ML20056H132
| ML20056H132 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/31/1993 |
| From: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C321-93-2152, NUDOCS 9309080267 | |
| Download: ML20056H132 (40) | |
Text
C GPU Nuclear Corporation Post Office Box 388 Q
f Route 9 South Forked River. New Jersey 08731-0388 609 971-4000
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Writer's Direct Dial Number:
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August 31,1993 C321-93-2152 l
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i U. S. Nuclear Regv'atory Commission Attn: Document Control Desk Washington DC 20555
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 l
10 CFR 50.59(b) Report In accordance with 10 CFR 50.59(b), the summaries of the changes to Oyster Creek l
systems and procedures described in the Safety Analysis Report (SAR) for the period January 1992 to March 1993 are attached to this letter. Attachment I addresses those activities which directly affected systems / components described in the SAR.
Attachment II addresses those activities for which a GPU Nuclear safety evaluation was performed due to the potential to adversely affect nuclear safety or safe plant operations, but which did not directly impact SAR systems / components.
If any further information is required, please contact Ms. Patty Arcaro, Administrator, at 201.316.7748.
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John J. Bar on Vi e Presid t
- nd Director yster Creek l
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Attachments 1
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Administrator, Region I Senior Resident Inspector f
g Oyster Creek NRC Project Manager
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Document Control Desk C321-93-2152 Page 1 Activities Directly Affecting Systems / Components Described in the Safety Analysis Report I.
Procedure / Document Changes Procedure:
Station Blackout Procedure S/E #402965-005, Rev. O Description of Chances Station Blackout itself is a new accident and malfunction that the NRC has mandated that all light water reactors must be capable of coping with for specific periods of times. This procedure provides the direction on how to use the combustion Turbines and those systems, equipment necessary to maintain the plant in a safe shutdown condition. A Station Blackout event was not part of the original design basis for Oyster Creek. The Unit substation tie breakers will not be operated until after a Station Blackout event has occurred.
No additional failures or malfunctions need to be considered or are postulated to occur before, during, or after a SBO event.
Therefore, this procedure and activity could not possibly increase the probability or the consequences of an accident or malfunction previously evaluated in the SAR.
Safety Evaluation Summary: This procedure has no adverse affects and will only be used after a station blackout event.
This procedure was written after reviewing the methodology contained in TDR 1099 and provides specific information for this event to be used together with the EOPs. If any situation develops that is not addressed by this procedure or if this procedure contradicts with EOPs then EOPs would take precedence. The operations defined in this procedure are those that are required to meet NRC 10CFR50.63.
PROCEDURE: Reactor Building To Torus Power Vacuum Breaker Test and Calibration Procedure #604.3.001 Description of chanae:
This one time change is being made to allow for rebaselining of V-26-18 after the actuator was replaced with a new larger actuator. The new actuator has almost twice the air volume as the old operator, but no changes were made to the air system or to the actuator. So it is expected that the valve will take longer to stroke with the new actuator.
Preliminary stroking of the valve indicates a stroke time of about 6 or 7 seconds.
Safety Evaluation Summary: This revision does not affect nuclear safety. The new time acceptance criteria maintains the valve within times to ensure the valve functions as required.
This revision does not increase the probability of a malfunction or an accident because this change to the stroke time acceptance criteria only af fects V-26-18 and it does not make the valve any more likely to fail when called upon to do so.
This change does not create a possibility for an accident or malfunction not i
identified in the SAR. By maintaining the opening stroke time at 10 seconds, the negative design pressure for the torus will not be exceeded.
This change does not reduce the margin of safety as defined in the basis for any Tech Spec.
The basis for Tech Spec 4.5 says that the closing time for l
containment valve is not critical, therefore this change does not reduce the l
margin of safety.
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C321-93-2152 l
Page 2 Procedure: Core Spray Pump Operability Test Procedure #610.4.002 i
Description of Chance This change deletes the requirement to verify that Core Spray System 2 is filled and vented prior to the performance of the pump i
l operability surveillance.
The requirement is in place due to original plant design which resulted in partially drained systems and led to damaging water i
hammer effects.
In the event of an actuation signal, the system would be required to start up and operate in the stand-by status; therefore, this change does not present an operating condition that does not exist already.
1 Safety Evaluation Summary:
This change does not increase the probability for equipment malfunction as evaluated in the SAR because the system will be operated i
l from the normal stand-by condition.
This system arrangement is identical to l
conditions found if the system received a safety actuation signal and the system l
started automatically. The Control Room Operator will be directed to check the Keep Fill Alarm prior to pump start and verify the alarm is clear.
Alarm response requires for the affected system to be filled and vented; therefore, alarm response take priority over the prerequisite.
The Tech Specs margin to safety are not reduced by this change because the change does not impact the Tech Spec operability requirements or system alignment.
Modification:
Rotameter Installation for Bleeding Air Into The Condensate i
l System S/E #312400-008 Descriptien of Modification: The purpose of this modification was to raise the dissolved oxygen level of the Feedwater/ Condensate System at the Oyster Creek l
Nuclear Generating Facility.
This was obtained by the installation of a i
rotameter, filter, and valve on a 1/2" line in the Condensate System. A decision i
l was made to bleed air into the system to increase the dissolved oxygen. The low dissolved oxygen levels are unfavorable due to erosion / corrosion problems experienced in the industry. This safety evaluation pertains to the installation I
of a rotameter, filter, and valve with associated fittings and tubing on a support piace of f of the condensate "A" instrument rack. They are tied into the 1/2" line with valve V-2-1007 at the suction of the "A" condensate pump.
Safety Evaluation Summary: The modification adds a rotameter, filter, valve, and i
l tubing with the associated fittings to the end of the 1/2" line (with valve V '
1007) on tne condensate pump
'A' suction. Air shall be bled into this line due l
to the low dissolved oxygen levels.
This modification will not cause an t
unfavorable environmental impact.
The portion of the Condensate System this i
modification deals with is neither required for safe shutdown of the reactor nor to mitigate the consequences of postulated accidents. Therefore, the activity will neither adversely af fect nuclear safety or safe plant operations nor reduce the margin of safety as defined in the UFSAR and basis of the Plant Technical Specifications. Hence, there is no unreviewed safety question in accordance with 10CFR50.59.
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Document Control Desk C321-93-2152 Page 3 Modification:
Feedwater and Recirculation Flow Control Systems Upgrade 1
S/E #402901-003 Description of Modification: Phase I was performed in the 14R refueling outage and covers the following work:
Control Room / Upper Cable Spreading Room / Mux Room
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Thirty (30) new cables were pulled (unterminated) using trays in the control room and new (upper) cable spreading room.
Control Panel 4F - relocate existing control rod drive system " settle" and
" permit" lights and " Admin Rod Block" keylock switch to new location on same panel. Reuse lights and switch.
Control Panel 17R - Install new conduit support above panel, modify control room ceiling to provide a 6" x 6" opening, and retag an existing
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sleeve on top of panel.
Feedwater Pump Room Install one (1) new conduit and pull one (1) new cable (leave unterminated).
Install new position transmitters ZT-625-004 and 005 for LFRV A and C.
Transmitters will not be tied in electrically at Control Room end.
Reactor Building Instrument Rack RK01 Install new narrow range reactor water level instrument LT-IDl3C with associated tubing and wiring, LI-ID13C will not be tied in electrically at Control Room end.
Site Emergency Building ISEB)
Install seven cables and other equipment for DFRCS/ Plant computer interface. These cables and equipment will not be tied in at the Control Room end.
Safety Evaluation Summary: This modification does not pose a safety concern or unreviewed safety question for the reasons cited herein.
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 between this modification and existing plant safety systems, Additionally, no new safety related digital components are installed. Although l
some existing cafety related components are being relocated, functionally they l
are unchanged.
New failure modes are identified and addressed, as is the justification for reduced potential plant transients attributable to feedwater system failures. Therefore, it is concluded that the proposed modification does not have any adverse effect on safety or environmental impact. This modification does not constitute an unreviewed safety question as determined by 10CFR50.59.
Document Control Desk C321-93-2152 Page 4 Modification:
Installation of Replacement RPV Surveillance Capsule S/E #402888-001 d
Description of Modification: The modification involveg the installation into the OCNGS during the 14R refueling outage of " Replacement Reactor Vessel Materials Surveillance Program (RVMSP) Capsule" and the BWROG Supplemental Surveillance Program (SSP) capsules, which are intended to provide test data to establish the RPV pressure-temperature curves to the end of life of the RPV.
The purpose of this shfety evaluation is to demonstrate that the replacement RVMSP and BWROG (SSP) Capsules which were installed in the OCNGS-RPV will not adversely impact nuclear safety, personnel safety, environmental protection or safe plant operation.
Safety Evaluation Summary:
This modification does not violate any existing license requirements or regulations.
No fluid or electrical system is functionally altered by this modification. The replacement RVMSP Capsule is a passive component in the reactor vessel very similar in design to existing specimen holder, basket No. 3.
The only design change of significance is the increased size and weight of the Capsule compared to basket No.
1 or 2,
previously removed. This modification does not alter the safety functions of any safety-related systems.
This modification does not change any radiological safety concerns under either normal operating conditions or design basis accidents. This modification will not cause additional radiological exposure to plant personnel during normal operation. No radioactive material is released to the environment either within the plant buildings or outside the plant as a result of this modification.
This modification will neither create any new plant effluents nor impact existing plant effluents; therefore, this modification will neither add to an existing environmental concern nor create a new environmental concern not evaluated in current environmental requirements documents.
This modification will have no adverse effect on personnel safety since the replacement RVMSP Capsule is entirely within the reactor vessel and has no effect on plant operations, effluents or radiation levels.
It is concluded that the installation of the replacement RVMSP Capsule will not result in an unreviewed safety question as defined by 10CFR50.59. No change is required in the OCNGS Technical Specifications. A minor change is require in the OCNGS-UFSAR to update the inventory of reactor vessel materials surveillance specimens present in the reactor vessel; however, the change satisfies the criteria of 10CFR50.59 Modification:
TBCCW, SFP and SDC Heat Exchanger DP Gauges S/E #312400-010 Description of Modification:
The purpose of this modification was to install dif ferential pressure indicators across the Reactor Building Closed Cooling Water (RBCCW) side of the Spent Fuel Pool Heat Exchangers, the Shut Down Cooling (SDC)
Heat Exchanger (C-17-003), and also across the Service Water / Circulating Water side and the Turbine Building Closed Cooling Water (TBCCW) side of the TBCCW Heat Exchangers, These indicators are used to monitor flow through the heat exchangers to support flow balancing the RBCCW and TBCCW systems and monitor tube fouling in the TBCCW Heat Exchangers.
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Document Control Desk C321-93-2152 Page 5 Safety Evaluation Summary:
The modification adds local differential pressure instrumentation across the shell side (RBCCW) of the Spent Fuel Pool Coolers.
This provides an indication of flow through these coolers so that the RBCCW system flow can be monitored and balanced. This modification has no impact on the Spent Fuel Pool Cooling System since the barrier between this system which is safety related and RBCCW remains the tubes and tube sheets of the heat exchangers.
This modification also adds local differential pressure indication across the tube side (Circulating or Service Water) and shell side (TBCCW) of the TBCCW heat exchangers.
This will provide some indication of tube fouling and allow the plant to better determine required cleaning frequency. This will also enable the plant to better control flow through the heat exchangers to minimize velocity induced tube vibrations.
This modification has no adverse effect on TBCCW, Service Water or Circulating water System availability or performance.
The modification also adds local differential pressure indication across the shell side (RBCCW) of the third Shutdown Cooling Heat Exchanger (Note, other two already have dp indication). This will not impact the Shutdown Cooling System since the barrier between this system and RBCCW remains the tubes and tube sheets of the heat exchanger.
Modification:
Retire Resin Storage Tank T-2-012 S/E #312400-011 Description of Modification: The purpose of this modification was to blank off and retire in place the Resin Storage Tank T-2-012 (also known as the 1-8 Demineralizer).
This modification is in response to the poor condition and lack of use of the Resin Storage Tank (T-2-012).
Six blanks were installed in between flanges upstream of the six valves going into the tank. This will permanently isolate the tank form the rest of the plant systems, thus retiring it in place. The air lines going to these valves were disconnected, removed back to a suitable location and capped.
Safety Evaluation Summary: The modification uses the existing vital power supply (VACP-1) to the condensate demineralizer control panel. There is a reduction in load on this circuit as loads are being removed. However, since these loads have not been used in years, no credit will be taken.
Consequently, there is no impact on the feeder, breaker, etc.
Therefore, there will be no effect on the safety functions of the vital a.c. power system. Additionally, this modification performs no safety functions nor does it replace, remove, or adversely affect any equipment which performs a safety function. This modification will only affect the Condensate Demin. System. The following safety concerns were considered and determined not be impacted by this modification; environmental phenomena, missile generation, high energy line breaks, containment isolation, environmental qualification, materials, compatibility and electrical loadir.g. It is concluded that the blanking off and retiring in place the Resin Storage Tank modification will not have an adverse impact on nuclear safety nor does an unreviewed safety question exist.
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Document Control Desk C321-93-2152 Page 6 i
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Modification:
Install High Point Vent (Hydrotest) on Containment Spray S/E #323560-009 Descrintion of Modification: The purpose of this modification was to allow the l
l 14" containment spray lines to vent air so that hydrostatic inspection is possible.
Previously, there was no feasible method to vent air out of the 14" lines. This inspection is dictated by the ASME BP&V Code,Section XI.
Safety Evaluation Summary: The modification will not affect the safety function of the Containment Spray System.
This modification utilizes nuclear safety related materials which are comparable or exceed the material specification required. This modification will not af fect the functionality of the containment I
Spray System.
i This modification provides a capability to perform a functional test on the containment spray system without altering designed functions of the containment spray system.
This modification does not constitute an unreviewed safety question. There is no environmental impact due to this modification and there I
are no changes required to the Technical Specifications.
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Document Control Desk C321-93-2152 Page 7 Modification:
Reactor Fuel Zone Level Indication System Modification S/E #000622-014 Description of Modification: The Fuel Zone Level Indication System (FZLIS) at Oyster Creek Nuclear Generating Station (OCNGS) provides compensated reactor vessel water level indications. The FZLIS is strictly a monitoring system and does not have any control functions. The system monitors level from 144 inches below the Top of Active Fuel (TAF) to 180 inches above the TAF.
The FZLIS has four (4) independent Channels A,
B, C,
and D.
Each. channel utilizes a Modicon 484 programmable controller for level computation using inputs from two differential pressure type level transmitters. One level transmitter monitors the narrow range level (L1) from 60 inches above the TAF to 180 inches above the TAF.
The second level transmitter monitors the wide range level (L2) from 144 inches below the TAF to 180 inches above the TAF.
PSC 92-005 was issued to document a concern by the Oyster Creek Simulator Project that the FZLIS will not perform adequately during accident conditions when the core region is voided and dynamic losses are occurring from steam or two-phase upflow across the level tape.
Subsequently, a design review was performed by GPUN E&D and SAPC. Based on this review, the modification defined in MDD-OC-622B was developed.
This modification made changes to the programmable controller logic to implement the design review recommendation to use the wide range level (L2) transmitter input for level computation in all conditions.
The scope of this modification includes the following:
1.
Revise the programmable controller logic for all four channels of FZLIS (software changes only).
2.
Perform surveillance test of the revised programmable controller logic.
Safety Evaluation Summary: This modification has been determined to involve no Unreviewed Safety Questions for reasons as follows:
1.
The activity will not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the SAR.
The FZLIS is a monitoring system only, has no interface with plant control functions, and as such the change has no impact on any accident or malfunction.
The FZLIS provides post accident trending information only.
This modification removes a potential source of nonconservative bias in level indication, and as such will not increase the consequences of an accident or malfunction previously evaluated in the SAR.
2.
The modification will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the SAR.
The modification is a software modification to a programmable controller used for plant monitoring and impacts no plant control functions.
3.
This modification will not decrease the margin of safety as defined in the Bases for any Technical Specification, since the FZLIS is not credited in the Technical Specification Bases, and no margin of safety is adversely impacted.
Document Control Desk C321-93-2152 Page 8 Modification:
Chlorination System Outside Piping & Valve Replacement S/E #323670-001 Description of Modification:
The chlorination system is provided to control biofouling of the heat transfer surfaces of the three main condensers, the Reactor Building Component Cooling Water (RBCCW) and Turbine Building Component Cooling Water (TBCCW) heat exchangers and containment spray heat exchangers through the circulating, service and emergency service water systems.
The modification replaces existing valves, piping and gaskets in the chlorination System due to continuous problems with leakage. The existing valves have f ailed due to a build up of sodium hypochlorite crystals. The existing piping has had numerous leaks due to the failure of its joints.
The original design specification for this system followed B31.1.
Apparently, the cure time for the joints was insufficient to achieve a plastic to plastic weld.
The new piping follows B31.3 (chemical Piping) for the bonding of the joints. The piping and valves replaced are outside on the discharge line from sodium hypochlorite tank one to the chlorination building and the annulus valve on tank one. Tank two is isolated and the jumper placed on the low level alarm under S&T #90-0292 is permanent.
The duplex strainer was eliminated since it was determined unnecessary after noticing that nothing had been collected since its installation.
Safety Evaluation Summary:
This modification will not adversely affect safe plant operations because it does not change the function of any plant systems or violate any plant Technical Specifications.
It does not degrade any plant systems. Normal plant operation will not be changed and equipment necessary for safe shutdown will not be adversely affected.
This modification will not increase the probability of occurrence or consequences of an accident because this system is not safety related and has no adverse impact on accident mitigating systems / equipment.
This modification does not increase the probability of occurrence nor the consequences of a malfunction because the modification is not safety related and does not interface with any NSR equipment.
This modification does not create a possibility for an accident or malfunction of a different type than any previously identified in the FSAR since it will not have a negative impact on any safety systems and has no safety function.
This modification does not create an unreviewed safety question as described in 10CFR50.59.
Modification:
Deletion of Containment Spray Auto Start Logic S/E #402894-001 Description of Modification:
The purpose of this modification was to address concerns associated with the present automatic start capability of the containment Spray System. Specifically, the automatic start capability of the Containment Spray and Emergency Service Water (ESW) pumps was deleted and replaced with manual operation capability only.
This provides for maximum operator flexibility in selecting Containment Spray and ESW pump combinations in a system.
Two new drywell pressure instrument loops are added to satisfy Regulatory Guide 1.97, Category I, Type A, design requirements. Drywell pressure will be used by the operator to initiate the CS and ESW systems manually. Also included is deletion of the Static Test capability which has no safety function and is not used.
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Document Control Desk C321-93-2152 Page 9 Safety Evaluation Summary: The modification proposed reduces the probability of evaporative cooling causing excessive negative drywell pressure and introduction of air into the primary containment.
Single failure concerns associated with Containment Spray System auto-start capability have been eliminated, and implementation of Emergency Operating Procedures (EOP's) will not be restrained.
i NRC approval of the Technical Specification Change Request for this modification is required. The modification does not contain an unreviewed safety question as i
defined by 10CFR50.59 and does not adversely impact nuclear plant safety or plant operations.
Modification:
Recorder Replacement - Phase I S/E #402910-001 Description of Modification:
To address the requirements of NUREG 0737, Iter.
I.D.1, regarding the Detailed Control Room Design Review (DCRDR), a human f acters I
review of the Control Room has been performed.
The DCRDR identified existing l
human factors deficiencies involving the existing sixty-six recorders.
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corrective action program has been developed to remedy the dericiencies.
l One conclusion of the DCRDR was that sixty six control room recorders required upgrading, This modification replaced or removed three of the recorders identified in this review. This modification also included the installation of the four class lE analog isolators, one on each of the four Main Steam Line Radiation Monitoring Channels.
Il Safety Evaluation Summary: This modification does not adversely affect nuclear safety because it does not degrade the integrity of the NSR systems with which
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it interfaces. This modification does not adversely af fect safe plant operations because it does not change the function of any plant systems nor violate any Technical Specifications.
I This modification does not reduce the margin of safety as defined in the basis j
of the Technical Specification or other licensing basis documents. This is true j
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because the safety function for existing safety related systems and the set points at which they are designed to actuate are not altered by this i
modification. Therefore, it is concluded the proposed modification does not have any adverse ef fect on safety or environmental impact. This modification does not l
constitute an unreviewed safety question as determined by 10CFR50.59.
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Modification:
Recorder Replacement for TR-IA02 and TR-RD46A&B l
S/E #402910-002 Description of Modification: The purpose of this modification was to relocate J
or replace recorders IA-02, RD-46A, RD-46B, RD-46C and IA14.
The following changes were also made as part of the modification:
1.
Control room recorder TR-IA14 was reconfigured for differential temperature of Reactor Vessel vs. Reactor ~ Vessel Flange in addition to Reactor Vessel Head vs. Reactor Vessel Head Flange. This was to meet the requirements of OCNGS procedures for plant heatup and cooldown.
2.
Control room recorder TR-IA0055 alarm circuit configuration was changed so that a recirculation pump motor generator set fluid drive coupler oil high temperature condition will actuate Annunciator window E-7-a on panel 3F which reads "MG BRG TEMP HI".
Currently this condition will actuate annunciator window E-6-a which reads "RCP MTR BRG TEMP HI".
Document Control Desk C321-93-2152 Page 10 Safety Evaluation Summary: This modification does not adversely affect nuclear safety because it does not interface or change any NSR systems.
This modification does not adversely affect safe plant operations because it does not change the function of any plant systems nor violate any Technical Specifications.
This modification does not reduce the margin of safety as defined in the basis of the Technical Specification or other licensing basis doer.ments. This is true because the safety function of existing safety related systems and the set points at which they are designed to actuate are not alterea by this modification.
Therefore, it is concluded the proposed modification does not have any adverse effect on nuclear safety, safe plant operations or the environment.
This modification does not constitute an unreviewed safety question as determined by 10CFR50.59.
i Modification:
J.
O. Backwash S/E #402914-001 Description of Modification: The purpose of the J. O.
Backwash modification is to provide a better method of cleaning condensate demineralizer resin beds than the current ABRO process.
The modification also provides for automatic sequencing for some resin transfer operations.
Safety Evaluation Sum:rary: The modification uses the existing vital power supply (VACP-1) to the condensate demineralizer control panel. There is a net reduction in load on this circuit, i.e., more loads are being removed than are being added.
These loads have been reviewed by EP&I and found acceptable. Consequently, there is no impact on feeder, breaker, etc. Therefore, there will be no effect on the safety functions of the vital a.c. power system. Additionally, this modification performs no safety functions nor does it replace, remove, or adversely affect any equipment which performs a safety function. This modification will only affect the regeneration system.
The following safety concerns were considered and determined not to be impacted by this modification; environmental phenomena, missile generation, high energy line breaks, containment isolation, environmental qualification, materials compatibility and electrical loading.
The J.
O.
Backwash modification replaces the ABRO resin cleaning process and automate some of the resin transfer sequence steps. It is concluded that the modification will not have an adverse impact on nuclear safety nor does an Unreviewed Safety Question exist.
Modification:
Protective Device Modifications S/E #402930-001 Description of Modification:
The purpose of this safety evaluation was to evaluate the impact on plant safety of replacing the existing GE molded case circuit breakers with the new Westinghouse C series Seltronics breakers.
The existing breakers had Long Tima Delay (LTD) and instantaneous trip functions whereas the new breakers have LTD and Short Time Delay (STD) trip function. The new breaker trip functions are required to improve the coordination between the upstream and downstream breakers and ensure the continuity of power supply to the safety and essential BOP loads. The affected MCCs are 1All, IB11, lA12, 1B12, IA21, vital MCC 1A2 and IB2.
Also af fected by this change is 120V vital AC Instrument Panel (IP) No.
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Document Control Desk C321-93-2152 Page 11 Safety Evaluation Summary: This modification consists of replacing the existing GE breakers in various MCCs, with new Westinghouse breakers and GE breakers in IP#4, to achieve better electrical coordination between upstream and downstream devices.
As a result, some existing compartments in some MCCs have been i
rearranged.
The setpoints of the new breakers provide the same overload protection as before and improved electrical coordination. The breakers will be commercially dedicated as Class lE devices for NSR use, will satisfy all applicable seismic requirements and the breakers in the NSR MCCs and Panel #4 will be installed under GPUN's QA plan.
Since these changes are only in the plant hardware which improves the reliability of electrical power supply to the NSR and BOP equipment and there is no change to the plant systems or their operating logic, it can be concluded that there is no impact on the existing plant safety margin, nor does it create any new unreviewed safety concern as described under 10CFR50.59.
Modification:
Air Compressor Unloader Valve Piping Mod S/E #402932-001 Description of Modification: Moisture has repeatedly gotten into the unloader i
valve cylinders of the plant air compressors.
This has caused the unloader piston to seize rendering the valve and associated air compressor inoperable.
The source of the moisture is the air receivers which currently supply actuation i
air to the unloader valves.
This modification supplies air from the instrument air system to the unloader valves. The supply piping connects to the existing piping downstream of the post filters. By sup71ying clean, dry air to the unloader valves, their service life and the reliability of the air compressors are increased.
The modification also replaces existing " point of use" filters on the air piping to the control switches and unloader valves of each air compressor.
I Safety Evaluation Summary: The modification involves providing clean, dry air to the unloader valves of the plant air compressors.
Its implementation will improve the reliability of the compressors and enhance safe plant operation. The air systems involved do not perform any active nuclear safety related function and are not required to mitigate tb consequences of postulated accidents.
It is determined no unreviewed safety question or environmental impact is involved with implementation of this modification.
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f Document Control Desk C321-93-2152 Page 12 Modification:
Corrective Change to Install Pressure Gauges Between The Torus to Reactor Building Vacuum Breakers corrective Change #130-92 Description of Modification:
This cc rective change installed two isolation valves, a capped tee, and a pressure gauge at the LLRT connections on V-26-15 and V-26-17.
Currently the LLRT connections are isolated with threaded Swagelok i
plugs, which act as the second primary containment boundary off the torus. The first boundary is provided by V-26-16 and V-26-18.
The new configuration has an isolation valve which acts as the second primary containment boundary.
This isolation valve is locked closed during normal plant operation. It will only be opened during the quarterly IST of V-26-15 and V-26-17 to allow a pressure reading to be taken. During this test primary containment is maintained by V 16 and V-26-18; there is a prerequisite in the IST procedure to ensure that V,
l 16 and 18 are closed.
Although the second valve, tee, and gauge are not qualified as part of the primary containment boundary, they can all withstand the torus design pressure of 35 psig.
Safety evaluation Summar.y:
This corrective change does not adversely affect nuclear safety. The first isolation valves will be checked for leak tightness prior to installation, and will be locked closed. This will provide at least as much assurance of isolation as is provided by the existing plugs.
The probability of occurrence of an accident or malfunction of equipment previously evaluated in the SAR is not increased.
The isolation valve is designed for pressures well in excess of the system design pressure, and this corrective change will not affect the operation of the Vacuum Relief System in i
i any way.
l This corrective change does not create the possibility of an accident or malfunction not previously evaluated in the SAR.
The isolation valve which acts as primary containment is tested / qualified to meet the requirements of the system.
The margin of safety in the Tech Specs is not affected by this corrective change.
There will still be two boundaries for primary containment from the torus.
1 Modification:
Appendix J Replacement Options l
S/E #402946-001 Description of Modification: This modification is intended to provide compliance with the following requirement of 10 CFR 50, Appendix J, paragraph III.C
" Type C tests shall be performed by local pressurization.
The pressure j
shall be applied in the same direction as that when the valve would be required to perform its safety function, unless it can be determined that i
the results from the tests for a pressure applied in a dif ferent direction will provide equivalent or more conservative results."
GPUN has determined that a number of drywell (containment) isolation valve installations do not comply with the above NRC criterion.
This modification provides the system design changes necessary to comply with the above design criterion.
There are also a number of isolation valves that have gasketed connections that cannot be tested in accordance with Appendix J Type B test requirements. Type B tests are required for:
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Document Control Desk C321-93-2152 Page 13
" Containment penetrations whrse design incorporates resilient seals, gaskets, or sealant compounds..."
Previously, these cases have been addressed by the performance of an Appendix J Type A test. However, these tests are expensive and time consuming. The changes installed by this modification permits local Appendix J testing which will save time, exposures and costs.
The purpose of this modification was to install additional testing isolation boundaries to pring the drywell isolation valve assemblies into compliance with the above criteria.
Safety Evaluation Summary: This modification will not adversely affect nuclear safety or safe plant operation, because the safety functions of safety-related systems are not altered by this modification. In addition, the integrity of the containment system has not been adversely affected by the modifications. Nuclear safety is improved by this modification because it will enable local testing of containment penetrations, and thereby improve confidence in containment performance.
This modification will not reduce the margin of safety as defined in the basis f
of the Technical Specification or other licensing basis documents. This is true because the safety function of existing safety related systems is not altered by this modification. Therefore, it is concluded the proposed modification does not constitute an Unreviewed Safety Question as determined by 10CFR50.59 and will not have any adverse effect on nuclear safety or the environment.
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Modification:
Cut and Cap Line 1"SA-249 S/E #000856-002 Description of Modification:
A leak developed in an abandoned portion of a service air line (l"SA-249) just outside of the New Sample Pumphouse.
The failure point appears to be at one of the pipe supports connected to the ORW Building. This line used to supply the New Sample Pumphouse with service air, however, when this system became contaminated this line was cut and capped inside the pumphouse.
This line still sees service air pressure because there is no isolation between it and the line that supplies ORW with service air (3"SA-150).
Service air is still required in ORW for the Fuel Pool Filter /Demin and the Filter Sludge Tank.
l The change cuts out and removes line 1"SA-249 from where it comes out of the NRW tunnel to its end inside the pumphouse. It removes line 1"SA-250 and valve SA-HV-244.
The open end of 1"SA-249 was capped.
Safety Evaluation Summary:
The safety concerns associated with this proposed change have been evaluated below and determined that this change will not adversely affect nuclear safety or safe plant operations, does not involve an unreviewed safety question, does not require a Teen Spec change and has no environmental impact.
1.
The change does not adversely affect nuclear safety because the system and i
structure involved in the change are not nuclear safety related nor are they connected to, or adjacent to any other nuclear safety related system, component or structure.
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Document Control Desk C321-93-2152 Page 14 2.
The probability of occurrence of an accident or malfunction of equipment important to safety that was previously evaluated in the SAR is not increased because the system and structure involved are not nuclear safety related or are they connected to, or adjacent to any other nuclear safety related system, component or structure.
3 1
3.
The consequences of an accident or malfunction of equipment important to safety that was previously evaluated in the SAR is not increased because the system and structure involved are not needed to mitigate the consequences of an accident or malfuncticn of equipment important to safety, nor are they connected to, or adjacent to any such systems, components or structures.
4.
The change does not create the possibility for an accident or malfunction of a different type than any previously evaluated in the SAR because loss of NRW Service air to the New Sample Pumphouse cannot initiate any
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accident scenario.
S.
A Tech Spec change is not required because this change does not conflict with any existing Tech Spec requirement.
j Modification:
Removal of the Drywell Cathodic Protection System S/E #402950-006 i
Description of Modification:
The drywell cathodic protection system (CPS) was I
installed to mitigate the corrosion of the drywell in the sand bed region. The l
design of the system assumed that the pre-existing inleakage of moisture into the i
sand bed could not be abated and would provide an adequate electrolyte for the impressed current system.
However, several modifications and/or maintenance j
activities were performed that seem to have eliminated the in-leakage into the sand bed. This resulted in a decrease in Dmpressed current such that the system was ineffective.
l With the drywell CPS deemed ineffective, it was determined that the anodes and reference electrodes could be removed.
This would facilitate sand removal through the holes utilized for their installation.
Hence the scope of the modification was to remove the hardware and cables associated with the sand bed region of the drywell CPS, to remove the cables associated with the power for the CPS monitoring equipment and rectifiers and to " spare" in place the remaining interconnecting cables / conduit and control panels.
Safety Evaluation Summary:
Thist modification removed the anodes, reference electrodes and interconnecting cables for the drywell cathodic protection Jystem.
There is no change to the operation of plant safety systems, Technical l
Specification requirements or limits or adverse impact on the plant environment.
[
No experiments or tests are perforned, which would adversely af fect plant safety.
i Hence, the modificatior, to remove the drywell cathodic protection system does not affect the margin of safety or create an unreviewed safety question as described under 10CFR50.59.
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Document Control Desk C321-93-2152 Page 15 Modification:
Removal of Isolation Condenser Shell Side Local Level Indication S/E #402953-015 Description of Modification:
Indicators LI-211-DOO1 and LI-211-DOO2 were intended to provide local shell side level indication of the Isolation condensers. These indicators were to be used by Operators when making up to the Isolation Condenser via the domineralized water system on elevation 95'.
However these indicators were unreliable.
This modification removed these indicators.
Safety Evaluation Summary: This modification removed a local-level indication of the Isolation Condenser shell.
This modification in no way affects level transmitters LT-IG-6A&B and associated instrumentation.
Therefore this modification does not affect nuclear safety or safe plant operation.
Modification:
Station Blackout Fire Protection Deluge Mod S/E #402965-003 Description of Modification: The modification installed a new deluge system with alarm provisions for the Station Blackout Transformer.
Safety Evaluation Summarva The new system is not required for the safe shutdown of the )lant nor does it adversuly affect the fire protection water system. The fire protection system remains functional such that a fire will not cause the loss of capability to safely shutdown the plant. Therefore the modification does not adversely affect nuclear safety nor reduce the margin of safety as defined in the UFSAR and basis of the Plant Technical Specifications. Hence, there is no unreviewed safety question in accordance with 10CFR50.59.
Modification:
Hardened Vent Modification S/E #402968-001 Description of Modific ation: The Hardened Vent System was installed as a result of NRC Generic Letter 89-16.
The requirement is to design and install a vent system that would permit a controlled depressurization of primary containment via the torus during a severe accident ' sequence that involve loss of decay heat removal capability.
Previously, the Oyster Creek Nuclear Generating Station (OCNGS) had a soft vent that performed the required venting operations beyond the design basis accident (DBA). The soft vent utilized the HVAC ductwork as the vent path which will fail when subjected to high pressures.
Cince the soft vent could result in significant contamination of the Reactor Building and a ground release of radioactive materials, this modification installed a. hardened vent system that will provide an elevated release point through the main stack.
Venting the primary containment was accomplished in accordance with OCNGS' Emergency Operating Procedures (EOPs).
Document Control Desk C321-93-2152 Page 16 The primary function of the hardened vent is to vent the pressure in the primary containment in the event that long-term decay heat removal capability is lost with heat input to the ::ontainment of 1% of rated thermal power and the containment pressure approaching the Primary Containment Pressure Limit (PCPL).
At this condition, venting the primary containment energy at the rate equivalent to 1% of thermal power will prevent the pressure from exceeding the PCPL.
The containment pressure will be ma' tained during this condition by repetitive venting as required.
Safety Evaluation Summary:
The Hardened Vent system which was installed as a result of NRC Generic Letter 89-16 to permit a controlled depressurization of primary containment during a severe accioent sequence that involves loss of decay heat removal capability will not adversely affect nuclear safety or safe plant operation; will not increace the probability of occurrence or consequences of an accident; will not create a possibility for an accident or malfunction of a different type than evaluated previously; and will not reduce the margin of safety as defined in the basis of Technical Specification. The removal of the rupture disc in the nitrogen purge system eliminates the potential for the bursting of the disc during the nitrogen purge operation.
This change will improve the safe operation of the nitrogen system. Therefore, the modification does not create any unreviewed safety question as defined in 10CFR50.59.
Modification:
ESW Chlorination Modification S/E #402972-001 Description of Modification: The purpose of the Emergency Service kater (ESW)
Chlorination modification was to reroute a portion of the chlorination lines and relocate the injection point to the service water cross-connection at the discharge of ESW pumps P-3-OO3B and P-3-003C.
The modification also adds isolation end check valves to the relocated lines at the intake structure.
The scope of the modification included excavation of an area south of the intake structure for access to the existing chlorination supply lines and the tie in with the 14 inch ESW lines. The chlorination lines were disconnected from the ESW pipe and re-routed above ground to the service water cross-connection above the intake structure deck.
The existing underground connection with the ESW piping was blanked with a blind flange.
Safety Evaluation Summary: Implementation of this activity will not adversely affect nuclear safety or safe plant operation.
The major portion of this modification is to the chlorination and Service Water Systems which are not safety related and perform no accident mitigation function. The tie in to the service water system will not affect operations since the affected line can be isolated without affecting the function of the system.
The existing ESW-Chlorination tie-ins were disconnected and blanked of f during plant shutdown when the ESW system is not required.
Implementation of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety Analysis Report.
The modification relocates an existing non-safety system line. Previous evaluations have demonstrated that the ESW system is capable of performing its safety related function assuming flow losses through a failed chlorination line.
The modification will mitigate the consequences of such a failure by removal of the direct connection between the chlorination system and ESW.
The new chlorination tie-ins are outside the ESW system boundary and include isolation capability so that a failure of the chlorination line will not result in any loss of ESW.
l Document Control Desk l
C321-93-2152 Page 17 The subject activity modifies the chlorination supply to the Emergency Service Water System by re-routing the chlorine supply piping to a point adjacent to the ESW pump discharge.
Based upon the above, this modification does not constitute an unreviewed safety question nor does it require a change to the Technical Specifications.
Modification:
Closed Cooling Water System Upgrades S/E #402986-004 Description of Modification: The Reactor Building Closed Cooling Water (RBCCW),
Turbine Building Closed Cooling Water (TBCCW), and Service Water Systems were modified as follows:
Local temperature gauges were installed on the inlet and outlet to the RBCCW shell side of the fuel pool heat exchangers.
A vent connection with isolation valve was added to the shell side of each fuel pool heat exchanger.
l Local temperature gauges were added on the suction of each RBCCW pump.
I The temperature gauge at TX-47 was removed.
i A drain connection with isolation valve was added to the RBCCW Surge Tank.
Inlet and outlet local temperature gauges were added to the service water i
side of the TBCCW heat exchangers.
Inlet local temperature gauges were added on the TBCCW side of the TBCCW heat exchangers.
This modification is classified as " Regulatory Required".
Safety Evaluation Summary: This mini mod adds local temperature gauges, drain j
and vent connections (with isolation valves) to the TBCCW, RBCCW and Service Water System. The TBCCW, and Service Water Systems are not safety related.
i i
The RBCCW system is only requirod to support a cold shutdown in case of an Appendix R fire, and to isolate on a containment isolation signal. The system
[
is not required for safe shutdown or to maintain the plant safety shutdown. This modification will not change the ability of the RBCCW system to meet these requirements.
In addition, all new fittings will be desi>oed and installed to meet seismic II
" Passive Integrity" criteria. Thereforn aclear safety will remain unchanged.
i This modification will have no affect on existing safe plant operation. All new j
fittings which will serve as a pressure boundary shall be designed and installed i
to meet or exceed existing system design pressures and temperatures.
This l
modification adds indication to monitor system performance.
These components will require no operator interf ace during normal plant operation. Existing plant l
operating procedures (beyond valve lineups) will remain unchanged.
This j
modification results in no unreviewed safety questions.
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Document Control Desk C321-93-2152 1
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Modification:
Core Spray System Relief Valve V-20-24 and V-20-25 Replacement
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S/E #402996-001 i
7 Description of Modification: The Core Spray System relief valves (V-20-24 and i
V-20-25) have a maximum capacity of 400 GPM.
Review of system operation shows j
that this capacity is excessive and may have contributed to several malfunctions in which these valves lifted and failed to reseat.
As a result of evaluation of these malfunctions the setpoint of the relief valves
{
was changed from 350 psi to 400 pai. Further evaluation has shown that the valve capacity of 400 GPM may have contributed to the failure of the valves to reseat.
j Evaluation also shows that these valves are not required to have an excessive capacity such as 400 GPM.
A reduced valve capacity of 50 GPM should result in
}
improved relief valve performance and core spray system reliability and capability.
This mini mod installed replacement core spray relief valves for V-20-24 and V-I 20-25 with a reduced valve capacity of 50 GPM.
The valve setpoint of 400 psig will not be changed. However valve setpoint tolerance was revised from 12% to 13% to meet ASME VIII testing requirements. Supports to the discharge line of V-20-24 were reconfigured in order to ensure that the new configuration meets seismic I criteria.
P This mini modification is classified as " Nuclear Safety Related".
Safety Evaluation Summary:
The Core Spray System provides 'a nuclear safety function in that it supports the removal of decay heat from the core following i
a postulated LOCA.
The purpose of relief valves _ V-20-24 and V-20-25 is to i
protect low pressure piping in the system, if the system isolation valves leak and allow the reactor to pressurize system piping and equipment. The new reduced I
valve capacity does not affect the relief valve's function since a capacity of
{
50 GPM is realistic with respect to isolation valve leakage. The reduced capacity
[
improves system reliability in that the reduced blowdown flow eliminates a possible contribution to failure of the valves to reseat, and improve core spray i
system capability in the case the relief valve ceas f ail to reseat.
There i
Nuclear Safety will remain unaffected.
j Review of FSAR Chapter 15 shows that this modification to the Core Spray System cannot increase the probability of any accidents evaluated. This modification l
does not change the Core Spray System's function which is to mitigate the r
I consequences of design basis accident such as a LOCA.
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Replacement of V-20-24 and V-20-25 in no way changes the NSR function of the Core l
Spray System. Therefore, nuclear safety is naaffected and no unresolved safety i
concerns exist.
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Modification:
Post LOCA H2/02 Monitoring System Modification S/E #402951-001-Descrirtion of Modification: The existing post-LOCA H2/02 Monitoring System has a history of operational and equipment problems. Part of the problems have been j
with existing solenoid operated containment isolation valves located in the sample lines of the H2/02 monitoring System. Valve problems have consisted of 1
solenoid rectifier failure, valve clogging and disc hangup. The root causes of these problems have been determined to include radiation level, ambient temperature, moisture, and airborne dirt and dust.
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Document Control Desk C321-93-2152 Page 19 The present H2/02 Monitoring System consists of two (2) redundant monitor channels. Each of the redundant channels is provided with a sample supply and return line to and from the Drywell.
Each sample line is provided with an inboard and outboard solenoid operated isolation valve.
During surveillance testing on October 26, 1989, the inboard isolation valve (V-38-41) associated with the "B" monitor supply line f ailed to open while the plant was in operation.
This failure constituted a Technical Specification Limiting Condition for operation as the failed valve could not be repaired or restored to operation without shutting down the plant. As a result, a temporary modification was made in which the existing sample supply and ret *2rn linee for the "B" line was cross-connected to the existing return line outboard isolation valve.
The sample return line was then provided with two (2) new outboard solenoid operated valves and connected up to the drywell via valve V-38-95 and penetration X-67.
This modification will return the H2/02 Monitoring System to its original sampling configuration except that the inboard solenoid isolation valves inside containment will be retired in place and replaced with new solenoid operated valves located outside the drywell.
The existing outboard solenoid operated valves will also be replaced so that the new configuration will consist of two (2) outboard solenoid operated valves in each sample supply and return line. The temporary modification will be removed.
Removal of the temporary modification will consist of removing solenoid valves V-38-93 and V-38-94, removal of their associated conduit and tubing runs, and capping off of the piping to containment isolation valve V-38-154.
Provision for leak testing the new sample line valve configuration will also be provided in order to comply with 10CFR50 Appendix "J" requirements. In addition, heat tracing of the sample supply lines is provided at 300 degrees F. nominal in order to maintain the integrity of the sample gas.
Safety Evaluation Summary: Primary containment post-accident atmosphere sampling is crucial to determine the formation of H2/02 gases; and to meet the single failure criterion, both trains "A"
and "B"
will be operational during normal plant operation.
Relocation of the new valves outside primary containment has resulted in a more favorable repair scenario. This should minimize the need to shutdown the plant since it allows corrective action to be taken within the 30-day time clock as required by the Technical Specifications.
The modification has been evaluated und determined not to present any adverse impact on nuclear safety or safe plant operations.
The modification has been evaluated and determined not to represent an Unreviewed Safety Question as defined in 10CFR50.59.
Modification:
Reactor Cavity Alternate Drain S/E #402994-001 Description of Modification: The present method for draining the Reactor Cavity requires that the Reactor Water Cleanup (RWCU) System be in service. The speed of Cavity draining is dependent of the capacity of the RWCU System.
This modification will install piping connections and valves that will permit the hookup of a temporary hose to drain the cavity by alternate means.
These connections will permit faster Reactor Cavity draining and allow draining when the RWCU System is out of service.
r Document Control Desk C321-93-2152 Page 20 Safety Evaluation Summary: This modification adds piping tie-in connections to the RWCU and SDC System piping. These modifications will not affect the normal operation of the RWCU or SDC System nor will it impact the safe shutdown of the reactor, nor will it be utilized during normal power operations. As a result, it has been concluded that the modification will not have any adverse effect on nuclear safety or the environment.
This modification is acceptable.
Modification:
Hydrolase Tap Connections for SSDSC and Reactor Cavity Drain Lines S/E #402988-001 Description of Modification:
The purpose of this Safety Evaluation was to evaluate the installation of two (2) hydrolase taps and one drain connection in the Spent Fuel Pool Cooling System.
I One 4" drain line form the Steam Separator and Dryer Storage Cavity (SSDSC) and j
another 8" drain line from the Reactor Cavity are the major contributors to the general area dose rate on elevation 75' of the Reactor Building. Numerous hot l
spots have necessitated installation of permanent shielding. For decontamination purposes, it is recommended to install two (2) permanent taps into 4" and 8" drain lines from SSDSC and Reactor Cavity.
Also an additional 1 1/2" drain connection was installed to provide a draining point during hydrolasing of the 8 inch line. The total dose reduction of 10-22 Han Rem / year is expected.
Safety Evaluation Summary: The activity of installing two (2) taps and one drain connection for hydrolasing is technically acceptable and does not constitute an i
unreviewed safety concern as defined in 10CFR50.59. This conclusion is based on:
l 1.
It does not alter SFPC System actuation, operational control or design of l
features.
J 2.
The change has been evaluated and shown that it meets the required piping, I
seismic and applicable design criteria.
i 3.
The change does not require any operator action during plant operation or
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emergency response.
i 4.
It improves the decontamination process on elevation 75'0" of the Reactor Building.
Modification:
Core Spray System Upgrade l
S/E #402996-002 t
Description of Modification:
Current Emergency Operating Procedures (EOPs) l direct the operators to cycle the core spray parallel injection valves (i.e., V-
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l 20-15, 21, 40 and 41) to control RPV water level within the specified limit during a Small Break Loss of Coolant Accident (SBLOCA).
Frequent cycling of these valves could result in valve motor failures and subsequently prevent core spray injection into the RPV or overfill the RPV.
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Document Control Desk C321-93-2152 Page 21 PSC 91-007 identified a range of small breaks (i.e., 0.05 to 0.41 ft2) in the primary coolant system which would require excessive cycling of the parallel injection valves in an effort to control RPV water level as described in the EOPs.
There is one control switch per core spray parallel injection valve pair. This plant configuration limits the ef fectiveness of controlling the RPV water level through cycling of these valves.
This modification provides independent valve control switches and indicating lights which allows alternate cycling of the valves thus increasing the actuator rest time and prevent valve motor failure.
It will therefore improve overall system reliability.
Safety Evaluation Summarva The margin of safety as defined in the safety analysis report and Technical Specifications is not reduced since the proposed modification does not affect Core Spray system ability to inject the required cooling water during a Design Basis Accident (DBA) to insure core temperatures less than 2200 degrees F.
This limit is defined in Section 3.10 of the Technical Specifications.
This modification does not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety - previously evaluated in the SAR.
It is designed to mest all the existing system design criteria. Since installation of independent control switches will increase th-rest time of the parallel injection valve motors during a SBLOCA, the probability of these valves to fail due to motor overheating decreases.
Installation of the pull-to-lock switches on the Core Spray main pumps to be used during an ATWS event has no impact on a pump (s) ability to start or the system automatic initiation during a LOCA. During normal plant operation the switches are in the " normal" position which, when required, allows the core spray pump to start automatically if the core spray pump initiation logic remains unchanged.
Therefore, the probability of equipment malfunction or its consequences as previously evaluated in the SAR, remain unchanged.
This modification to the Core Spray System does not increase the probability of the occurrence of any accident and does not affect ' safe plant operation.
Further, it does not reduce the margins of safety as defined in the FSAR and the bases for Technical Specifications. Therefore, no unreviewed safety questions as defined by 10CFR50.59 are involved.
Modification:
125V Station Batteries A & B Replacement.
S/E #403009-001 Description of Modification: This project replaces station batteries A and B.
Recently, hairline thin cracks were discovered on jar covers of some cells on both A and B batteries. At present there is no visible electrolyte leak.
The-i battery capacities are well above the minimum level required for replacement.
l However, in the absence of data regarding growth of crack, it. cannot be I
determined if the batteries would have made it through Cycle 15.
Th refore,.
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these batteries were replaced during 14R outage. The replacement batteries (AT&T round cells) will have greater - load capacity (ampere hours) to accommodate i
present load and future load growth.
The total scope of the project is as follows:
1
Document Control Desk C321-93-2152 Page 22 Replace existing A & B batteries and racks with AT&T circular cells and racks.
Remove roll-up steel door and associated columns that separates A & B batteries.
Install new conduit and conduit supports required for the new battery configuration.
Safety Evaluation Summary: The new cells are larger in capacity and have longer life expectancy. Both the cells and the racks for batteries A and B have been upgraded to seismically qualified.
Therefore, overall system reliability and availability is enhanced.
There are no changes to the electrical system.
Therefore, there are no unreviewed safety questions as defined by 10CFR50.59 or technical changes requitsd as a result of this modification.
The FSAR has been revised to reflect new rating of the station batteries, removal of the roller door and new cell configuration.
Plant surveillance / test procedures have been revised to reflect higher discharge current and higher equalization voltage required to perform surveillance / tests on the new cells.
The replacement of the batteries does nat. increase the possibility of occurrent e or consequences of an accident or malt metion, or reduce a safety margin.
Modification:
NRW Service Water System Piping Mod S/E #408800-002 Description of Modification: The above ground portion of the New Radwaste (NRW)
Service Water System piping was displaying signs of corrosion beyond repair.
This system supplies cooling water to the NRW and the Augmented Offgas System (AOG) heat exchangers.
In order to maintain cooling water flow to these heat exchangers, the NRWSW system piping must be intact.
Therefore, the original piping required replacement.
The purpose of this modification was to replace the above ground portion of this piping with materials compatible with sea water. As recommended in BRC Report No.
3731-063, "NRW Service Water System Piping Modification Pipe Material Evaluation," the piping material for this modification is Fiberglass Reinforced Plastic (FRP).
Safety Evaluation Summary:
Since this modification does not affect safety related equipment or systems, this modification will not adversely affect the plant margin of safety as defined in the Technical Specification or other licensing basis documents.
This modification replaces existing piping and valves, adding only a passive component that does not change the function of the system. Therefore, it will not adversely affect safe plant operations because it does not' change the function of any plant system nor violate any technical specification, as listed in the OCNGS License and Technical Specifications.
Control Desk C321-93-2152 Page 23 Since this modification does not affect safety related equipment or systems, this modification will have no adverse ef fects on nuclear safety or the environment.
Modification:
Installation of Test Plugs for ADS Actuation Circuit Test &
Calibration S/E #408853-003 Description of Modification:
This modification eliminates installation of contact blocks of relays and installation of jumpers on Terminal Blocks in Panels l
ERBA, ERlBA, ER18B, ER542-078, 112 through 135 to simulate the conditions when surveillance Procedure 602.3.005 on ADS Actuation circuit test and calibration is performed.
I Safety Evaluation f;ummary:
This modification provides a capability to perform l
ADS Actuation Test and Calibration Surveillance without installing jumpers and/or I
lifting leads. This modification will not alter designed functions of ADS and l
Core Spray Systems.
This modification will minimize the possibility of human error when performing the related surveillance.
It is determined that this modification does not constitute an unreviewed safety question. There is no environmental impact due to this modification and there are no changes required to the Technical Specifications.
Modification:
Installation of Test Plugs for Turbine Load Reject Scram Test at Less Than 40% of Load S/E #408853-004 Description of Modification:
This modification eliminates installation of I
control blocks on Relay Contact 1 and 2 of Relays 1K112A, 1K112B, 2K112A, and 2K112B and installation of jumpers on Relay 1K17,1K18, 2K17, and 2K18 in Control Panels 6R and 7R respectively to simulate the conditions when the Surveillance Procedure ;;o. 618.3.018 on Turbine Load Reject Scram Test at less than 40% load is performed.
Under this modification, the Weidmuller Test Plugs Type SADK-10 will be installed in the Core Spray and Containment Spray Systems Circuitry which will eliminate jumper installation and lifting leads when performing this surveillance.
The original intent of the Core Spray and Containment Spray and Turbine Control Systems design is not altered.
Safety Evaluation Summary:
This modification provides a capability to perform Turbine Load Reject Scram Test at less than 40% load surveillance in Panels 6R and 7R without installing jumpers and/or lifting leads. This modification will minimize the possibility of human error when performing this surveillance.
With the above evaluation, it is determined that this modification does not constitute an unreviewed safety question. There is no environmental impact due to this modification and there are no changes required to the Technical Specifications.
a Document Control Desk C321-93-2152 Page 24 Modification:
Gate Valve in the Bypass Line of the CRD System S/E #408853-006 Description of Modification:
The purpose of this safety evaluation was to justify installation of one new gate valve (TAG #V-15-237) and replacement of the existing leaking V-15-30 globe valve in the CRD System.
The V-15-30 valve has a 7 gpm leak at 1020 psig reactor vessel pressure.
Development of leakage through this valve can result in a significant change in CRD hydraulic balarace, and can require a plant shutdown. Replacement only of the V-15-30 valve will not completely resolve the problem.
A new double disk gate valve was installed upstream of the existing V-15-30 valve and the V-15-30 was replaced to minimize leakage through the bypass line.
Safety Evaluation Summarva The activity of installing one gate valve upstream of the V-15-30 is technically acceptab.e and does not constitute an unreviewed safety concern as defined in 10CFR50.59. This conclusion is based on:
1.
It does not alter CRD Hydraulic System actuation, operation control or design of features.
2.
The change has been evaluated and shown that it meets the required piping, seismic and applicable design criteria.
3.
This change requires some additional operator action to provide the water supply through the bypars line. This will not constitute any problem to the plant operation.
Modification:
Mux Room Corridor Sprinkler Installation S/E #408853-008 Description of Modification:
The modification installs a new branch from sprinkler system #15 with alarm provisions for the area. The new addition is not required for the safe shutdown of the plant nor does it adversely affect the fire protection water system. The new branch line into the Mux Room corridor includes (5) sprinkler heads (bulb type 175 degrees rated) and will receive its water supply from sprinkler system #15 which protects the New Cable Spreading Room.
Although the Mux Room Corridor sprinklers and the NCSR share a common supply, these areas are separated by rated fire barriers. Since it is not necessary to postulate more than one fire occurrence at the same time, this addition will not degrade the existing suppression capability of Sprinkler System #15.
Safety Evaluation Summarva The Mux Room does not include any safety related equipment required for safe shutdown of the plant. Presently all penetrations into this enclosure are sealed and as part of this installation these will be inspected and re-sealed if necessary to minimize any water impingement concerns.
In addition, due to the close proximity to the control room, (which receives a flow alarm from the system) any system activation can quickly be investigated and isolated if necessary. If at any time sprinkler system #15 is out of service the
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operations Department is required to post a firewatch and pro ide back-up suppression capability to the NCSR.
This change does not raise any fire protection concerns that can cause the loss of capability to safely shutdown the plant.
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Document Control Desk
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C321-93-2152 l
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Therefore the modification does not adversely affect nuclear safety nor reduce the margin of safety as defined in the UFSAR and basis of the Plant Technical Specifications. Hence, there is no unreviewed safety question in accordance with 10CFR50.59.
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Modification:
Outage Support Bldg. on T.
B. Heater Bay Roof (Structural)
S/E #409763-001 Description of Modification:
This modification installs three 14 x 24 foot pref abricated buildings and one 8 x 10 foot vestibule on the east Heater Bay roof at Elevation 46'-6".
These new structures are a permanent addition to the Turbine Building that serves as an accessory area for staging personnel prior to entering the Turbine Building during outage work. They also house computer and electronic equipment to control personnel access into the building and to monitor outage work on the turbine operating floor.
Safety Evaluation Summary: This modification provides permanent accessory use i
area on the Turbine Building east Heater Bay roof for staging personnel who will i
perform outage work on turbine operating floor.
All work will comply with current building codes and Oyster Creek's licensing documents.
The stiginal design conditions of the Heater Bay roof were not reduced by the addition cf the modular buildings. The installation of three modular buildings and a vestinale on the Heater Bay roof does not involve an unresolved safety question or new environwntal impact, nor does this activity adversely affect nuclear safety.
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Modification:
Domestic Water Hot Tap & Valve Installation S/E #900100-001 Description of Modification:
The purpose of this modification was to hot tap into the underground domestic water line to provide water to the toilet trailer in the ESSF.
A valve was installed with a reach rod to control the valve from the ground level. Tubing was run from the main 6" header to the trailer.
The main header was originally designed with reserve capacity with the understanding that there may be future tie-ins. The domestic water system is the only system affected.
Safety Evaluation Summary:
The proposed activity hot-taps into an existing I
domestic water line, adds two valves and run tubing from the main header to the north yard toilet trailer.
This modification will not cause an unfavorable environmental impact.
The Domestic Water System is neither required for safe shutdown of the reactor nor to mitigate the consequences of postulated accidents.
Therefore, the proposed activity will neither adversely affect nuclear safety or l
safe plant operations nor reduce the margin of safety as defined in the UFSAR and basis of the Plant Technical Specifications.
Hence, there is no unreviewed safety question in accordance with 10CFR50.59.
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Page 26 Modificetion:
Tie-In of FRCTs for use as an AAC Source to Mitigate a Station a
Blackout Extent S/E # 40296b-002 Description of Modification: The purpose of this modification is to install the equipment necessary for the interconnection of the two (2) Forked River Combustian Turbines (FRCTs) to OCNGS as an AAC power source in the event of a Station Blackout.
A Station Blackout (SBO) is defined as a Loss of Off-site Power (LOOP) in conjunction with both Emergency Diesel Generators (EDGs) unavailable to supply the OCNGS electrical buses. The general requirements for j
an AAC source for use in this application are given in NUMARC 87-00.
The FRCTs are owned and operated by the Jersey Central Power & Light Company (JCP&L) and i
are utilized as " peaking" units for the JCP&L power distribution network.
l Safety Evaluation Summary:
This rodification installs the tie-in equipment i
between the Forked River Combustion Turbines (FRCTs) and Oyster Creek non-safety related switchgear 1B.
The SBO System does not impact the operational boundary of the on-site plant electrical distribution system and is only required as a
" defense in depth" power stipply to be utilized for a Station Blackout event.
There is no change to the operation of plant safety systems, Technical Spccification requirements and limits or adverse impact on the plant environment.
No experiments or tests are performed which would adversely affect the plant's safety. Hence, this modification to utilize the FRCTs as an AAC does not affect the margin of safety or create an unreviewed safety question as decribed under 10CFR50.59.
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Modification:
Refueling Platform Upgrades S/E #323653-001 Description of Modification:
Several minor modifications were implemented to resolve problems with the refueling bridge prior to the 14R outage.
In the original design, the main load cell strain gauge had a tendency to drift under load and consequently needed frequent re-zeroing, removal of the doors from either the control or power cabinet panels resulted in a loss of power to the entire bridge, and the slack cable response was too slow so that too much of the cable became loose on the drum after the signal was initiated.
These minor modifications include:
replacement of the main hoist load cell, main hoist weight indicator / controller, and the main hoist indicator, the elimination of the refueling platform control and power panel door interlock switches and associated change to the platform control logic, and installation of the slack cable limit i
switch and the mast lug.
i Safety Evaluation Summary: These changes do not have the potential to adversely affect nuclear safety or safe plant operation. These changes do not affect the control rod and refueling platform movement interlocks.
The continued functionality of these interlocks will be proven during post installation testing. These changes do not alter the safety features designed into the bridge that mini.mize the potential for dropping or damaging f uel bundles. There are no changes to the operator controls or alarm displays.
The probability of occurrence of an accident previously evaluated in the SAR is not increased since there is no change in the refueling operation logic or design characteristics associated with the refuel bridge.
This modification will not reduce the margin of safety as defined in the basis for any technical specification.
The probability of occurrence or the i
consequences of an accident previously evaluated in the FSAR is not increased since the refueling operation is not changed and the modifications to the bridge and the refueling mast will not change existing electrical interlock functions.
l Thus, the FSAR assessment of the inadvertent criticality transients is not i
impacted by these refueling bridge modifications.
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1 Modification:
Installation of Test Plugs for Rx Protection M-G Set Generator Output Breaker Trip Test l
S/E #408853-005 Description of Modification:
This modification eliminates lifting leads to connect Voltmeter, Variacs and frequency generator and counter meters in the EPA l
panels (located in lower cable spreads ag room) when performing Surveillance Procedure No. 619.2.019.
Under this modification the Weidmuller Test Plugs and Test Terminal blocks were installed in the Rx Protection M-G set generator circuit which eliminated the i
lifting of leads and connecting the instrument lead over the existing
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terminations when performing this surveillance. The original intent of the Rx Protection M-G Set System design is not altered.
Safety Evaluation Summary: Under this mod, the margin of safety as defined in the bases of Tech Spec is not reduced because no changes in operating limits of the plant or the set points of the system instrumentation are required due to I
this modification.
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Document Control Desk C321-93-2152 Page 2 This modification provides a capability to perform Rx Protection M-G Set Generator Output Breaker Trip Test Surveillance in EPA Panels without lifting leads. This modification will not alter designed functions of RX Protection M-G Set control System.
This modification will minimize the possibility of human error when performing the related surveillance.
It is determined that this modification does not constitute an unreviewed safety question. There is no environmental impact due to this modification and there are no changes required to the Technical Specification.
Modification:
Replacement or Removal of Snubbers on Selected 79-02/14 Fire Systemo S/E #402970-001 Description of Modification:
The purpose of this modification was to replace snubbers with rigid struts or the removal of snubbers completely from selected 79-02/79-14 pipe systems. This safety evaluation justifies the removal of these snubbers or the replacement of the snubbers with rigid struts.
Safety Evaluation Summary:
Nuclear safety or safe plant operations are not adversely affected by this modification since the changes to the pipe support configuration incorporate passive restraint devices of high reliability requiring little or no maintenance and existing levels of safety are being retained.
This modification results in no increase in the probability of occurrence or l
consequences of an accident previously evaluated since no operational changes are being made to the plant and the existing design criteria is not changed, j
l This modification results in no increase in the probability of occurrence or l
consequence of a malfunction of equipment important to safety since the existing design criteria is still satisfied and the existing snubbers which require in-l service inspection are being removed or replaced with struts not requiring I
inspection and are inherently more reliable due to their passive design.
Modification:
Scram Discharge Volume Hydrolasing S/E #315302-059 Description of Modification: The purpose of hydrolasing of the Scram Discharge Volume (SDV) and Scram Discharge Instrument Volume (SDIV) is to provide
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decontamination of the 4",
6" and 10" piping and 1 1/2" and 3" instrument piping located on elevation 23'-6" in the Reactor Building.
System contamination has contributed significantly to personnel exposure.
The evaluation was performed to determine the potential impact of SDV hydrolasing that this process may have on existing safety or environmental cciditions as governed by the FSAR, Technical Specifications or other applicable documents.
l The scope of the work involves the analyses of the hydrolasing itself and the l
possible consequence of this activity, and other activities required to support the hydrolasing effort.
Safety Evaluation Summary: Hydrolasing is provided to decontaminate the part of the CRD System which is located on elevation 23'-6".
1.
The design function and performance of the CRD system are not affected.
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Document Control Desk C321-93-2152 Page 3 2.
The QA requirements are consistent with the existing system.
3.
No SAR or Tech Spec requirements are affected.
4.
The probability of a system malfunction is not increased.
5.
Area radiation fields will be measured during the, process to identify increases in the HCU area and secure hydrolasing if necessary.
Modification:
MOV Limit Switch Modification (Phase 3)
S/E #323616-002 Description of Modification: The scope of this modification includes:
1.
Relocating the "open" indication switch from (switch #7, rotor #2) to (switch #15, rotor #4) or (switch #9, rotor #3).
2.
On MOV's with two train geared limit switch assemblies:
(V-1-12, V-1-13, V-3-16, V-3-2 2, V-3-2 3, v-3-2 5, V-3-2 6, V-3-2 8, V-3-29, V-3-31, v-3-32, V-1-99 and V-1-lO3 the existing two train switch assembly has been replaced with four train switch assemblies.
3.
The relocated open indication light has been adjusted to turn OFF when the valve is between 97% and 99% closed. Due to differences in valve stroke time and location, Plant Engineering has reviewed and dispositioned valves with switch setpoints less than 97% closed.
4.
MOVATs current and switch signatures as a minimum will be taken to ensure item 3 is satisfactorily completed.
Safety Evaluation Summary: Relocating the "open" indicating switch to another switch within the valve operator compartment and replacing two train limit switch assemblies with four train will not change system or component function nor will it change operational characteristics. Therefore, the proposed modification will not have any adverse affect on nuclear safety or safe plant operation, existence of an unreviewed safety question or a need for a Technical Specification change.
This modification will provide a more accurate indication of valve position.
Modification:
EOP Jumpers and Lifted Leads Modification - Phase II S/E #323653-006 Description of Modification:
Oyster Creek Nuclear Generating Station (OCNGS)
Procedure 312.1 provides procedural instructions and administrative controls required to bypass isolation interlocks and automatic scram signals under emergency conditions or as required by Symptom Based Emergency Operating Procedures (SBEOPs).
To implement this procedure operators have to lift lead wires and/or install jumpers across terminal blocks or relay terminals in control panels located in Control Room and 480V Switchgear Room.
There is a concern regarding the ability of the test clips, used to install the jumpers, to remain attached to the heads of the screws normally found at the relays and terminal blocks. Other concerns include the risk of personal injury since the operators have to work around energized circuito, and the probability of inadvertent system or component actuation due to operator error.
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The modification described in DC-MM-323653-006 installs terminal Document boxes with Weidmuller shorting terminals in several control room panels. With this, the jumpering and lifting leads can be achieved by inserting or removing l
shorting plugs between the Weidmuller terminals.
Safety Evaluation Summarv: The subject modification does not reduce the margin of safety as defined in the basis of the Technical Specification or other l
licensing basis documents. This is true because the safety function for existing safety related systems and the set pointe at which they are designed to actuate i
are not altered by this_ modification.
i The subject modification does not adversely af fect nuclear safety because it does j
not change any NSR systems.
This modification does not adversely affect safe plant operation because it does not change the function of any plant systems nor violate any Technical Specifications.
The Weidmuller shorting terminals installed by the modification will reduce the time required to bypass isolation I
logic and scram signals during emergency conditions. In addition, the concerns about the risk of electrical shock to operators and probability of inadvertent f
actuation due to operator error will be eliminated. Therefore it is concluded the subject modification does not have any adverse effect on nuclear safety, safe j
plant operations, or the environment. This modification does not constitute an unreviewed safety question as determined by 10CFR50.59.
i Modification:
Core Spray S'ystem Temporary Monitoring Instrumentation I
S/E #328316-002 i
Descriction of Modification:
During the past several years some Core Spray l
System II supports have been found damaged, presumably due to water-hammer j
transients initiated during surveillance testing.
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Temporary installation of a monitoring instrumentation scheme on the Core Spray i
System II is proposed to support the root cause analysis of these support disturbances.
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Safety Evaluation Summary: This modification will not adversely affect nuclear safety because it does not degrade the integrity of the NSR systems with which it interf aces. This modification will not adversely af fect safe plant operations because it does not change the function of any plant system nor violate any l
Technical Specifications.
This modification will not increase the probability of occurrence or consequence of an accident. This modification does not adversely affect the performance nor degrade the integrity of the safety related systems with which it interfaces.
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This modification will not-reduce the margin of safety as defined in the basis i
of the Technical Specification or other licensing basis documents. This is true because the safety function of the existing safety-related systems and their automatic actuation. are not altered by this modification.
Therefore, this modification does not constitute an unreviewed safety question.
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Modification:
Core Spray Pump Casing Vent Line Upgrade S/E #328328-001 l
Description of Hodification: The Core Spray Main and Booster pump casing vent I
l lines have an inappropriate configuration to support easing venting. The vent l
lines are isolated by root valves (V-20-64 through V-20-71) which have 3/4" socket weld ends with no downstream fittings.
l Operators must install an unsuitable hose to the valve ends when venting is required. This arrangement is not ideal and often results in a minor leakage.
This mini mod installs at each valve a " pipe to swagelok connector" and a hose adapter" suitable for 1/2" tygon hose. This makes pump venting more convenient to the operators.
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This mini mod is classified as " Regulatory Required".
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Safety Evaluation Summarva This mini mod simply installed fittings downstream j
of the Core Spray Pump vent line root valves. These fittings support pump casing i
l venting.
The fittings will in no way affect core spray system and its safety l
function.
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H2/02 Post Accident Monitoring System Mod i
S/E #328328-002 t
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Description of Modification: The Comsip Post Accident Hydrogen / Oxygen Monitoring System at Oyster Creek is maintained by a routine surveillance which frequently l
results in deviation reporte due to either out-of-specification analyzer flow or out-of-calibration meter and recorder indications.
In addition, the amount of i
time the monitors reliably operate after maintenance is sporadic.
The following modifications increase the reliability, decrease the recurring maintenance problem, and ease the calibration, testing, troubleshooting, and
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repair of the H2/02 Post Accident Monitoring System.
i 1.
Installation of an isolation valve at the inlet to the volume chamber to i
facilitate system testing and troubleshooting.
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Elimination of dual scales on the hydrogen and oxygen analyzers (IT-l 0001A/B) in order to simplify cell calibration and maintenance.
This modification includes the installation of single range scales for the j
analyzer indicators and control room (16R) indicators and recorders as i
well as the removal of the existing range switch.
The oxygen analyzer range is 0-10% full scale and the hydrogen analyzer range is 0-30% full j
scale. The af fected plant computer points has been changed to the proper ranges.
.i Safety Evaluation Summary:
This modification will not ' reduce the margin of safety as defined in the basis of any technical specification. The plant margin l
of safety will be enhanced by the provisions of positive indication of containment atmosphere parameters to the control room operators. Eliminating the 0-10% H2 and 0-25% 02 scale does not decrease the accuracy or limit the capability of the system to provide H2 and 02 concentration information..The system meets the range requirements of 0-10% 02 and 0-30% H2 set forth in Reg.
Guide 1.97.
r This modification will not endanger the safety or health of the general public.
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This modification will improve the operational effectiveness and consistency of the H2/02 Post Accident Manitoring System.
This modification does not constitute an unreviewel safety question as determined by 10CFR50.59 Modification:
RAGEMS Purge Solenoid Valve Operator Replacement S/E #328331-001 Description of Modification: The RAGEMS system purge valves open to cleanse the RAGEMS system immediately prior to cartridge changeout or maintenance.
The valves utilize room air to " flush" the piping of possible airborne contaminants.
However, the possibility exists that a purge valve may remain stuck open.
If this were to happen, the cartridge would become " diluted" with room air and would not successfully represent the required sample of analyzed effluent.
This modification changes out the existing ASCO solenoid valves with ASCO " Proof-of-closure" solenoid valves.
These fail-safe valves are equipped with limit switches that monitor valve stem travel. The limit switches are wired in series with the Allen Bradley Unit and alarm to the Control Room should one of the valves fail to close.
j Safety Evaluation Summary:
This modification insures that the purge valves operate as required and remain closed in the event of an accident.
There is no change to the system function as a result of this modification.
This evaluation precludes the potential for an Unreviewed Safety Question or the need for a Technical Specification change.
Modification:
Hydrogen Water Chemistry System Flow Element Replacement S/E #402840-003 Descriction of Chance:
The purpose of this modification was to install new hydrogen flow elements in the existing Hydrogen Water Chemistry (HWC) System.
l The existing thermal type mass flow elements have not functioned properly since their installation and have not given accurate measurements when functioning.
Included also in the modification was the installation of new flow controllers and the replacement of valve trim in the existing flow control valves.
The previous valve trim was oversized and had a flos characteristic which is inappropriate for this application.
The flow element installation consisted of two phases, namely 1) factory testing of a thermal type flow element which has been used successfully at the Fitzpatrick Nuclear Plant and other BWR's throughout the country, and 2) l installation of the new elements into the hydrogen supply piping if the testing proves successful.
Safety Evaluation Summary: This modification corrects a significant deficiency in the existing hydrogen flow measurement system and enables the HWC System to be operated in the manner for which it was originally designed.
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Page 7 The modification has been evaluated and determined not to present any adverse I
impact on nuclear safety or safe plant operations or safe plant operations.
The modification has been evaluated and determined not to represent an unreviewed i
l safety question as defined in 10CFR50.59.
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l Modification:
Service Water Radiation Mor.itoring (SWRM) Suction and Return Line Upgrade S/E #402857-006 i
Description of Modification:
The purpose of this modification was to install l
several upgrades to the SWRM system to improve operational performance of the system.
l This modification relocates the supply / suction point of the SWRM from the seal well to a new 3" inch branch connection on the 20" service water line above and to the north of the seal well.
This is intended to reduce the amount of entrained air which is introduced into the SWRM system at the existing suction / supply point due to the high turbulence which occurs in the seal well.
The air which is drawn into tha SWRM system accumulates in system high points j
(which result in friction loss) and in the SWRM pump housing (which results in pump air binding). This situation is believed to be contributing to lower than expected flows which also results in low velocities which reduces the ability of the system to retard bio-fouling. The new suction / supply branch connection point i
is believed to have less entrained air which should result in less pump air i
binding. Additionally, since the new branch connection is physically above the l
SWRM there should be no high point air pockets since the only high point would be at the 20" line.
The SWRM system return point was also relocated to an existing 4" flange on the 30" line out of the seal well.
This new arrangement provides a static head to the system which should provide an additional force for flow.
t This modification also installs isolation valves at both new connection points
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l in order to provide isolation between the Service Water System and the SWRM i
system.
l Finally this modification replaces the existing duplex strainer with a larger l
strainer. This should decrease the frequency in which operators must clean the l
strainer baskets.
l Safety Evaluation Summary:
Since the SWRM and the seal well have no nuclear saf ety function, and are not required for safe plant shutdown, or to maintain the plant safely shutdown, this modification has no effect on nuclear safety.
This modification is intended to improve operability and maintainability of the SWRM system. No unreviewed safety questions, environmental impacts, Tech Spec changes, or any impact on nuclear safety will result from this modification.
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t Modification:
Final Rinse and Storage Tank Bypass Mod S/E #402914-002 l
Description of Modification:
This modification adds three manual saran lined diaphragm valves and associated piping.
To prevent resin from filling up the i
empty line going to the Cation Tank (which could cause plugging problems and i
increased personnel exposure during the Japan Organo (J.O.)
backwash modification), two spectacle flanges were installed. The stainless piping and saran lined valves are materially compatible with the existing saran lined piping.
Safety Evaluation Summarv This modification only adds three manual valves and l
two spectacle flanges to add flexibility in the transfer of condensate Demin Resins. There are no power or air requirements. Additionally, this modification performs no safety function nor does it replace, remove, or adversely affect any l
equipment which performs a safety function.. The condensate Demin/ Regen System is classified as "OTHER".
This modification will only affect the Condensate Demin/ Regen System. The following safety concerns were considered and determined i
not to be impacted by this modification; environmental phenomena, missile i
generation, high energy line breaks, containment isolation, environmental i
qualification, materials compatibility and electrical loading.
Therefore, it is concluded that installing three manual valves within the l
original piping design envelope to the Condensate Demin/ Regen System will neither adversely affect Nuclear Safety or Safe Plant Operation nor reduce the margin of i
safety as defined in the UFSAR and Plant Technical Specifications.
l Modification:
CPM /DW Isolation' Valve Mod S/E #402936-001 l
Description of Modification: The purpose of this modification was to eliminate l
an existing concern wherein resetting of a containment isolation signal could.
l result in the automatic opening of containment isolation valves associated with the containment particulate monitor. Specifically involved are solenoid operated j
valves V-38-9, 10, 16 and 17 which function to isolate the containment l
particulate monitor (CPM), drywell oxygen analyzer, and the post accident i
sampling system form containment atmosphere.
Upon reset of the containment isolation signal, the valves are configured to return to the pre-isolation signal position, which is open.
This is a noncompliance with NUREG 0737 and was reported to the NRC in LER 88-029-00, dated November 21, 1988, with a commitment to schedule a modification to meet the requirements of NUREG 0737 in accordance with GPUN's Integrated Schedule.
The modification involved installation of a three-position spring return to l
j normal control switch for each of the four valves with an associated control l
relay for each of the four valves. The four spring return-to-normal switches and i
the four relays were mounted on Panel 10F in the Control Room.
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Safety Evaluation Summarva The modification eliminates the probability of inadvertent opening of the CPM isolation valves when reset of the containment isolation signal occurs. This meets the GPUN commitment stated in LER 88-029-00 and provides for a greater degree of plant safety. In addition, the modification does not create an unreviewed safety question as defined by 10CFR50.59.
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Modification:
Drywell Rigging System l
S/E #402939-007
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Description of Modification: This modification adds the following equipment to i
the drywell rigging system for removal and installation of the EMRV's and MSSV's during plant outages (1) replacement booms to the four existing drywell jib cranes located 123 degrees, 150 degrees, 210 degrees and 234 degrees; (2) extension booms and cross boom to jib cranes at 123 degrees. This modification l
minimizes the mid-air transfer between jib cranes when removing or reinstalling valves and, as a result, minimizes the radiation exposure of the workers. This modification services MSSV's NR-28D, E,
F, G, H,
J, K, L and M and EMRV's NR-108A, B and E.
EMRV's NR-108C and D are not covered by this modification.
f Prior to plant re-start, the cross booms are to be removed and secured on the grating floor at Elevation 46'.
Chain hoists shall be removed. Trolleys which are used for lifting are to be removed from jib cranes at 123 degrees and 150 i
degrees. Trolleys may be left in place and additional trolleys may be stored on jib cranes at 210 degrees and 234 degrees with no more than two trolleys per jib crane. The tie rod, boom and extension boom can remain in place and tied to the i
adjacent jib cranes during plant operating condition.
Safety Evaluation Summary: This modification does not affect the nuclear safety of the plant and safe plant operation since the additional booms added are passive components that are designed for the safe shutdown earthquake (SSE) load and does not alter the performance / functions nor degrade the integrity of any
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safety related systems.
f The structural design of this modification is in compliance with all safety i
requirements codes and regulations and will not affect the safety function.
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Therefore, it is concluded that the proposed modification will not have any.
adverse effect on nuclear safety or the environment.
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Modification:
Installation of Control Air Trip Valves on V-26-16 and V-26-18 S/E #402953-012 Description of Modification:
The V-26-16 and V-26-18 control air trip valves have become obsolete.
f This mini mod replaces the existing obsolete trip valves with new trip valves and modify control air tubing and valves.
Safety Evaluation Summarv: This modification replaces existing V-26-16 and 18 l
control air trip valves which have become obsolete. This Document Control i
replacement required modification of the control air system. Modification was I
mounted seismically and will be thoroughly tested to ensure that the requirements 1
of V-26-16 and 18 are met. Therefore, this modification does not effect nuclear i
safety or safe plant operation or the margin of safety.
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e Document Control Desk C321-93-2152 Page 10 Modification:
SBO - Installation of Electrical Duct Bank and Transformer -
Trenching of Backfilling S/E #402965-001 Description of Modification:
In order to provide an additional source of off-site power for DCNGS in case of a station blackout (SBO), i.e.,
a simultaneous loss of both emergency diesel generators (EDG) and of normal of fsite power, two existing combustion turbines (CT) located at the Forked River site were connected to the OCNGS emergency busses.
This safety Evaluation addresses trenching and backfilling activities for installing an underground 13.8 KV electrical cable and duct bank system from the CT's to the new transformer in the vicinity of the Turbine Building. This safety evaluation also addresses the transport and placement of the transformer.
Electrical cable placement and termination were addressed separately. Control of work was governed by the implementation requirements and procedures as outlined in the specification. These documents have been reviewed and concurred with by Plant Engineering and Site Services divisions to ensure compliance with station requirements for control of construction within the plant protected area.
Some of the electrical, cable and duct bank work was under the security fence and in the isolation zone and requires disconnecting and reconnecting security systems.
As delineated in specification, all work in the security area was performed under the guidance and supervision of the OCNGS Security Department.
Disconnecting and connecting security systems was performed by the Security Department.
The safety aspects of the fully completed SBO modification was addressed in a separate safety evaluation.
Safety Evaluation Summary: The margin of safety as defined in Licensing Basis Documents are not being reduced during the construction phase of this l
modification because the safety functions of the nuclear safety related (NSR) systems will not be altered in any way and the structural integrity of all systems will be preserved.
The activities during the construction phase of this modification have no negative impact on nuclear safety or safe plant operation.
Based on the results of this evaluation, it is concluded that the construction phase of installing an underground electrical cables and duct bank system from the Forked River CT's to the new on-site SBO transformer, including the placement of the transformer does not adversely impact plant nuclear safety.
The construction phase of this modification does not introduce an unreviewed safety question, will not i.mpact any critical plant structures or prevent the operation of NSR or RR systems.
3 Modification:
Environmental Monitor Connection to the Plant Computer S/E #402986-002 I
Descrirtion of Modification: The purpose of this modification was to connect the environmental monitor inputs to the plant computer and to eliminate the Hewlett j
Packard processor and multiprogrammer.
The HP was replaced due to its poor l
maintenance history and spare parts are no longer produced for this unit. These problems can result in difficulty in meeting the environmental technical specifications.
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e Document Control Desk C321-93-2152 Page 11 The scope of this modification was to install cables from the control room to the plant computer via the mux room and connect the canal water temperature monitor and CWP and Dilution pump running indication to the plant computer.
This included reprogramming the PCS to provide the appropriate alarms and reports and accept inputs to the PCS from Canal water temperature monitoring system and pump running circuits for the circulating water and dilution pumps.
Safety Evaluation Summary: The environmental monitoring system verifies plant compliance to its environmental permit requirements and the requirements of Technical Specifications section 3.15 for the Service Water Radiation Monitoring System operability. It is not required for safe shutdown of the reactor nor to mitigate the consequences of an accident.
Implementation of this modification does not change any control room alarms or the information provided to the operators. Therefore, this modification does not adversely affect nuclear safety or safe plant operation. This modification does not propose an unreviewed safety question or a Technical Specification change and will not adversely affect the environment.
Modification:
EOP Jumpers / Lifted Leads Subpanel 3F Mod S/E #402986-003 Description of Modification:
Currently, the OCNGS Procedure 312.1 " Bypassing Isolation Interlocks and Automatic Scram During Emergency Conditions", requires that jumpers /lif ted leads be installed in Panel 3F to bypass the isolation of the Reactor Water Cleanup System or Shutdown Cooling System.
This requires an operator to work in close proximity to energized electrical equipment in a confined area. There is also concern regarding the ability to attach the jumpers to the pan head screws found at the respective relay terminal points.
The purpose of this modification was to install 12 Weidmuller "SAKC-lO", Nuclear Grade, shorting terminals so that in the event that Procedure 312.1 is invoked, the operator would only have to swap a shorting plug from normal to bypass for each isolation valve. The physical difficulty of location, proximity to other relays and incompatible screw head /siligator clip leads is eliminated.
The purpose of this safety evaluation was to evaluate the impact on safety as a result of this modification.
Safety Evaluation Summarv: The margin of safety as defined in the SAR and in the Technical Specification is not reduced because the installation of the EOP - 3F Isolation Bypass Subpanel does not change the bypass logic or normal system logic.
The EOP - 3F Isolation Bypass Subpanel is an enhancement to aid the operator in the performance of EOP support evolutions.
Nuclear Safety or Safe Plant Operations will not be affected by the installation of this modification because the Emergency Operating Procedure's Isolation Bypass function and purpose remain the same and this modification does not make any changes to the systems functions.
This modification only facilitates installation of isolation bypass jumpers / lifted leads during an emergency condition.
It will reduce installation time, the risk of electrical shock to operators, and insure connection integrity.
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Document Control Desk i
C321-93-2152 Page 12
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This modification to provide Weidmuller shorting terminals for RWCU/SDC Isolation 1
Bypass will not decrease the margin of safety as defined in the SAR or in the 1
Technical Specifications.
Also, it will not increase the probability of j
occurrence or consequence of an accident or malfunction of equipment important j
to safety, will not create a possibility for an accident, and will not involve j
j any radiological or environmental effluent. Therefore, this modification will i
not create any unreviewed safety questions as determined by 10CFR50.59, and will j
not cause a potential environmental impact.
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I Modificatio7:
Rod Worth Minimizer Lightning Protection Modification i
S/E #402986-005 l
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Description of Modification:
The purpose of this modification was to provide j
lightning protection for the Rod Worth Minimizer mux select circuit from the DAS
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cabinet in the old cable spreading room to the RWM3 cabinet in the SEB.
The RWM j
i has been knocked out of service by lightning strikes at OC because the select l
circuit is not protected. This mod provides protection to the mux select circuit
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by utilizing a surge protected power supply.
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The scope of this modification was to install the mux select relay cabinet in the i
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DAS cabinet in the OCSR and rewire the select circuit to use a 28v signal, which is lightning protected.
f Safety Evaluation Summary: The modification to the RWM mux select circuit was I
to prevent lightning-induced transients from adversely affecting system j
operation. This modification installs a relay panel to protect the system from j
being adversely affected by lightning by utilizing a power supply which is
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j transient protected.
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Based on this evaluation, this modification does not propose an unreviewed safety i
question, have an adverse effect on nuclear safety or safe plant operations and i
does not involve a Technical Specification change.
I Modification:
Isolation condenser Makeup Valve Fuse Installation S/E #402986-007 l
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Description of Modification:
The CLX alarm relay and the open-closed limit l
s switches currently installed in the Isolation Condenser Makeup Valve control 4
circuits are unqualified for use in a harsh environment. In the event that any
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}j of these components experienced an electrical fault there exists the possibility for the loss of power to the Solenoid for Isolation Condenser Makeup Valves', V-l 11-34 and V-11-36.
There would also be a loss of valve position indication in j
the control Room and Remote Shutdown Panel (if activated). A blown fuse would l
cause false "open" makeup valve alarm window annunciation, i
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Installation of fuses in the Isolation Condenser Makeup Valve Circuits eliminates j
the loss of the Isolation Condenser Makeup Valves in the event that CLX alarm i
i relay, or the open-closed limit switches fail.
In the event of an electrical l
failure of one of these devices, the fuse would blow and there would be a loss of valve position indication and the associated alarm would annunciate. In the 1
event of a loss of valve position indication, the Control Room operators could j
determine valve position by monitoring Isolation condenser Level; local valve indication would also be available.
In the occurrence of a blown fuse, the l
annunciator point (K-6-d) and the computer point would falsely alarm.
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<= o Document Control Desk C321-93-2152 Page 13 The purpose of this safety evaluation was to evaluate the impact on nuclear safety and safe plant operation due to the installation of this modification.
Safety Evaluation Summary: The margin of safety as defined in the SAR and in the Technical Specifications is not reduced by the installation of the fuses for Isolation Condenser Makeup Valve Fuses Installation Modification.
This modification is an enhancement to the Makeup circuits for protection from f ailure of an unqualified alarm relay and limit switches.
Valve control will be protected. Nuclear safety or safe plant operations will not be affected by the installation of this modification because the Isolation Condenser Makeup Valves function and purpose remain the same and does not make any changes to the systems functions. The modification adds protective features to the circuitry for the makeup valves.
This modification will not involve any radiological or environmental effluent.
Therefore, this modification will not create any unreviewed safety questions as determined 'y 10CFR50.59, and will not cause a potential environmental impact.
I Modification:
Elimination of Potential Hot Shorts in MOV's S/E #402986-013 Description of Modification: The purpose of this modification was to address an oversight in the original design of the Appendix R analysis with respect to the i
control circuit of motor operated valves (MOV). In typical MOV applications, the contacts of the torque and limit switches are the normal power cutoff to the MOV contactors when the full stroke position is reached.
If these switches are j
electrically located above the MOV control room switches in the control logic ladder, a hot short in the control room will bypass the contacts of the control room switches and the torque / limit switches and cause the valve to stroke. With the torque and limit switch contacts bypassed, the MOV has no means to cutoff control power and the valve will continue to operate until either the thermal I
overload / breaker trips or mechanical damage may occur. With the possibility of l
hot shorts occurring during a control room fire, several MOV's required for i
alternate / remote plant shutdown are susceptible to failure for reasons described i
above. This modification rewires af fected MOV (V-14-32, -33, -37, V-17-19, -54, i
V-37-54) torque and limit switch contacts immediately below their contactors in their logic ladder to eliminate possible valve failure due to a hot short occurring during a control room fire.
Safety Evaluation Summary: This modification was to electrically move the torque and limit switch contacts to a position immediately below the valve operators (contactors) in the valve logic ladder.
In the new configuration, the valve's torque and limit switches will retain its normal capability of isolating power to the valve at completion of a valve stroke; even during a postulated hot short event in the control room. Because the margin of safety is actually increased, performance of this modification will not adversely affect plant safety or operations nor violate the Technical Specifications or any other licensing 4
agreement.
No unreviewed safety question exists as a result of this modification.
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