating) during the operating cycle. This would allow easier and quicker access for future maintenance on this equipment. The scaffolding which will be left in place will become a permanent plant configuration.

Safety Evaluation Summaty; The scaffolding has been erected ir. such a fashion that no parts ofit will fail under normal dead loading as well as transient Icading including SSE. The grating above the Shutdown Coo:ing System is restrained by bracing made of scaffold poles from sliding off and impacting on the equipment below. This safety evaluation concluded that the scaffolding willin no way effect nuclear safety and safe plant operation nor will it efTect plant reliability and availability. Therefore, the plant's ability to safely shutdown remaim unaffected.

Stodification - 51DD-OC-6648, Div.1, Rev. 2 SE #403005-001, Rev. 2 Regulatory Guide 1.97 Instrument Enhancement Ductiplion of Change:

The detailed control room design review conducted at OCNGS to mi.;t NUREG 0737 Item 1.D.1 requirements, had identified human factor deficiencies pertaining to the design of reactor fuel zone level recorder on control room panel SF/6F. This modification will address these deficiencit.s. In addition, this modification will replace the thermocouple-to-current (TC/I) converters in fuel zone level instrumentation system (FZLIS) to correct instrument drift problems of the existing units.

Safety Evaluation Summarv:

This modification does not adversely efTect safe plant operations because it does not change the function of any plant systems nor violate any Technical Specifications. The subject modification does not have any adverse effect on nuclear safety, safe plant operations, or the environment.

This modification does not constitute an unreviewed safety question as determined by 10 CFR 50.59.

CFR 50.59 Repon April,1993 - March,1995 Page 27 of 58 Modification - OC-CCD-40103G2, Rev. O SE #403037-001, Rev. O

" Reactor Vessel Shroud Repair" Descriptio, of Change:

- This modification permanently repairs cracking of any and all circumferential welds in the Oyster Cret : core shroud.

Safety Eyahlation Summaty; The addition of tie rod assemblies to the Oyster Creek reactor vessel shroud does not significantly increase the probability of occurrence or consequences of an accident or equipment malfunction previously evaluated in the UFSAR. Further, this change will not effect plant margins of safety.

No new potential radiological release paths are created. Accordingly, it is concluded that the installation of the tie rod / radial support assembles to permanently repair cracked circumferential welds in the Oyster Creek reactor vessel shroud does not constitute an unreviewed safety question.

Modification - OC-MM-403-011 001, Rev. O SE #403011-001, Rev. 4

" Core Spray Pumps Recirculation Line Upgrade" Descriotion of Change:

This modification relocates the Core Spray pump minimum flow recirculation lines to a higher elevation such that an overflow path is created for the fill pumps while maintaining the minimum flow capability for the main and booster pumps when they operate. The existing recirculation piping and valves V-20-92,93,94, & 95 shall be removed with this modification. The valve interfaces both electrical and mechanical will be altered, deleted or capped to facilitate the new installation.

Safety Evaluation Summaty; Alteration of the fill pump alarm does not reduce the readiness of the core spray system since monitoring activities by Plant Operations assures that the fill pumps operate satisfactorily to keep the piping full ofliquid The installation of a manual vacuum breaker on System 11 enhances the operator's ability to recognize a" ' detect fill pump failure by assuring that the upper portions of piping are relieved of vacuum conditions. The vacuum breaker and associated consequences of its installation have been evaluated and determined not to have a negative imp'act on Core Spray System 11 or other safety systems. This evaluttion demonstrates that an Unreviewed Safety Question does not exist. The modification has no adverse impact on nuclear safety or safe plant operations.

Attachment I 1

CFR 50.59 Report April,1993 - March,1995 Page 28 of $8 4

Modification - OC-MM-402990 010, Rey,1 SE #402990-009, Rev.1 "Recirc Pump Suction Temperature Instrumentation" Descriplion of Changq; This modification streamlines and upgrades the circuits and electronics for the recirculation pump suction temperature monitorirg loops. This willinclude the installation of connectors between the Ronan modules and the thermocouples for improved maintainability. This modification will also affect the loop for the feedwater header thermocouple TE-47. Transient voltage surge suppressors will be added across shutdown cooling valve permissive relays.

Safety Evaluation S_nmmary:

System performance will not be degraded, since the modification will not alter the functions of existing temperature monitoring and valve control circuits. The modification will not increase the probability of occurrence or consequences of an accident since it does not effect the performance of any safety related systems. The proposed modification will not have any adverse effect on safety or the environment and does not constitute an unreviewed safety question as specified in 10 CFR 50.59.

Modification - OC-MM-402990-007, Rev. 0 SE #402990 010, Rev 0

" Retirement of Main Steam Line Radiation Monitor Scram & Isolation Function" Description of Change:

This modification removes the main steam line high radiation signal input into the reactor scram and reactor isolation logic based upon TSCR NO. 214 and NEDO-31400A safety evaluation for eliminating the boiling water reactor main steam line iso!ation valve closure function and scram function of the main steam line radiation monitor. The modification serves to: reduce the chance ofinadvertent reactor scrams, isolations and challenges to safety systems; improve the availability of the main condenser for decay heat removal; not significantly increase the potential offsite dose from any cesign basis accident; lessen calibration / surveillance requirements and eliminate risks associated with half scrams and isolations during these surveillances.

Safety Evaluation Summary:

This modification achieves the following without significantly increasing the potential offsite dose from any design basis accident nor adversely effecting nuclear safety or safe plant operations:

reduces the chance ofinadvertent reactor scram /isolations and challenges to safety systems; improves the availability of the main condenser for decay heat removal; does not significantly

CIR 50.59 Report April,1993 - March,1995 Page 29 of $8 increase the potential ofTsite dose from any design b. isis accident: and lessens calibration / surveillance requirements and eliminates risks associated with half scrams and isolatians during these surveillances.

Modification SE #000211-013, Rev.1 "in-Line Snubbers for 1805/11 Instruments" Dsfgtiption of Changt i

This modification to the Emergency (Isolation) Condenser Pipe Break Sensor in line snubbers defeats the snubber function by removing the sintered disc that performs the dampening function.

This eliminates the delay in the instmment response time caused by snubbers in this application.

Safety Evaluation SummaIyj The removal of the snubber internals does not affect the tensors ability to initiate the isolation of an emergency condenser upon detecting r. high flow rate in the emergency condenser piping.

Without its ir.ternals, the snubber still maintains its pressure boundary. The only impact of the modification, which is increased noise in the process lines reaching the sensors, has been shown to be negligible. There is no unreviewed safety question, and no environmental impact is involved.

Modification - OC-M M-402990-011, Rev. O SE #402990 014 Rev. O

" Replacement of UV Protection Status Lights for 1C & ID Buses" Ducription of Change The modification shall replace the undervoltage protection system status lights with neon lights.

Characteristically, neon lights have long bulb life, low heat dissipation, and low power requirements. Fusing of these lights will isolate the lights from the UV protection circuit in the event of an electrical shon of the light. Resistors will be added to drop bus voltage to an ideal operating range across the lights.

Safety Evalt:ation Summary:

The modification replaces the incandescent UV protection status lights for the 4160V IC and ID buses with neon lights. Fusing and resistors will be added in the lighting circuit to protect the main UV protection circuit from electrical shorts of the new lights. There is no detrimental effect on safety or the environtnent and the modification does not pose an unreviewed safety question per 10 CFR 50.59.

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l Atuchment I CFR 50.59 Reprt April,1993.- March,1995 Page 30 of 58 Modification i

SE #403014-002, Rev. O

" Cycle 15 LPRM Reasr*gnment" D_cistiption of Change:

j This safety evaluation analyzes operation without Local Power Range Monitor (LPRM) 20-49 l

j and 36-41 and the reassignment of all detectors from LPRMs 20-41 to APRMs 1,2,5, and 6.

These APRMS were previously fed signals from LPRMs 20-49 and 36-41. It also addresses other I

issues associated with the LPRM reassignment: the effect on the Plant Computer Systems (PCS),

the reassignment of LPRM output to the control room indicators and the necessary changes to the Traversing incore Probe (TIP) system.

Safety Evaluation Summary; The reassignment of LPRM 20-41 levels A and C to APRM 2, LPRM 20-41 levels B and D to APRM 6, LPRM 44-41 levels A and C to APRM 1, LPRM 44-41 levels B and D to APRM 5, will not result in decreasing Critical rtaver Ratio (CPR) and the mechanical and thermal overpower margins because: LPRM 20-41 is in close proximity and in the same quadrant as the replaced LPRM 20-49; LPRM 44-41 is in close proximity of and in the same quadrant as the replaced LPRM 36-41; APRMs I and 2, and 5 and 6, LPRM detector level assignments will not be altered.

1 The Cycle 15 reload withdrawal error analysis bounds the reassignment of the LPRMs. The results showed there was no reduction in CPR margin since the maximum delta CPR did not require a rod block Therefore, since the rod block was not required, *.he LPRM assignments are not critical to the reload analysis. The original rod withdrawal error analysis was confirmed by an additional analysis using the best estimate control rod patterns. The delta CPR for this analysis was less than the original analysis. Therefore, there is no reduction in the CPR margiu, in addition, the analysis confirmed the mechanical and thermal overpower limits were not exceeded.

The replacement of LPRM 20-49 and 36-41 with 20-41 and 44-41 in APRMs 1,2,5, and 6 will not adversely impact the function of the neutron monitoring em ieactor trip, or rod block systems. For these reasons, nuclear safety is maintained and nu oceviewed safety questions exist.

Modification - OC-CCD-402991-004, Rev. 0 SE #402991-007. Rev. I

" Emergency Service Water Pipe Inspection" Description of Change:

This modification will add a 2" Stainless Steel (SS) tee nipples and 2" SS ball valve a', the 10" Emergency Sersice Water (ESW) elbow at the intersection of the 2" keep fill line. This

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Attachinent i CFR 50.59 Report Apn!,1993 - March,1995 l

Page 31 of 58

{

configuration change will allow the inspection of the 10" ESW pipe for marine growth in the non-chlorinated portion of the 10" discharge pipe from the ESW pump.

Safety Evaluation Summary:

This configuration change will allow for inspection of marine growth at the 10" ESW pipe from the outlet of the ESW pipe. This change will not adversely effect the function of the Service Water, Emergency Service Water System or Containment Spray Systems, since it is not required to perform any safety function, nor willit be used during system actuathn. There are no unreviewed safety questione Modification SE #000621-008, Rev. O "APRM Scram Clamp Setpoint Change" Ducription of Change; This Safety Evaluation proposes a reduction of the APRM scram clamp setpoint which is checked during the weekly surveillance by Plant Procedure 620.3.003. The purpose of this: reduction is to allow widening the acceptance band of the weekly calibration. Lowering the setpoint will also aid in readabihty of the meters during the surveillance.

Safety Evaluation Summary; Lowering the A?RM Scram Clamp ietpoint from 114.7% to 114.0% of rated reactor power will keep the APRM scram function within the limits imposed by Technical Specification 2.3. The APRM scram will occur slightly sooner during a large power excursion above 100% rated power.

There will be no impact on plant operations other than to improve the weekly APRM Surveillance Test, For these reasons, there is no adverse impact on nuclear safety or safe plant operations. No unreviewed safety question is raised by this setpoint change.

Modification - MDD-OC 622 D, Rev. O SE #402996-001 Hev. O

" Replacement of Recirculation Flow Monitoring Electronics" Description of Change; This modification replaces the existing recirc flow monitoring electronics. This replacement includes the 10 flow transmitters (2 for each of the 5 recirc loops) and electronics on control room panels 3R and SR. Control room equipment being replaced by this modification includes the transmitter power supplies (2), spare root converters (10), summers (4) and the Average Power Range Monitor (APRM) flow units (2 flow converters, 2 power suppliers).

Attachment I CFR $0.59 Report April,1993 - March,1995 Page 32 of 58 SafetyAaluation Summary:

This modification replaces the existing recirc How monitoring system with new electronics. This i

will result in increased signal accuracy and reduced setpoint drift. Enhancements provide further separation between Division 1 and Division 2 flow systems. This modification does not involve an unreviewed safety question as dermed by 10 CFR 50.59, and does not adversely effect nuclear safety or safe plant operations.

Procedure Revision - PSC-91-007 Rev. 0 SE #315403-025, Rev. 0 Emergency Operating Procedur.: Change DnqIiption of_Chaest A concern exists with using core spray to provide makeup for a range of small breaks (0.05 to 0.4i ft.' )in the primary coolant system. The concern is associated with excessive cycling of the parallel injection valves in an effort to control Reactor Pressure Vessel water level as described in the EOPs. In order to reduce the required cycling frequency of these valves, it is recommended that an expanded operating band (100" TAF to 175" TAF) be established for the core spray system.

Safely fJa!uatt0n Summary:

The expanded level control band will ensure that the core spray parallel injection valves will not be operated too frequently. It provides a reasonable valve motor operator rest time while reducing the operator burden when controlling RPV water level with the core spray system. The water level control band was selected to ensure that neither the Lo-Lo nor the Lo-Lo-Lo level limits are exceeded. Therefore, it is concluded that the implementation of this procedure change will not adversely effect nuclear safety or safe plant operations.

Modification - JCP& L #21079004182-01, Rev. O SE #402965-001, Rev.1 l

Silo - Installation m Electrical Duct Bank and Transformer Dmriplion_oLCha_nst Two existing combustion turbines (CT) wil: oe connected to the emergency busses to provide an additional source of off-site power in case of a station blackout (SBO). This modification is regulatory required but the components are commercial grade and the installation is classified as "other"

Attachment I CFR 50.59 Report l

April,1993 - March,1995 Page 33 of Sit Safsty Evaluation Summary; The safety evaluation concluded that the construction phase ofin ig an underground electrical cables and duct bank system from the Forked River CT'.i to the nev. n site SB0 transformer, including the placement of the transformer does not adversely impact plant nuclear safety. The construction phase of this modification does not introduce an unreviewed safety question, will not impact any critical plar.t structures or prevent the operation of NSR or RR systems.

Corrective Change 345-92, Rev. O SE #000621006, Rev. O "IRM 51eter Face Change (127%)"

Rsicription of Changru Corrective Change 345-92 describes the addition of a tick mark and self adhesive label to the meter faces of the Intermediate Range hionitor (IRht) drawers. The meters are on the Control Room rear panels 3R and SR. The tick mark and level will indicate 127% and will enable widening the acceptance band of the weekly surveillance procedure to i 2%. The current acceptance criteria is +0%, -2%. because there are no gradations above 125%.- This results in needless deviation reports and out-of-service time. The meter face marking will be perfonned during the performance of the refueling bench calibration.

Safety Evaluation Summary:

This meter face change will allow an improvement of the IRhi system. Placing tick marks at 127% will allow an improvement of the IRhi system as well as allowing widening a calibration tolerance and thus eliminating needless recalibrations and out of service time when the IRhi front panel test. This meter face change does not constitute an unreviewed safety question. There is no environmental impact or change required to the Technical Specifications.

Stodification - OC-5151-402950-009 Rev. 2 SE # 402950-009. Rev. 2

" Set up for Boring of Alanway lloles in Shield" Description of Change:

hiodifications include installing the equipment and senices needed to bore manway holes in the drywell shield wall near the base of the drywell. These holes are to be used to allow workers to locally access the sand bed area to complete sand removal, as well as to clean and coat the drywell exterior (in the area of the sandbed). The modification includes installing permanent scaffolding between the torus and the shield wall to support work activities.

Attachn nt I CFR $6,$9 Repon April,1993 - htarch,1993 Page 34 of $8 Safety Evalugl' n. Summary:

s An analysis of Safety Systems and functions affected by equipment installation has not revealed any new safety concerns introduced by this work. Specifically, the plant's ability to mitigate the effects of all Design Basis Events (DBE's) is not changed during or subsequent to equipment installation. The plant's position on fire protection and Appendix R are not adversely effected by equipment installation.

l As a result, this modification does not involve an umeviewed safety question or environmental impact.

i Modification - OC-MM 403004 00l, Rev. O SE #403004 001, Rev 1

" Chemical Waste Tank (WC-T-l H) Mixer" Dnniplion of Changt b

The purpose of this modification is to install a mixer in the Chemical Waste Tank (WC-T 18) to achieve mixing of solids prior to processing the contents of the tank. The chemical waste tanks accumulate sludge from various drains due to inadequate mixing.

Safety Evaluation Summary:

The purpose of this modification is to provide an alternate means of mixing the contents of the Chemical Waste Tank (WC-T-18). Since the modification does not change the system function in any way, there is no effect on nuclear safety or safe plant operations. There are no unreviewed safety questions.

Modification - OC-MM 402987-001, Rev. O SE #402987 001, Rev.1

" Condenser /lleater Bay Walkover Platforms" Description of Changt The purpose of this modification is to design and install walkover platforms, stainless steel rungs, and access ledges in the Condenser Bay Heater Bay, and Condenser Water Box areas of the Turbir e Building.

j

Attachment I CFR 50.59 Report April,1993 - March,1995 Page 35 of $8 Safety Evalaticalummux These installations are used to facilitate passage of equipment and personnel in frequently traversed areas and to aid maintenance activities during plant cutages. The structural design of __

these installations will comply with a!! safety functions. The proposed modification will not have any adverse effect on nuclear safety or the environment. There are no unresiewed safety questions.

Modification - OC MM-402986-012, Rev. O SE #402986-013. Rev. I "Flimination of Potential llot Shorts in MOV's" Dnstipti'on of Change; The purpose of this modification is to address an oversight in the original design o.'the Appendix R analysis with respect to the control circuit of motor operated valves (MOV). This modification shall rewire elTected MOV (V-14 32, 33, 37, V-17-19, -54, V-37-54) torque and limit switch I

contacts immediately below their contactors in their logic ladder to eliminate possible valve failure due to a hot short occurring during a control room fire.

Safety Evalation Summary; This modification requires electrically moving the torque and limit switch contacts to a position immediately below tlie valve operators (contactors)in the valve logic ladder, in the new configuration, the valves torque and limit switches will retain its normal capability ofisolating power to the valve at completion of a valve stroke; even during a postulated hot short event in the control room. Because the margin of safety is actually increased, performance of this modification will not adversely effect plant safety or operations nor violate the Technical Specifications or any other licensing agreement. No unreviewed safety question exists as a result of this modification.

Modification - OC-MM 402986-011, Rev. O SE #402986-012, Rev. 0 "LLRWSF Fire Protection / Detection Changes" Dncription of Changes This modification will simplify and increase the reliability of the fire protection system in Low Level Radwaste Storage Facility (LLRWSF). The modification will allow the LLRWSF fire detection and fire suppression systems to operate independently. This will prevent spurious actuation of the fire suppression system which will prevent dilution pump trips.

CFR 50.59 Report April,1993 - March,1995 Page 36 of 58 Sdl;

.duation Summary; This modi ation to provide equipment and rewnfigure the LLRWSF Fire Protection /S6ppression does not d-crease the margin of safety as dermed in the SAR or in the Technical Specifications, does not increase the probability of occurrence or consequence of an accident or malfunction of equipment important tot safety, does not involve any radiological or environmentalimpact.

Therefore this modification dou not create any unreviewed safety questions as determined by 10 CFR 50.59, and does not involve a potential environmentalimpact.

Modification - OC-MM-323653-008, Rev. 0 SE #323653 008, Rev. O

" Chem Waste Filter Bypass Modification" DnqIiption of Change:

This modif; cation will install flexible hose to provide a flow path directly from the filter sludge tank (T-22-0010) to the batch tank (SL-T-008) and permit bypass of the Chem Waste Filters.

The tie in will cut 1 1/2" line SL-303, install a cap on the tee between HV-389 and HV-390, and install a hose connection downstream of WC-HV-411. Currently a hard piped flow path exists, however use of this flow path requires several entries into high radiation areas to perform initial and restoration valve line-ups. The installation of this modification will eliminate the need for these entries and subsequently reduce radiation exposure and simplify operations, Safety Evaluation Summary:

The solid radwaste system does not support or impact any system or component needed to ensure nuclear safety. Review of FSAR f 'spter 15 demonstrated that the modification cannot increase the probability or consequences of accidents previously evaluated This modification will not efTect nuclear plant safety or safe plant operation There are no unreviewed safety questions, Modification - OC-MM-323560-008, Rev. O SE #323560 007, Rev. -0

" Change of Fuses for 480V U/V Trip Unit Circuits" Dnstiption of Change; This modification changes the fuses for the Under Voltage Trip (UVT) device circuits. These fuses will provide better protection for the UVT solid state time delay unit.

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Attachment i CFR $0.59 Report April,1993 - March,1995 Page 37 of $8 i

l Safety Evaluation Summary:

This change provides better protection of the UVT timing device and improves the fusing coordination of their powered feed circuit. An improved coordination will prevent the loading of non-safety loads onto the Emergency Diesel Generator and its possible overloading for a DBA scenario. There is no detrimental effect on the safety or the environment and this modification does not pose an unreviewed safety question per 10 CFR 50.59. No change is required to the Oyster Creek Technical Specifications as a result of this modification.

Procedure #619.2.019, Rev.12 SE #000733 004, Rev. O

" EPA Time Delay Setpoint Tolerance" D_cictiption of Change; The Electrical Protection Assemblies (EPA) provide overvoltage, undervoltage and underfrequency protection for the Reactor Protection System. Associated with each trip function is an adjustable time delay, which can be utilized to avoid nuisance tripping. The existing time setpoints for the EPAs (except for EPA No. i and No. 6 overvoltage time delay) are delineated in Station Procedure 619.2.019 as being 5 200 ms. However, according to GE Vendor Manual GEK 83433 (VM OC-0221), the minimum setpoint (200 ms) has a tolerance of 100 ms.

11ence, the purpose of this revision to the subject station procedure is to include the GE specified time delay setpoint tolerance.

Safety Evaluation Summary:

The existing time delay setpoint and ass ciated tolerances are consistent with RPS design and OEM recommendations for system operation. There is no change to the operation of plant safety systems, Technical Specification requirements and limits or adverse impact on the plant

~

emironment. No experiments or tests are performed which could adversely effect plant safety.

lience the clarification of the OEM specified time delay setpoint tolerance does not effect the nwrgin of safety or create an unreviewed safety question as described under 10 CFR 50.59.

Modificat!o.

SE #000700-902, Rev. O

" Installation of Combustion Turbine at Forked River" l

l l

Attachment I CFR 50.59 Report April,1993 - March,1995 Page 38 of $8 Qciqtipilon of Change; in order to add peaking power to the southern portion of the JCP&L service region, two Combustion Turbine (CT) generating plants have been constructed on the Forked River site adjacent to the Oyster Creek Nuclear Generating Station. The CT's are outside the plant protected area approximately 1/4 mile from the nearest safety related structure at Oyster Creek.

Safety Evaluation Summarv:

Based on the results of this evaluation, it is concluded that the CT installation does not adversely impact plant nuclear safety, cannot fail in a manner which will impact any critical plant structures or prevent the operation of safety related systems, nor does the installation introduce an l

unreviewed safety question.

l Modification - OC MM-403022 00h Rev. O, SE #403022 001. Rev. I

" Containment Spray Low Flow Alarm Signal Manual Reset Switches" Description of Change:

The present containment spray low flow alarm logic contains an automatic reset of the alarm signal if the alarm is due to a pump trip. The alarm signal duration can be too short to allow the alarm window to seal in. Additionaily, no alarm signal is received if the pump trip occurs within 40 seconds of system initiation. The purpose of this modification is to reduce the potential of an operator missing an actual low flow condition in the containment spray system during operation.

This is accomplished by the following changes to the low flow alarm circuit:

The automatic reset feature of the low flow alarm signal is replaced with manual reset push buttons.

The low flow alarm circuit is enabled following a time delay after system initiation.

Safety Evaluation Summarv; This modification does enhance operator effectiveness by reducing the probability of an operator missing a low flow condition in the containment spray system. The modification does not have any adverse impact on nuclear safety or safe plant operations. Therefore, this modification does not contain or introduce an unreviewed safety question as defined by 10 CFR 50.59.

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Anachment 1 CFR 50.59 Report 2

April,1993 - March,1995 Page 39 of 58 hlodification OC-htal-403020 002 Rev. I SE #403020 002, Rev. I

" Alternate Service Water Supply from the New Radwaste Service Water System" Description of Change:

The modifica' ion provides a temporary alternate service water supply from the New Radwaste Service Water (NRW SW) System. This modificaticn can be used to supply cooling water to the Reactor Building Closed Cooling Water (RBCCW) Heat Exchangers after the plant has been in cold shut down for at least threa days.

The modification allows replacement of the service water piping at the intake structure or future repair / replacement of underground Service Water piping. A small portion of this change will remain permanent. This will be the tie-in point to the NRW SW System at the Radwaste Building Closed Cooling Water lieat Exchanger inlet.

Safety Evaluation Summary:

This modification is acceptable for providing coolinF water through the RBCCW heat exchangers at least 3 days aller achieving cold shutdown in order to allow a portion of the SWS piping to be isolated for aeplacement. This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations (3) the probability or consequences of an accident or malfunction either evaluated in the FS AR or not, (4) the margin of safety defined in the Technical Specifications, and/or, (5) the emironment. It is important to note the provisions necessary to be in place depending upon how the NRW SW system is operated via this modification. Provisions for re-establishing decay heat iemoval must be available should a fire render NRW SW system inoperable. Also, if serv;ce water is unavailabl.e at the intake structure and in the Reacter Building, a means to re establish the ultimate heat sink via TBCCW fer decay heat removal must be in place in the event ofloss of station and auxiliary power.

Stodification - OC-SIM-402991-001, Rev. O SE #402941, Rev. O

" Replace / Upgrade Differential Pressure Switches DPS 66A/B and DPS-RV40A, H.C.D" Dngription of Change:

The purpose of this modification is to install replacement switches for differential pressure switches DPS-66A/B and DPS RV-40A, B, C and D. This modification will also install transient surge suppressors on the relays which DPS-66A/B actuate in order to prolong switch contact life.

l Attachment i CFR 50.59 Report April,1993 - March,1995 Page 40 of $8 Safetv EvaluatioJLSilmmam This modification replaces existing differential pressure switches with new switches and installs transient surge suppressors. This modification makes no change in any existing function performed by these switches or circuits. This modification introduces no new single failure which wou11 prevent initiation of required safety functions. The probability of occurrence, consequences, or type of an accident or malfunction other than previously described in the FSAR has not been changed. There is no environmentalimpact resulting from this modification. Based on the above evaluation, it is concluded that this modification will not have an adverse effect on nuclear safety or an adverse environmentalimpact. This Modification does not create an unreviewed safety question as described in 10 CFR 50.59.

Modification - OC-MM-402991-002, Rev. O SE #402991-002, Rev. 0 "RWCU Valve Nest Hatch Entry" pescripfton of Change:

During the 14R outage a 6'-O" x 4' 6" hole was cut in the reinforced concrete floor of the Reactor Building 7F.3" elevation to provide quick access for personnel to enter the Reactor Water Cleanup (RWCU) System Valve Nest located on the floor below A removable hatchway with lead shielding was installed to cover this hole and to shield the high radiation from the Valve Nest.

To facilitate removal of the hatchway shielding, a new 3 picce concrete plug will be installed to replace the existing lead shielding.

Safety Evaluation Summary; The purpose of this modification is to replace the existing steet/ lead brick hatch covers with concrete covers. The structural design of the concrete covers comply with all requirements. In addition, the proposed modification will not efTect the normal operation of the RWCU system, nor impact the safe shutdown of the reactor during movement of the concrete covers. In the ualikely event the RWCU piping is damaged by failure of the crane, the handrail or the gate, isolation of the system can still be achieved, thus maintaining primary containment. The RWCU has no other safety related function. As a result, it has been concluded that this modification will not have any adverse effect on nuclear safety nor the environment.

Document FCN-C078860, Rev. O SE #000106-002, Rev,1

" Justification for Scaffolding in Condenser Bay to Remain Permanent"

Attachment i CFR 50.59 Repon April,1993 - March,1995 Fage 41 of $8 Dsseription of Changt l

During refueling outages, a great deal of maintenance is performed on equipment (i.e., valves) in the vicinity of the moisture separators, reheaters, feedwater heaters and the condenser water boxes. Most of this maintenance work requires '.he erection of scaffolding. In past outages, this l

scaffolding has been erected prior to work and then removed after work is complete. This erection and removal takes time and resources. This safety evaluation documentsjustification for i

l scafTolding in the Condenser Bay to remain in place (including fiberglass grating) during the operating cycle. This would allow easier and quicker access for future maintenance on this equipment.

Safsty Evaluation Summarv:

This safety evaluation documents justification to permanently install scaffolding in the condenser bay. This scaffolding does not interfere with or effect equipment important to safety, it is located in areas of the condenser bay where it is not directly above important to safety equipment. In the unlikely event that the scaffolding collapses during an SSE, it will not damage equipment important to safety. Therefore, the plant's ability to safely shutdown and remain shutdown is unaffected.

Modification - OC-MM-402910-004, Rev. O SE #402910 004, Rey, i

" Control Room Recorder Upgrade - Cycle 14" pescription of Change:

The scope of this modification is to remove or replace Control Room recorders. In addition, the subject modification will revise the ranges for Control Room nitrogen purge and makeup flow recorder 12XR-0004. Main Turbine eccentricity and speed recorder will be converted from a single pen to a two pen recorder. The nitrogen makeup flow transmitter (FT-009) will also be replaced.

Safety Evaluation Summary:

The proposed modification does not have any adverse effect on nuclear safety or safe plant operations or the environment.' This modification does not constitute an unreviewed safety question as determined by 10 CFR 50,59.

Attachment I Cf'R $0.$9 Report April,1993 - March,1995 Page 12 of $4 ConHguration Change Document - O C-CCD-402991002, Rey, O SE #402991004, Rev. O

" Chem Waste Demineralizers in Series" DngIjption of Change:

The Chem Waste system was originally designed for a flow path through one filtei, one evaporator, and one demineralizer and then to the High Purity system but due to p blems with operation of the evaporators, a temporary demineralizer system operated by Chem Nuclear (from now on referred to ALPS) was installed and is being used instead of the evaporator and demineralizer, in an attempt to reduce the operating cost paid to Chem Nuclear, this configuration change will allow the existing Chem Waste demineralizers to be operated in series to process water from the filters instead of using ALPS.

Safety Evaluafton Summary:

The installation of thi configuration change will allow the existing Chem Waste Demineralizers to be operated in series. This will allow them to be used instead of the evaporators or ALPS to process liquid radwaste. The piping and components being used meet the material, pressure, and temperature requirements of the system. The installation is all within the seismic " bathtub" of the NRW Building. Installation of this change complies with the requirements of Reg Guide 1.143 and B31.1 There are no safety or operational concems involved in the completion of this change.

This configuration change does ot impact on nuclear safety and does not result in any unreviewed safety questions.

Modification - OC-MM-402986-005, Rev. O SE #401986-005 Rev. I

" Rod Worth Minimizer Lightning Protection" D.escription of Changeu The purpose of this modification is to provide lightning protection for the Rod Worth Minimizer (RWM) mux select circuit from the DAS cabinet in the old cable spreading room to the RWM3 cabinet in the SED. The RWM had been knocked out of service by lightning strikes at CC because the select circuit is not protected. This mod will provide protection to the mux select circuit by utili7hg a surge protected power supply. The scope of this modification is to install the mux select relay ubinet in the DAS cabinet in the OCSR and rewire the select circuit to use a 28v signal, which is lightning protected.

l

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Attachment I CFR 50.$9 Repon l

April,1993 - htarch,1995 Page 43 of 58 Safety Evaluation Summary:

The modification to the RWM mux select circuit is to prevent lightning induced transients from adversely affecting system operation. This modification will install a relay panel to protect the system from being adversely affected by lightning by utilizing a power supply which is transient protected. This modification does not propose an unreviewed safety question, have an adverse efTect on nuclear safety or safe plant operations and does not involve a Technical Specification change.

Modification - OC CCD-402991-007, Rev. O SE #402991 010, Rev. O "Startup Transformer (Bank 5 & 6) Control Room Meters - Panel 8F/9F Drsription.oLCham:

This modification provides voltage indication of startup banks 5 and 6 on control panel 8F/9F.

This information will be evaluated into the degraded voltage assessment program of the plant electrical distribu jon system.

Salbylyaluation Summary:

This modification does not have any adverse effect on nuclear safe plant operation or the environment. This modification does not constitute an unreviewed safety question as defined by 10 CFR 50.59.

Modification - OC-MM-409725-002, Rev. O SE #409725-002, Rev. I Fire Protection Connection for New Administeation Building" Description of Chang The purpose of this evaluation is to determine the effects on nuclear safety by the addition of a fire protection supply to the new Administration Building from the plant fire protection, vater supply system. This document also evafuates the effects on nuclear safety ofinstalling fire hydrants in the vicinity of the new Administration Building and extending this line along the north-south access road.

Safety Evaluation Summary:

The proposed modification will install a connection from the existing plant fire protection system to supply fire protection systems for the new Administration Building and to provide fire hydrants along the plant access road. The new connection is not required for the safe shutdown of the yis wy--

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Attachment I CFR 50.59 Report April.19% - March.1995 Page 44 of $8 plant nor does it adversely affect the fire protection water system. The fire protection system remains functional such that a fire will not s ause the los:: of capability to safely shutdown the

]

plant. Therefore, the proposed modification does not adversely effect nuclear safety nor reduce the mar 7e1 of safety as defined in the UFSAR and the basis of the plant Technical Specifications.

Hence, there is no unreviewed sr.fety question as defined by 10 CFR 50.59.

Modification - OC-MM-996900-001, Rev i SE #996900-001, Rev. I M

" Redundant Fire Protection Restoration" O

pescription of Change:

The purpose of this safety evaluation is to document the restoration of the Redundant Fire Protection System. The Redundant Fire Protection System was damaged due to a crane accident in December 1992. This activity is intended to entirely restore the system consistent with the original design. In addition, this safety evaluation documents the addition of a new 6x6x6 inch tee with a blind flange on the 6 inch pump recirculation line; and a new fire hydrant with an isolation valve on the pump discharge line east of the enclosure.

Safety Evaluation Summarv; The recoration of the Redundant Fire Protection System is inten ~

' provide a means of filling the Rcdundant Fire Protection Tank (if required), provide a mearis a Gaw water from the tank should it be required for a fire, and restore the system consistent with original design. No unreviewed safety questions, environmental impacts, technical specification changes, or any impact on nuclear safety will result from this modification.

Modification - Plant Procedure 656.4.001, Rev. O SE #408788-003, Rev. 2

" Refueling Bridge Test Weight Installation" Description of Change-.

A stainless steel weight has been fabricated to test the grapple hoist limit switch on the new refueling bridge. This test weight will be placed on the bottom of the spent fuel pool. The purpose of this safety evaluation is to demonstrate that the placement of this test weight system can be accomplished in a safe manner and Qat it does not impact nuclear safety.

S;ife:y Evdua. tion SummaII Movement of the test weight has been evaluated. NUREG 0612 requirements have been used as a guideline for review of heavy load control measures. The risk of a load drop is considered to be i

CFR 50.59 Report April,1993 - March,1995 Page 15 of 58 acceptably small due to crane design features and design margins when lining a load of 1438 lbs.

The use of a load lift procedure that incorporates the recommended guidelines minimizes the probability of a drop. The movement of this test weight is in accordance with GPUN commitments for compliance with the requirements of NUREO 0612. The safety evaluation concludes there are no unreviewed safety questions.

Modification - OC-MM-402990-004, Rev. O SE #402990-006, Rev. O "RBCCW Pump Return Header Venting Upgrade" Description of Chance:

This modification installs two automatic " float type" air purge valves on the RBCCW pump return headu. Also the addition of hose adapters on three existing RBCCW piping high point vent valves (V-5-523, V-5-612, and V-5-6909) are required to allow easier access to venting system piping during system fill. No other systems will be affected by this modification.

Safety Evaluation Summarv; This mini mod will improve the venting capability of the RBCCW system by adding hose adapters to existing vent valves and add two new automatic vents on the pump return header. This mini mod in no way effects the safety ftmetion of the system and does not result in unreviewed safety questions.

Modification - OC-MM-402988-001, Rev. O SE #402988-001, Rev. O "llydrolase Tap Connections for SSDSC and Reactor Cavity Drain Lines" Dncription of Change:

This Safety Evaluation evaluates the installation of two hydrolase taps and one drain connection in the Spent Fuel Pool Cooling system. One 4" drain line from the Steam Separator and Dryer Storage Cavity (SSDSC) and another 8" drain line from the Reactor Cavity are the major contributors to the general area dose rate on elevation 75' of the Reactor Building. Numerous hot spots have necessitated installation of permanent shielding. For decontamination purposes, it is recomr ended to in tall two permanent taps into 4' and 8" drain lines from SEDSC and Reactor Cavity. Also an additional 1 1/2" drain connection shall be installed to provide a draining point during hydrolasing of the 8" line.

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CFR 50.59 Report Aprit,1993 - March,1995 Page 46 of 58 Safety Evaluation Summarv:

The activity ofinstalling two taps and one drain connection for hydrolasing is technically acceptable and does not constitute an unreviewed safety concern as defined in 10 CFR 50.59 This conclusion is based on: 1) it does not alter SFPC System actuation, operational control or design of features, ?) the change has been evaluated and shown that it meets the required piping, seismic, and applicable design criteria. 3) the change does not require any operator action during plant operation or emergency response,4) it improves the decontamination process on elesation 75'-0" of the Reactor Building.

Modification OC-MM-328332-004, Rev. O SE #328332-004, Rey,0

" Containment Spray Heat Exchanger Drain Reroute" Description of Change:

This Safety Evaluation examines the modification which reroutes the drain line of the Co: tainment Spray Heat Exchangers. This Mini-Mod will provide a drain header from the drain isolation valves of the heat exchangers. A pump with a section header, isolation valves and hoses will be provided. A welded sock-o-let with an isolation valve will be installed on the 14" ESW line, downstream of the heat exchangers. This conSguration will provide an eflicient means for draining the heat exchanger.

Safety Evaluation Summary:

This modification will not effect the function of the Emergency Service Water System and the Containment Spray Heat Exchanger. The safety evaluation concluded there are no unreviewed safety questions.

Modification OC-MM ',23721-o01, Rev. O SE #323721-001, Rev, I l

" Spent Fuel Cooling Piping Relocation" l

Descrintion of Change:

A section of 6" underground aluminum piping has been found damaged by galvanic corrosion due to external coating failure. The 6" pipe is the retum line to the spent fuel pool cooling system from the filter and demineralizer which is located in the Old Radwaste Building. The degraded section of piping is located between the Reactor Building south wall at elevation 18'-l1" ond the pipe tunnel. Because of uncertainties as to th2 condition of the piping within the reactor building penetration and lack of accessibility for welding, the line will b. rerouted through a new penetration which will be installed in the reactor building south wall approximately 8' east of the existing penetration. No other systems will be effected by this modification.

Attachment i CFR 50.59 Repon

- April,1993 - Mar:h,1995 4

i Page 47 of 58 Sdety Evaluation Sunmary:

This modification in no way effects the functions of the Spent Fuel Pool Cooling system and the

- reactor building. The safety evaluation conclodes there are no unreviewed safety questions.

Modification - OC-MM-408853-016, Rev. O SE #408853-011, Rev. O

" Install Ground Fault Interrupters at the Intake" Description of Change The scope of this modification is to replace the existing circuit breakers with ground fault interrupter circuit breakers for the heat trace and receptacle circuits on PD-4, LP-1A31 and LP-IB31.

Safety Evaluation Summarv:

This modification will install ground fault interrupter circuit brealwrs for personnel protection at the intake canal. = This will increase the safety from shock hazards of personnel working at the intake. This modification does not adversely effect nuclear safety or safe plant operations, and does not involve an unreviewed safety question or require a Technical Specification change.

Modification OC-MM-409725-003, Rev. I SE #409725-003, Rev.1

" Domestic Water Connection for New Omce Building" Descriotion of Change:

The scope of this modification includes a tie-in to the existing 6-inch underground domestic water line from the North Well. A 6 inch line will be run from the North Well tie-in to the NOB's domestic water connection. This modification also includes a tie-in to the existing 3-inch underground domestic water line between the Flant Engineering Building and the Site Emergency Building. A 4-inch line will be run from this point due east beyond the security fence and routed south to the NOB's domestic water connection. isola: ion valves will be provided where required.

Safety Evaluation Summary:

The modification described above will not adversely effect nuclear safety or safe plant operation since the domestic water system performs no safety function and the change has no impact on any system which has a safety function.

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CFR 50.59 Report April,1993 - h'v h,1995 Page 48 of 58 Modification OC-MM-601001-001, Rev.0 SE #601001-00L Rev. 0 "Old Radwaste Building Partial Decommissioning" Description of Change:

i fhe purpose of this modification is to partially decommission the Old Radwaste Building at Oyster Creek by the removal and disposal of selected abandoned in place components.. All of the equipment that is being removed is located in the processing area of the Old Radwaste Building.

The equipment may be removed and 'sent to offsite processors for decontamination and release, or volume reduction, or sent directly to burial.

Interconnecting piping and ancillary components attached to the equipment, such as instrumentation, should also be removed and disposed of. Breached pressure boundaries will be reestablished in accu-dance with the applicable codes.

Safety Evaluation Summary:

The modification will remove some of the abandoned equipment in the Old Radwaste Building.

The modification will leave the plant in the same functional condition it is currently in. The plant margin of safety is not reduced by the removal of this equipment. Nuclear safety and safe plant operation are not adversely effected. No radiological safety concerns exists. The modification is -

acceptable under 10CFR50.59.

Modification - DCN C081382 SE #402840-005 "HWC FE Test Vent Line" Descriotion of Change:

This modification adds a small vent line for the periodic calibration of the HWC flow elements.

Safety Evaluation Summary:

This modification does not adversely effect nuclear safety or the environment. No unreviewed safety question is generated by this modification and it can be implemented under 10 CFR 50.59.

CFR 50.59 Report April,1993 - March,1995 Page 49 of 58 Modification - Dwg. 3D-187-38-005, Rev. O SE #402940-003, Rev. O

" Shielding for Reactor Vessel 2" Drain Line" pescrintion of Change:

Installation of temporary shielding during each outage is required for reactor vessel 2" drain line in order to provide a workable condition for maintenance and repairs. This modifi.ation is to install permanent shielding supports for tempu ary shielding for the reactor vessel 2" drain line in order to reduce worker radiation exposure and to improve worker productivity in the drywell.

Safety I t;aluation Summary:

This modification does not constitute an unreviewed safety question as determined by 10 CFR 50.59 and will not have any effect on nuclear safety, safe plant operations or the environment.

Modification - OC-MM-402991-003 SE #402991-003, Rev. I "ESW Keep Fill Throttle Valve" Descrintion of Modification:

This modification addresses the premature failure of the Emergency Senice Water (ESW) " Keep Fill" line check valves (V-3-131 and V-3-133). Historically the internals of these check valves j.

have been damaged by the high flow rates that they have been subjected to. This modification will install throttling globe valves upstream of the check valves to reduce the flow and the erosion damage associated with it.

Safety Evaluation Suminary; This modification will alleviate the flow erosion problem that has led to premature ESW check valve failure and will not adversely effect the function of the Service Water nor the Emergency Senice Water System. There are no unresolved safety concerns.

Corrective Change 518-93, Rev. 0 SE #000532-013, Rev. O

" Evaluate Material Change for ESW Pumps" Descriotion of Change:

This document evaluates the safety significance of changing the casing material for Emergency Senice Water pumps to allow the use of stainless steel. There will be no change in system

Attach.nent I CFR 50.59 Report April,1993 - March,1995 Page 50 of 58 performance. The design of the putnp is unchanged and therefore the horsepower requirements are unchanged. The material change is to provide for better corrosion resistance to sea water attack and prolong the life of ESW pumps.

Safety Evaluation Summary; The change of material from mechanite (cast iron) to 316 stainless steel will not change the performance of the ESW pumps. Hydraulically and mechanically they will perform the same.

g Nuclear safety or safe plar.t operations will not be effected. The safety evaluation concluded there s

were no unreviewed safety questions.

O Modification - OC-MM-409772-001, Rev. O SE #409772 00i, Rev,1

" Installation of Access Card Readers / Alarms on Security Doors" Description oiChangt The purpose of this modification is to increase security levels in the area of the control room to comply with the regulations of 10 CFR 73.55. This will effect the three access doors to the corridor between the control room and the outage coordination center. The doors are normally locked requiring a key for entry. The three doors are door 1: entrance door to the corridor left of the control room; door 2: opposite door of the new cable spreading room (entry to the oflice building roof), and door 3: rear door between the outage coordination center and corridor next to the control room. The modification shall alarm all three doors to prevent forced entry or improper door closure. For doors I and 2, security access card readers will be provided on both sides of each door that will allow entry or exit from the hallway. The hallway will not be monhored by security and will become part of the Oyster Creek Physical Security Plan.

Safety Evaluation Summarv:

This modification is to provide "IN" and "OUT" card readers for doors I and 2. Alarms will be provided for all three doors for intmsion purposes or failure of a door failing to fully close. This tightened security measure will secure the corridor between the outage coordination center and the control room to better secure the control room. This modification will meet GPUN's commitment to 10 CFR 73.55 and be administered under 7413-PLN-1530.01 - Oyster Creek Physical Security Plan. There is no detrimental effect on safety or the environment and the modification does not pose an unreviewed safety question per 10 CFR 50.59.

Attachment I CFR 50.59 Report I

April,1993 - March,1995 Page 51 of 58 Modification SE #407955-001, Rev. I "79-14 Hanger Mods / Upgrades Phase 11

Description of Change:

The purpose of this modification is to ensure that the 4 NSR pipe systems under the NRC Bulk. in t

79-14 program meet ASME/ ANSI B31.1 requirements by modifying supports to take the dead weight plus thermal and seismic leads developed in the NRC P.ulletin 79-14 analysis.

Safety Evaluation Summary:

The purpose of the proposed activity is to ensure compliance with the FS AR Design Basis including ASME/ ANSI B31.1 code. Therefore, these modifications willincrease the support capacities for natural phenomena protection and as such will not increase the probability of occurrence or consequences of an accident due to the natural phenomena and normal operating loads.-

Modification SE #408853-001, Rev. O

" Installation of Test Plugs for Diesel Generator Auto Actuation Test" i

Description of Change; This modification eliminates installation of contact blocks on relays and jumpers on thermal blocks in panels ER8A, ER88, ERI8A and ERl8B to simulate the conditions required for performing surveillance procedure 636.2.001 on Diesel Generator Auto Actuation Test. Under this modification, the Weidmuller Test Plugs type SAKC-10 will be installed in the Core Spray and Containment Spray System circuitry which will eliminate jumper installation and lifting leads when i

performing this surveillance.

Safety Evaluation Summary:

This modification provides a capability to perform Diesel Generator Auto Actuation Surveillance in Panels ER8A, ER88, ERl8A and ERl8B without instJling jumpers and/or lifting laads. This modification will not alter designed fimetions of Core Spray, Containment Spray and Emergency Diesel Systems. This modification will minimize the possibility of human error when performing the related surveillance. It is therefore determined that this modification does not constitute an anreviewed safety question. There is no environmentalimpact due to this modification and there are no changes required to the Technical Specifications.

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CFR 50.59 Report April,1993 - March,1995 Page 52 of 58 Modification - OC-CCD-403035-001, Rev. O SE #403035-001, Rev. O "TB Operating Floor Shielding" Description of Changt The six steam lines to the turbine will be shielded to lower radiation exposure during operating cycles and during the Turbine Building roof replacement work,. Reinforced concrete inverted "U" shape shields will be set above and around the steam lines.

Safety Evaluation Enmmary:

The six steam lines to the turbine will be shielded to lower radiation exposure using reinforced concrete inverted U-shape structures. The shield structures are designed to withstand their dead load during lifling, placement, as well as during operation. The Turbine Building floor capacity to withstand the shield load has been verified. No impact on any safety-related equipment or i

structures will occur due to this modification and no change to any plant procedure or operation l

will be required. The radiation shields are designed for the same loads and to the same criteria as the Turbine Building structure and the non-NSR equipment and piping in the vicinity of the shields. Therefore this modification will not have any adverse etTects on Nuclear Safety or the environment.

Corrective Change SE #000231-007, Rev. O

" Installation of AOG Blower Motor Housing Drain / Vent Line" Description of Change:

This proposed change wiU be in two parts. The first part will be to add a drain line with valves and fittings to the existing plugged drain connection at the bottom of the Augmented Off Gas (AOG) blower motor housing, This drain line will be used to drain excess moisture that cohcts in the blower motor housing. The second part will be to add a vent path in the discharge IL.e of the blower motor to aid in the measurement of the level in the blower motor housing. This line will be added so that a PVC type tubing sightglass can be valved into service to measure the level of water in the blower motor housing. The level will be measured by valving in the PVC tubing, to be used as a sightglass to the bottom of the blower motor housing via the drain line installed as part I and the tap into the blower motor housing discharge pipe ia the existing hydrogen sample line and the new vent / level piping being installed by part 2 of this activity. These two parts together will comprise the combination blower motor housing drain system and level indication system as described in Engineering Evaluation 604-93. All of the new valves being added will be closed during normal operation of the AOG train and will be used to drain the blower motor housing or to gauge the level in the housing. These valves may be opened during normal l

CFR 50.59 Report April,1993 - March,1995 Page 53 of $8 operation of the AOG train if they can be used for troubleshooting purposes and are controlled in accordance with plant procedures.

Safety Evaluation Summary:

This activity does not create an unreviewed safety question because: 1) the piping being added will not effect the normal operation of the AOG skid; 2) the AOG system is not relied upon to mitigate the consequences of any design basis accident; 3) the AOG system does not have any safety related equipment; and 4) the proposed activity will not reduce the margin of safety as defmed in the basis for any Technical Specification.

Modification - OC-MM-403019-001 SE #403019-001, Rev. I "Reacter Vessel Water Level Instrumentation (GL 92-04)"

Description of Change; The modification replaced the existing condensing chambers (IA15 A&B) with condensing chambers where the steam inlet and condensate return flow paths are separate. The excess condensate is returned to the Reactor Pressure Vessel via the variable leg which carries dissolved non-condensables. The new condensing chambers are also provided with an auxiliary reservoir / accumulator. The reservoir will drain into the condensing chamber and makeup for the reference leg column following a rapid depressurization.

Safety Evaluation Summary:

The modification will not adversely effect nuclear safety or safe plant operations, does not involve an unreviewed safety question, nor require a change to the Technical Specifications.

Modification SE #408895-001, Rev. 0

" Replacement of Valve V-5-165" Desenption of Change:

' V-5-165 is the inboard drywell isolation valve for the Reactor Building Closed Cooling Water System. The valve, a wafer body duel plate check, was leaking excessively, and was replaced with a wafer body swing check.

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Attachment I CFR 50.59 Report April,1993 - March,1995 Page 54 of $8 Safety Evaluation Summary; This modification does not adversely effect nuclear safety nor safe plant operations, does not involve an unreviewed safety question, and does not require a change to the Technical Specifications.

Procedure - Setpoint Change, Rev. O SE #00661-015, Rev. O "Setpoints for Domestic Efiluent Radiation Monitor" Qgscription of(. g The radiction setpoints for RE-661-1778 were lowered to conform with the requirements of the revised 10 CFR 20 Appendix B Table 3. The monitor has two setpoints, the higher setpoint, the pump trip alarm, is revised to 2105 cpm or less. This is equal to the Table 3 limit reduced by the uncertainty of the monitor calibration and instrument surveillance. The lower setpoint, the High i

Alarm, is revised to 1080 cpm or less. This is equal to half the Table 3 limit reduced by the uncertainty of the monitor calibration and instrument surveillance. The pump trip alarm is

- conservative because it will terminate discharge of domestic effluent if, even momentarily, the efiluent contains a concentration of Cobalt 60 that is equal to the acceptable limit for discharge as -

a n onthly average.

Safety Evaluation Summary:

The function of the radiation monitor is to differentiate between plant releases and medical releases to the sewer system in accordance with DMGS Sewer Tie-In Agreement (7/27/92) with the Ocean County Utilities Authority. This monitor is regulatory required, but does not perform a safety function. The safety evaluation determined that the setpoint change will not adversely effect nuclear safety nor safe plant operations; does not involve an unreviewed safety question and does not require a change to the Technical Specifications.

Modification - OC-MM-402990-002 SE #402990-003, Rev. O

" Alternate Water Supply to Emergency Condensers" Description of Change:

This modification provides for an alternate flow path to the shell ride of the emergency condensers. The modification adds a throttle valve and 2-1/2" hose adapter to an existing drain / overflow line from the emergency condenser. An existing threaded elbow downstream from V-14-96 is replaced with a threaded tee to accommodate the throttle valve and hose adapter. The alternate flow path uses fire protection hose station #46 located on the Reactor Building 95'-3" elevation northeast quadrant. A storage chest is chained nearby for storage of the 2-1/2" hose.

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CFR 50.59 Report -

April,1993 - March,1995 Page 55 of 58 This alternate flow path will only be used if the existing flow paths fror ' e condensate transfer, deminera:ized water and the fire protection systems are unavailable a change does not effect the existing drain /overf'ow function.

Safety Evaluation Summary; The modification will not adversely effect nu: lear safety nor safe plant operations, because the change does not affect the existing drain / overflow function. Since the sa ty function of the c

Emergency Condensers was not changed, and the margin of safety defined in the Technical Specification is not reduced.

Modification SE #000523-009, Rev. O

" Add Isolation Valve V-12-348 and Replace V-12-1000 and V-12-7001" Descrintion of Change:

The purpose of this modification is to replace two valves (V-12-1000,1001)in the Demineralized Water Transfer System. To ease future maintenance of these valves, Oyster Creek added a new isolation valve (V-12-0348).

Safety Evaluation Summarv:

The Safety Evaluation determined that the valve replacement and the addition of an isolation valve did not jeopardize the integrity of the system nor does it change the operation of the system as described in the FSAR. The change does not aaversely effect nuclear safety nor safe plant operations; does not involve an unreviewed safety question and does not require a Technical Specification change.

Procedure - Surveillance Change SE #00666-002, Rev. O "Drywell H2/02 Analyzer Surveillance" Description of Change:

The Drywell H2/02 analyzer monitors the drywell atmosphere post accident for potentially hazardous concentrations of hydrogen and oxygen. This activity proposed the use of the anfound tolerance of up to +/-2.5% for both channels, and also proposed a change in the surveillance frequency fro:a weekly to monthly.

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CFR 50.59 Report April,1993 - March,1995 Page 56 of $8 Safetv Evaluation Summary:

An as found tolerance of +/ 2.5% is acceptable for the hydroger -

^9 ring channels. Assuming an error of-2.5%, the 1.5% action level for the EOP's will be reacrn when the actual hydrogen concentration is 4%. OCNGS analyzed the surveillance data for the H2/02 analyzer over a three year period and determined the measurements did 'not drift sigmncantly with time, and an extension of the frequency was justified.

The change to the as-found tolerance and a revision to the surveillance frequency does not adversely effect nuclear safety or safe plant operations. The procedure changes do not involve an unri. viewed safety question or require a change to the Technical Specifications.

Modification SE #000871-005, Rev. O i

" North Yard Well Water System Modification" Description of Change:

- This modification adds a sodium hypochlorite feeder, particulate filter, pressure regulating valve and a set of raw water by-pass valves to the North Yard Domestic Water Treatment Facility. The domestic water system performs no safety relatec' betion. The purpose of the modification is to ensure high water quality and to meet the required state and local water quahty standards, hfety Evaluation Summary:

This modification does not adversely effect nuclear safety or safe plant operations; does not involve an unreviewed safety question and does not require a change to the Technical Specifications.

Procedure Change SE #315302-058, Rev. O

" Defeat of Stack RAGEMS Isokinetic Flow Default" Description of Change:

This activity changes the normal operating configuration to defeating the automatic bypass function. The normal flowpath uses an isokinetic flow control (proportional to stack flow) and routes the sample through particulate and iodine filters and then the high and low range radiation monitors.

Attachment I CFR 50.59 Report April,1993 - March.1995 Page 57 of 58 i

Safety Evaluation Summary:

Operation in this mode was included in the original design capabilities. This change does not adversely effect nuclear safety or safe plem operation, involve an unreviewed safety question, or require a Technical Specification change.

Modification SE #000838-004, Rev. 0 l

t "Drywell Fans Upgrade to Direct Drive" Description of Change:

The modification changed the existing belt driven Drywell Cooling Fans to direct drive with flexible couplings between the motor and the fan belt.

Safety Evaluation Summaty; The function of the Drywell Cooling Fans is not altered by this modification. The modification does not: adversely effect nuclear safety or safe plant operations; involve an unreviewed safety question, or require a Technical Specification change.

1 Document Change SE #945100-071, Rev. O "FSAR Drawing Elimination" Dncription of Change:

2 The Final Safety Analysis Report (FSAR) contains figures which enhance the description of the facility or depict the results of transient or accident analyses. A significant number of these FSAR figures are copies of the latest revision of a GPUN controlled drawing.

This activity removed these tigures from the OCNGS FSAR for Update 9, and substituted a reference to the GPUN controlled drawing number for the FSAR figure n.imber in the text.

Safety Evaluation Summary:

g T here is no change to any safety system or component, no change to any accident or transient analysis, and no mcrease in the probability of occurrence or consequence of a malfunction of equipment important to safety. The activity does not: adversely affect nuclear safety or safe plant operations, involve an unreviewed safety question, or require a Technical Specification change.

Attachment i CFR 50.59 Report April,1993 - March,1995 Page58of58 Modification SE #000561-003, Rev. 0 "New Radwaste Service Water Chlorination System" Drgdption of Change:

The modification eliminates one regulater and associated tittings and a " tee" from the New Radwaste Service Water Chlorination System. Most small leaks which previously have occurred were at fittings. The reduction of fittings from four to two reduces the probability of system malfunction.

Safety Evaluation Summary:

The change does not: adversely effect nuclear safety or safe plant operations, involve an unreviewed safety question, or require a Technical Specification change, Procedure Change SE #945100-058, Rev. 2

" Revision to Fire Protection Inspection and Audit Procedure" Description of Change:

The changes reflect organizational changes in Technical Functions. Managers of System Engineering are now responsible for providing system engineers for audit teams, FPPC is responsible for providing a Fire Protection Engineer and reviewing the technical content of the Audit Plan.

Safety Evaluation Summary:

f The activity does not change any procedural requirements. The change is limited to the concurrence signature and titles from the corporate reorganization. The activity does not:

adversely effect nuclear safety or safe plant operations, involve an unreviewed safety question, or require a Technical Specification change.

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