ML20249C333
| ML20249C333 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/31/1997 |
| From: | Roche M GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 1940-98-20319, NUDOCS 9806290077 | |
| Download: ML20249C333 (48) | |
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GPU Nuclear. Inc.
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U.S. Route (9 South l
NUCLEAR Post Office Box 388 Forked River, NJ 08731-0388 Tel 609-9714000 l
June 15,1998 1940-98-20319 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Dear Sir;
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 10 CFR 50.59(b) Report - April 1995 through May 1997 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 (SAR) for the period April 1995 to May 1997.
If any additional information is required, please contact George Busch at (609) 971-4643.
Very truly yours, Michael B.
che Vice President and Director Oyster Creek Attachment cc: Administrator, Region I NRC Project Manager
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NRC Resident inspector
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i 10 CFR 50.59 Report - 04/95 - 06/97 L
Safety Evaluation No.: 000222 001, R0 16R RVI Inspection Scope: Continued Use of Core Support Plate (CSP)
The CSP prosides a safety related function. That is, the CSP must restrict lateral movement of the core during the combined loading condition of nonnal operating pressure and a safety shutdown carthquake (SSE) in order to maintain coolable core geometry by insuring control blade insertion.
The material condition of the design reliant features of the CSP cannot be determined. The structural welds are not inspectable because of access limitations. The hold-down bolt pre-load cannot be practically verified. Aging da.nage due to IGSCC is possible in the CSP structural welds. Irradiation induced stress relaxation is possible in the hold-down bolts.
The purpose of this evaluation is to show that reliance on the CSP to perform its design function can be maintained.
for an additional operating cycle (Cycle 16) assuming that substantial aging damage has already occurred. This can be shown assuming only that the CSP remains intact and taking credit for the spherical seat on the hold-down bolts, a feature which provides self-tightening should the CSP shift laterally. This safety evaluation is written to
- address the impact on nuclear safety of a degraded condition for which there is not factual esidence. There is no season to believe that the material condition of the CSP has actually degraded.
Based on Oyster Creek actual experience with reactor internals cracking, cracking mitigation currently in effect, and CSP design features, it is concluded that the CSP is acceptable for continued operation for one additional cycle
_ Cycle 16) without further action. Additional senice beyond an operating cycle will be addressed before 17R.
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This safety evaluation has determined that the continued operations of the CSP for one cycle has no adverse impact on (1) nuclear safety, (2) safe plant operation, (3) the probability or consequences of an accident or malfunction -
either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications. Since this safety evaluation has determined that no unreviewed nuclear safety question has been created and that no emironmental impact is involved, continued operation of the CSP for one cycle is acceptable.
Safety Evaluation No.: 000574-005, R0 F
' Alternate Replacement: Alternate Replacement to lastall High Temperatum Sump Pumps x
n The 1-12 Sump pumps; P-022-033A & B, will be replaced with Flygt wann liquid pumps. These alternate pumps are similar to the existing pumps in function, but will be more reliable because they will be capable of operating at higher temperatures than the existing pumps. The new pumps will produce a higher head, still within the design parameters of the system, and contain a higher horsepower motor.
The system being changed is not safety related, does not affect safe shutdown and does not affect the operation of
- the system. The motor horsepower increase will not affect electrical loading or isolation. Therefore, an unreviewed safety question does not exist.
l Safety Evaluation No.: IMMl411-017, R0 Alternate Replacement: Low Pressum Main Steam Line Switches The purpose of this safety evaluation is to analyze the Alternate Replacement with respect to 10 CFR 50.59 criteria and determine the impact on nuclear safety and if an unresiewed safety question exists.
The existing Main Steam Low Pressure switches are to be replaced with hermetically scaled switches. The replacement will serve the same form, fit and function as the original. The safety of the plant is not affected by this replacement. The new RE23 switches will operate during all modes of plant operation. This replacement does not adversely affect nuclear safety, does not reduce the margin of safety, does not involve an unresiewed safety question and will be performed in accordance within existing electrical / instrument installation requirements.
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Safety Evaluation No.: 000411-017, R0 (Cont'd.)
This replacement will not affect the safety functions of any interfacing systems and does not affect the FSAR accident analysis; therefore, this replacement does not involve an unresiewed safety question or new environmental impact and does not adversely affect nuclear safety.
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Safety Evaluation No.: 000423-003, RO l
Configuration Change: Abandon Low Conductivity Overflow Tank There currently is a leak in the Regeneration Waste Drain Tank. This tank connects to the Low Conductivity l
compartment of the Regeneration Waste Tank (Low Conductivity tank) by a 6" diameter pipe. Rather than repair l
the leak, the drain tank will be abandoned because it is no longer needed. In the past, the Low Conductivity tank l
was used during the AABRO process but currently it is not. The only time the Low Conductivity tank is used is l
during refueling outages to drain the Condensate System or to collect seal leakage during run cycles. The capacity of the Low Conductivity compartment, without the drain tank, is sufTicient for these uses.
Configuration Change 0270-96 will remove the 6" pipe that connects the two tanks. The suction piping between the Low Conductivity tank and the pumps will be restored, and the fianges on the drain tank will be covered. No other work on the drain tank is required since there are no inputs to the tank, other than the 6" flange.
The syster.: being changed is not safety related, does not affect safe shutdown and does not affect the operation of the system. The pumpsvill have sufficient capacity to pump the contents of the tank to radwaste. Therefore, an amreviewed safety question does not exist.
Safety Evaluation No.: 328373-001, R0 Configuration Change: Chemical Waste Tank Desludging i
l The purpose of this document is to evaluate the safety of modifying the suction to the Chemical Waste Floor Drain Pumps so that the contents of a Chem Waste Collector Tank can be desludged following the addition of a coagulating agent. This document will also evaluate the safety of the removal of the sparger in Chem Waste Tank C. The new suction point allows the direct reprocessing of supernate which will significantly extend the life of the Chem Waste Dewatering filter and significantly reduce the generation of radwaste from this source. Settled solids can then be directly transferred to a solid waste treatment vessel for dewatering or solidification.
After the backwashing of a Condensate Demineralized, the high loading ofiron oxide, resin fines and crud results in short radwaste filter run time, which in turn causes an increase in filter sludge generation. This modification
- will significantly extend the life of the radwaste filter.
This modification will add an upper connection to the Chemical Waste Collection Tank (WC-T-lC) so that the supernate can be processed separately after solids are allowed to settle in the bottom of the tank. Additionally, this modification will modify the tank inlet so that the contents of the tank can be more effectively mixed with the coagulating agent or other additives. The sparger will be replaced by an eductor within the tank.
This modification does not adversely effect nuclear safety or the emironment. No unreviewed safety question is generated by this modification and it can be implemented under 10CFR50.59.
Safety Evaluation No.: 000770-001, R0 Configuration Change: Disposition MNCR 96-0142 This safety evaluation will disposition MNCR #960142 which addressed a plant configuration change that removed Radiant Energy Heat Shield (REHS), REHS-0078 in the Turbine Building, elevation 3' 0". This REHS was removed between ISR and 16R and is identified on GPl]N Drawing 3D-770-14-005, Revision 1.
As part of a plant Preventive Maintenance program all REHS were identified and are inspected at the end of every refueling outage. Each REHS is used to insure that Class IE electrical separation is maintained in accordance with separation criteria established by the modification in which they were installed.
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Safety Evaluation No.: 000770-001, R0 (Cont'd.)
This configuration change removed an existing REHS-0078 from the turbine building that provides no Class IE separation for safety related circuits or from high energy piping. It was concluded that the proposed configuration change does not constitute an unreviewed safety question and will not have any adverse effect on safety or the environment.
Safety Evaluation No.: 000532-018, R2 Configuration Change: ESW System i Vent Connection The purpose of this document is to evaluate the safety ofinstalling a vent connection and root valve in ESW System 1 just downstream of orifice R.O 21 A. The connection allows the installation of a hose that will vent the system outside of the Reactor Building. This will permit verification that venting of the pipe downstream of the orifice will reduce cavitation and flow induced vibration that is now present. This safety evaluation also addresses the temporary installation of a hose to the Reactor Building Ser ice Air penetration to allow the ESW System to be vented to the outdoors and also addresses the method ofinstalling the connection to the existing pipe so that secondary containment remains intact.
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.
Safety Evaluation No.:*000542-010, R0 Configuration Change: Plug Sampling in TBCCW Sy stem The change made to the three sample lines outlined in this configuration change involves adding a plug to each of the three isolation valves in each of these lines. The valves in these lines are normally closed and tagged, since the sample lines are no longer used. Adding plugs in the valves will ensure that no inadvertent leakage of TBCCW process water containing corrosion inhibitor and other chemicals will find its way to the 1-5 sump. Samples from i
the system are taken elsewhere. This system has no interface with nuclear safety related equipment. No impact is possible.
There are no unreviewed safety questions resulting from this modification. No Technical Specification change is required and these changes can be implemented under 10CFR50.59.
Safety Evaluation No.: 000531-021, R1 Configuration Change: Replacement of ESW Keep Full throttle Valves The purpose of this document is to evaluate the safety of replacing tue ESW keep full throttle valves with valves that have better characteristics for maintaining fluid velocity within the capability of the check valve. The existing 2-inch valves, V-3-940 and V-3-941, will be replaced with 1 1/2 inch globe valves. The new valves are specifically sized for this application.
This modification does not adversely efTect nuclear safety or the environment. No unresiewed safety question is generated by this modification and it can be implemented under 10 CFR 50.59.
Safety Evaluation No.: 000430-001, R0 Configuration Change: Replacement of V-1-99 The purpose of this document is to evaluate the safety of replacing V-1-99. The existing valve is a standard pattern i
l globe valve with a bypass valve (V-1-247). This valve will be replaced with a "Y" pattern globe valve. Also, the bypass will be eliminated.
This modification does not adversely effect nuclear safety or the environment. No unresiewed safety question is generated by this modification and it can be implemented under 10CFR50.59.
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Safety Evaluation No.: 000444-007. R0 Configuration Change: Restoration of Auxiliary Boiler Deacrator Vent (CH-HV-142)
The purpose of this document is to evaluate the safety of re-establishing a constamt vent path from the Auxiliary Boiler Deacrator Manual Vent Valve (CH-HV-142). A constant vent from the Deacrator is required to remove and release oxygen from the demineralized feedwater that supplies both boilers. This is required to control chemistry and preclude boiler tube failure.
This modification returns the function of the deaerator to its as-designed purpose..The replacement valve is a constant vent as required by the equipment manufacturer. This modification does not adversely effect nuclear safety or the emironment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 328367-001, R2 Core Spray Fill Valve Replacement & Discharge Pressure Indicator The purpose of this document is to evaluate the safety of replacing valves V-20-0072 and V-20-0073 and discharge valves V-20-0100 and V-20-0102. Additionally, a tee and root valve will be placed in the discharge line of each Core Spray Fill Pump. This will provide information for verifying performance of the respective fill pump.
The suction valves will be replaced with valves having a lower pressure loss. The existing valve is a standard L
pattern globe valve. Calculations have shown that the pressure drop through the valve, at the present fill pump efiow rate, creates a negative pressure at the fill pump suction. This causes gas (possible nitrogen) to be separated from the water. This causes the pump to lose suction. The replacement valve will be a gate valve. This valve will maintain a positive pressure at the pump suction.
The discharge valves will be replaced with globe valves. This will allow the flow from the pumps to be adjusted.
Such adjustment may be required to keep the fill pumps in an acceptable operating band (prevent excessive runout).
Pump NZ-04B will be relocated to a position approximately five feet from the deck. The pump will be orientated horizontally with the discharge pointed up. This will allow any gas that is present to self-vent from the pump. The pump will be supported by a clamp type support on the main pump discharge pipe. The power cable will be lengthened as required. A drain connection will be added to the piping to facilitate maintenance. To minimize the amount of piping that will require modification, the design temperature for the piping will be reduced to 170 Miegrees F. This will preclude a larger change to the pipe support arrangement. Design temperature for core spray I
' piping was established for the maximum torus water temperature used in the Mark 1 Containment System I
evaluation. MPR performed the analysis and used 170 degrees F water temperature for evaluating the torus and torus attached piping. A GPU calculation verifies that peak temperature is approximately 159 degrees F.
Additionally, the need has been identified to determine pressure at the fill pump discharge for diagnostic purposes.
This requires the installation of a tee in the fill pump discharge. An instrument root valve, threaded nipple and pressure gauge will be added to the tee.
The changes made are intended to improve the operation of the Core Spray Fill Pumps. These modifications will be used with a surveillance test program to determine if further changes are required.
l This modification does not adversely effect nuclear safety or the environment. No unresiewed safety question is generated by this modification and it can be implemented under 10CFR50.59.
l Safety Evaluation No.: 00(1532-013, RI 1
Corrective Change: Evaluate Material Change for Emergency Service Water (ESW) pumps The purpose of this safety evaluation is to evaluate the significance of changing the casing material for ESW pumps to allow the use of stainless steel.
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Safety Evaluation No.: 000532-013, R1 (Cont'd.)
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 perfonn the same. The only change will be the corros4n resistance of the pump, which will be better. A scismic cvaluation was performed for the new material and h was detennined that there is no change to the scismic qualification of the ESW pumps as a result of this material change. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000225-024, R0 Test: Control Rod Operability for 16-U-2 Startup The control rod drive system should be considered operabic to support the 16-U-2 startup after the replacement of the SSPV diaphragms with Viton AB, followed by the successful completion of cold scram time tests on all control rods, and the ARI surveillance. Prior to exceeding 40% reactor power, technical specification requirements for control rod scram times will be verified.
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Control rod scram time concerns will not adversely affect nuclear safety or safe plant operations during the 16-U-2 startup. All SSPV Viton A diaphragms will be replaced with Viton AB diaphragms, which are qualified as an alternate replacement and are designed to alleviate the sticking problems that resulted in slow 5% scram times. Cold scram timing and the ARI surveillance will be performed prior to cartup to ensure contr61 rod scram capability. Power will be limited to 40% during startup until scram time testing is performed at normal operating pressure to verify compliance with tech spec limits. Additional
' performance monitoring will be implemented during the operating cycle to ensure performance does not degrade with the new diaphragms.
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Tech Spec Section 3.2.B.3 provides control rod scram insertion time requirements. The SSPV diaphragm replacements and cold scram tests will provide reasonable assurance that control rods will scram normally within expected values. Insertion time requirements will be verified prior to exceeding 40% power, and future performance will be monitored to detect any degradation. Also, analyses have shown that the
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maximum anticipated degradation in 5% scram time performance would have a negligible impact on the MCPR Safety Limit. Therefore, the consequences of any accidents or malfunction of equipment imponant I
to safety will not be increased.
The alternate replacement of Viton A SSPV diaphragms with Viton AB will not increase the probability j
of SSPV failure or the failure of any other control rod drive system components. Also, cold scrams were considered inthe design basis of the CRD System. and will not significantly degrade system components according to a GE Sersice Infonnation Notice. Also, cold scrams from the fully withdrawn position have much less of an impact than scrams from 00. Therefore, the CRD system is expected to perform as designed, and the probability of an accident or malfunction of equipment previously evaluated in the SAR is not alTected.
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The alternate replacement of SSPV diaphragms will maintain normal scram performance, and the subsequent cold and hot scrams are pan of the system design basis. Therefore, the possibility of an accident or malfunction of a difTerent type than presiously evaluated in the SAR is not created.
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The alternate replacement of SSPV diaphragms will not decrease the margin of safety as defined in the basis for any technical specification. The new diaphragms are expected to restore nonnal 5% scram performance while meeting scram time requirements at all insertion levels. The cold scram tests will provide assurance of acceptable scram performance, and compliance with Technical Specification Section 3.2.B.3 scram time criteria will be verified prior to exceeding 40% power. Also, the ARI surveillance will be performed prior to reactor startup. Any concerns for the 5% scram time will not decrease the margin of safety, because analyses have shown that the maximum anticipated degradation in 5% scram time perfonnance has a negligible impact on the MCPR Safety Limit.
All Technical Specification requirements for the CRD system will be maintained. This modification does not adversely efRct nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are regttired.
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Safety Evaluation No.: 000571014, R0 Document: Covered Floor Drains Two floor drains in the Turbine Building have plastic covers oser them to prevent a radiological concern. Either contaminated water can splash out or airborne material from the ofTgas system could potentially leak out. This safety evaluation addresses the fact that flow into these floor drains is restricted due to being covered.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000732-018, R0 Document: GE Thermal Overload Relay Testing The purpose of this safety evahtation is to review the use of a new or changed method of testing Motor Control Center thermal overload relays in the field by Maintenance personnel. The present method of testing overload relays was previously reviewed and determined not to have affected equipment operability even though the method was not derived directly from present day General Electric literature. The proposed method of testing overload relays is derived directly from GE literature, curves. tabics, and discussions with GPUN engineers. This proposed method is being accepted as an appropriate method and is being directed for use by the 125.1 #92-97.
This change in method for testing of GE overload relays will not adversely affect nuclear safety or safe operations because it was determined and demonstrated that there is no significant change involved from previous methods.
There is no increase to the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety because the new method for thermal overload testing will not change thermal relay setpoints from their present design tolerance band.
There is no possibility for an accident or malfunction of a difTerent type because there has been no change to any system. component, operating mode or setpoint as a result of this change.
Thermal overload testing or setpoints are not specifically discussed in the Technical Specification. Since there ha>
been no change to components or setpoints for any system, no Technical Specification system or equipment margins are reduced by this change.
Safety Evaluation No.: 000153-016, R1 Document: Reactor Building Internal Pressure The UFSAR describes blow out panels in the Reactor Building, which relieve building internal pressure at 0.25 psi.
An effort to verify the existence of the blow out panels either by document review or by visual inspection of the Reactor Building was unsuccessful. A calculation was performed showing that without blowout panels the Reactor Building internal pressure will reach 0.95 psi during a high-energy line break outside of primary containment.
This pressure level could potentially damage the torus, cause failure of the masonry block walls adjacent to the southeast corner room, damage the Reactor Building structure, and adversely afTect equipment qualification profiles.
A structural calculation performed to assess the impact of this condition has demonstrated that the Reactor Building will relieve internal pressure at 0.20 psig due to lateral torsional buckling of girts which support siding panels. Girt buckling and subsequent panel separation relieve internal pressure that could cause damage to the load carrying members of the building superstructure.
The purpose of this document change is to correct inadequate documentation of existing plant conditions in the UFSAR. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
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Safety Evaluation No.: 000911-001, R0 Evaluation: Evaluation of Monitor & Change Room Mas <mry Block Wall The purpose of this safety evaluation is to address the acceptability of the existing configuration of a masonry block wal! as a fire zone boundary instead of a rated fire barrier. This boundary is located above the computer room entrance in the Oflice Building and extends up to the ceiling. The dimensions of the boundary in question are approximately 59" wide x 66" H. This wall was installed to comply with commitments to Branch Technical Position APCSB 9.51, Appendix Z (Appendix A to BTP 9.5-1), " Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976." The wall was originally designed to be a 1 1/2-hour fire barrier. The Oyster Creek Fire llazards Analysis Report (FHAR) currently identifics this boundary as a one hour rated fire barrier. The boundary is part of the south wall separating Fire Zone OB-FZ-10A from Fire Area OB-FA-9. The boundary construction consists of masonry block walls. Penetrations through this barrier are scaled up with non-combustible material. The reason for reclassifying this boundary is that it does not appear that it was installed as
' originally intended (type of blocks are not what was specified, penetration grout does not appear uniform thickness). This was reported on MNCR 97-001. Because of this, the installed configuration's fire rating cannot be compared to a tested configuration and the fire rating is therefore indeterminate. Due to both the lack of a fire rating of the boundary and an inspection of the boundary, it is concluded that the masonry block wall discussed above is not one hour rated as documented in the FHAR or the original design documents (21/2 hour) and will be evaluated acceptable as is in lieu of reconstructing the boundary to a one hour rating.
This evahration concludes that the wall meets the requirement of a Category A zone boundary barrier as defined by the FHAR. It is adequate to prevent the spread of flame, smoke and hot gases across it regardless of which side the fire originates. The block wall is therefore considered acceptable as-is. This evaluation does not adversely effect nuclear safety or the environment. This evaluation does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 402880-002. R2 Independent Spent Fuel Storage Installation The purpose of this safety evaluation, which is for the operation and design of the ISFSI, is to demonstrate that the loaded canisters can be transferred to, and stored in the ISFSI in a safe manner in accordance with all licensing requirements and commitments.
This safety evaluation is for when the loaded canister, secured within the transfer cask and on the trailer, is outside the closed Reactor Building Truck Bay Airlock Inner Door and the Reactor Building Truck Bay Airlock Outer Docr is open.
The OCNGS ISFSI is suitable for use of the NUHOMS system under the general license granted under 20 CFR 50.59 and 10 CFR 72, Subpart K. The NUHOMS system CSAR design bases bound the applicable design bases of the OCNGS FSAR; however, the NUHOMS system er be used under a general license only in accordance with written procedures.
Safety Evaluation No.: 328286-001, RO Inspection & Coating: Isolation Condenser "B" - Inspection and Recoating The purpose of this safety evaluation is to resiew and assure that the inspection, cleanup and application of a new protective coating on the carbon steel surfaces ofIsolation Condenser "B" internals as outlined in specification SP-1302-06-010, RO, will not compromise the integrity of the vessel or safety of the equipment or plant.
l OCNGS is equipped with two isolation Condensers to provide cool down of the reactor when the main condenser is not available. The two Isolation Condenscrs, A and B, were designed and manufactured by Foster Wheeler Corporation for passive cooling. Both condensers were hydrolaze c! caned during the 10R outage after removal of chromates, which served as a corrosion inhibitor in the past. Then the inside part of both shells was coated with Rust-Olcum Corporation's Damp Proof Red Primer #769. This coating failed on both condensers since it was unsuitable for ctmtinuous immersion senice.
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Safety Evaluation No.: 328286-001, R0 (Cont'd.)
During Cycle 12 octage, Isolation Condenser "A" internals including shell ID was grit blasted and coated with K&L Kolor-Poxy Primer 3200. Isolation Condenser "B" coating was deferred for a later outage and morpholine was added to the shell side water for pH (and therefore corrosion) control.
The 16R activities on Isolation Condenser 'B" will not adversely impact the performance of the affected systems, affect the safety functions of these systems, increase the probability of occurrence or consequence of an accident, create the possibility of an accident, decrease the margin of safety as defined in the bases of the Oyster Creek Technical Specifications, siolate any licensing requirements, cause a radiological concern nor affect environmental conditions. Therefore, it is concluded that this work will not have any negative impact on the nuclear safety nor does an unreviewed safety question exist.
' Safety Evaluation No.: 402880-006, R1 ISFSI Facility PIDS This modification shall install a Perimeter intrusion Detection System (PIDS) for the ISFSI facility, w hich will meet the applicable requirements of 10CFR72,10CFR73, and related Regulatory Guides. The Horizontal Storage Module (HSM) Temperature Monitoring System (TMS) shall satisfy the requirements of the NRC SER and Vectra Generic Temperature Monitoring Specification. In accordance with the GPUN Operational Quality Assurance Plan, this modification is classified as " Regulatory Required."
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The ISFSI/PIDS and HSM TMS do not impact the operational boundary of any plant safety-related system. The PIDS is required to prevent the possibility of radiological sabotage of the ISFSI facility. The HSM TMS is required to monitor the HSM temperature to ensure that their design basis temperatures are not exceeded. There is no change to the operation of plant safety systems, Technical Specification requirements and limits or adverse impact on the plant environment. No experiments or tests are performed which could adversely affect the plant's safety.
Hence this modification to install the ISFSI/PIDS and HSM TMS do not afTect the margin of safety or create an unresiewed safety question as described under 10 CFR50.59 n
Safety Evaluation No.: 000225-022, RO i
Job Order Maintenance Work: Ilydraulie Control Unit 26-31 Multiple Fireze Seal 16U2
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There is some concern that a flow restriction exists in the two lines feeding drive water to HCU 26-31, and it is necessary to perform an internal inspection of these lines. The proposed maintenance activity will require these valves to be disassembled.
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The Control Rod Drive (CRD) System provides the hydraulic pressure and control logic, via HCU, to move Control Rods into and out of the reactor core. Valves V-305-101 and V-305-102 are manual isolation valves (one on cach side of the HCU), which, when closed, isolate HCU 26-31 from the reactor. There is no installed way to isolate these valves from the reactor, Therefore, a freeze will be applied to the reactor side of these valves, which will allow the needed inspections. This safety analysis will prove that no safety issues exist by application of a freeze seal to each of the two lines.
Job Order 51530 im olves the internal inspection of portions of HCU 26-31, namely valves 101 and 102, to scarch for any flow restrictions present in the line, evidence ofimproper valve operation, and damaged components. This safety evaluation concludes that the application of a freeze seal upstream of these valves as part of this work i
activity does not effect safe plant operations, nuclear safety or the emironment. No unreviewed safety question is l
generated by this activity, and it can be implemented under 10 CFR 50.59.
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Safety Evaluation No.: 403036-008, R0 Measurement / Procedure: Trolley Weight Measurement of Reactor Building Crane at Oyster Creek l
The purpose of this safety evaluation is to assess the potential affect on nuclear safety of placing the Reactor l
Building overhead crane trolley on jacks which will suspend the trolley a maximum of 1/8" above the crane trolley rails. The trolley will bejacked at three locations. Oncejacked oft the rails, direct measurement and calculation j
will determine the weight and center of gravity of the trolley The potential impact ofjacking and lowering the trolley and measuring the trolley weight is addressed in the safety determination of Job Order 509267. This safety evaluation addresses the configuration w hen the trolley is held 1/8" above the rails while various measurements are being taken.
The reactor may be in any mode during the measurement. The measurement will not be conducted over the spent fuel pool or the reactor cavity. The temporary condition of the Reactor Building trolley during the measurement of trolley weight has no impact on plant safety or safe plant operations. The Reactor Building crane and support structure can safely withstand a free drop of the trolley from its maximum lift height. There is no other potential impact on plant safety, i
l The procedure does not increase the probability or consequences of an accident or malfunction of equipment, cause the possibility of a difTerent type of accident or malfunction or reduce margins in the Technical Specifications.
There is no impact on the FSAR and no radiological or environmental consequences. No unreviewed safety questions resub and this procedure can be performed under 10CFR50.59.
f rc Jiafety Evaluation No.: 403024-001, R1 Miscellaneous SQUG Support Modifications in December,1980, the Nuclear Regulatory Commission staffinitiated an unresolved safety issue, USI A-46, "Scismic Qualification of Equipment in Operating Plants," related to scismic adequacy of mechanical and electrical equipment in oldar nuclear plants.
To address the unresolved safety issue at the OC Nuclear Facility, a Safe Shutdown Equipment List (SSEL), in accordance with Generic Implementation Procedure (GIP) guidelines, was developed, followed by a walkdown and evaluation of the equipment on this list. During the walkdown evaluation, items of equipment that did not comply with all the screening guidelines provided in the GlP were identified as outliers. Those outliers requiring modification were grouped into four different categories based on the type of scismic concern: Missile Hazard, Interaction Hazard, Inadequate Anchorage, and inadequate Structure.
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'The purpose of this safety evaluation is to assess the affects, if any, on nuclear safety of the completed modifications' configuration to the outliers in cach of the categories.
The modifications covered by this safety evaluation are limited to increasing the scismic capabilities of the affected equipment. No changes will be made to any equipment or system function. All work is done to applicable codes, standards and materials, and in accordance with GPUN Operational Quality Assurance Plan for NSR and RR systems, subsystems and components. The modifications do not increase the probability of occurrence or consequence of an accident presiously evaluated in the FSAR. The margin of safety as defined in the basis for Technical Specification is not reduced. There is no unreviewed safety question resulting from these modifications.
No Technical Specification change is required and this change can be implemented under 10CFR50.59.
Safety Evaluation No.: 000214-015, R0 Modification: Shutdown Cooling Relay Modification This modification will outline control circuitry changes needed to 1) prevent nuisance trips of the Shutdown Cooling System (SCS) pumps uhile configured in the recirculation mode; 2) provide protection for the SCS pumps should the discharge valve (V-17-54) fail to close under certain conditions; and 3) remove isolation valve (V-17-19 and V-17-54) position contact as a permissive from the pump start logic.
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1 Safety Evaluation No.: 000214-015, R0 (Cunt'd.)
Trips of one or more SCS pumps occur during their start sequence, when the fluid hydraulics, created by inertia, produce a momentary low-pressure condition in the suction piping. The magnitude of the pressure transient is j
enough to dip below the suction pressure switch setpoint of 4 psig. The response of the switch is sufficiently fast to l
responds to the transient, since it is a solid fluid system. However, the duration of the momentary, low-pressues transient is very small, perhaps less than a second.
In the event that a containment isolation signal occurs during Shutdown Cooling pump operation, and the suction isolation valve (V-17-19) closes, but the discharge line isolation valve (V-17-54) does not, for u hatever reason, the pump (s) would damage themselves in the second and a half, which is now part of the circuitry. An interlock, based on the position of the suction valve is added to the pump trip logic to prevent this from occurring. To facilitate weekly bumping of the pumps, a momentary action, bypass switch will be installed to bypass this interlock.
Changes to the start control circuitry of the pumps will also be made. A starting permissive for closure of one, or I
both, of the loop isolation valves (V-17-19 or V-17-54) will be removed. There is no logical reason to have this i
restriction. A scarch of the design and licensing basis literature cannot determine that this issue has any basis.
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Furthermore, the Nine Mile Point Unit I station does not have these interlocks. Their circuitry does possess,
pump trip on valve position.
i The replacement of the original relay with a time delay relay in the 480-Volt Substation I A2 and IB2 as described 4above will not adversely affect plant safety and does not involve an unreviewed safety question.
The change in the pump START and STOP logic circuits as described above will not adversely affect plant safety and does not involve an unresiewed safety question. This modification does not adversely effect nuclear safety or the enviromnent. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 403069-001, R1 Modification: Stator Coolant Generator Run-back Circuit upgrade The purpose of this configuration change is to modify the stator coolant runback logic to a two-out-of-three logic.
This system initiates an automatic power reduction on either high stator outlet coolant temperature or low supply coolant pressure. A two-out-of-three initiating logic will climinate the possibility of a single sensor failing and
, causing an erroneous runback and subsequent plant trip. This modification was initiated in response to a BWR Owners group SCRAM reduction committee recommendation.
Additionally, there is a pipe interference, which prevents proper installation and removal of stator coolant conductivity ecli CC-713-002. The header, which contains valves Y-77, Y-78, and Y-79, interferes with the straight line clearance required for the proper installation and removal of the cell. The plant currently operates with valve Y-77 open. These three valves currently serve no purpose and will be removed and replaced with pipe to resolve the interference problem. This modification is recommended in GE TIL 1098-3R2.
This inodification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 315403-030 RI Modification : Iligh Pressure Trip Delay for Recirculation Pump Trip The purpose of this safety evahiation is to evaluate changing the trip setpoint for two reactor recirculation pumps from a high pressure of 1051 psig to a persistent high pressure of 1051 psig for up to 12 secon.ls.
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Safety Evaluation No.: 315403-030, R1 (Cont'd.)
In summary, this safety evaluation concludes that two recirculation pumps trip on persistent high pressure of 1051 psig for up to 12 seconds does not increase the consequences of postulated accidents. These pumps must be in the recirculation loops not connected to the emergency condensers. Compliance to the ATWS Rule is met by this change. Having two recirculation pumps operational following a scram with a temporary pressure spike (< 9 seconds) is desirable as it provides better core cooling with level and temperature indications representative of the core region. In addition, thermal overstress of the CRD stub tubes and incore housing welds will be prevented and there will be reduced thermal cycling of the plant.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000215-013, R0 Modification / Document: Disposition of MNCR 96-0135 MNCR 96-0135 documents the fact that the as-left body guide / disc guide slot clearance for valve V-16-14 is not in compliance with the vrdve overhaul procedure.
The evaluation determines the impact of not meeting the recommended body guide / disc guide slot c!carance.
GPUN Calculation C-1302-215-E310-051, Rev.1, evaluates the impact on valve capability to isolate a RWCU HELB within the required stroke time considering the minimal clearance, u hich exists. As a result of the minimal guide clearance, the potential exists for the valve to experience very high loads while isolating a HELB in the
'RWCU system. Further, the evaluation of MOV capability concludes that the potential would exist that the motor TOLs would trip preventing the valve from complet;ng its stroke. As a result, the TOLs are being bypassed. This safety evaluation evaluates the impact ofinstalling thejumpers that bypass the motor TOLs for V-16-14.
Furthen, bypassing the TOLs lowers the circuit resistance resulting in a higher voltage available to the motor. This increase in motor voltage allows for increasing the motor torque available.
This evaluation documents that the existing condition of V-16-14 as documented in MNCR 96-0135 does not l' ave an adverse impact on Nuclear Safety. The MOV maintains its capability to isolate a HELB in the appropriate time frame, in accord;mcc with its GL 89-10 Design Basis. Bypassing the subject valve's TOL relay during plant operation ensures the TOL does not jeopardize the valve's safety function. The TOL function is procedurally controlled by having a qualified person at the starter monitoring he current with the responsibility ofinterrupting
-the circuit should a motor stall occur.
It is concluded the subject evaluation does not have any adverse efTect on Nuclear Safety or Safe Plant Operations or the environment. The MNCR does not constitute an unresiewed safety question as determined by 10 CFR59.59.
Safety Evaluation No.: 000731-007, R1 Modification: "E" Reactor Recirculation Pump Conduit Reroute On February 12,1997, Reactor Recirculation Pump Power feeder to MG Set 'E' was apparently damaced by an electrical fault. The purpose of this safety evaluation will be to evaluate the installation of a new conduit run from the Feedwater Room to the 'E' Reactor Recirculation MG Set. This conduit run will climinate the anticipated problems with using the existing conduit routing passing through the Feedwater Room to the Reactor Building.
The work will include the installation of a new power feed (3-1/C #1/0 with #6 AWG 600 volt ground cables) from 4160 Volt Switchgear A to MG Set 'E'(circuit #14-3).
This configuration change does not introduce my new accident or malfunction not presiously evaluated, nor does the modification increase the likelihood of r.ratrence or consequences of any accident as analyzed in the UFSAR.
This modification does not decrease the margin of safety as desenbcd in the Technical Specification because the modification does not impact any system safuy functions. This evaluation concludes there is no unresiewed safety question per 10 CFR 50.59 and no changes to the Technical Specifications are required.
Safety Evaluation No.: TMH 1571-008, R0 Modificathm: 1-1 Sump Level Control This modification will improve the reliability of the 1-1 Sump high level alarm and will proside a means, independent of this alarm, to indicate that the sump should be pumped down.
i A level switch on a captive air tube actuates the current high level alarm. A FLYGT float switch will replace this arrangement. The mechanical alternator, which is currently unused, will be removed. The float for the alternator will remain in place and will be used to indicate a level control band that the operators will maintain by manually operating the sump pumps.
This modification is contained completely within the Turbine Building basement and has no impact on nuclear l
safety or safe plant operations. This evaluation is required for the sole reason that system and component descriptions in the UFSAR will require resision on completion of the modification.
Safety Evaluation No.t IMHl523-017. R0 Modification: Addition of Test and Home Connections The modification is the addition of a surveillance test connection to the Demineralized Water System, downstream of valve V-12-178 on elevation 23' 6" and the documentation of an "as found" hose connection with a drain valve (V-12 328). Both of these changes are in the Reactor Building.
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4 This modification does not adversely effect nuclear safety or the environment. This modification does not involve b :any unresiewed safety question and no changes to the Technical Specification are required.
' Safety Evaluation No.: (HH1422-008, R1 Modification: Additional Valve to Pump 1 A Suction Drain Line The purpose of this document is to evaluate the safety of adding an additional valve to the feedwater Pump 1 A suction drain line. The additional vah c will be used to add a zine injection skid to the system at a later date while the plant is operating. The future addition of the zine injection sicid will be covered separately. This evaluation only deals with adding a valve to the suction drain line.
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.
p M Safety Evaluation No.:4 312400-026, RG
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Modification: Alternate Replacement of Liquid Poison Tank Piping Heat Tracing The purpose of this safety evaluation is to evaluate the modification to replace the heat trace in the Standby Liquid Poison System (SLCS). This heat trace is on the pipe from the Poison Tank (T-19-(XX)l) to the suction of the two pumps (P-19-001 A and P-19 001B). Heat tracing the pipe prevents crystallization of the boron solution internal to the SLCS piping on the suction side of the pumps. Temperature controller TIC-1106-32 controls the temperature of the suction side piping. The heat tracing will also include the three suction valves (1106-5A (V-19-6),1106-5B (V 19-5) and i106-12 (V-9-4) and branches to first stop valves 1106-13 (V-19 11),1106-15 (v-19-9) and 1106-16 (V-19-22)). A temperature indicator will be added to aid in the reading of the temperature of the solution in the suction piping to the pumps. The configuration of the existing heat trace does not include the pumps.. This modification will not add heat tracing to the pump.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
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Safety Evaluation No.: 000571-009 R0 Modification: Alternate Replacement to Install liigh Temperature Sump Pump This alternate replacement will improve the reliability of the 1-2,1-3, and 1-4 Sumps by installing pumps designed to withstand high temperatures. The pumps currently installed in these sumps have a design temperature limit of 105 degrees F. The water in these sumps has been and continues to have the potential to be well in excess of this temperature. Wann liquid versions of these pumps, with a design temperature limit of 160 degrecs F, will be installed in these sumps to increase the reliability of operation.
The basic purpose and operation of the system will be unafTected. This activity will be performed completely within the Turbine Building north basement and mezzanine and will have no impact on nuclear safety or safe plant operations. This evaluation is required for the sole reason that system and component description in the Updated FSAR will require revision on completion of the replacement. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 328389-002, R0 Modification: Alternate Senice Water Supply from the NRW SW Sptem and Differential Pressure gauge Installation This configuration change provides a temporary alternate senice water (ASW) supply from the New Radwaste Sen-ice Water (NRW SW) System. This temporary piping can be used to supply cooling water to the Reactor Building Closed Cooling Water (RBCCW) Heat Exchangers after the plant has been in shutdown for a period of time required to achicyc sufficient decay heat removal with the temporary system (as determined by analyses for the specific outage). The ASW may also be used in an cmergency. The temporary piping allows any future repair, replacement, or change of Senice Water (SW) piping upstream of the tic-in location. A portion of this configuration change will remaisi pennanent. This will be the installation of a differential pressure gauge (FI-531-0013) and re-configuring the SW tic-in piping. The gauge will be located near the nonnal SW system differential pressure gauge (F1-531-1033), however, the additional gauge will be a lower range to allow more accurate flow measurement when the NRC SW System is the source of cooling water to the RBCCW heat exchangers. An additional isolation valve (V-3-976) with a valve box will be installed between valve V-3-711 and the existing blind fiange at the SW tie-in location that is approximately 4 feet below grade. The piping from this valve will then be routed up closer to grade and be terminated with a Victaulic cap and coupling. This will minimize digging in future outages when the ASW System (via NRW SW) is required. The additional valve provides double isolation to prevent leakage and potential freezing within the deadleg.
Note that this document identifies provisions that must be in place to accommodate events such as fires or a loss of station and auxiliary polver while construction on senice water piping is ongoing. This document does not resiew the acceptability of these provisions, but identifies the need for them as construction sequencing may make the plant vulnerable to a loss of decay heat removal if the aforementioned events occur during construction.
The temporary portion of this modification is acceptable for providing cooling water through the RBCCW heat exchangers after achieving cold shutdown in order to allow a portion of the SWS piping to be isolated for replacement or repair. The pennanent portion of this modification will meet as a minimum the original plant design, fabrication, crection and test / inspection requirements. The pennanent changes are downstream of the normal SWS isolation valve and will not affect nuclear safety or safe plant operation.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
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Safety Evaluation No.: 3283924Hil, no Modification: Cation Tank Underdrain Replacement The Neva-Clog screens used in the seven Condemins, Cation Tank and Final Rinse and Storage Tank have been known to fail at other plants as well as one failure in November,1994, in the Cation Tank at Oyster Creck.
Therefore, a new screen retention underdrain system has been designed. This new underdrain system will be installed by this modification into the Cation Tank. This modification to the Cation Tanks will 1) provide experience in the installation of this new underdrain design which will be proposed for the seven Condemins in an upcoming outage; and,2) provide better removal of resin fines (under 300 microns) in order to obtain more assurance that the Condemins post strainer (approx.160 microns) will catch all the resins should a Neva-Clog screen fail in a Condemin - thus protecting the Reactor from a very large conductivity spike. The new materials being added to the Cation Tank have been tested by GPUN, Reading Labs, and determined to be acceptable for this application.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 328312-002, R2 Modification: Change Overload Heaters Required by Calculation This safety evaluation addresses the changing of thermal overload heaters (TOLs) recommended by Calculation Cl302-730-5350-005, R9/"O/L Heater Sizing for NSR OCNGS GL 89-10 MOVs" Changing of the thermal overload heaters for MOVs V-16-0002, V-20-0003, V-204X)o4, V-20-0032, V-20-0027, and V-20-0033, will not degrade the abilitics of the MOVs to perform their safety functions ornffect safe plant operations. No modes of operation will be affected as a result of these changes. TOL sizes were calculated per an approved engineering standard and the results were design verified. Therefore, changing the TOLs is considered to be acceptable.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 328373-002, R0 Modliication: Chem Waste Monitoring This modification will installinstrumentation to monitor the contents of the Chem Wastc/ Floor Drain Collection i
- " tanks. This instmmentation will proside information necessary to support processing decisions and to determine l
the amount of coagulating agent to add to the collection tanks; Modifications will also be performed to increase the reliability of the sample lines and to improve the sample sink itself.
This modification is contained completely within the NRW building and has no impact on nuclear safety or safe plant operations. This evaluation is required for the sole reason that the detailed system description of Process Sampling in the UFSAR will require revision on completion of the modification.
It is therefore, concluded that this safety evaluation has detennined that there is no adverse effect on nuclear safety and/or safe plant operations, increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety presiously evaluated in the SAR, creation of the possibility for an accident or malfunction of a different type than any presiously identified in the SAR or, decrease in the margin of safety as defined in the basis of any Technical Specification. Since this safety evahiation has determined that no unresiewed nuclear safety question has been created and that no environmental impact is involved, this assessment is acceptable.
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Safety Evaluation No.: 403070-001, R0 Modification: Chlorination System Upgrade The change involves removal of the hypochlorite metering pumps and the installation of eductors. This will be the new method of chlorine injection into Service Water, Emergency Senice Water and Circulating Water Systems.
Chlorine is injected into these systems to control befouling of the heat transfer surfaces of the three main condensers, RBCCWfrBCCW heat exchangers and the Containment Spray heat exchangers through the Circulating, Senice and Emergency Sonice Water Systems. The balance of the change suppons the educator installation. This system has no direct interface with nuclear safety related equipment. Senice/ESW Chlorination is classified as *Other" because it is outside the NSR ESW check valves V-3-131/135.
There are no unresiewed safety questions resulting from this modification. No Technical Specification change is required and these changes can be implemented under 10CFR50.59.
Safety Evaluation No.: 403049-001, R1 Modification: Circulating Water Pipe Replacement at intake Structure General corrosion has caused pipe wall thinning on the discharge piping of each CW pump. In early cycle 15, a through wall leak developed on the 78" CW pipingjust downstream of valve V-3 11. This is the butterfly valve on the discharge of the 1-4 CW pump. The location of the leak is on a portion of pipe that is approximately 6" long just before it penetrates the intake tunnel wall. Additional through wall leaks subsequently occurred in the same general area. During cycle 15, a temporary encasement was installed around the 78" spool piccc that was leaking
-(Reference 2.2.15 and 2.2.16). This temporary repair was designed to last until 16R outage. In addition, pressure switch PS-119 and associated valves were removed to accommodate installation of the pipe encasement.
Associated circuitry and control room alarms were disabled as wcll as the input to the sequence of events recorder.
The scope of this safety evaluation is to evaluate the replacement of the 78" piping between the CW pumps and the discharge butterfly valves. The replacement is inclusive of the pump expansion joints for pumps 1-1,1 3, and 1-4 Valves V-3-8 through V-3-Il will be procured per BA 408957.
The scope also includes evaluating the replacement of piping from valve V-3-11 to and within the intake tunnel wall. A pipe tap and root valve will be installed on this segment of pipe. This will allow re-installation of PS-119 if future CW Pump / System changes make this alarm function necessary. In addition, piping downstream of valves V-3-8. V-3-9 and V-3-10 (including tunnel penetration) will be inspected and repaired as required.
This safety evaluation has determined that this modification does not (1) adversely affect nuclear safety and/or safe plant operations (2) increase the probability of occurrence or the consequences of an accident or malfunction of
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equipment important to safety presiously evaluated in the SAR, (3) create the possibility of an accident or malfunction of a different type than any previously identified in the S AR or (4) decrease the margin of safety as defined in the bases of any Technical Specification. Since this safety evaluation has determined that no unresiewed nuclear safety question has been created and that no environmental impact is involved, this modincation is acceptable.
The CW System does not have any safety functions. This configuration change will not impact the operation of the CW System or any other system.
Safety Evaluation No.: 000535-010. R2 Modification: Circulating Water Sptem Starting Imgic The purpose of this modification is to proside a mechanism by which the circulating water pumps can be restarted in a controlled manner after trip. The existing configuration, following a loss of offsite powcr, will cause the pumps to restart automatically w hen power is restored. This is a challenge to the operators because of the potential to trip the bus uhen all the operating pumps restan simultaneously as the power is restored.
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. Safety Evaluation No.: 000335-010. R2 (Cont'd.)
An interlocking relay will be added in the control circuit of the discharge valve for each pump. The purpose is to prevent the automatic restart of the pumps. The physical location of the four new relays will be in the control room i
panel 5F/6F where other safety related wiring and components exist. This safety evaluation addresses three specific requirements: 1. Electrical svaration betwecn IE and non lE circuits. 2, Scismic anti-falldown to prevent impact on operation of other safety components in the vicinity; 3, Diesel Generator loading.
The proposed modification incorporates new control relays in the circulating water pump staning logic. These relays implement a necessary design feature: The prevention of automatic pump restart following a loss of ofTsite l
power.
The circulating water system does not perform any safety function and is not required during any accident
' scenario. Safe plant operation will not be affected in the event of malfunction of any equipment added by this modification. Necessary controls have been placed on the installation of this modification. The relays are to be mounted in control room panel 5F/6F. There are safety-related components in this panel. Scismic requirements j
are that the equipment maintain passive integrity in a seismic event. This safety evaluation has determined that there is no unreviewed safety question or environmental concern involved with this activity.
i Safety Evaluation No.: tHH1106-002, R2 Modification: Condenser Bay Permanent Scaffolding w*<m s
- During refueling outages a great deal of maintenance is performed on equipment (i.e., valves) in the vicinity of the
" moisture separators, reheater, feedwater heaters, and the condenser water boxes. Most of this maintenance work requires the crection of scaffolding, in past outages this scaffolding has been crected prior to work and then removed after work is complete. This erection and removal takes time and resources.
This safety evaluation documents justification for scaffolding in Condenser Bay to remain in place (with wood planking removed) during the operating cycle. This would allow easier and quicker access for future maintenance on this equipment. The scaffolding, w hich will be left in place, will become a pennanent plant configuration.
Resision 2 of this safety evaluation documentsjustification for permanent installation of 24 scaffolds (including fiberglass grating) that were erected in the Condenser and Heater Bay areas of the Turbine Building during 16R outage and labeled as permanent scaffolding.
The structural design of this modification is in compliance with all applicable safety requirement codes and regulations and will not affect the safety function. This safety evaluation has determined that this modification a does not (1) adversely affect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important to safety presiously evahtated in the SAR, (3) create the possibility for (a) an accident or (b) a malfimetion of a different type than any presiously identified in the S AR or (4) decrease the margin of safety as defined in the bases of any Technical Specification.
Since there is no adverse affect on nuclear safety and/or safe plant Operations, no creation of an unresiewed nuclear safety question and no environmental impact, this modification is acceptable Safety Evaluation No.: tH)0212-034 R0 Modification: Core Spray Pipe Support Addition The purpose of this safety evaluation is to evaluate the configuration change (support addition) to the Core Spray drain line connecting to the 12" main line off pump NZ01-A.
f This safety evaluation has determined that this modification does not (1) adversely affect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important to safety presiously evaluated in the SAR, (3) create the possibility for (a) an accident or (b) a malfunction of a difTerent type than any presiously identified in the SAR or (4) decrease the margin of safety as defined in the bases of any Technical Specification.
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j Safety Evaluation No.: 000212-034, R0 (Cont'd.)
Since there is no adverse affect on nuclear safety and/or safe plant Operations, no creation of an unresiewed nuclear safety question and no environmental impact, this modification is acceptable Safety Evaluation No.: 403011-001, R5 Modification: Core Spray Pumps Recirculation Line Upgrade This modification relocates the core spray pump minimum flow recirculation lines to a higher elevation such that l
an overflow path is created for the fill pumps while maintaining the minimum flow capability for the main and
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booster pumps when they operate. The existing recirculation piping and valves V-20-92,93,94, and 95 shall be
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removed with this modification. The valve interfaces both electrical and mechanical will be altered, deleted or capped to facilitate the new installation. This is necessary to mitigate core spray system piping drain down prior to
' ' system initiation when a small break loss of coolant accident (SBLOCA) coupled with an open failure of valves V-l 20-92,93,94, and 95 as a result of a loss of offsite power (LOOP) or single failure occurs. Drain down causes L
system voiding which can lead to water hammer upon pump start. A water hammer and resulting scismic event 3
may render damage to the system and preclude the system from performing its safety function. This concern is j
y possible with the existing configuration and has previously been identified in PSC 90-005. Locating the l-rceirculation lines to a higher elevation and establishing an overflow path for the fill pumps assures the main piping remains full such that voiding is climinated.
b This modification deletesthe existing recirculation line piping and valves along with the associated instrument air Rand control interfaces. This eliminates the drain down concern identified in PSC-005 and assures the main piping E
%cmains full, eliminating voiding and water hammer that could occur with the existing configuration and climinates the structural deficiencies identified in Desiation Repons94-043 and 94-160. 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 manual vacuum breakers on System I and 11 enhances the Operator's ability to recognize and detect fill pump failure by assuring that the upper portions of piping are relieved of vacuum conditions. The vacimm breakers and associated
- consequences of their installation have been evaluated and determined not to have a negative impact on Core Spray System I and II, primary containment integrity, or other safety systems. This safety evaluation has determined that this modification does not adversely affect nuclear safety and/or safe plant operations; increase the probability of occurrence or the consequences of an accident or a malfunction of equipment important to safety previously evaluated in the SAR; create the possibility for an accident or a malfunction of a different type than any previously identified in the SAR or, decrease the margin of safety as defined in the bases of any Technical Specification.
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- Since there is no adverse affect on nuclear safety and/or sa[ plant operations, no creation of an unreviewed nuc safety question and no emironmental impact, this modification is acceptable.
Safety Evaluation No.: 408858-001, R1 Modification: Cover Plates for Drywell Seal Bulkhead Platform The purpose of the modification is to replace the existing carbon steel seal cover plates at elevation 94' 8" in the drywell seal bulkhead platform with aluminum plates. Vertical handles for hand canying will replace the existing horizontallifting lugs. With the lighter material used and the handles prmided the cover plates can be easily
" moved by hand instead of using the overhead crane. As a result of this improvement, the radiation exposure to the workers will be reduced and the leakage probicm will be reduced during outages.
The cover plates are watestight and designed to resist the water pressure plus the associated design scismic load during refueling outages when the reactor cavity is flooded. They will be removed and stored in a secure location before plant restart. Water lost due to smallleaks will be made up from the condensate transfer system.
The proposed modification will not reduce the performance of the affected systems, affect the safety functions of these systems, increase the probability of occurrence or consequence of an accident, create a possibility for an i-accident, or malfunction of a difTerent type, decrease the margin of safety as defined in the bases of Oyster Creek Technical Specifications, violate any licensing requirements, cause a radiological concern, or afrect the environmental permit.
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f Safety Evaluation No.: 403060-001, R1 Modification: Descrating Steam System Retirement The Deaerating Steam System was intended to assist in the condenser deacration process when the hotwell condensate is below saturation temperature. This was to be accomplished by diverting steam from the 20 inch Bypass Steam header directly to the Main Condensers. The system has never been needed and consequently has never been used. The work being performed will isolate and remove the Deacrating Steam System from senice.
The isolated piping, valves, and instruments will be abandoned in place.
The Deaerating Steam System from the Main Steam Line up to and including valves V-1-0047 V-1-0051, V 0055, V-10376, V 1-0151, V-1-0152, and V-10153 falls within the scope of the ASME Section XI Code, Class 2.
Consequently the ISI program will have to be resised to delete the Deacrating Steam line and vah es from the program.
This modification physically separates a system that has never been used from the primary plant systems.
This safety evaluation concludes that because this activity does not involve a significant increase in the probability or consequences of accidents presiously considered and noes not involve a significant decrease in a safety margin, this activity does not involve a significant hazards considerations; there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and plant operations will be conducted in compliance with the NRC's rules and regulations. Therefore, an unreviewed safety question does not k.
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" Safety Evaluation No.: 328383-001, R0 Modification: DG Air Plenum Drain The Diesel Generator (DG) air plenums are susceptible to rain water intrusion. The rain water intrusion comes from the engine cooling radiators in the ceiling area to the plenum area. The rainwater that enters the plenums currently drains along the DG skid floor and the beams, which has created a corrosion problem. In order to mitigate the corrosion problem and the maintenance associated with it, drainage pans will be installed at the bottom of the plenums.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
$afety Evaluation No.: 403024-003, R0
' Modification: Diesel Generator Alternate Fuel OilSupply Line The purpose of this safety evaluation is to evaluate the configuration change that resises the alternate fuel supply line in the Main Fuct Oil Storage and Transfer System. The aforementioned line is routed in close proximity to the Diesel Generator switchgear unit. The concern is that a seismic crent may cause this pipe to bang the switchgcar unit and have an adverse affect on the function of the switchgear internal components. The configuration change brings the 2-inch pipe up approximately l'3" and then comes down again to meet the original pipe. This was don to avoid an obstruction from the concrete wall and allows adequate distance from the switchgear unit. This pipe is protected by an angle iron shield located just above. The configuration change incorporates a similar shield design.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety de the Technical Specifications. Since this safety evaluation has determined that no unresiewed nuclear safety question has been created and that no emironmental impact is involved, this modification is acceptable.
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4 Safety Evaluation No.: 408925-001, R1 Modification: Drywell High Level Pressure Indication During Containment Flooding l
' During an unisolable LOCA below the Top of the Active Fuct (TAF) (RB, El. 69'2"), the Emergency Operating l
Procedures (EOP's) require primary containment floodup above TAF level but below the level of maximum venting l
capability. EOP EMG-3200.02 requires a drywell pressure reading at an elevation greater than TAF as part of l
calculating primary containment water level during containment floodup. There are no remote daywell pressure indications at levels greater than RB, El. 48'. Therefore, an operator must be sent to the H:02 cabinet (RB, El. 75')
l to obtain a local drywell pressure reading from a drywell penetration at El. W. During an accident conditions in the area would preclude sending anyone to this location. Without the drywell pressure indication, cutainment water level cannot be accurately calculated. In addition, NUREG-1358, Supplement I cautions specifying the use of equipment not available during an accident scenario that is addressed within the EOP's.
71us modification shall construct a new instmment loop that will provide drywell pressure indication at a level greater than TAF to the Control Room. The reading will be obtainable during an accident, senice the purposes of Emergency Operating Procedures and climinate the concern within NUREG-1358, Supplement 1.
A new drywell pressure indication at a level above TAF with provision to indicate at a remote location improves the computational ability of containment water level. Having a drywell level indication available during the EOP's, climinates any weakness within the procedures as cautioned by NURREG-1358, Supplement 1. The proposed modification has been evaluated and determined not to represent an unresiewed safety question as defined in 10 CFR 50.59. The modification does not have any adverse efTect on nuclear safety, safe plant operations, or the environment.
y Safety Evaluation No.: 328323-003, R0 Modification: EDG Louser Control System Upgrade This modification will replace the existing EDG #1 and #2 louver blades, bushings (as necessary), and actuating arms. The current control system shall be replaced with a split blade control system. This will include the replacement of all louver blades, bushings, actuators arms, turnbuckles, and lever arm assemblics. This configuration change does not change out the actuator. However, it may be necessary to reposition the actuator to install the new split motion actuating arms.
In addition, the 1 1/2" tube steel cross ties which are located diagonally on the east and west side of both EDG hoods will be removed, and reinforcing legs shall be installed under the EDG hoods. This work will be performed prior to the louver work and is necessary to climinate the interference that the crosstics create.
This configuration change uill not change or affect the electrical portion oflouver control system that provides control signal to the louver actuator.
This configuration change will also install a cable-extension position transducer that will proside position indication to the EDG Data Acquisition System. The scope of this activity shall be to mount the transducer, physically connect the transducer cable (lanyard) to the louver actuating arm assembly pin, and terminate the transducer to the EDG Data Acquisition System.
This safety evaluation concludes that because this activity does not involve a significant increase in the probab or consequences of accidents previously considered and noes not involve a significant decrease in a safety m this activity does not involve a significant hazards considerations; there is reasonable assurance that the heahh a safety of the public will not be endangered by operation in the proposed manner; and plant operations will be conducted in compliance with the NRC's rules and regulations. Therefore, an unresiewed safety question does not exist.
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Safety Evaluation No.: 328383-002, R3 Modification: EDG Relay Replacement / Upgrade This modification serves to upgrade the aging components with new model types having a high degree of longevity, sturdiness, repeatability, and availability. The layout of the EDG cabinets will be rearranged to create more panel space, minimize wire bundle size, and make the EDG #1 and EDG #2 tmits physically more similar.
Control logic enhancements will be made to improve EDG control. More data inputs to the EDG Data Acquisition System (DAS) will be provided. Finally, this modification addresses the seismic qualification requirements of several relays in the EDG system for resolution as identified in the NRC's Unresolved Safety Issue (USI) A-46.
The modif5 cation increases EDG reliability with an upgrade in the material condition of the EDG system. The modification also achieves gains in EDG control without changing existing operator procedure. The modification also creates a means for additional system performance monitoring. The modification meets these objectives without constituting an unreviewed safety question as determined by 10 CFR 50.59.
Safety Evaluation No.: 403057-001, R0 Modification: Emergency Diesel Generator (ED/G) Day Tank Level Snitch Replacement This safety evahiation shall address the replacement of the existing obsolete level swiths and the upgrade of associated components for each ED/G fuel oil day tank.
The existing fuel oil day tank level switches are obsolete and there are insufficient sparc switches available. The existing installation does not easily allow for maintenance of the switches because of their location. This modification shall install new switches and relocate the level switch mounting from the side of the tank to the top.
Because of the current draw of the primary and backup fuel oil transfer pump motors starters, the starters shall be replaced to prevent damage to the level switch contacts. The removal of the existing motor starters will require the installation of new motor overload protection devices.
This safety evaluation has determined that this modification does not (1) adversely affect nuclear safety and/or 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 S AR, (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 defined in the basis of any Technical Specification. Since there is no adverse affect on nuclear safety and/or safe plant operations, no creation of an unresiewed safety question, and no environmental impact, this modification is
, acceptable.
Safety Evaluation No.: 000215-015, R0 Modification: Electrical B3 pass of V-16-0001 and V-16-0061 TOL Contacts Deviation Report (DVR) 96-1097 identified. " Break flow from a 6 inch line break in the Reactor Water Cicanup System is now thought to be higher than previously analyzed." The impact of this evaluation requires review of the Environmental Qualifications of the safety-related equipment in the Reactor Building. This detailed evaluation remains ongoing, however, the thermal overload (TOL) relays for two isolation valves (V-16-1 and 61) have been identified as restrictive components due to the increased ambient temperature. These two valves are required to isolate this specific event, so although the DVR is still under evaluation, it is believed prudent to bypass the thermal overloads for these two MOVs.
TOLs are protective devices that sense the current flow of the motor. The motor current heats the TOL element which can interrupt the circuit contacts (disconnect the motor from power) at a predetermined point. Since TOLs are heat-sensing devices, changes in the ambient temperature have an impact on their settings.
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Safety Evaluation No.: 000215-015, R0 (Cont'd.)
The redundant isolation valve to V-16-1 is V-16-14, the change in the Reactor Building environment is not expected to affect this valve's performance. Note: its overloads are currently bypassed. V-16-2 Inlet isolation valve to the Cleanup Auxiliary Pump, in series with V-16-1 is closed during normal operation which is its safety position. The redundant isolation valve to V-16-61 is V-16-62, check valve inside the drywell, this new condition does not impact its operation. Bypassing the subject valves TOL during plant operation ensures the TOL does not jeopardize the valves safety function. The TOL bypass will be removed during periods of surveillance, periodic, or maintenance testing.
This modification does not adversely efTcct nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 402901-003, R4 Modification: Feedwater and Recirculation Flow Control Systems Upgrade The Digital Feedwater Control System (DFS) portion of the DCS design will provide control and monitoring capabilities similar to the existing control system including automatic reactor water level control, feedwater pump runout protection flow control, automatic reactor level setdown, and steam leak detection. Changes in the system hardware design will be the replacement of related control room located operator interface devices (M/A stations, indicators, control switches, electronic modules, indicating lights) and field instrumentation (transmitters,1/P converters) as further d Pucd in GPUN Modification Design Description MDD-OC-625Bk. Feedwater system final drive elements (control valves) will not be replaced. The DFCS will regulate feedwater flow by controlling valves associated with the feedwater trains including the Main Feedwater Regulating Valves (MFRV), the Low Flow Regulating Valves (LFRV), and the MFRV Block valves. The installing of the MFRV Block Valves by MDD-OC-422A and the installation of the total Feedwater Flow Element by OCMM-402945-001 was performed during the 13R outage. The new DFCS will provide signals to the existing plant monitoring computer system. In addition, a new third reactor level signal is required for validating the A and B re. ctor :evel signals. This new signal will serve as a "referce" signal and will be installed by this modification. Th, %cdwater pump controls, minimum feedwater flow controls, and heater string outlet valves are not included in the DFCS. The reactor level setdown logic is resised to Post Scram Level Control (PSLC) logic.
The new DCS requires two separate power sources. Power from existing 120 VAC vial power panel CIP-3 will be used as one source. The second source will be from 120 VAXC vital power panel PSP.
This modification does not pose a safety concern or unreviewed safety question. This modification interfaces with safety related components through devices that do not contain digital electronics such as fuses and circuit breakers.
There is no software interface bctween this modification and existing plant safety systems. Additionally, no new safety related digital components are installed. Although some existing safety related components are being l
relocated. functionally they are unchanged. There is no impact on safe plant operation due to the relocation of the i
safety-related equipment not related to DRFCS. New failure modes are identified and addressed, as is the justification for reduced potential plant transients attributable to feedwater system failures.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 402991, R0 Modification: Fire Protection System for Heater Bay Roof Outage Support Buildings This modification installs dry pipe automatic sprinkler fire protection in three 14' x 24' prefabricated buildings and j
I one 8' x 10' vestibule on the Heater Bay Roof of the Turbine Building at El. 46'6". These structures are permanent additions to the Turbine Building that serve as an assembly area for personnel working in the Turbine Building during major outages. They also house computer and electronic equipment to control personnel access into the j
building and to monitor outage work on the Turbine operating floor. Only the Fire Protection Mem, its water i
supply and supervisory circuitry portions of this modificat on are discussed in this safety evaluation.
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No significant changes are made to the Fire Protection Water System or the non-vital electrical distribution system.
both systems will function the same after the modification is installed. This evaluation precludes the potential for an unreviewed safety question or the need for a change to the Technical Specifications.
Safety Evaluation No.: 403079-001, R3 Modification: GL 89-10 MG Set harger B Autostart l
The purpose of this modification is to auto start the M-G Set Charger B thereby providing greater available valve actuator motor torque by increasing the voltage available at the motor terminals. This will ensure that the valve completes its isolation function during design basis events.
This modification. improves the ability of the MOVs to isolate HELB and the corresponding reactor response. The auto starting of the M-G Set Charger B will allow credit to be taken for the higher voltage supply of the DC BUS B r under loss of offsite power conditions. Modifications to bus undervoltage alarms and voltmeters will enhance operator response to degraded conditions of the bus during normal operation.
This modification converts the M-G Set B start from manual to automatic upon Emergency Diesel Generator start in order to achieve higher DC-B bus voltage during LOOP events. The higher bus voltage which results in higher available motor torque and maintains valve stroke time. This will ensure that these MOVs complete their isolation function during the design basis events. Changes to alarms windows and placement of voltmeters will reduce operator burden and enhances system perfonnance monitoring.
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- This modification does not have any adverse effect on nuclear safety or safe plant operations of the environment.
The modification does not constitute an unresiewed safety question as determined by 10CFR 50.59.
J Safety Evaluation No.: 403079-002 R1 Modification: CL 89-10,16R Mods - Cable & Conduit The purpose of this modification is to provide greater available valve actuator motor torque by increasing the voltage available at the motor terminals to assure the valve completes its safety function during design basis events.
This modification does not introduce any new accident or malfunction not presiously evaluated nor does the modification increase the likelihood of occurrence or consequences of any accident as analyzed in the UFSAR.
This modification does not decrease the margin of safety as described in the Technical Specification because the modification does not impact any system safety functions. This evaluation concludes there is no unreviewed safety question per 10 CFR 50.59 and no changes to the OC Technical Specifications are required.
Safety Evaluation No.: 403088-001, R0 Modification: IL P. Turbine Shield Wall The two pre-fabricated shield walls being placed around the H. P. turbine will proside radiation shiciding to personnel in the area. The shield structures are designed to withstand their dead load during lifting, placement, as well as during operation. The Turbine Pedestal floor capacity to withstand the shield load has been verified. There is no impact on any safety-related equipment or structures because:
The implementation of this modification will not adversely affect nuclear safety or safe plant operation. No changes in plant equipment and plant operating procedures are required due to the shiciding installation. The probability of occurrence of the consequences of an accident or malfunction of equipment important to safety presiously evaluated in the Safety Analysis Report is not increased because this modification does not interface with any NSR equipment. The possibility for an accident or malfunction of a different type than any evaluated presiously in the Safety Analysis Report is not created. The Margin of Safety as defined in the basis for any Technical Specification is not reduced. Any equipment or piping damage that could be caused by a collapse of the shields is enveloped by presiously analyzed design basis accidents. The shields are designed to the same criteria and allowables as those used to design the Turbine Building and the subject equipment. The plant Technical Specification is not impacted due to the fact that the shiciding does not afTect the integrity and operation of equipment covered in the Technical Specification. This modification does not impact any license requiremen regulations. No changes are being made to the function of any plant equipment and no revision to plant operat will result.
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Safety Evaluation No.: 402840-001, R7 Modification: Hydrogen Water Chemistry Sptem and Generator Cooling Sptem This modification will provide a permanent Hydrogen Water Chemistry System to mitigate Intergranular Stress Corrosion Cracking in the reactor piping during all reactor power levels at which the system is operating. This system will provide a capability in accordance with the EPRI Report, " Guidelines for Permanent BWR Hydrogen Water Chemistry Installations, " 1987 Resision, which will enhance the plant availability without any safety concerns. In addition, this modification will provide a safe and reliable supply and makeup of hydrogen to cool the main generator.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000561-008, R0 Modification: Installation of PVC Liner in TankT-33-002 This safety evaluation has been completed for the installation of a 3/32-inch (94 mil) PVC liner into Sodium Hypochlorite tank T-33-002, install PVC liners through the 3 inch overflow nonic, pump suction nonic and drain noule, and hang the PVC liner in the tank T-33-002. It has been shown that PVC is inert to 12-15% sodium hypochlorite and the PVC liner should last to end of plant life.
This modification does notadversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000611-001, R0 Modification: Installation of RAC Work Station in the Control Room A Radiological Assessment Code (RAC) workstation located in the Computer Room is relecated to the Control Room. it will be installed between pancis 15R and 17R. The RAC workstation consists of a computer, monitor, printer and modem secured inside a workstation (cabinct). The workstation will be anchored to the concrete floor.
This safety evaluation will assess the effects (if any) on Nuclear Safety of the completed installation of the workstation in the Control Room.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) 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 environment.
Since there is no adverse afTect on nuclear safety and/or safe plant operations, no creation of an unresiewed nuclear safety question and no emironmental impact, this installation is acceptable.
Safety Evaluation No.: 408975-001, R0 Modification: Installation of SEG Supplied RW Equipment This Configuration Change provides permanent equipment to:
Dewater spent resin and filter media in preparation for disposal. The portion of this safety evaluation 1.
associated with the dewatering equipment is limited to its use in the New Radwaste Building (NRW) truck bay; Perform ultrafiltration (UF) on the water removed from the resin or filter media. The portion of the safety 2.
evaluation associated with the UF equipment is limited to its use in the NRW fill aisic; Collect the concentrates from the UF equipment in a tank for further processing. The portion of this 3.
Safety Evaluation associated with the concentrates equipment is limited to its use in the NRW fill aisle.
The new equipment is to functionally replace existing ultrafiltration and dewatering units at Oyster Creek. This configuration change will not adversely affect SAR or Technical Specification safety margin (s), nucicar or plant safety, radiological safety, the emironment, or the probability or consequences of an S AR evaluated accident. It will not create a new SAR related accident. The configuration change does not violate license requirements. This configuration change is safe and may be implemented.
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Safety Evaluation No.: 000641-020, R1 Modification: Isolation Between Scram Brush Recorder and RPS I
Material conformance report MNCR 97-0026 is issued to identify lack of fusing of the scram brush recorder I
associated with Reactor Protection System (RPS). The MNCR also highlighted discrepancy of this brush recorder circuitry not reflected on plant as-built drawing. The purpose of this configuration change is to resolve issues identified on the subject MNCR.
This modification adds a class IE fuse in the RPS to isolate scram rod test circuit from RPS. This modification will prevent false 1/2 scram in Group 1 and prevent fire in CR Panel 6R due to a fault in the test circuit.
This modification does not adversely efTect nuclear safety or the environment. This modification does not involve
'any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 402986-007, Rt Modification: Isolation Condenser Makeup Vahe I use Installation This safety evaluation addresses a modification to the Isolation Condenser NE-01 A and NE-01B makeup valves V-11-0034 and V 1-0036 indication and alarm circuits. This modification will increase the potential of the makeup valves to perform their function from the control room and at the same time preserve the funct on of V-11-0034 at i
the 460-Volt switchgearroom remote shutdown p:mel.
The new configuration will not change the existing safety function of the makeup valves. Because the margin of safety was not changed, performance of this modification will not adversely affect 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.
Safety Evaluation No.: 000616-003, R0 Modification: Main Battery Chg. "h" Trip Alarm Modification The alarm window U-2-e engraved "Batt. Chg. B Trip" annunciated when Run/Off control switch (8F-185) is placed in Off position. This is not acceptable to Operations. The purpose of this configuration change is to modify the alarm circuit to prevent trip alarm when control switch is in Off position.
This modification replaces control switch to obtain required contacts development for modification to the alarm circuit. It modifies alarm (U-2-c) circuit to prevent alarming w hen control switch is in Off position.
v This modification does not adversely efTect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 409791-001, R1 Modification: OC Security Intrusion Protection Upgrade The Nuclear Regulatory Commission (NRC) has amended 10 CFR Part 73, " Physical Protection of Plant and Materials," to include the use of a four wheel drive land vehicle by adversaries for transporting personnel and their hand-carried equipment to the proximity of vital areas and to include a land vehicle bomb. The amendment requires the installation of vehicle control measures, including vehicle barrier systems (VBS) to protect vital equipment from damage due to a design basis explosion (DBX) at the point of vehicle denial. These two threats are considered separately.
The proposed modification is required by 10 CFR 73. During the evaluation for the implementation of this regulation, Regulatory Guide 5.68 and NUREG/CR 6190 are followed. The safe shutdown systems required for protection follow safe shutdown for Loss of Offsite Power. The resulting modifications are shown in Drawing 3 120101-003 and are summarized as follows:
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Safety Evaluation No.: 409791-001, R1 (Cont'd.)
Supplement existing gates with hardened gates powered by existing security gate power.
Supplement existing roll up door in the warehouse with a hardered rising bollard system powered by existing security door power.
Place rigid passive barriers in proximity to the outer nuisance fence except at locations uhere the Warehouse, North Gate Guard House, Intake Canal, Low Level Radwaste Storage Facility, and Discharge Canal are utilized as part of the barrier.
These proposed modifications do not interfere with station safety functions directly or indirectly and they improve the security function by preventing unauthorized vehicle intrusion into the protected area. In addition, these l
proposed modifications provide protection to safe shutdown components against the effects of the DBX. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 315302-075, R0 Modification: Partial Removal of Loop Discharge Bioshield Doors The purpose of this activity is to minimize personnel radiation exposure and streamline operations in the Drywell The bioshield penetrations for the Recirculation Loop discharge piping uses 18 layers of 1/2-inch thick steel plate shielding. As demonstrated by this safety evaluation, only one of the half-inch layers is needed and the others will
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not be re-installed after removal for wcld inspections in 16R. This will reduce personnel radiation exposure and support scope for future inspections.
The bioshield wall is classified "NSR" but the doors do not fulfill any "NSR" requirements. This change removes shielding and resultant radiation fields will be within acceptable levels. The outer 171/2 inch layers are not needed for shiciding, drywell temperature control or alTect bioshield structural integrity. The single layer of 1/2 inch plate that is reinstalled, is attached to the bioshield wall in the same manner as before and poses no ad missile risk. The doors are not mentioned in the Technical Specifications.
Since there is no adverse afTect on nuclear safety and/or safe plant operations, no creation of unresiewed nuclear safety question and no environmental impact, this modification is acceptable.
I Safety Evaluation No.: 403044-001, R0 Modification: Platforms for C.S., RBCCW and a TBCCW IIcat Exchangers The purpose of this modification is to design and install platforms that will be used to perform maintenance t on the Containment Spray, RBCCW, and TBCCW heat exchangers. In the present configuration, whenever end cap removal, tube cicaning and/or anode replacement maintenance is to be performed, platforms are cre after completion of the maintenance task, they are removed. Installing and removing platforms is costly an consuming and often causes delays in restarting the systems. Installing permanent platforms, at each end o horizontal heat exchangers and at the top end of the vertical heat exchangers, will reduce cost and time for the above activities resuhing in timely restart of the out of service systems. In addition, having the platforms will reduce possible violations of Technical Specification clocks for the Containment Spray system and improv performance and availability of the Containment Spray, RBCCW and TBCCW systems.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 328333-001, R1 Modification: Reactor Fuel Zone Level Temperature Instrument Loop The purpose of this modification is to streamline and upgrade the circuits and electronics for the fuel zone temperature monitoring loops. This will improve the system maintainability and reliability and reduce th out of senice time.
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Safety Evaluation No.: 328333-001, R1 (Cont'd.)
This safety evaluation concludes that the proposed modification will not have any adverse effect on safety or the emironment and does not constitute an unresiewed safety question as specified in 10 CFR 50.59.
Safety Evaluation No.: 000641-018, R0 Modification: Reactor Protection System Separation Improvements This configuration change Mms to improve separation between the Reactor Protection System (RPS) control rod scram load groups' circuitry, control rod groups I through 4. A hot short that simultaneously contacted conductors from redundant load groups in the Control Room pancis or interconnecting cables could disable the scram function. The probability of this scenario is heightened by the extreme proximity of the redundant load group circuits in Control Room Panels 4F,6R and 7R. The existing configuration does not violate the single failure criterion as defined in OCNGS licensing documents, but this modification is undertaken to improve upon this configuration. This configuration change impacts the RPS scram pilot solenoid status indication lamp circuits in Pancis 4F,6R and 7R. These indication circuits are presently wired in parallel with each of the RPS Trip Systems' scram solenoid load groups. The new fuses will be installed in Pancis 6R and 7R. This modification individually fuses cach of the scram pilot solenoid status indicating lamps on Pancis 4F,6R and 7R.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications and/or (5) the environment. Since there is no adverse affect on nuclear safety and/or 5
safe plant operations, no unresiewed safety question or emironrnental concern is introduced.
Safety Evaluation No.: 402821-002, R0 Modification: Reactor Recirculation Pumps Vibration Monitoring System The purpose of this modification is to install a pennanent on-line continuous vibration monitoring system for all five RRP at OCNGS.
This modification does not adversely efTect nuclear safety or the emironment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000622-016, R0 Modification: Reactor Vessel Thermocouple Abandonment j
The purpose of this safety evaluation is 19 provide justification for abandoning Reactor Vessel thermocouple that are currently inoperable. The non-working thennocouples identified in the FSAR, Table 7.6-3, covered by this safety evaluation are as follows: TE-15, TE-16, TE-19, TE-21, TE-27, TE-28, TE-31, TE-32. TE-34, TE-37, TE-
- 39. A!! the mentioned thennocouples are connected to recorder TR-I A02 in the Reactor Building, El. 51'-3" As discussed in the safety evaluation, these thermocouple can be abandoned since the function of these thermocouple is provided by the other re.naining thermocouple.
This safety evaluation has detennined that this modification does not (1) adversely affect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important te rafety previously evaluated in the SAR, (3) create the possibility for (a) an accident or (b) a malfunction of a different type than any presiously identified in the SAR or (4) decrease the margin of safety as defined in the bases of any Technical Specification.
Since there is no adverse afTect on nuclear safety and/or safe plant Operations, no creation of an unresiewed nuclear safety question and no environmental impact, this modification is acceptable l
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Safety Evaluation No.: 328395-001, RI Modification: Relocate Nitrogen Makeup I
In order to meet the requirements of Technical Specification 4.5.H the nitrogen makeup to the primary containment is monitored to ensure that the leak rate does not excessively degrade. The normal make-up flow is through line GN-1, a small amount of nitrogen also enters the drywell from pneumatic controls fed from nitrogen j
compressors. The nitrogen to the compressors is supplied from a branch of GN-1 upstream of flowmeter FE-010.
The flow to the compressors is not measured and is calculated using the compressor capacity and run times. The purpose of this configuration change is to relocate the nitrogen make-up flow orifice to allow the existing instrument to measure the nitrogen flow to the compressors. This will allow the tracking of nitrogen make-up to the drywell/ torus from one instrument that can be read from the control room. This change will increase the certainty and accuracy of the calculated primary containment leak rate.
This modification does not adversely efTect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 403045-002, R0 Modification: Remote Automatic Positioning System The purpose of this safety evahiation is to evaluate the Remote Automatic Positioning (RAF) System. This modification involves software and hardware enhancements to the current Oyster Creek Refueling Platform control system that will increascits cfliciency by providing semi-automatic positioning of the bridge and trolley. This will reduce the time required to move the fuel assemblies and climinate the time consumed during final positioning of the bridge and trolley for insertion or removal of fuct assemblies and blade guides. In addition, the safe operation of the system is enhanced significantly as this type of modification greatly reduces the probability of misplaced assemblics by reducing the dependency on each operator's efliciency and skill. These system enhancements will be l
designed to allow the Refueling Platform to receive move coordinates, one move at a time, directly into the Refueling Platform control system from a control room computer using software and hardware developed and supplied by GPU Nuclear.
Based on the results of the evaluation presented, the implementation of the Refueling Platform Remote Automatic Positioning System will ensure safe plant operation. All protective boundary zone functions currently provided by the Refueling Platfonn's programmable logic controller (PLC), as defined in GPUN Vendor Document #VM-RB-1710, Refuel Bridge Ladder Diagram Manual. will remain in place after implementation of the new PLC programming. The operator on the platform will continue to perform the same verification, fine position J
- adjustment, and grappling /ungrappling of the fuel, as presiously done. Manual override of all automatic function will be available at all times. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 000212-035, R0 Modification: Removal of Core Spray Drain Valves (V-20-47, V-20-17)
The purpose of this document is to evaluate the safety of removing valves V-20-47 and V-20-177 and the threaded 4x2 reducer upstream of them. A threaded cap will replace the threaded reducer. The valves are drain valves located on the suction line to Core Spray Pump (Core Spray System 1) P-20-001 A. Although the design temperatme for this pipe is listed as 350 degrees F the design temperature for the piping was reduced to 170 degrecs F. These valves will be removed to reduce the scismic loading on the piping. A support will be placed on the remaining pipe.
This modification does not adversely alTect nuclear safety or the emironment. No unreviewed safety question is generated by this modification and it can be implemented under 10CFR50.59.
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Safety Evaluation No.: 403063-001, R2 Modification: Removal of High Recirculation Flow Scram Function The purpose of this modification is to climinate the scram function on high recirc flow. This upscale function from the recirc flow monitoring electronics currently results in the following actions when recire flow exceeds 114% of rated: 1) reactor scram; 2) rod block; 3) rod block display (APRM INOP or FLOW BIAS); 4) alarm - FLO BI AS OFF NORMAL (G-5-f)
The modification consists of replacement of a trip status and reset module in each system and some minor rewiring within cach system. When the recirc flow menitoring electronics are modified, the upscale function will be retained. However, the setpoint will be reduced to 100% of rated, and the following actions will occur: 1) rod block; 2) rod block display (APRM INOP or FLOW BIAS).
The scram function will be removed since the Technical Specifications no longer require this function, and the alarm will be removed since flow above 100% of rated is not an off normal condition. This configuration will 4
climinate a current administrative setpoint for a rod block for recirc flows in excess of 15.7 x 10 GPM (98.125%
d of rated). This setpoint will also ensure compliance with Tech Spec 2.3.B for recirc Flows in excess of 61 x lo LB/hr (100% of rated).
This safety evaluation concludes that because this activity <locs not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, this activity does not imelvc<a significant hazards considerations; there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and plant operations will be conducted in compliance with the NRC's mies and regulations. Therefore, an unresiewed safety question does not exist.
Safety Evaluation No.: 000661-017, R2 Modification: Removal of Isocondenser Vent Rad & Containment Spray /ES Water Heat F.schanger Monitoring System The purpose of this safety evaluation is to evaluate the removal of the following two systems, Isolation Condenser Vent Radiation Monitoring System and Containment Spray / Emergency Senice Water Heat Exchanget Radiation Monitoring System, from the plant configuration as these monitoring systems presents the potential for operator confusion.
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l oThe scope of the modification for the Isolation Condenser Vent Radiation Monitor ng ystem nvo ves
' " disconnection and physical removal oflog ratemeters RN0007A3,4, 5 and 6 from Control Room Panel 2R and replacement with blank panels, disconnection of four (4) input signals to Recorder RN006B on Control Room Panel 10F, disconnection of signal circuitry to Alarm Window 10F-1 c ( A Vent Hi) and to Alarm Window 10F-2-e (B vent Hi) on Panel 10F and replacement with two (2) blank windows, and physical removal of all associated abandoned wiring within the Control Room. Detectors RE-RN0004A3,4,5 and 6 and their associated shielding and cabling are to be physically removed from Reactor Building Elevation 95'3" The scope of the modification for the Containment Spray / Emergency Senice Water Heat Exchanger Monitoring System involves disconnection and physical removal oflog ratemeters RN0040Al,2,3 and 4 from Control Room Panel 2R and replacement with blanks, disconnection of four (4) input signals to Recorder RN006B on Control Panel 10F, disconnection of input signals to Alarm Window 10F-4-g ( Arca/ Vent /Efil. Dnscl.) on Panel 10F, disconnection of signal circuitry to Alarm Window 10F-t g (ESW A/B Hi) and to Alarm Window 10F-2-g (ESW C/D Hi) on Panel 10F and replacement with two (2) blank windows, and physical removal of all associated abandoned wiring within the Control Room. Detectors RN0038Al,2,3 and 4 and their associated shiciding are to l
l be abandoned in place at Reactor Building Elevation 23'6",
The proposed changes will not have any adverse effect on plant safety, and do not represent an unreviewed safety question. The changes will however require a change to the Oyster Creek FS AR as described in Section 3.5.
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Safety Evaluation No.: 3283544H)2. R2 Modification: Removal of RPV Internals Vibration Brackets l
The purpose of this Safety Evaluation is to describe the efTect on safety, system perfonnance and plant design l
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 tic-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 detennining the vibration response characteristics of key reactor internal components and i
recirculation loops due to dynamic forces generated by the coolant flow and pump rotation. There is no need for the vibration system presently or in the future. Only the interfering brackets and conduits will be removed.
Conduits lef) in place will be cut at the nearest remaining bracket. The instrumentation brackets and associated j
conduits are non-safety related and do not support, directly or indirectly, any safety or operational plant function.
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 wcll as personnel performing the task will be qualified as to prevent loose pieces or damage to the shroud structure or the recirculation system or the reactor internals.
This safety evaluation has detennined that this modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications and/or (5) the environment. Since the safety evaluation has determined that no unreviewed nuclear safety question has been created, this modification is acceptable.
Safety Evaluatios No.: (HH12154116, R0 Modification: Removal of RWCU Flow Integrator FQ-lJ0061 The Reactor Water Cleanup (RWCU) System flow integrator (FQ-ljo061) has not been functional since refueling outage 14R. In addition. this flow integrator is introducing noise into the flow loop. This flow loop provides indication as well as control function.
This modification removes the flow integrator FQ-lJ0061 from the RWCU System flow loop. The existing flow recorder along with digital display will be used to monitor flow and system perfonnance. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
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' Safety Evaluation No.: 403065-001, R0 Modification: Replace A/B Battery Room Fire Protection System This configuration change will replace existing obsolete control panel and equipment in the halon fire protection system for the A/B Battery Room. The existing ionization detectors will be replaced with a combination of ionization detectors in one zone and photoelectric detectors in the second zone. This modification will serve to increase the reliability of the Fire Detection S) stem, which also serves to activate the halon suppression system for the A/B Battery Room, Tunnel and Electrical Tray Room in the event of a fire.
This modification does not adversely efTect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
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i Safety Evaluation No.: 328376-001, RI Modification: Replacement of Corroded Underground Aluminum Piping In 1980, underground Aluminum piping located in the vicinity of the Condensate Storage Tank (CST) and the Demineralized Water Storage Tank (DWST) was excavated, inspected and replaced. The Aluminum piping was installed during original plant construction and was in senice for ~12 years. In 1991 and 1994, two (2) sections of l
underground Aluminum piping adjacent to the Condensate Storage Pump House r i excavated, inspected and replaced. These sections of Aluminum piping were installed in 1980, as replacemen, piping, and were in senice for (~11) years and (~l4) years. Inspection results indicated deterioration of the protective pipe insulating material, pitting corrosion, gah anic corrosion and through wall pipe leaks. These degradation mechanisms compromised the leak-tight integrity and reliability of the Condensate Transfer System and the Demineralized Water Transfer System. This Configuration Change reroutes Aluminum piping within the Condensate Transfer System, Condensate Transfer Building, as well as, replaces Fire Protection Water System Valves V-9-9 and V 10.
l This modification does not adversely effect nuclear safety or the emironment. This modification does not involve
. any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 403057-002 R0 Modification: Replacement of EDG Fuel Oil Transfer Pump Strainer
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The existing EDG fuel oil transfer pomp strainers (S-39-001,002,004 and 005) cause a high differential pressure that creates low pump suction pressure as the strainers become dirty. The low suction pressure has resulted in transfer pump cavitation. Larger strainer with more surface area should reduce the differential pressure and l
improve pump performance.
Also the fuel transfer pump / motor (P-39-13,14,15 and 16) base plates are susceptible to bending and warpage.
This condition prevents the pump and motor from being properly aligned which results in high vibration. A new support plate, w hich is more rigid, will allow for proper motor to pump alignment and reduce pump vibration.
This configuration change (reference 3.1.3) will replace the existing i %" EDG fuel oil transfer pump suction strainers with larger 1-1/2" strainers. Due to the larger size of the replacement strainers, they will be relocated and the piping between the strainers and the pumps will be rerouted. In addition, the existing EDG fuel oil transfer pump / motor base plate will be replaced with base plates, which are more rigid. Also, Unit #1 EDG Sump drain
- valve V-39-5 will be reconfigure with an upstream tee and a second valve. This new configuration on EDG unit l
- 1 will be match existing arrangement for valves (V-39-12 and V-39-1011) on EDG unit #2.
This modification docs' not hdversely effect nuclear safety or the emironment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 408895-001. R0 Modification: Replacement of V-5-165 The purpose of this document is to evaluate the safety of replacement of valve V-5-165. Presently, the valve is a wafer body, duel plate check valve as manufactured by Clow. A wafer body swing check valve as manufactured by CAS Valve will replace it.
V-5-165 is the inboard Drywell Isolation Valve for the Reactor Building Closed Cooling Water System (RBCCW).
I The present valve leaks excessively. Part of the problem is that this particular type of check valve is not as tight a scaling valve as would be necessary for a containment isolation valve. The replacement valve was selected to be functionally equal to or better than the original u hile being able to be installed between the existing pipe flanges.
Except for the replacement of the valve, no other clumge will be made.
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i This modification does not adversely effect nuclear safety or the environment. No unresiewed safety question is j
l generated by this modification and can be implemented under 10 CFR 50.59.
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Safety Evaluation No.: 4030614)01, R1 Modification: Retirement of Reheat Protection S stem Supply Side 3
The supply side of the Reheater Protection System was intended to supply warmed, dry air to the shcIl and tube sides of the Reheaters to prevent corrosion during layup or long outages. The system consists of a supply and an exhaust side. The exhaust side is used during maintenance activitics to maintain a negative airflow (into the Reheaters) and will remain in senice. Due to component and operational control problems the Reheater Protection supply side has not been in senice for over twelve years. The advent of shorter refueling outages and past inspections has shown that this part of the Reheater Protection System is no longer required.
The work being performed will isolate and remove the supply side of the Reheater Protection System from senice.
The components will be abandoned in place. This will be done by removing a short section of pipe at the four locations where the Reheater Protection supply piping ties into the Reheat piping. Pipe caps will be welded on the resulting Rcheat Piping branch.
This modification does not adversely alTect nuclear safety or the environment. No unreviewed safety question is generated by this modification and it can be implemented under 10 CFR 50.59.
Safety Evaluation No.: 409583-001, R0 Modification: Roof Replacement at Main Office Building Fan Roof The Main Oflice Building (MOB) roofis a reinforced concrete slab that is protected from the environment by a built up roof. This built up roof was installed when the plant was built and has deteriorated by weathering and pedestrian traffic associated with inspections and maintenance activities. The built up roof has exceeded its design life and is no longer capable of providing protection to the reinforced concrete slab and, therefore, it must be replaced. The purpose of this modification is to replace the existing built up roof with a new built up roof that satisfics current building codes.
The design and installation was evaluated to ensure that the new built up roofing has no alTect on the ability of any plant system to perform its intended function during or afler a design basis event. This modification does not increase the probability of occurrence or consequence of an accident not previously evaluated to reduce the margin of safety as defined in the technical specification. This modification will not have any adverse effect on nuclear safety or the environment.
Safety Evaluation No.: 000531-019, R0 l
Modification: Service Water Pipe Relocation
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1 The purpose of this docunient is to evaluate the safety of relocating a portion of the buried Senice Water pipe west of the Turbine Building. The pipe is suspected ofleaking. The pipe is approximately fourteen feet below grade.
Excavations to that depth are very costly. The purpose of this configuration change is to relocate a portion of the
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pipe so that it is closer to grade. A valve will also be added. This will facilitate future repairs if they become l
necessary.
j This modification does not adversely affect nuclear safety or the environment. No unreviewed safety question is generated by this modification and it can be implemented under 10CFR50.59.
Safety Evaluation No.: 403037-003, R1 Modification: Shroud IIcad Bolt Reduction j
l The purpose of this safety evaluation is to evaluate the acceptability of plant operation with twelve (12) equally
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spaced and latched shroud head bolts (SHB) and twenty-four (24) removed shroud head bolts. The twelve (12) l l
SIIB were installed during the ISR outage are the new type (IGSCC resistant) bolts that were designed and I
constructed for the new overhead core spray sparger (not installed) and modified for use with the original core spray sparger.
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Safety Evaluation No.: 403037-003, R1 (Cont'd.)
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 based on evaluations that Oyster Creek can be safely operated with only twelve (12) equally spaced and latched SHB's. Operational experience over the past cycle 15 (21 months of l
operation) has resuhed in no plant abnormalities and no steam cutting of thejoint mating surfaces have been observed. Therefore, it has been proven that the reduced number of SHBs (from 36 to 12) has no effect on plant operation.
This safety evaluation has determined that this modification does not (1) adversely affect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important to safety previously evaluated in the SAR, (3) create the possibility for (a) an accident or (b) a malfunction of a different type than any presiously identified in the SAR or (4) decrease the margin of safety as dermed in the bases of any Technical Specification.
Since there is no adverse affect on nuclear safety and/or safe plant Operations, no creation of an unresicurd nuclear safety question and no emironmental impact, this modification is acceptable Safety Evaluation No.: 328333-006, R1 Modification: SL-P-3A/3B Motor Replacement in NRW, two holdup tanks are provided for 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 3 hp 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 Area), radiation exposure to personnel is higher than desired due to the high maintenance activity. This modification shall replace the existing motors with 5 hp, AC motors. These new motors shall provide a higher torque to aid in stating and operating the pumps. Tne need to control flow rate is no longer required because the polymer solidification process has been abandoned.
This safety evaluation concludes there is no detrimental effect on safety for the emironment and the modification does not pose an unresiewed safety question per 10 CFR 50.59.
Safety Evaluation No.: 403024-002, R0 Modification: SQUG Diesel Generator Seismic Battery Stops in December,1980, the Nuclear Regulatory Commission staffinitiated an unresolved safety issue, USl A-46, "Scismic Qualification of Equipment in Operating Plants," related to seismic adequacy of mechanical and electrical equipment in older nuclear plants.
To address the unresolved safety issue at the Oyster Creek Nuclear Facility, a Safe Shutdown Equipment List (SSEL), in accordance with GIP guidelines, was developed followed by a walkdown and evaluation of the equipment on this list. During the walkdown evaluation, items of equipment that did not comply with all the screening guidelines provided in the GIP were identified as " outliers." Those outliers requiring modification were identified as M-39-001, Diesel Generator Batteries, EDG 1 (System 741) located in the west room of the Diesel Generator Building and M-39-002, Diesel Generator Batteries, EDG 2 (System 741) located in the cast room of the Diesel Generator Building. Batteries are located in the floor vault of the Diesel Generator Unit. The scope of the modification is to provide spacers to prevent batteries from moving during a scismic event.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications. Since this safety evahiation has determined that no unresiewed nuclear safety question has been created and that no emironmental impact is involved, this modification is acceptable.
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4 Safety Evaluation No.: tHH1823-006, R0 Modification: SQUG Support Upgrade As part of the resolution of the USNRC Unresolved Safety issue (US1) A-46, GPUN developed a Safe Shutdown Equipment List (SSEL) as defined in the Generic ! implementation Procedure (GIP) developed by the Seismic Qualification Utilities Group (SQUG).
A walkdown and evaluation of equipment on the SSEL in accordance with the GIP was performed. During the SQUG walkdown evaluation, items of equipment that did not comply with all the screening guidelines provided in the GIP were identified as " outliers". The purpose of the modification is to resolve specific " outliers" by upgrading equipment anchorage support to ensure scismic adequacy in conformance to plant design bases. The modification only involves equipment anchorage and support. Equipment is not moved, added, or replaced and pressure boundaries are not impacted. Loads on the roof will not be increased beyond the allowable stress limits specified in the Oyster Creek License Basis Documents. Anchor bolts are installed in accordance with site procedures with QA involvement to assure that reinforcing is not cut and building structural integrity is not affected.
There are no unresiewed safety questions resulting from these modifications. No Technical Specification change is required and these changes can be implemented under 10CFR50.59.
Safety Evaluation No.: 0(H1713-002, R0 Modification: Stator Cooling Pumps Control Logic Modification
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n This modification bypasses the stator cooling pump trip from the following lockout relays for the Stator Cooling Water Pumps SCP-A and SCP-B control logic:
230 kV Bus Secondary Lockout relay Generator Master Lockout relay Generator Backup Master Lockout Relay Main Transformer Lockout Relay i
Auxiliary Transformer Lockout Relay.
Jumpers shall be provided in the Motor Control Centers I AII (B05) and 1B11 (B05) to bypass the above lockout l
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relays contacts provided in the control logic of the stator coolant pumps.
This modification will prevent long term degradation of stator caused by thermal transients that occur following a stator coolant pump trip. - Also, water conductivity will not be impacted due to shutdown of the cooling flow 4
- through the filters. This modification bypasses all lockout relay trips of the stator coolant pumps. Jumpers provided in the motor control centers l Al1 and 1Bl1 will bypass the 86 lockout relay contacts.
This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 403050-001, R0 Modification: Thermal Overload Relay Replacement Protection Motors on the A & B 4160V Buses The purpose of this modification is to replace the 49/50 (thermal overload /instantar.cous) relays protecting motors on the A & B BOP 4160V buses with 50/51/83 (instantaneous, time overcurrent, high drop out (DO) instantaneous) relays having more reliable and predictable operating characteristics along with a shorter reset time.
The long reset time of the 49/50 relay poses longer out of senice time for plant equipment following relay pickup.
Other 49/50 relay drawbacks such as spurious alarms, performance questions. and extraneous calibration time warrant the change. Also, the "83" and "50" desice setpoints will be changed. The "83" setpoints will be lowered l
f from 300% to 200% of FLC to allow for more time for an operator action in case of overloads. The "50" device l
instantaneous setting will be raised (for several relays) in order to allow for more margin between the trip current and the starting in rush current, to prevent any spurious trips.
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Safety Evaluation No.: 403050-001, R0 (Contd.)
The relay replacements are specifically designed for ruotor protection, have more reliable and predictable operating characteristics, and a shorter reset time. Plant safety equipment is not adversely affected because the modification does not affect safety-related equipment / logic and is confined to BOP loads. The proposed modification has been evaluated and determined not to represent an unresi:wed safety question as dermed in 10 CFR 50.59. The modification does not have any adverse effect on nuclear safety, safe plant operations or the environment.
Safety Evaluation No.: 000911-002, R0 Modification: Thermolag 1/4" Drain The purpose of this safety evaluation is to address the acceptability ofinstalling a drain on a raceway fire barrier constructed on Thermo-Lag. The barrier is a pull box for two 4 inch conduits locate:1 in the Turbiac Building basement (Fire Zone TB-FZ-11D). The barrier configuration has a fire rating of 50 minutes as determined by evaluation documented in electronic database TLDB-OC-814-1 for element no. 328.
The 1/4-inch drain on the underside of Thermo-Lag barrict element 328 does not negate or lower the 50-minute rating of the barrier. NEl testing of such a configuration has demonstrated that a gap of this size has no effect on the fire rating of a barrier because the hole will scal during fire exposure when the Thermo-Lag reacts, expands and begins char layer formation. Therefore, a maximum 1/4-inch drain located at least three inches from a joint is acceptable for this barrier.
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This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
j Safety Evaluation No.: 328382-001, R1 Modification: Top Guide Specimen Removal
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This safety evaluation addresses the pennanent removal of two specimens from the top guide structure, inside the reactor vessel at the Oyster Creek Nuclear Generating Station. This work is being done to better understand the material condition of the top guide since cracking has been fonnd on the top guide beams over past inspections.
The specimens will be examined to determine a root cause of the cracking, as well as to assist GPUN and the industry in developing inspection criteria and long tenn actions.
The purpose of this safety evaluation is to address the safety aspects associated with the removal of two specimens from the top guide structure insiA: the reactor vessel. The specimens will be used to determine the metallurgical l
properties of the irradiated tracrial and to detennine the root cause for the cracking of the top guide beams that l
has been obscrTed in tim recent past. This safety evaluation covers the structural assessment of the top guide as a result of the pennanent removal of the specimens from the top guide cross beams.
1 This safety evaluation has determined that this modification does not adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications and/or (5) the environment. Since this safety evaluation has determined that no unreviewed nuclear safety question has been created and that no environmental impact is involved this modification is acceptable.
Safety Evaluation No.: 000411-014, R0 Modification: V-1 M08 Conversion to Live Loaded Chesterton Packing i
The purpose of this activity is to convert the packing gland on V-1-0008 from a conventional stufling box to a " live I
loaded" stufling box. The implementation of this corrective change will make the gland leak-offline ineffective and unnecessary.
The packing configuration clumge and elimination of the glad leak-offline will not affect nuclear safety or safe plant operation. This activity does not create an unresiewed safety question. The technical specifications will not require revision; however, the FSAR will require resision. There is no emironmental impact associated with this l
activity.
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Safety Evaluation No.: 403049-001, R1 Modification: Circulating Water Pipe Replacement at Intake Structure The purpose of this safety evaluation is to evaluate the configuration change per OC-CCD-403049-002, which replaces circulating Water (CW) piping downstream of the CW pumps.
General corrosion has caused pipe wall thinning of the discharge piping of each CW pump. In early cycle 15, a through wall leak developed on the 78" CW pipingjust downstream of valve V-3-II. This is the butterfly valve on the discharge of the 1-4 CW pump. The location of the !cah i: en a pertion of pipe that is approximately 6" long just before it penetrates the intake tunnel wall. Additional through wall leaks subsequently occurred in the same general area. During cycle 15, a temporary encasement was installed around the 78" spool piece that was leaking.
This temporary repair was designed to last until the 16R outage. In addition, presure switch PS-l19 and associated valves were removed to accommodate installation of the pipe encasement. Associated circuitry and control room alarms were disabled as well as the input to the sequence of events recorder.
The scope of this safety evaluation is to evaluate the replacement of the 78" piping between the CW pumps and the discharge butterfly valves. The replacement is inclusive of the pump expansion joints for pumps 1 - 1, 1 -3, 1 -4.
The scope also includes evaluating the replacement of piping from Valve V-3-11 to and within the intake tunnel wall. A pipe tap and root valve will be installed on this segment of pipe. This will allow re-installation of PS-119 if future CW Pump / System changes make this alarm function necessary, in addition, piping downstream of valves V-3-8, V-3-9, and V-3-10 (including tunnel penetration) will be inspected and repaired as required. The platform between CW pumps 1-3 and 1-4 will be modified to climinate interference with the installation of pump l-4 discharge piping from valve V-3-ll to the intake tunnel wall.
This safety evaluation has determined that this modification does not: 1. Adversely alTect nuclear safety and/or safe plant operations; 2. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety presiously evaluated in the SAR; 3. Create the possibility of an accident or malfunction of a different type than any presiously identified in the SAR; or, 4. Decrease the margin of safety as defined in the bases of any technical Specification. Since the safety evaluation has determined that no unresiewed nuclear safety question has been created and that no emironmental impact is involved, this modification is acceptable.
The CW System does not have any safety functions. This configuration change will not impact the operation of the CW System or any other system.
Safety Evaluation No.: 403052-001, RI Modification: IRM and SRM Drynell Cable Replacement The purpose of this safety evaluation is to analyze the modification with respect to 10 CFR 50.59 criteria and determine the impact on nuclear safety and if an unresiewed safety question exists.
The existing instrument cables under the reactor vessel for the IRM and SRM neutron monitoring system inside the drpvell will be replaced with a removable section ofinstrument cable. This will be accomplished using qualified instnnnent connectors and cable. This section of cable will be removable during refueling activities.
The safety of the plant is not affected by this modification. This modified instrumentation cable will not alTect the existing function of either the IRM or SRM systems and will operate during refuel and startup modes of plant operation. This modification does not adversely affect nuclear safety, does not reduce the margin of safety, does l
I not involve and unresicwed safety question. Since there is not adverse affect on nuclear safety and/or safe plant l
operations, an unresiewed nuclear safety question does not exist. In addition, no environmental impact results from this modification. No Technical Specification change request is needed to implement this change.
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Safety Evaluation No.: 403075-001, R0 l
Modification: LP Turbine Monoblock Rotor Installation l
This modification is being implemented to reduce concerns with the existing low pressure (LP) turbine "C" rotor l
whcci bore and dovetail cracking. A monoblock rotor design consisting of a forged rotor with integral u heels with l
dovetails of an improved radii design which are shot pcened prior to bucket installation will replace the original l
LP-C rotor. The new monoblock rotor will also be equipped with wider 38" last stage heavy buckets (LSHB) to allow for future greater flow capacity. However, this modification does not address increased LP turbine flow capacities. The OCNGS originally supplied low-pressure turbine rotors were fabricated using " shrunk on" wheels with keyways and keys used to tmnsmit wheel loads to the rotor shaft during design overspeed conditions. These original whcci bores and keyways are susceptible to intergranular stress corrosion cracking (IGSCC) and will bc l
climinated with the monoblock rotor installation in the LP-C turbine.
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The performance of the Turbine Generator and the efTects of failures of components on the rest of the plant
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associated with this modification have been evaluated in detail. The installation of the monoblock rotor and 1
associated components will not adversely afTect the operation of the Turbine Generator or the associated subsystems. The implementation of this modification will not adversely afTect nuclear safety or safe plant operations.
A failure of the turbine generator does not directly impact any safety-related equipment. The only potential affect of a turbine failure on plant safety is due to the generator of a missile, which impacts on an adjacent structure. As described in GE memo dated 06/07/96, the probability of turbine overspeed makes such as event not credible. In addition, the consequences have been evaluated in the extraordinary event a missile did occur. The result of this evaluation has determined there is no increase to the consequences that could ads ersely efTcct nuclear safety or any reduction in the margin of safety as a result of the monoblock LSHB rotor in the LP-C turbine installation from either high or low trajectory turbine missiles. The probability of occurrence of an accident or a malfunction of equipment is not increased and the possibility of a ditTerent type of accident is not introduced. There fore, there is no impact on nuclear safety and no unresiewed safety question exists. This modification can be performed under 10 CFR 50.59.
Safety Evaluation No.: 403091-001, R0 Modification: Tnmnion Room Fan RF-1-6 and RF-1-7 Replacement On February 7,1996, the inmnion room cooling unit RF-1-7 fan and motor failed. This placed the trunnion roorn cooling unit RF-1-7 out of ser ice.
I A temporary blower was installed to the cooling unit discharge duct. The temporary blower circulates trunnion room air across the 10-7 cooling coil which is still operational. The temporary blower is powered from the existing 1-7 motor power sc,urce: MCC 1B12-460VE01.
Since the installation of the temporary blower, trunnion room temperatures are being maintained at or below presious levels. The temporary blower is easier to inaintain and more reliable since it has a direct driven motor which can casily be replaced. The permanent fans are belt driven and have had poor reliability. Also the permanent fans and motors are dillicult to access when maintenance or repairs are performed.
As a result of the positive operating experience and improved reliability, the decision has been made to convert the temporary blower to a permanent configuration and to install a new blower to replace the RF-1-6 cooling unit fan and motor. The blowers will be permanently mounted to the floor. Eight-inch (8") flexible metal type duct will be used to connect the cooling units to the blowers. Electrical power will be supplied from the existing power supplies for the existing fan motors.
This configuration change to make the temporary blower on RF-1-7 permanent and to install a similar unit on RF-1-6 will result in no unresiewed safety questions as defined in 10 CFR 50.59.
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Safety Evaluation No.: 403053-001, R1 Modification: Upgrade of the MSIV Limit Switches The Oyster Creek Main Steam isolation Valve (MSIV) limit switches have been subject to numerous failure-2 cause of the majority of the failures has been attributed to the mounting configuration of the switches and th. type of operating lever used on the switch. The switch mechanism itself has performed well in this application. In addition to the switch failures, the configuration of the switch maunting plates and the lack of quick disconnect electrical connectors create an increased maintenance burden when the switches require removal and reinstallation.
The purpose of this modification is to: I. Reduce the probability oflimit switch failures. 2. Allow casy installation and removal of the switches and mounting plates. 3. Simplify adjustment of the limit switches uhile installed.
In order to fulfill the purpose as described above, an upgrade of the MSIV limit switches is planned that will consist of the following changes. 1. Removal of the old switches, Patel seals, mounting plates, wiring (back to the local terminal box), and flex con <luit (back to the rigid conduit fittings). 2. Installation of two new mounting plates on each valve. The new plates run nearly the full length of the valve yoke rods and are secured to the yoke rods with U-bolts. 3. Mounting and setup of two switches on each mounting plate (one for scram, one for position indication). Each switch is configured with a prewired quick disconnect electrical connector. 4. Mounting and wiring of the quick disconnect electrical plug and new ficx conduit for tie in back to the local terminal boxes for cach switch. 5. Welding of extensions onto the spring seat in four locations to provide actuation of the limit switches as the valve opens and closes, y r
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This modification upgrades the mounting configuration of the MSIV limit switches and incorporates a quick disconnect electrical connector. This safety evaluation has determined that this modification does nql_(1) adversely affect nuclear safety and/or safe plant operations, (2) increase the probability of occurrence or the consequences of an accident for malfunction of equipment important to safety presiously evaluated in the S AR, (3) create the possibility for an accident or malfunction of a different type than any presiously identified in the SAR or (4)
. decrease the margin of safety as defined in the basis of any Technical Specification. Since there is no adverse affect on nuclear safety and/or safe plant operations, no creation of an unreviewed nuclear safety question, and no
'emironmental impact, this modification is acceptable.
Safety Evsluation No.: 403024-004, Re Modification: Vent Line on Diesel Generator Fuel Oil Tank The purpose of this Safety Evaluation is to evaluate the configuration change that allows movement of the vent line
- on the Diesci Generator Fuel Oil Tank within the Diesel generator Fuel Oil Storage and Transfer System.
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The aforementioned vent line is embedded in the wall of DG Building Fuel Oil Tank Compartment. The concern
' is that this configuration does not allow any movement in a seismic event and may cause this pipe to rupture at the
. tank. The configuration change chips the DG building Fuel Oil tank Compartment wall at the penetration to allow movement of the 6-inch vent line. A plate will be installed within the penetration under the pipe. This will allow the pipe to slide on the plate and also perform as a support for deadweight loads. The penetration will be filled with a fire resistant material to prevent debris or rain water from entering the building.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications. Since this safety evaluation has determined that no unreviewed nuclear safety question has been created and that no emironmental impact is involved, this modification is acceptable.
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Safety Evaluation No.: 000211-017, R0 Modification: ECS Tubing Support Modification The purpose of the safety evaluation is to evaluate the configuration change (support addition) to the Emergency Condenser System (ECS) tubing connecting the 16" ECS lines to the transmitters PT-lG04 A/B. Calculation C-1302-211-5320-095 wais generated tojustify that the support addition satisfies the design criteria for the power plant.
This safety evaluation has determined that this change does not (1) adversely afTect nuclear safety and/or 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 UFSAR,3) create the possibility of an accident or malfunction of a different type than any previously identified in the UFSAR, or (4) decrease the margin of safety as defined in the design criteria.
Since this safety evaluation has determined that no unreviewed nuclear safety question has been created and that no environmental impact is involved, this configuration change is acceptable.
Safety Evaluation No.: 403036-001, R1 Modification: Upgrade of the Reactor Building Crane The Reactor Building crane was originally purchased and installed as a commercial-grade off-the-shelf system.
The seismic classification was Class 11 w;hich imposed only local building code criteria. Subsequently, a seismic anti-fall-down criteria was established by GPUN for modifications.
This mod 7::ation will replace the control system for the Reactor Building crane with a state-of-the-art system for controlling nearly all crane ftmetions. The existing motor / generator (MG) sets and control' that provide variable DC power for motor speed control will be replaced with variable frequency (trolley and bridge) or flux vector (main and aux hoist) drives. Radio controls will be added for crane control from outside the cab. Safety features are added that monitor crane performance and take corrective action as necessary. Electronics for the crane control system will be mounted in new cabinets on the crane walkway. New 480V runway conductors and a bridge to trolley festoon system will improve the integrity of circuit connections between moving components.
Additional changes involve crane cab access, cab shielding, operator comfort, and lighting. Also, the south bridge bumpers are being replaced to allow the bridge to travel an additional two-inch further south prior to docking.
.The upgrade of the Reactor Building crane is intended to replace an antiquated control system with a state-of-the-art system for controlling crane function and menitoring crane performance. This safety evolution has detennined that this modification does not adversely affect nuclear safety and/or safe plant operations; increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety presiously evaluated in the S AR; create the possibility for an accident or malfunction of a different type than any previously identified in the SAR or; decrease the margin of safety as defined in the basis of any Technical Specification.
Since there is no adverse affect on nuclear safety and/or safe plant operations, no creation of an unreviewed nuclear safety question, and no environmental impact, this modification is acceptable.
Safety Evaluation No.: 403024-004, R0 Modifications to Reactor Building Trolley for Fixed Link System i
GPUN has developed a redundant (fixed-link) support system for handling of the NUHOMS spent fuel transfer j
cask. Modifications are being made to the Reactor Building trolley structure to allow installation of the fixed-link l
system. The purpose of this safety evaluation is to document that the preposed modifications will not impact the l
structural integrity of the trolley structure nor create any safety concerns or impacts to nuclear safety.
The scope of work covered by the safety evaluation includes the installation of eight 1/2 inch stiffener plates to the trolley beam; removal of existing stitch welds between the top of the trolley beam and the 6/8 inch cover plate; cut / grind access holes and slots in the 6/8 inch cover plate.
Safety Evaluation No.: 403024-004, R0 (Cont'd.)
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The modifications to the Reactor Building tmiley stmeture can be completed safely and will not adversely affect safe operation of the plant. As a result, this modiScation does not imolve an unresiewed safety question nor does it have any impact on nuclear safety or safe operation of the plant.
Safety Evaluation No.: IMHl4114116, R0 NS038 (V-14MMl8) Leak Repair Valve Body to Bonnet The purpose of this safety evaluation is to evaluate the safety of a temporary repair to Main Steam Isolation Valve NS03B (V-14XK)8). The repair will be to the body to bonnet flanges of the valve. The repair will be made by injecting scalant into a clamp placed around the cin.umference of the bonnet and body flanges bridging the gap between them. The injection will be made through holes drilled into the clamp for this purpose.
A temporary repair will be made to valve V-14)003. The repair is to the body to i onnet scal. By the foregoing evaluation it is demonstrated that this modificatico does not adversely effect nuclear safety or the environment. No unresiewed safety question is generated by this modification and it can be implemented under 10 CFR 50.59.
Safety Eraluation No.: 402N904Kil, R3 Permanent DrywcH Scaffolding The purpose of this modification is to design and install a permanent drywell scaffolding basic framework that can be left in the dryucli during plant operation. This modification will climinate the repeated work of crecting and dismantling scaffolding frames during each outage.
v This modification provides the basic framework (skeleton) of scaffolds only Other temporary items which are required to form a senice platform such as planks, additional posts, bracing, handrails, ladders, and lead blankets for temporary shielding, will be added onto the pennanent scaffolding basic framework during each outage in accordance with OC Procedure 105.2 and is not covered under the scope of this modification. Those temporary items shall be removed prior to plant restart.
The same design criteria that had been used to design the original daywell permanent scaffolding is used for the modified design.
The structural design of this modification is in compliance with all safety requirement codes and regulations and will not affect the safety function. It is concluded that the proposed modification will not have any adverse affect on nuclear safety or the environment.
Safety Evaluation No.: 315403-035, R0 Pinnt Specific Technical Guidelines for SBEOPs The modification of the OC EOPs to address ATWS stability is based upon evaluations conducted by GE and the BWP.OG. GE performed the evaluations using a modified version of the TRACG computer code. The numerical solution scheme used in the code was specifically modified for this application to eliminate numerical damping of the oscillations.
Implementation of these changes to the Plant Specific Technical Guidelines and to the Emergency Operating Procedures developed from the PSTGs will not adversely affect nuclear safety or safe plant operation. These changes have been developed by the BWROG to prevent or mitigate large power oscillations that can develop during an ATWS situation and cause damage to fuel assemblies.
The changes to the EPGs are consistent with the guidance found in Rev. 4 of the EPGs and act only to simplify, clarify, or prmide additional reasons / indications for pedorming an action aheady included and approved in Rev. 4.
Bypassing of MSL high radiation interlocks does not adversely affect nuclear safety or safe plant operation. The function of the high radiation interlocks is to prevent an ofTsite release in the event that there is fuel damage.
l Prmided that there is no failure in the main steam line this function is maintained by the off gas system. Since j
man steam line integrity is a criteria associated with this action nuclear safety and safe plant operation is assured.
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4 Safety Evaluation No.: 315403-035 R0 (Cont'd.)
The overall risk to the health and safety of the public and potential public dose are actually decreased if the MSL high radioactivity isolation interlocks are bypassed when the Main Condenscr is available for use as a heat sink.
The requirement of terminating and preventing injection in to the RPV is a vital part of the existing EPG strategy for controlling and mitigating an ATWS condition. The only system at Oy ster Creek that requires interlocks to be defeated is the Core Spray System. These interlocks are defeated by use of the Core Spray override push buttons and the pull-to-lock features on the comrol switches for the main pumps. In all cases, the Core Spray System can still be manually operated and detailed guidance on the use of this system is contained within the EOPS.
Additionally, operation of the Core Spray System during ATWS cvents is outside the design and licensing basis for Oyster Crcck. Therefore this will not adversely affect nuclear safety or safe plant operation.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety presiously evaluated in the Safety Analysis Report has not been increased by these changes.
Safety Evaluation No.: 000641-017, R0 Procedure Resision: Anticipatory Scram Bypass Setpoint (PSil Switches)
Increasing reactor power results in an increasing third stage pressure. As pressure increases PSH-A, B, C & D contact opens and the Anticipatory Scram Signal bypass is removed. This opening of the pressurc switch will be referred to as the "setpoint" of the switch, Decreasing power results in a decreasing and third stage pressure. As pressure decreases, PSH-A, B, C and D contact closes and the Anticipatory Scram Signal bypass is actuated. This closing of the pressure switch will be referred to as the " reset" of the switch.
The Technical Specification and the Safety Analysis Limits for the Anticipatory Scram Signal bypass removal are 40% and 45% reactor thermal power respectively. With the heaters out of senice, the third stage pressure values corresponding to Technical Specification Limit (40%) and Safety Analysis Lituit (45%) are determined by a GPUN Calculation No. C-1302-641-5350-008. Thus the respective calculated third stage pressures are 137.628 and 162.136 psig. This determination was based on actual test data. Another GPUN Calculation, No. C-1302-641 5350-001 determined the tolerance for the PSH pressure switches based on the instrument as-found data. This was factored in:o the PSil setpoint per Calculation O. C-1302-641-5350-008. Accordingly, the proposed setpoint with its tolerances will conform to the Technical Specification limit under the limiting conditions of all three strings of HP and XP feedwater heaters out of senice.
Although the setpoint and reset values are at the lower end of the instrument calibrated span, historically the switches have functioned well. Hence, there is no reason to doubt their ability to perform intended safety function which is to remove the Anticipatory Scram Signal bypass on increasing power before reaching the Technical Specification limit of 40% reactor thermal power.
All of the affected procedures will be resised some of which deal with the following: Plant Startup and Shutdown; Turbine Warm-up and Start-up; Feedwater Heaters; Turbine Dypass Valve 1
Implementation of the proposed setpoint will not adversely alTect nuclear safety or safe plant operation.
l The new setpoint modified by the instrument tolerances for accuracy, drift, etc., will conform to the required
)
Technical Specification Limit of 40% reactor thermal power. The tolerances are based on statistical analysis of the
)
as-found data. Thus the calculated tolerances have a 95% confidence level which is acceptable per Reg. Guide 1.105 (February,1986).
New setpoint assures that the assumptions in the Safety Analyses are preserved. The specific cases are as follows:
Turbine Trip Without Bypass (UFSAR Chapter 15) i Technical Specification Limit (40%) and Safety Analysis Limit (45% as described in the bases of Technical Specification Section 3.1.
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Safety Evaluation No.: 000212-036, R0 Procedure: Core Spray System Operability Criteria Resision The purpose of this document is to evaluate the effect on safety of changing the acceptance criteria for the Core Spray System operability. The proposed new operability acceptance criteria are:
System 1 Loop A - 3400 gpm at 230 psig System 1 Loop C - 3400 gpm at 230 psig System 2 Loop B - 3640 gpm at 230 psig System 2 Loop D - 3640 gpm at 230 psig The safety evaluation demonstrates that an unresiewed safety question does not exist. This change has no adverse impact on nuclear safety or safe plant operations.
Safety Evaluation No.: (HK)666-002, RI Drywcil110 Analyzer Surveillance Procedure:
2 2 This safety evaluation supports the use of an As-Found tolerance of up to +/- 1.5% for both the hydrogen and oxygen channels in OCNGS Procedure 604.3.020. Drywell H;O Analyzer Surveillance. It also supports changing 2
the frequency of surveillance from weekly to monthly, 1
This safety evaluation shows that the estimated performance is sufficient for these monitors to perform their intended function and that it is reasonable to adjust the As-Found tolerances in the procedure to accommodate the estimated monitor's performance at 31 days after calibration.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) 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.
Safety Evaluation No.: (HHl542-014 R0 Prwcedure: Generator Hydrogen Gas System (336.3)
Operation of the TBCCW system with the V-5-198 valve in automatic control has camsed swings in the cooling of other end users, when the TBCCW system experiences flow or cooling transients. Because of this problem Plant Operations has deemed it necessary to operate the valve in manual, adjusting the valve position, and, therefore, the cooling as temperatures in the generator change. Until such time as the transient problem is solved, or an alternate way is found to place the valve in senice without severcly impacting the remaining users, the valve will be operated in mamial. The Operations staff monitors the generator temperatures regularly to ensure that its temperatures are controlled, that safe plant operation is preserved, and any nuclear safety concerns are avoided.
This safety evaluation has determined that this modification does not (1) adversely alTect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important to safety presiously evaluated in the S AR, (3) create the possibility for (a) an accident or (b) a malfunction of a different type than any presiously identified in the SAR or (4) decrease the margin of safety as defined in the bases of any Technical Specification.
Since there is no adverse affect on nuclear safety and/or safe plant Operations, no creation of an unresiewed nuclear safety question and no emironmental impact, this modification is acceptable Safety Evaluation No.: 945100 072, R0 Procedure: Management of Potential Safety Concerns This safety evahiation supports Revision 7 to GPU Nuclear Procedure 1000-ADM-7330.01, " Management of Potential Safety Concerns." Resision 7 deletes the procedure.
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i Safety Evaluation No.: 9451004)72. R0 (Cont'd.)
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This safety evaluation has determined that this procedure resision, which deletes the PSC process, does not (1)
{
adversely affect nuclear safety and/or safe plant operations; (2) increase the probability of occurrence or the l
consequences of12m accident or malfunction of equipment important to safety presiously evaluated in the SAR; (3) create the possibility of an accident or malfunction of a different type than any previously identified in the SAR; or, l
(4) decrease the margin of safety as defined in the bases of any Technical Specification.
I Since the safety evaluation has determined that no unres icwed nuclear safety question has been created and that no environmental impact is involved, this procedure resision, which deletes the PSC process, is acceptable.
Safety Evaluation No.: tHH16224117, R0 Procedure: 603.3.tH12. R19, Recirculation Loop Flow Calibration Frequency Extension The purpose of this safety evaluation is to demonstrate that the results of the loop crror calculation are acceptable for operation of the recire flow monitoring electronics for up to 30 months between calibrations (24 month fuel cycle plus 25%). This will allow the technical specification calibration requirement to be changed from 1/20 months to each refueling outage in support of TSCR #203. This change is necessary since calibration of the electronics by Procedure 603.3.002 can only be performed while the plant is shut down.
This safety evaluation has determined that this modification does not (1) adversely affect nuclear safety and /or safe plant operations, (2) increase the probability of occurrence of the consequences of(a) an accident or (b) a malfunction of equipment important to safety previously evaluated in the SAR, (3) create the possibility for (a) an accident or (b) a malfunction of a difTerent type than any previously identified in the SAR or (4) decrease the margin of safety as defined in the bases of any Technical Specification.
Since there is no adverse alTect on nuclear safety and/or safe plant operations, no creation of an unreviewed nuclear safety question and no environmental impact, this modification is acceptable Safety Evaluation No.: tHH1234-005, R1 Process Control Plan for Biocide Injections Configuration change to build the biocide injection skid and the Process Control Plan for biocide injection will be done to eliminate the methane gas-producing bacteria found in the processed waste. By the destruction of the methanc-producing bacteria, this will ensure that the waste will be safely transported, stored and buried without harm to persons transporting, handling and disposing of such waste. The final waste fonn will not be changed.
Since Glutaraldehyde is a surface reacting biocide, there will probably be a little residue left behind in the filter sludge. This material is not an EPA listed hazardous waste and is approved for burial at the Barnacle Facility.
This configuration change and the new procedure for operating the biocide injection skid have been evaluated to j
not cause any Nuclear Safety concerns, Safe Plant Operations concerns and does not constitute an unresiewed
)
safety question. An FSAR change must be submitted to address biocide treatment of filter sludge as an alternative to solidification.
Safety Evaluation No.: 4030364H15. R0 Reactor Building Crane Load Test i
The purpose of this document is to evahiate the impact of a planned load test of the Oyster Creek Reactor Building Crane on safe plant operation. GPUN has developed a fixed-link (redundant) support system for handling the NUHOMS spent fuel transfer cask. The fixed-link support system (FLSS) will be attached to the Reactor Building Trolley structure to provide redundant support for the fuel transfer cask while the cask is on the 119' 3" El. of the Reactor Building. The FLSS consists of two support arms that hang from the main trolley that uhen engaged with the cask, provide redundant support. If the crane cable, hook, brakes or other load-carrying component of the main crane fails, the FLSS will support the cask without incident.
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i Safety Evaluation No.: 403036-005, R0 (Cont'd.)
Modifications have been completed to allow the FLSS to interface and attach to the trolley. Prior to the load test, FLSS bracket and tic rods will be added to the cranc and the tie rods will be torqued to a total pre-load of 255 KIPS as specified on LES drawing 70-05 01, Sheet 2. In addition, the angles holding the main trolley beam to the trolley and trucks will be replaced, and various modifications to the auxiliary hoist will be completed. The additional components of the fixed-link system will be tested independently. The test will be performed in three phases:
1.
A load test of the main hois to 125% ofits rated capacity 2.
A load test of the FLSS brackets and tie rods including their ability to deliver loads adequately to the main trolley structure 3.
A load test and functional test of the auxiliary hook.
A test load with a final assembled weight of 250,000 lbs. will be placed in the Reactor Building at El. 23' 6". This load will be lifted with the Reactor Building Crane as specified by the load test procedure. Following the load test, the crane will be certified in accordance with CMAA #70-1988. A simila, iteration will be conducted for the auxiliary hoist on the main crane, utilizing a test load of 10,000 lbs.
The Reactor Building Crane load test as described in load test procedure TP 422/3 has no impact on plant safety or safe plant operation. The Reactor Building is capable of witlistanding the impact of a load drop during the main hoist load test without damage to safety related equipment. No other aspects of the test have a credible potential affect on plant safety.
Therefore, the load test does not increase the probability or consequences of an accident or malfunction of equipment, cause the possibility of a different type of accident or reduce Technical Specification margins. There is no impact on the FSAR or the Technical Specifications. No unresiewed safety questions result from the test of the Reactor Building Crane or FLSS.
Tiafet) Evaluation No.: 328089-001, R4 R-actor Coolant System (RCS) Piping Repair - Weld Overlay Repairs The purpose of this safety evaluation is to document and justify that the use of wcld overlays for repairing stainless l
stect weldments containing IGSCC will not jeopardize plant safety. Wcld overlay may also be used to repair through-wall cracks, whether produced solely by IGSCC or when produced by performing IHSI on mostly through-wall IGSCC. The systems evaluated are 1) Recirculation,2) Isolation Condenser,3) Shutdown Cooling. 4) Core Spray, and 5) Reactor Water Cleanup.
The NRC as an interim repair has accepted Weld overlays. Approval to operate with wcld overlays for more than one operating cycle have been obtained on a cycle-by-cycle basis. NRC per NUREG 0313, Rev. 2 and Generic Letter 88-01 specifics the wcld overlay design criteria. Wcld overlay repair ofIGSCC in stainless steel weldments will be performed to reduce the potential for piping leakage and failure in the Reactor Building. This repair method is adequate since the wcld overlay provides a barrier that is structurally sufficient by itself without taking credit for the remaining pipe ligament.
This modification does not adversely effect nuclear safety or the environment. This taodification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 945100-076, R0 Reorganization: GPU Nuclear Reorganization l
A GPU Nuclear Technical Functions Division reorganization became effective on January 1,1994. This evaluation addresses the nuclear safety impact of this reorganization and prmides the basis for Licensing Basis Document changes.
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Safety Evaluation No.: 945100476, R0 (Cont'd.)
This safety evaluation concludes that because this activity does not involve a significant increase in the probability or consequences of accidents presiously considered and does not involve a significant decrease in a safety margin.
This activity does not involve a significant hazards considerations; there is reasonable assurance that the heahh and safety of the public will not be endangered by operation in the proposed manner; and plant operations will be conducted in compliance with the NRC's rules and regulations. Therefore, an unresicued safety question does not exist and implementation of the reorganization is acceptable.
Safety Evaluation No.: 9451tHbO79, R0 Reorganization: Transfer of Site Senices to Plant Division A partial GPU Nuclear reorganization became cfrective February 1,1995. The content of Chapter 13 of the Oyster Crcck FSAR must be changed to reflect the new organizational alignment. This evaluation addresses the nuclear safety impact of the reorganization.
This safety evaluation concludes that because this activity does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin.
This activity does not involve a significant hazards considerations; there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and plant operations will be conducted in compliance with the NRC's rules and regulations. Therefore, an mireviewed safety question does not exist and implementation of the reorganization is acceptable 1
Safety Evaluation No.: 000822-023. R0 Repair of SGTS Duct at the Stack On August 6,1996, the 14" suction duct to the Standby Gas Treatment System (SGTS) blower EF 1-8 was found to have corrosion. The corrosion is in an area on a 14" duct that rises from the stack tunnel to the pad outside the stack. As a result, Operations declared SGTS I out of senice. The 14" duct rises through the tunnel ceiling, which is l'6" thick, then through a 5'6" section of sand, and then through a l'3" concrete pad, EF-1-8 is supported off this l'3" pad.
The purpose of this modification is to repair the 14" riser on System I by installing an inner sleeve in the riser.
Also, since SGTS System 2 could be subjected to the same corrosion mechanism, this modification will install the same type inner sleeve in the riser for System 2.
This modification has no adverse impact on (1) nuclear safety, (2) safe plant operations, (2) the probability or l
consequences of an accident or malfunction either evaluated in the FSAR or not, (4) the margin of safety defined in the Technical Specifications and/or (5) the emironment. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
I Safety Evaluation No.: 000823-006, R0 SQUG Support Upgrade l
As part of the resolution of the USNRC Unresolved Safety Issue (USl) A-46, GPUN developed a Safe Shutdown l
Equipment List (SSEL) as defined in the Generic Implementation Procedure (GIP) developed by the Scismic Qualification Utilities Group (SQUG). A walkdown and evaluation of equipment on the SSEL in accordance with the GIP was performed. During the SQUG walkdown evaluation, items of equipment, which did not comply with all the screening guidelines prosided in the GIP, were identified as " outliers". The purpose of this modification is to resolve certain specific outliers by nuking changes to the hardware to ensure equipment scismic adequacy in
.anformance to plant design bases.
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Safety Evaluation No.: 000823-006, R0 (Cont'd.)
Provide horizontal scismic restraints for the IIVAC fan, motor and components, including the sheet metal structure housing the plenum, and modify existing inadequate supports without alTecting operational vertical vibration isolation for SF-1-20 Supply Fan and EF-1-20 Exhaust Fan on the Omcc Building Roof, El. 61'. Prmide scismic support framing to support ilVAC ductwork connecting the SF-1-20 supply fan system with the EF 1-20 exhaust fan system for SF-1-20/EF-1-20 connecting HVAC ductworks support frame on the Omcc Building Roof at El. 61'.
Replace existing access stairway platform on the Omcc Building Roof at El. 61'. Modify existing corroded supports for SF 121 Supply Fan, EF-1-21 Exhaust Fan, SF-1-12, SF 1-13, and SF-1-14 Supply Fans, SF-1-16, and SF-1 17 Supply Fans, EF-1-26 and EF-1-27 Exhaust Fans, the conduits for Chilled Water System and Heating Steam System. All are located on the Omcc Building Roof at the 61' El.
The purpose of this modification is to resolve specific " outliers" by upgrading equipment anchorage support to ensure scismic adequacy in conformance to plant design bases. The modification only involves equipment 3
anchorage and support. Equipment is not moved, added or replaced and pressure boundaries are not impacted.
Loads on the roof will not be increased beyond the allowable stress limits specified in the Oyster Creek License basis Documents. Anchor bolts are installed in accordance with site procedures with QA involvement to assure that reinforcing is not cut and building structural integrity is not affected. There are no unresiewed safety questions resulting from these modifications. No Technical Specification change is required and these changes can be implemented under 10CFR50.59.
- Safety Evaluation No.:.000215-oll, R1 Temp. Modification: Sealant injection of V-16-63 The purpose of this document is to evaluate the safety of a temporary repair to Reactor Water Cleanup (RWCU) return to the Reactor Recirculation Loop Manual Isolation Valve (V-16-63). The repair will be to the pressure seal l
area of the valve. The repair will be made by injecting scalant into the scal cavity at the body to bonnet seal interface. The injection will be made through holes drilled into the body for this purpose.
This modification has no adverse impact on (1) nuclear sa'fety, (2) safe plant operations, (3) the probability or consequences of an accident or malfunction either evaluated in the FS AR er not, (4) the margin of safety defined in the Technical Specifications and/or (5) the emironment. No unresiewed safety question is generated by this modification and it can be imp!cmented under 10CFR50.59.
Safety Evaluation No.: 000622-020, R0 Temporary Modification: TR-1 A0014 Alarm I
i This safety evaluation is being writtenio evaluate the impact of control room alarm G-6-c not functioning properly and the defeating of the alarm. TR-I A0014 is the Reactor Metal Differential Temperature recorder on Panel 3F in the main control room. This recorder monitors two differential temperatures. Reactor head to head flange i
temperature and vessel to vessel flange temperature. Following modification in 14R and the subsequent reactor l
heatup, it was discovered that this recorder does not reflect the differential temperatures properly. This recorder will alarm on a high temperature differential and is the input to alarm G-6-c. Operations bypassed this alarm in July of 1995 due to spurious alarms, using Section 6.0 of Procedure 108.8 (Defeating Alarm Circuits / Bypassing Recorder inputs / Removing Recorders from Scrsice). A 10 CFR 50.59 Safety Determination was not performed to determine if this temporary modification involved an unresiewed safety question. Operations then resised Procedures 201.2 and 203.2 to change the locations of the temperatures monitored during a heatup or cooldown.
The temperatures are now monitored on TR-I A0002 in the Reactor Building SI-foot elevation. The NRC cited not i
performing a 10 CFR 50.59 safety evaluation to specifically determine the impact of losing the alarm function or of defeating the alarm circuit as a violation in Inspection Report 96-009. This evaluation is one part of the corrective actions in response to this violation.
A Plant FSAR Update will be submitted for changes to two sections,5.4.11.2 and 7.6.l.1.3, of the FSAR due to recorder TR-I A0014 not indicating or alarming properly.
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Safety Evaluation No.: (XX)622-020, R0 (Cont'd.)
Bypassing the alarm (G-6-c) for high differential temperature betueen the reactor head and head flange and vessel and vessel flange will not constitute an unresiewed safety question because the same temperatures can be monitored by Operations to ensure that no procedural limit or Technical Specification or FSAR limit will be exceeded during plant heatups and cooldowns. Also this parameter is not required during a design basis accident.
This modification does not adversely cfTect nuclear safety or the environment. This modification does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 403036-002, R0 Test: Reactor Building IIHi-Ton Crane Load Test The purpose of this document is to evaluate the impact of a planned test of the Oyster Creek Reactor Building Crane on safe plant operation. The crane has recently been modified and will be tested to satisfy the requirements of CMAA #70-1988.
A test load with a final assembled weight of 250,000 lbs. will be placed in the Reactor Building at El. 23'6". This load will be lifted with the Reactor Building cranc as specified by the load test procedure. Following the load test, the crane will be certified in accordance with CMAA #70-1988. The test load and various smaller test loads will then be used to conduct additional functional tests of the crane in accordance with the load test procedure. A similar iteration will be conducted for the auxiliary hoist on the main crane, utilizing a test load of 12,500 pounds.
9-The Reactor Building Crane Load Test as described in load test procedure TP 422/2 has no impact on plant safety or safe plant operation. The Reactor Building is capable of withstanding the impact of a load drop during the load test without damage to safety related equipment. The probability of a load drop during the functional test is acceptably small and the consequences are within acceptable limits with regard to safe plant operation and protecting public health and safety. No other aspects of the test have a credible potential affect on plant safety.
Therefore, the load test does not increase the probability or consequences of an accident or malfunction of equipment, cause the possibility of a different type of accident or reduce Technical Specification margins. There is no impact on the FSAR or the technical Specifications. No unresiewed safety questions result from the test of the Reactor Building crane.
Safety Evaluation No.:,328403-001, R1 Thermal Dilution Gates Restoration Thermal Dilution (Sluice) Gates are installed (six in number) in the Recirculation Tunnel located at the west side of the intake Structure. These gates control the discharge of hot water at the intake of the circulating water pumps.
The hot water is conveyed from the circulating water discharge tunnel via the recirculation tunnel on which the gates are mounted. The injection of this water prevents ice buildup during the winter months.
Presently the six gates are in various stages of disrepair due to corrosion and/or buildup of silt in the gate guides.
This change is to remove the existing center gates in cach half of the intake structure and to replace them with new l
gates similar to the originals but with enhanced corrosion resisting materials. A portable gas powered operator is I
also provided to facilitate operation. The operator is not part of the configuration because it is not permanent and not required to operate the gate.
In order to allow work to be performed "in the dry", a temporary dewatering enclosure is required and the crection and removal of this structure is included in this change. No separate SE was prepared for the construction phase of l
this project.
I The following items are discussed in this Safety Evaluation:
1.
Removal of two of the six Dilution Gates and replacement with two new gates of the same size and construction but made of materials of enhanced corrosion resistance.
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Safety Evaluation No.: 328403 001, R1 (Cont'd.)
2.
The option of using a self-contained, portable gasoline powered operator to assist in operation of these j
gates in lieu of manual operation.
3.
The crection and removal of the temporary dewatering enclosure for the construction phase.
l-This restoration does not adversely effect nuclear safety or the emironment. This restoration does not involve any unresiewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No.: 402991-tH17,RI l
Title:
Emergency Senice Water Pipe inspection This configuration change will allow inspection of the 10" (ESW) Emergency Senice Water pipe for marine growth in the non-chorinded portion of the 10" discharge pipe from ESW pump (1-2,1-3).
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 affect the function of the Senice Water, Emergency Senice Water System or Containment Spray Systems, since it is not required to perform any safety function, nor will it be used during system actuation. This modification does not adversely effect nuclear safety or the environment. This modification does not involve any unreviewed safety question and no changes to the Technical Specification are required.
Safety Evaluation No. 945100-101, R1 Document: Operational Quality Assurance Plan The primary changes incorporated into Resision 10 of the Operational Quality Assurance Plan are the combination of Licensing and Regulatory Affairs (L&RA) and Nuclear Safety Assessment (NSA)into a single department and an expanded list of responsibilities for the position of Director, Nuclear Safety Assessment. The additional responsibilities concern Licensing and Regulatory Affairs. Revision 10 also provides that the Manager, Assessments has the authority and responsibility to resolve disputes with other non-QA functions that report to the Director, NSA. A separate evaluation of the change has been prepared in accordance with 10 CFR 50.54 (a),
which concluded that the change does not constitute a reduction in commitment from the program description previously accepted by the NRC. This safety evaluation addresses the nuclear safety impact of the changes in the separate analysis, this change has been resiewed under the provisions of 10 CFR 50.54(a). This safety evaluation concluded that: (1) because this activity does not irwolve an increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to safety presiously evaluated and does not involve a decrease in the margin of safety; (2) does not create the possibility of an accident or malfunction of equipment not presiously evaluated; and does not reduce the margin of safety defined in the basis of the Technical Specification; an unresiewed safety question does not exist. Therefore, the change is acceptable.
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