ML20138H366

Cycle 18 Operating Rept

ML20138H366
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	05/02/1997
From:	VERMONT YANKEE NUCLEAR POWER CORP.
To:
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ML20138H358	List: ML20138H354 ML20138H366
References
NUDOCS 9705070152
Download: ML20138H366 (63)

Text

1 VERMONT YANKEE CYCLE 18 OPERATING REPORT l i

Between May 3,1995 and November 2,1996 Vermont Yankee implemented a number of changes requiring evaluation in accordance with 10CFR50.59(a)(2). This report includes the safety evaluation summaries of one Basis for Maintaining Operation (BMO), twenty-seven Engineering Design Change Requests (EDCRs), eight Plant Design Change Requests (PDCRs),

eleven procedure changes, thirteen Special Test Procedures (STPs), sixteen Temporary Modifications (TMs), one Installation and Test Procedure, four FSAR changes, three changes to the Core Operating Limits Report, one core reload, two Safety Classification changes, and the following additional subjects: thermal performance of RHR heat exchangers, pressure locking of a HPCI valve, RHR minimum flow valve position change, recirculation discharge bypass valve position change, Regulatory Guide 1.97 revision, RHR and CS Systems' Primary Containment boundary change, and Safe Shutdown Capability Analysis revision. In addition, two summaries inadvertently omitted from the Cycle 17 Operating Report are also provided. There were no  ;

safety relief valve failures during this operating cycle.

1. Changes in Facility Design ,

l A. Between May 3,1995 and November 2,1996, there were no changes made which '

required authorization from the Commission.

B. The following changes did not require Comndssion approval. They were reviewed by the I Plant Operations Review Committee and approved by the Plant Manager. It was l determined that these changes did not involve unreviewed safety questions as defined in  :

10CFR50.59(a)(2).  !

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Basis for Maintaining Operability (BMO) 96-05 Increase in Maximum Torus Temoerature Post-LOCA (Loss of Copjant Accident)

General Summary During the 1995 performance of the FSAR Cycle Update, it was determined that elements of the Technical Specification Amendment 88, (increase of initial torus temperature from 90 F to 100 F), had not been completely incorporated into the appropriate FSAR sections. Further evaluations during the incorporation of this information revealed that this amendment potentially impacted a number of additionalissues from those identified during Amendment 88 submittals.

It was determined that Loss of Coolant Accident (LOCA) analyses may be nonconservative for i determination of torus temperature. Specifically, modified assumptions for off-site power used l currently by General Electric have been identified that could potentially increase the maximum ,

torus temperature to 176 F with an initial torus temperature of 90 F. This is contrary to the  !

current design basis which predicted a maximum torus temperature of 166 F with an initial temperature of 90 F. Currently, the Vermont Yankee Technical Specifications allow the torus to reach 100 F. Vermont Yankee has issued Standing Order #19 requiring that the torus temperature be kept at or below 90 F until the issues identified are resolved.

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l Safety Evaluation Summafy 1

A safety evaluation was prepared for the increase in maximum torus temperature from 166 F to l

176 F assuming an initial torus temperature of 90 F. Acceptability of this evaluation is based l on the standing order which requires that the torus remain below 90 F.

The potential increase in maximum torus temperature post LOCA is not an initiator of any FSAR Chapter 14 accidents. Assessment of the effect of the change in maximum torus temperature post LOCA on all accident and transient analyses concluded that there is no adverse impact from the change to the higher maximum torus temperature. The radiological consequences associated with Control Rod Drop, Refueling, and Main Steam Line Break Accidents are independent of primary containment and ECCS performance because the release path bypasses the primary containment and ECCS does not mitigate releases from these accidents, therefore a change in maximum torus temperature has no impact. Evaluations addressing items such as ECCS pump NPSH, structuralintegrity, and equipment qualification concluded that no adverse impact results from the change to the higher maximum torus temperature.

The safety parameters potentially affected by an increased torus temperature are related to containment integrity and fuel clad integrity. Containment design temperature and pressure are not significantly impacted and ECCS performance is not adversely affected. Therefore, this change does not reduce the difference between a system failure point and accepted safety limit or reduce the margin of safety, as defined in the basis for any Technical Specification.

Technical Specification Bases 3.7.A addresses an initial torus temperature of 100 F based on Amendment 88, and demonstrates adequate condensation margin if the initial temperature is 100 F. A maximum temperature in the torus of 176 F may be reached with an initial torus temperature of 90 F based on different assumptions. However, this tech spec basis is based on complete condensation of the blowdown, which would occur prior to the maximum torus temperature. Therefore, the margin identified in TS Basis 3.7.A is not reduced by the potential increase of-maximum torus temperature to 176 F.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. The change to maximum torus temperature did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and 1 there is reasonable assurance that the health and safety of the public was not endangered. l EDCR 92-407 ECN2 Turbine Buildina Ventilation Svstem j l

GeneraLSummary This design change revised the Turbine Building ventilation system by eliminating the roof ventilators and exhausting all Turbine Building air to the primary vent stack, which allows radiation monitoring of the turbine building exhaust air using the existing sampling system in the stack. Monitoring capability is provided for normal plant operating conditions as well as accident and post-accident conditions in accordance with the requirements of NUREG-0737.

The new exhaust system consists of two 25,000 cfm centrifugal fans. A single suction is taken from the east wall of the turbine building and routed outside, where it enters the HVAC room and discharges into a plenum common to both fans. Two discharge ducts tie into the HVAC duct which continues to the main stack. Each fan is powered by a 50 hp motor. A manually 2-

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operated isolation damper was installed on the suction side of each fan. A gravity operated backdraft damper was installed on the discharge side to prevent backflow when the respective fan is not operating. Each fan is interlocked with an inlet louver to draw make-up air into the turbine building from outside. There is a local air flow measurement tap on the downstream side of each backdraft damper.

During a review of the final closeout package for this design change, it was determined that the safety evaluation required further description of effects on the control rod drop accident evaluated in section 14.6 of the FSAR. ECN-2 was written to revise the safety evaluation accordingly.

Safety Evaluation Summyy The old turbine building roof ventilators had the potential to exhaust unmonitored air from the turbine deck directly into the atmosphere. This design change disabled those ventilators and re-routed the exhaust air to the main stack. This does not alter the amount of radiation discharged from the plant; routing to the stack provides both an elevated release to minimize radiation exposures, and permanent monitoring to ensure that releases do not exceed the 10CFR100 guidelines.

The consequences of the control rod drop and the main steam line break could be affected by this exhaust system. Because this design change increases the exhaust rate to the stack, there is a small (less than a factor of two) increase in the reported consequences of the accident as presented in FSAR Section 14.6. However, the reported consequences in Section 14.6 are 3 to 5 orders of magnitude below those in Section 14.9. This occurs because Section 14.9 assumed a ground level release of fission products. The control rod drop accident reported in Section 14.9 is, therefore, the licensing basis event. With respect to the Section 14.9 analysis, the consequences will not be affected by this design change because the turbine ,

building exhaust system is assumed to be unavailable following the accident. Therefore, the consequences of the design basis control rod drop accident are not increased or affected by this design change.

The old turbine roof vents would immediately shutdown if high radiation levels were discovered in the turbine building; there was an emergency stop switch in the control room. With the new exhaust system, the fans will not be shutdown as quickly because the emergency stop switch is disabled. The additional 50,000 cfm of air exhausting to the environment does not increase the radiological consequences of the main steam line break accident for two reasons. The FSAR description of the main steam line break assumes that the pressure buildup inside the turbine building causes the blowout panels to function. The steam from the pipe break escapes through the blowout panels. The release path, therefore, is independent of either the existing roof ventilators or the new exhaust system. Also, the accident description in the FSAR assumes that all of the activity released from the reactor vessel to the turbine building escapes to the environment. The maximum possible amount of radiological material is, therefore, already assumed to be released.

l There was no increase in the probability of occurrence or consequences of an accident or malfur.ction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

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EQCR 93-405. ECN1 Modifications to Non-Seismic Branch Lines of the Service _ Water l System '

General Summary This design change modified the supply lines to the traveling screen sprays, the supply line to the chlorination system, the supply line to the circulating water pump cooling system, and the backwash lines on each of the two service water strainers. These lines had been identified as not being automatically isolable following a design basis seismic event. The licensing basis l took credit for prompt operator action to isolate leakage from damaged non-seismic lines; this l

design change relieved the operator of this burden.

The solenoid operated valves (SOVs) of the supply lines to the traveling screen sprays were replaced at each flow control valve (FCV) with a safety class electrical (SCE), seismically qualified SOV, ensuring that they operate to close the FCVs when power is lost. A restricting orifice was installed on the supply line to the chlorination system inside the intake structure downstream of valve V70-189A. A modulating globe valve was installed on the supply line to the circulating water (CW) pump cooling system, inside the CW pump room. The backwash valves on the service water (SW) strainer backwash lines were replaced with valves that are safety class 3 and seismically qualified; the actuator was replaced with a qualified single acting model that uses air to open and springs closed on a loss of air. Each SOV was replaced with a 3-way SCE, seismically qualified SOV. l Safety Evaluation Summar.y I

l These modifications limit service water leakage from previously non-seismic branch lines that may have failed during a seismic event. These modifications indirectly enhance the reliability of the station emergency diesel generators by increasing the availability of a continuous cooling water supply.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered EDCR 94-405 Vermont Yankee Meteorological Tower Upgradf1 GeneralSummary This design change replaced the instrumentation of the primary meteorological (Met) tower and the backup tower. Previously, the instrumentation installed on the primary and backup towers were supplied by different manufacturers; this design change replaced both towers' instrumentation with new instrumentation supplied by the same manufacturer. This change increases the reliability of all sensors and associated indications, and reduces the amount of spare parts needed to maintain both towers. A new variable (barometric pressure) was added to the primary tower and ERFIS computer. The solar radiation variable was removed. The strip chart recorders located in the relay house were replaced, and the associated signals combined, with two video graphic recorders; one for the primary tower, and one for the backup tower. A snow shield was installed on the rain gauge to increase the measurement accuracy during the winter months. The ERFIS DAS-E was relocated from the Met system rack, located in the relay house, to ERFIS DAS-C located in the cable vault.

Safety Evaluation Summarv i The Met system providas plant operators with local atmospheric indications and is used in calculating off-site doses following a nuclear accident, but does not provide any inputs into plant systems. The Met system is an information only system that does not provide any automatic inputs into any safety or non-safety system. The consequerees of failure of equipment installed with this design change is no worse than the consequences of failure of Met system equipment before implementation of this design change.

There was no increase in the probability of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is

! reasonable assurance that the health and safety of the public was not endangered.

J ED_CR 94-407 Reactor Water _ Cleanup (RWCU) Filter Demineralizer Control Panel MR9fadt J

General Summary This design change upgraded the RWCU filter demineralizer control logic system from one that used mechanical timers and rotary stepper switches to one that is based on a Pro ;mmable

. Logic Controller (PLC). The PLC is programmed to mimic the existing logic to provide similar i functions and outputs as the original controllogic, e.g. valve openings and closings. The l controller program incorporates additional features that allow the system to sense arect conditions that are conducive to inadvertent system spills, automatically halt the current program, and return the system to a " SAFE" condition. The PLC program generates status and error messages of the current program status and conditions which have caused the program to halt.

Since the PLC internal program emulates the existing control logic, externnt logic relays are not needed; the majority of control cabinet relays were removed. An operator interface display was installed in the control cabinet to provide associated program status and error messages. The PLC was also programmed to generate these displays.

Safety Evaluation SDmmary The RWCU filter demineralizer control system logic does not provide any automatic inputs into any safety system. The logic system is only used when a RWCU filter requires replacement.

Although the valve operation sequence has been modified to provide more precise control of the associated valves, the overall operation of the control system was not changed. During the times that this system is in use, the RWCU filters are isolated from the reactor by closed NNS isolation valves. Consequently, a failure of the controllogic cannot create a malfunction of any safety system.

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There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present I significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

EDCR 95-402. EDCR 95-402 ECN-1 Feedwater Heater E-5-1 A & E-5-1B Replacement General Summary This design change replaced the two low pressure feedwater heaters E-5-1 A and E-5-18.  ;

These heaters are located on top of the main condenser and are the first heaters in the feedwater flowpath from the condensate system. The heaters receive energy from the low l pressure stages of Vermont Yankees low pressure turbines. Inspections and evaluations  !

indicate that erosion / corrosion required the replacement.

The replacement heaters are fabricated from a combination of carbon steel, stainless steel and chrom-moly materials to address the erosion / corrosion problems. The new design will increase i

the drain flow capacity to improve operability and also help mitigate erosion / corrosion. These  ;

heaters are intended to support plant operation for at least 40 years. l Portions of the Instrument Air System and the Condenser Spray Header were also affected by

this design change. l During the course of evaluation of the condenser for the replacement of the feedwater heaters, l 4 design deficiencies in the condenser bracing was discovered. These design deficiencies were l also corrected by this ECN-1 of this EDCR.  ;

i Safety Evaluation Summary j The components installed by this design change and components associated with this l installation are neither accident initiators or mitigators. This design change will not degrade the

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level of confidence in the integrity of any barriers, affect the proximity to any safety limits, nor will it adversely affect any process variables relative to previously analyzed variables.

The analysis for the licensing basis Loss Of Feedwater Heating is an instantaneous 100 F. This

] remains essentially the same for the replacement heaters. There are no margins of safety l directly attributable to the feedwater heaters; therefore no margins of safety are reduced. l There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present

. significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a 4

reasonable assurance that the health and safety of the public was not endangered.

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l EQCR 95-0403 APRM Flow Converter Reolacement General Summan This design change addresses the inability of the APRM flow converter and power supply to maintain a stable output which periodically causes them to drift out of the Technical Specification limits. The replacement of the APRM Flow converters and power supplies would not necessarily guarantee the stability of the instrument channels as other associated components are of the same vintage. The design replaced the two recirculation flow l instrumentation channels in their entirety. This includes the APRM flow converters, power supplies and associated components. This change willimprove instrument loop reliability and allow the instruments to maintain a stable output to preclude instrument drift outside of the  !

Technical Specification requirements.

Loop recorder indicators and computer points, initially considered as Safety Class Electrical, will be equipped with safety class electrical isolators and the components maintained as Non-Nuclear safety.

Safety Evaluation Summary The APRM recirculation flow measurement loops are not accident initiators but do support accident mitigation functions. No limiting accidents rely on an APRM scram. The modifications as described within this design change were implemented to improve system reliability, accuracy and operation, and as such operational limits will be maintained in an equivalent or improved manner. Malfunctions of this equipment are not made more probable by this design as the material and construction of the components are equal to or better than those previously used. This change does not adversely affect the performance of the safety functions of the instrument loop as there are no failure modes that are different from those that existed. Due to improved system reliability, accuracy and operation, the margin of safety has increased.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EQCR 95-404. ED0R 95-404 ECN-2 Feedwater Chesk Valve Replacement General _ Summary This design change replaced the 16 inch Y-pattern lift check valves V2-278 and V2-968 with two new Anchor Darling 16 inch 900# special class carbon steel swing check valves. The l replacement was made because of cracking on V2-278 and V2-968. To allow access for

! maintenance on valve V2-278, RRU-17A was relocated two feet above its prior location. Two l structural steel struts on the main steam and feedwater anchor / restraint in the steam tunnel were modified to accommodate the width of the new swing check valves. ECN-1 was not issued due to major changes during its review cycle. ECN-2 permanently installed bypass piping from the Reactor Water Cleanup System to the RHR system, for use during shutdown which bypassed the Clean-up System inlet to the Feedwater System.

.n Safety Evaluation Summary This change did not alter the manner in which the feedwater system operates, nor did it change system design bases. The primary difference between the new swing check valves and the old Y-lift check valves is the specific mechanism which prevents reverse flow. Under system operating and design conditions there is no difference in the function of either type of valve.

Both types will seat under feedwater system operating and design pressures to prevent backflow into the feedwater system during HPCI or RCIC injection. The failure modes and consequences of a failure are identical for both the old check valves and the new check valves.

Modification of the restraint structure did not alter the function or reduce the capacity to resist loads for a pipe break.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This des!gn change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

EQCR 95-406 Core Shroud Modifications General Summary The core shroud is a stainless steel cylindrical assembly which helps maintain core geometry and provide a partition to separate the upward flow of the coolant through the core from the downward recirculation flow. Cracks developed in the circumferential core shroud welds and Vermont Yankee decided to repair the shroud during the 1996 refueling outage.

This design change is a modification which replaces the structuralload path of the welded cylinder with a system of tie rods and lateral bumpers. This design will function even if one or more of the circumferential welds fail following the installation.

Safety Evaluation Summary The core shroud is not an accident initiator for any of the design basis accidents nor can it initiate any abnormal operational transients. The cord shroud modification does not adversely affect the ability of the core shroud to maintain a coolable core geometry or achieve control rod insertion during a main steam line break or loss of coolant accident. This modification has no effect on any plant systems or equipment which prevents or mitigates failures of the four radioactive material barriers.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 95-407. EDCR 95-407 ECN-1. EDCR 95-407 ECN-2. EDCR 95-407 ECN-3 and EDCR 95-407. ECN-4 Motor Operated Valve (MOV) Design improvements including Appendix R Hot Short Modifications General Summary NRC Generic Letter 89-10, " Safety-Related Motor Operated Valve (MOV) Testing and Surveillance", requested licensees to develop a comprehensive program to ensure that MOVs will perform their safety function for design basis conditions. As part of this program, design basis reviews have been performed to evaluate valve operability for worst case accident conditions as well as degraded bus voltage conditions. EDCR 95-407 includes MOVs which have been shown by design basis reviews to require hardware modification to improve functional margin. These changes ensure that the required thrust for worst case accident and degraded bus voltage will be achieved and the valves will perform their safety function.

Changes included the re-wiring of limit and torque switches; re-sizing of overload heaters; removal of overload alarm circuits; replacement or adjustment of circuit breakers; installation of new motor control center (MCC) cubicles; installation of EGS QDCs (quick disconnect connectors); installation of test sensors; modifications to motors, actuators, and valves as required; and implementation of Appendix R modifications as well as pressure locking modifications and Smartstems.

Safety Evaluation Summarv implementation of this design change reduces the chance of a malfunction of equipment by  !

ensuring that valve motor-operator switch settings are set and maintained to assure valve I operability. The Appendix R wiring modifications help to ensure the operability of the Appendix j R valves from the Alternate Shutdown Panelin the event of a Control Room fire. Installation of i EGS connectors and standardized wiring reduces the probability of a malfunction to simplifying I electrical maintenance activities associated with the valve. The modifications to drill holes in the valve disk or bonnet reduce the probability of valve malfunction due to pressure locking.

The new line from the bonnet to the steam piping has been designed to the same design basis cor.ditions as the original piping. The modifications do not reduce the difference between a system failure point and accepted safety limit, nor in any way affect the margin of safety provided, as defined in the basis for any technical specification.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

EDCR 95-408 HPCI Pushbutton General Summary This design change replaces the present HPCI turbine trip push-button with a push-button two  !

position selector switch that allows the turbine to be tripped with an initiation signal present {

during times when water level control is not a problem. One position will be AUTO / PUSH TO TRIP which functions the same as the existing switch. The other position will be INHIBIT.

When the switch is selected to the INHIBIT position, the switch duplicates present operator

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• 1 procedural action which is to depress and hold the push-button for several minutes until the turbine has coasted to a full stop. The new switch position will relieve the operator of this task and allow him/her to perform duties elsewhere in the Control Room.

Safety Evaluation Summary The HPCI System is not an accident initiator but is used to mitigate an accident. Installation of this switch cannot cause any accident analyzed in the FSAR or any accident of a different type.

This installation does not increase the probability of a malfunction as the postulated failures of the switch are the same as those already analyzed for the original switch. The biccked operation of, or the blocked normal manual shutdown of the HPCI system are the only potential l affects of this design. However, when the switch is selected to INHIBIT an immediate control room annunciation occurs. The technical bases for this design materials and construction of this modification are equal to or better than the standards used for the existing components.

There was no increase in the probability of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

I EDCR 95-40fLEDCR 95-409 ECN-1 Service Water - Water Hammer I General Summary This design change installed vacuum breakers to protect the safety-related portions of the ,

Service Water System and the Reactor Building area from the adverse affects of water hammer I and potential flooding during postulated Station Blackout (SBO) and Appendix R fire scenarios.

The vacuum breaker assemblies consist of two series swing check valves along with maintenance and test valves and are installed in five locations at the high point of the Service Water pipes.  !

During an SBO or Appendix R fire, power to the Service Water pumps could be lost. If flow is )

not restored in approximately 1 minute the Service Water upper elevation piping could drain creating a vacuum in the piping. Upon pump restart, water hammer could result. The vacuum  ;

breakers admit air into the piping providing a cushion of air against potential water hammer and l subsequent rupture.  :

1 ECN-1 changed the safety classification requirements for the piping supports for the SCH-2 I vacuum breaker from SC-3 to NNS. It also incorporated piping and support upgrades associated with the RBAC vacuum breakers.

Safety Evaluation Sm11 mary The vacuum breakers are not required to open during Design Basis Accidents or Abnormal Operational Transients. They were added to assure the ability of the Service Water System to perform its safety function following a Station Blackout or Appendix R fire. This modification significantly reduces or eliminates manual operator actions following these events. All equipment is of equal or better quality than the original.

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Although not required, redundant series check valves were installed to assure that inadvertent flooding would not occur due to valve failure.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health eno safety of the public was not endangered.

EDCR 95-410 Intake Gates General Summary This design change replaced the Limitorque motor operators with hydraulically operated actuators on the "B" and "C" Circulating Water System intake gates at the intake structure.

This change was implemented to reduce overall maintenance costs and allow optimization of plant performance.

Safety Evaluation Summary Hydrauiic actuators have proven to be more reliable and require less maintenance than motor-driven actuators. This modification results in a more reliable Circulating Water System. It does not increase the probability of a loss of circulating water and therefore does not increase the probability of an Abnormal Operational Transient. The probability of a malfunction of the intake Gates is reduced as a result of tNs modification. The effects of Circulating Water System failures on other equipment w.ss not altered, since this modification does not alter the manner in which the system operates.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

EDCR 96-401. ECN-1 Appendix R - Alternate Shutdown Redundant Fuses General SMmmary This design change modifies the Alternate Shutdown Systems that are used to achieve hot shutdown conditions by the permanent installation of backup control circuit fuses. IE Information Notice No. 85-09 alerted recipients of potential deficiencies in the electrical design of isolation transfer switches installed outside of the Control Room. The specific concern was that fire damage, which occurred before switching to local control, could blow fuses in the motor control centers or local panels which would make equipment inoperable until the fuses were replaced.

In these cases, troubleshooting / repair would have to occur before hot shutdown could be achieved or maintained. The addition of these fuses and associated circuit address this concern.

4 Saletylvaluation summarv.

The backup control fuses installed by this design change are not accident initiatom but affect systems used to perform accident mitigation. The operation of the affected equipment has not changed and the response to a Control Room / Cable Spreading Room fire has been enhanced

' with the addition of the backup control fuses which increases availability of the circuits. No new -

accidents are created by implementation of this change a;.d the equipment installed is equal to or better than the present equipment. As this change enhances the response to a fire the margin of safety is not reduced.

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There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee F6AR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDOR 90-402 Vernon Tie Upgrade for Appendix R and Loss of NDImal Power (LNP)

Circuit Testing General Summarv in the event of a Control Room / Cable Spreading Room fire and a Loss of Normal Power (LNP),

I the Vernon tie line will be used to energize Bus 4 and supply power to alternate shutdown 2 loads. This design change was installed to allow expedited control of the 4kV switchgear associated with the Vernon tie line during the alternate shutdown scenario. This EDCR modified the 4kV switchgear buses 3 and 4 by installing a remote / local switch and permanent redundant fuses in the control circuitry to allow the expedited uso of the Vernon tie line in the event of a Control Room / Cable Spreading Room fire and an LNP. This allows more rapid restoration of power during the alternate shutdown scenario, with fewer operator actions and less reliance on support systems.

The LNP signal to Vernon tie breakers 3V and 4V was rewired to a parallel LNP relay in buses 3 and 4. This allows functional testing of the load shed trip to the Vernon breakers at the same a

time the load shed to the RHR, core spray and RHRSW pump breakers are functionally tested.

1 Safety Evaluation Summary The 4kV switchgear are used to perform accident mitigation functions or support accident mitigation functions. The changes to the load shed circuit to the Vernon tie breakers do not change the LNP logic or automatic operation of any equipment during an LNP event.

Therefore, this design change does not degrade or prevent any automatic operation of equipment required to mitigate the consequences of an occident. Installation of the transfer switch and redundant fuses provides operator flexibility to enable the Vernon tie line during the alternate shutdown scenaric, and is not involved in the automatic action circuitry of safety systems required to mitigate the consequences of an accident.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or impiicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

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l EDCR 96-403 Annendix R - SRV Hot Short ProtectlOD i

General Summary This design change was installed to address safe shutdown concerns associated with a fire in

! the Control Room, Cable Vault or Reactor Building Fire Zone RB-3, which caused a single hot l short that could energize a Safety Relief Valve (SRV) circuit. If this were to occur, the reactor vessel would be inadvertently depressurized making the RCIC (Reactor Core Isolation Cooling) system unavailable for makeup during a Control Room or Cable Vault fire. -

This change replaces the existing Automatic Depressurizing System INHIBIT switch in the Control Room with one that willinterrupt both the positive and negative legs of the circuit, isolate each SRV solenoid circuit in dedicated conduit and reroute two SRV circuits out of Reactor Building Fire Zone RB-3 and provide alternate power and control from the RCIC room.

To facilitate this change, RPS circuits were rerouted to penetrations on the opposite side of the drywell.

Safety Evaluation Summary Systems affected by this change are used for accident mitigation and are not accident initiators.

Operation of the affected equipment in the Control Room has not been altered. The installation ,

of transfer switches for two SRV's from an alternate shutdown station provides an operator with the ability to control depressurization when required. There is no change to the automatic operation of any system affected by the implementation of this change.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered. ,

t EDCR 96-404 Appendix "R" Drvwell TemoeIRLure Indication General Summarg 1 This design instahed additional drywell temperature monitoring for use during Appendix "R" safe shutdown strategies snd fire scenarios. The new equipment will monitor primary containment area temperature, using a dual element RTD, in the general area of the vessel level reference legs and SRV pilot solenoid valves. One indication was installed on the Alternate Shutdown ,

Panel in the RCIC room and the other indication was installed in the control Room on Panel 9-25.

The control room indication will be utilized for fire scenarios in the RB-3 zone and the RCIC i indication will be utilized for a fire in the control room or cable vault.

1 S;nfety Evaluat!on Summary This equipment will only be used by the operator responding to a fire emergency and has no other function than indication. This modification cannot initiate an accident nor does it provide any function to mitigate any accident. The power sources for the new meters are isolated from their safety class supplies with Safety Class Electrical fuses. This design change enhances the mitigation effort in the event of an Appendix R fire scenario by adding the capability to monitor j primary containment temperatures from either the Control Room or the RCIC room.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present l significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a 1 reasonable assurance that the health and safety of the public was not endangered, j i

EDCIL91-406 Vermont Yankee Main Septic System Modification i

General Summary l

This design change increased the capacity of the main septic system and implemented )

enhancements to increase its overall efficiency. A 3,500 gallon septic tank was installed in series with the existing 10,000 gallon septic tank to increase the storage capacity of the overall )

system. A third leach field (pressurized) was installed in the area to the north of the protected area. A pumping station and valve distribution box were installed to distribute the septic effluent j to all three fields. Automatic distribution controls were installed that distribute the septic effluent l equally to all three fields during peak usage times, as well as allow only one field to be used  ;

during normal usage times (allowing the remaining two fields to rest and clean themselves).

Also installed were two distribution boxes with increased capacity for the existing gravity leach fields.

Safety Evaluation Summarv l The Vermont Yankee Sewage Disposal System is classified and non-nuclear safety related.

The system is not part of the Technical Specifications but is described in the FSAR. This system is not an initiator or mitigator of any accidents or malfunctions analyzed in the FSAR nor

- does it affect the radiological consequences identified in section 14.9 of the FSAR. There are no margins of safety identified with this system. Therefore, the margin of safety is not affected.

There was no increase in the probability of occu..*ence or censequences of an accident or j malfunction as previously evaluated in the FSAR. This design change did not present j 1,

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a l reasonable assurance that the health and safety of the public was not endangered. I i

! EDCR 96-407 Fire SuppressjoR3ystem ExtRDalon j GencIal Summary This design change modifies the Reactor Building Pre-Action fire suppression system and the

! Recire MG set foam deluge system to ensure full compliance with the Appendix R program.

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1 The water supply to the Reactor Building Pre-Action System was increased by replacing the existing four inch main in the Turbine Building with an eight inch feed main which supplies the Reactor Building header, and by increasing the number and size of sprinklers above the j electrical trays in the Reactor Building. This satisfies the 0.30 gpm/ft2 required discharge j density. l The existing foam system was replaced with a larger system. A 150 gallon foam concentrate I tank replaced the existing 100 gallon tank and uses a ratio controller instead of a proportioner.  :

Twenty-eight new sprinkler heads in a more hydraulically efficient array replaced the existing twenty heads. This meets the required 0.16 gpm/ft' discharge density.

l Safety Evaluation Summarv Neither the Pre-Action fire suppression system nor the MG Set Foam system are accident initiators. The fire protection modifications are all Non-Nuclear safety changes and are additions or improvements to the existing systems which do not impact or change any safety class boundaries or safety classifications.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design changa did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

l EDGB 96-408 Batterv B-1-1 A and B-1-1B Rack Modifications GentiaLSUInmaIV This change modified two of the Station Battery Room Walls to restore them to conform to their seismic design basis and NRC approved criteria. Station Battery racks used two of the masonry walls in the battery room as anchor points to resist lateral seismic loading. During a l design basis earthquake the wall stresses would exceed the applicable code allowances. The anchor point for the battery racks were removed from the masonry walls to bring the walls in compliance and the racks were strengthened with new structural steel members attached to the floor and ceiling which provide an alternate seismic anchorage.

1 Safety _EyaMation Summarv '

This modification and the equipment which it interfaces with are not accident initiators. The I

, installation provides a passive structural function to ensure the batteries and walls will remain in  :

place, following a seismic event, and perform their required safety function. This modification  !

does not interfere or change any electrical features of the station batteries.

l There was no increase in the probability of occurrence or consequences of an accident or i malfunction as previously evaluated in the FSAR. This design change did not present I significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a j reasonable assurance that the health and safety of the public was not endangered. i i

M EDCR 96-410 RHR. CS and RBCCW Appendix "J" Mods General Summary This design change will install test connections and block valves for specific primary containment isolation valves to facilitate Appendix "J" testing requirements. The affected systems are the Residual Heat Removal (RHR), Core Spray (CS), and the Reactor Building Closed Cooling Water (RBCCW). The majority of the modifications involve the addition of %"

test connections. Exceptions to this are Containment Spray isolation valves which will be fitted with a test connection attached to the valve bonnet as a means of pressurizing the space between the valve discs, and the RBCCW modification which willinclude the installation of 8" block valves on either side of the RBCCW supply and return isolation valves. These changes do ,

not alter the normal operation of these systems. j Safety Evaluation Summary The RHR and CS systems connect directly to the reector coolant pressure boundary such that any failures associated with the new test connections would be classified as a LOCA. However,

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the design of the connections mitigates the possibility of the occurrence of a LOCA in that all  !

existing safety class boundaries and classification requirements are preserved, the installation l and material standards are equal to or better than existing equipment, the design is seismically adequate and each connection has a redundant isolation device. The safety class / seismic evaluation of the RBCCW loop within the containment further ensures the isolation function of that system.

The valves are all manual valves. The test connections will remain closed and the RBCCW block valves will rernain open during all operating modes such that there can be no active failures of these valves.

There was no increase in the probability of occurrence or consequences of an accident or

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malfunction as previously evaluated in the FSAR. This design change did not present I significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a  !

reasonable assurance that the health and safety of the public was not endangered.

EDCR 96-411 RHR Min Flow V10-16A/B & RRU 7/8 Appendix R Circuit Modifications )

General Summary. j This design change will ensure that the RHR/RHRSW loops are available for shutdown, if a fire I were to occur in either the West or East Switchgear Rooms. l l

This change rewires the breaker position interlocks from RHR pumps B & C to the RHR i Minimum Flow Bypass Valves A/B open circuit and the RHR Corner Room Coolers RRU 7/8 j start circuit. This modification will achieve Appendix R separation to ensure that both the "A" 1 RHR/RHRSW loop is available for a fire in the West Switchgear Room and the "B" RHR/RHRSW loop is available for a fire in the East Switchgear Room.

1 Additionally, the operability of RHR pump "A" for Alternate Shutdown will be preserved, in case of a fire in the Control Room /Cabe Vault, by placing the pump control switch in the PULL-TO-l LOCK position prior to leaving the Control Room and taking local control of the pump breaker at 1 4kV Bus 4. A new cable was installed in conduit, for the trip circuit, to provide added integrity of the PULL-TO-LOCK signal such that the pump will not start as a result of spurious signals.

i Safety Evaluation Summarv i

The circuitry associated with RHR pump "A", the Minimum Flow valves, and the corner room coolers (RRUs 7/8) are accident mitigators, not accident initiators. The manual and automatic l functions of this equipment in response to an accident was not altered by this design change.

The technical bases, materials and construction are equal to or better then the existing standards and components. Physical separation and electrical isolation between Si and Sil divisions is provided with SCE relays and fuses.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 96-412 SEPV and SDV VeJ1t & Drain Pilot Valve Replacement General Summary This design change repinced the original pairs of separate Scram Solenoid Pilot Valves (SSPVs) and the pairs of separate Scram Discharge Volume Vent & Dr$!n Pilot Valves (SDVPVs), with a new design consisting of a single valve with two solenoid r.cils. This change was initiated to address concerns with the fact that rod scram times had tended to degrade requiring replacement after a relatively short period of time. The new design uses a valve with an internal metallic shuttle where the previous design used a non metallic diaphragm which was the source of the degradation in scram times. The net result is a simpler and more reliable '

design.

This design change also replaced the SSPV air header supply isolation valves, which are NNS, as a one-for-one replacement.

Safety Evaluation SummaIy The replacement SSPVs and SDVPVs meet or exceed the material design requirement of the original valves. Small differences in power demand, capacity factor and weight were evaluated and found acceptable. The replacement valves were evaluated as more reliable than the existing valves and were tested to show improved performance. It was concluded that the replacement valves will enhance the scram function (in bo'h the normal and ATWS modes).

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the pubhc was not endangered.

i EDCR 96-413 FCV 2-20 Actuator Replacement General Summary This design change replaced the existing actuator on the upstream reactor head flange seal leak detection control valve (FCV 2-20). This replacement was undertaken because it was discovered that the existing failure mode on loss of air to the operator was to fail open while the original design required the operator to fail closed, The replacement actuator will cause the valve to fail closed on a loss of air per the original design.

Safety Evaluation _ Summary The reactor flange leak detection system is installed as a means to identify leakage via the head flange for information purposes only. Both FCV 2-20 and the downstream control valve (FCV 2-21) are maintained as safety class components as they would maintain the primary l coolant pressure boundary in the event of a failure of the inner flange seal. Per the design, FCV l 2-20 is not required to act as an isolation valve and therefore the functionality of the actuator has no safety significance. The new actuator does provide a passive safety function as part of the seismic support for the valve just as the old actuator did.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 96-414 Appendix R Cable Reroute and Combustible Free Zone Redefinition l General Summary l

This design change was initiated to correct several deficiencies in the separation of equipment used to satisfy Appendix R safe shutdown criteria. The changes included relocation of equipment, rerouting of cables, protection of cables and redefining Combustible Free Zones (CFZs).

Safety Evaluation Summary There was no change in the equipment installed in the plant as a result of this chane e. Some equipment was moved, some cables replaced with similar cable in new routing paths and the location of several CFZs modified. There was no change in equipment functions or capability and new/ replacement materials were equal to or better than the previously existing materials.

l There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

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2 EDCR 96-415 Torus Vacuum Breaker Piping Support Modificallona i l

General Summary. '

This design strengthened two supports in each of the eight torus vent lines so that they met the intent of the GE Mark l Containment Program. The 16 supports were modified so that they now resist all Mark i loads and the Torus Vacuum Breaker lines can perform their design basis l function. The modifications consisted of the addition of structural steel stiffeners to some j supports and replacement of U-bolts with larger diameter bolts on other supports. I

Safety Evaluation Summary The modifications were accomplished with materials which are equal to or better than the j

existing materials and have no cffect on any active safety functions. The modifications provide added strength to the supports so that the Primary Containment structure is more resistant to 1

structural and seismic loads.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 96-416, EDCR 91-416 ECN 1 and EDCR 96-416 ECN-2 Overprenture Protection of Drvwell Piolna General SummRIY This design change provides overpressure protection for the Main Steam Line Drain, RHR, Reactor Sampling, RBCCW and Radwaste Systems which penetrate the drywell. Protection is I provided by the installation of check valves and relief valves located inside the drywell. The intent is to provide overpressure protection of isolated piping systems whose pressure could increase due to thermal expansion which could occur as a result of temperature increases in the drywell. This could lead to failure of the piping inside or outside the drywell with a resulting breach in primary containment. Additionally, the drywell RRUs will be modified to preclude the need to provide overpressure protection by removing the ability to close their isolation valves remotely.

The bonnets on both RHR suction valves were be provided with a means to relieve excess pressure due to their susceptibility to pressure locking.

Relief valve discharges will be designed to limit the consequences of discharged water.

SAfttylYRIMatlOD.SilmmRIX The systems affected by this design change are not initiators of any analyzed accidents in the FSAR. The addition of relief protection on these systems preclude ruptures and breaches of the primary containment which eliminates additional containment leakage paths from these systems in the event of a LOCA. The design is equal to or better than the existing design; therefore the probability of failura is not increased for the existing installation. A stuck open relief valve or check valve would constitute a single failure but the drywell penetration piping i

i would remain intact. Inclusion of these valves into the IST program provides assurance that the margin of safety is not compromised.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a  :

reasonable assurance that the health and safety of the public was not endangered.

I EDCR 96-418 Core Spray Minimum Flow Bypass Logic Modification I General Summary i

This design change modified the control logic for Core Spray Minimum Flow Valves (V14-5A and B) so that they can now be manually closed from the control room when the core spray l

pump is not running. Previously the logic allowed the valve to close tut the valve (s) would immediately reopen unless the system flow was high enough for safe pump operation regardless of the pump status (on or off). This prevented the control room from using these valves to establish primary containment isolation when the pump (s) were not operating. l Safety Evaluation Summary This design change had no effect on the normal valve position or operation of the system. The change allows the operator to effectively take action which was previously prevented by the system logic. The original pump (s) protection logic will still be available when ever the pump is running and the system performance is unchanged. The change did add some cabling which is  ;

now reflected in the EQ and Appendix R programs. The new cable did not require any changes l to the basis, scenario's or analyses of either program.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 92-012 Single Cell ECCS Battery 3h8Iger GRacial Summary This design change installed a single cell battery charger to be used to provide an equalizing change to a single battery cell in either 24 Volt ECCS Battery System. This provides the capability of returning a single cell to a full charge without requiring the battery to be taken out of service it eliminates up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of equalizing charging on each battery which will significantly reduce the impact on an outage.

The charger is only connected to the battery system when a change is required and only one cell can be charged at a time.

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l j Safety Evaluation Summarv 1 i The ECCS 24 Volt DC System and the systems that it supplies are not accident initiators and do not affect the response times of the systems they supply There is no increase in the  !

I potential for a malfunction, as with all the postulate failures the system voltage will be I

maintained within the acceptable range. This design supports the original bases and reduces

. the time a battery has to be out of service for charging. Thus the margin of safety is not J reduced.

There was no increase in the probability of occurrence or consequences of an accident or i malfunction as previously evaluated in the FSAR. This design change did not present j significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a l reasonable assurance that the health and safety of the public was not endangered.

l_ P_QCB_93-008 Moisture SeparatoIMestade 1

l General Summarv

, This design change provided the necessary modifications to the internals of the main turbine 1

moisture separators to allow the use of an improved vane design. The improved design results in higher quality of outlet steam, reduced pressure losses and increased liquid removal. All changes were within the NNS system.

Safety Evaluation Summary I This change required an adjustment in the heat balance calculatior' and increases the moisture 4

separator drain flows. The results of these effects have been reviewed and no negative impacts j were found. An FSAR change was required.

i i There was no increase in the probability of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This desir,n change did not present significant l hazards not described or implicit in the Vermont Yankee FE AR, and there is a reasonable assurance that the health and safety of the public was not endangered. l 1

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. 1 PDCR 94-003 RefueLBridge Modifications - Phase _ll 4

! General Summary This design change installed a personal computer (PC) driven supervisory system in conjunction with a programmable logic controller (PLC) for control of most functions on the refuel platform. Drive systems for the bridge and trolley were replaced with variable speed AC drive systems. The main grapple assembly was replaced. A video system including a mast- l mounted camera and accompanying lights were installed to provide a close-up view of the  ;

. grapple during operation. A cable-track cable support system replaced the " clothesline" that  !

supported the cabling to the trolley as it moved. New solenoid operated valves (SOVs) replaced the unserviceable SOVs that were part of the air system.

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1 Safety Evaluation Summary The primary basis for the grapple replacement was to provide additional visual verification of  ;

positive engagement of the grapple on the bail handle. This feature, coupled with multiple  !

indications of a closed grapple, reduces the probability of a misgrapple. The modifications l provide additional operator enhancements including additional system interlocks, boundary l zones, slow zones, enhanced bridge / trolley positioning system, human factored operators '

controls, and grapple position interlocks designed to reduce the chance of operator error and to help ensure smooth, safe handling of fuel within the defined perimeter boundaries. Fuel hoisting continues to be accomplished only by manual operator control.

The new refuel bridge equipment does not alter refueling interlock logic inputs, conditions, sequences or outputs. In the unlikely event of an equipment failure which results in unexpected operation of the refueling machine, an Emergency Stop pushbutton, independent of the PLC, is maintained on the operator's console, which stops all movement of the bridge, trolley, and main hoist drives by disconnecting the 480 volt power supply. Also, in the event of a PLC failure, a keylock override switch is installed which bypasses the PLC to allow a grappled fuel bundle to be safely stored.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

PDCR 94-006 DCN1 Low Pressure Turbine Storace Facility General Summary This design change constructed a weather tight storage facility for the Low Pressure Turbine components which were replaced during the 1995 outage. Storage of the turbine components meets the criteria established for maintaining power generation through the license period. The storage facility is located adjacent to the 345 kV switchyard and consists of four fiberglass-reinforced enclosures.

This DCN was issued to detail the final conditions for storage of the components and incorporated a safety evaluation addressing the final conditions of the storage.

S.afety_Evaluatio1LSummary The enclosures are physically separated from all other plant buildings, are not Fire Control areas, and are not vital to plant safety or operation. Failure of these structures will not result in a challenge to a safety system. This change does not directly or indirectly impact the barriers credited in the FSAR for limiting the consequences associated with equipment malfunctions of safety-related components. As this change did not alter the function or performance of a safety system, it therefore did not alter the challenges the barriers were designed to mitigate. This DCN demonstrates that there are no radiological concerns associated with this change.

There was no increase in the probability of occurrence or consequences of an accident or ,

malfunction as previously evaluated in the FSAR. This design change did not present significant

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a hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that tne health and safety of the public was not endangered. l I

EDCR 94-006. DCN-1 Low Pressure Turbine Storane Facility -

1 General Summarv 3

The original safety evaluation covered changes made to the facility that resulted from buildir.g  ;

I the storage site; an additional safety evaluation, summarized below, was written to cover the '

storage of contaminated turbine components in the storage facility.

Safety Evaluation Summarv The old turbine components and the storage facility are not associated with any safety limits, or any systems where the margin of safety could be Huced. The turbine components were 3 connected to the grounding system to prevent the aildup of any static charges that may result due to the proximity of the components to the switch yard.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. Storage of contaminated turbine components in the storage facility does not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public is not endangered.

EDCR 94-009 Radwaste System Uogradas General Summary This design change permanently installed the (Rapid Dewatering System) RDS-100 unit and addressed additional radwaste system concerns with seven steps:

Temporary Modification 91-053 was made permanent. This TM replaced the radwaste  ;

centrifuge / hopper resin dewatering system with a new Rapid Dewatering System, RDS- l 1000. The primary function of this system is to reduce the percentage of water volume in resin filled casks to be shipped off-site for disposal.

The obsolete arrangement of centrifuge / hopper dewatering components were removed I from the plant and disposed of.

i A circular support for the cask lid above the blower skid was installed, for the reduction i of cask lid maintenance an

• environmental noise, and improved ability to monitor the  !

level in the High Integrity Container (HIC) while filling. l 1

Temporary Modification 93-007, which replaced the filter influent pump P95-1 A with Condensate Phase Separator (CPS) decant pump P32-1 A, was made permanent.

PDCR 94-009 replaced the air operated P95-1B with an electric drive pump.

Corresponding valves and piping were installed to support these changes.

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Floor drain collector tank (FDCT) blowdown valve LRW-562 was replaced with a two-way ball valve and a pneumatic / hydraulic rack and pinion actuator. The cartridge (cone) type strainers on the floor drain collector and waste collector pump suction lines were replaced with duplex strainers.

Temporary Modification 93-013 was made permanent This modification installed a 4" gate valve in the waste surge tank (WST) suction line to facilitate maintenance of the condensate storage tank.

Auto solenoids were installed on the P-3'i-1 A and P-35-1A seal flush lines. This allows the pump seal cooling water to be automatically initiated / terminated upon pump start /stop.

The spent resin tank (SRT) was modified to prevent resin and water from overflowing.

The 2" vent line on the SRT was replaced witn a 4" SS line. The strainer basket was removed. The hinged tank cover was gasketed and bolted to the lip of the tank.

Following removal of the centrifuges and hoppers, the empty rooms were converted over to calibration labs and radiation protection storage. ,

The monorail superstructure on top of the radwaste building was enclosed, thus eliminating the hazards of performing work in inclement weather. This also eliminates the possibility of contsmination from radwaste components being washed down the roof drain.

Safety _ Evaluation Summary i

The radwaste system was designed to process liquid and solid radwaste following transfer of the waste to a remote location. Once transferred, waste can be processed without regard to the operational mode of the reactor and primary coolant system. Since there was nothing altered in this design change which deals with the actual transfer of waste from the primary l system to the radwaste system, this change was not capable of affecting the probability of occurrence of a malfunction resulting in any abnormal operational transients.

This design change did not affect the radwaste capacity / inventory of radioactive material; this change neither increased nor decreased the potential off-site release of radioactivity due to a l failure of either the radwaste building or any component in the radwaste system. l l

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present j significant hazards not described or implicit in the Vermont Yankee FSAR, and there is l reasonable assurance that the health and safety of the public was not endangered.

EDCR.94-019 CEDtrol Room /Tufhine Loading Bay Fire Pane 1 Enhancements Generallummary This design change modified the Control Room Fire Protection Panel (CP-115-3). The steady-state load from the battery charger was removed. A Pyrotronics enclosure was installed adjacent to the fire panel. A 24VDC power supply (PS-35 module) was installed to provide l

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normal power to the Control Room Fire Protection Panel and the outputs of other plant fire protection panels supplied through fuses. This was done to prevent the frequent cycling of the 4

battery charging system caused by the previous wiring confiauration.

To prevent Turbine Loading Bay (TLB) Fire Protection Panel false detections of fire, which a

cause the deluge system to operate, the system was upgraded by replacing the UV detectors

with an Ultraviolet / Infrared (UV/lR) detector combination, which allows for work in the area  !

without removing the system from service. A new fire detection controller unit was also ,

1 installed.

4 A new 24VDC battery charger / power supply unit was installed to provide power to the Turbine Loading Bay Fire Protection Panel (CP-115-12). A 24VDC,55AH battery backup was installed to provide power to the panel should normal power be interrupted.

J l A new Weidmuller terminal strip with sufficient points was installed to replace the old terminal strip.

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Three-way valves were installed to allow for independent isolation of pressure switches (PS- l 115-1 and PS-115-2) for deluge valve DV-76-301, to assist surveillance testing and calibration.

l j . The remote alarm horn for the Control Room Panel was moved from a location inside the l j Secondary Alarm Station (SAS) to a location along the North wall within the Control Room.

i Safety Evaluation Summary The change from a UV detector to a UV/lR detector allows for improved discrimination between 1 true ignition sources / fires and casual non-fire activities such as welding and grinding. The  !

detector replacement enhances the ability of the system to detect actual fires, thereby l minimizing the need to bypass this system when UV-generating activities (i.e., weldinygr.inding, l vehicle moves) are to occur in the area. The new UV/lR sensors perform the same functions in l_ a similar fashion without an increase in the response time or a loss in sensitivity compared to the old detection system. The only system which the detectors interface with is the Turbine Loading Bay Fire Detection System. This system in no way would prevent the initiation of FSAR 1.6.2 or 1.6.3 equipment, if required, nor is the response time of any equipment important to safety impacted by this design change. 4 i I

There was no increase in the probability of occurrence or consequences of an accident or ,

malfunction as previously evaluated in the FSAR. This design change did not present l

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is l i reasonable assurance that the health and safety of the public was not endangered. I l

PDCR 94-020 RHRSW Pump Motor Coolina SMRDiy Source i General Summary This design change converted Temp Mod 93-065 as reported on page 6 of VY's supplemental Cycle 16 Operating Report (BVY 95-110, dated 4/22/94) to a permanent change. No additional )

modifications were required to make the change permanent and the original summary requires '

no cprrection.

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t Safety Evaluation Summary The original evaluation was reviewed and no revisions were required.

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There was no increase in the probability of occurrence or consequences of an accident or i 3

malfunction as previously evaluated in the FSAR. This design change did not present significant l hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Procedure OP2180 Revision 30, Circulating Water / Cooling Tower Operation GRDeral Summary The latest issuance of the plant's National Pollution Discharge Elimination System (NPDES) permit deletes the thermal limit for closed cycle. The previous limit for cooling tower blowdown was 93 F maximum. FSAR section 11.6.3, Description, states that "A spray pond is used to limit blowdown temperature during closed cycle operation." Since the NPDES permit no longer requires limiting blowdown temperature, the blowdown spray pond will no longer be used to limit blowdown temperature. The FSAR states that a spray pond is used to limit temperature, a safety evaluation was prepared to assess the safety impact of deletion of spray pond operation

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from OP2180 Revision 30. j Safety Evaluation Summary The blowdown spray system does not contribute to nor mitigate any accident previously evaluated in the FSAR. Thus, the blowdown spray system st;tus cannot affect the consequences of occurrence of any accident previously evaluated in the FSAR. The blowdown spray system is not an initiator for any of the abnormal operational transients in the FSAR, nor does the blowdown spray system interface with any safety systam nor any system that could inhibit successful performance of a safety function. This change does not modify any system, structures, or components. This change simply stops operation of an installed system. The system is essentially in the condition it would be in during cold weather periods. Tbn this change cannot create the possibility of an accident occurring which is different from FSAR analysis.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Procedure OP3E5 Revision 18. EmitgBDGyfteparedness Exercians_andatilla (lmplementing Procedure to the_V. Y E-Plan)

. Generals.ummary This procedure was revised to reflect the change in management responsibility for emergency planning and interface with local and state govemments from the Director of External Affairs to the Operations Support Manager. The relocation of this function provides a more direct access f

to additional technical resources to enhance the level of planning and maintenance for implementation of the Emergency Plan.

4 Safety Evaluation Summary This change ....ered an administrative procedure contained in the FSAR. This change to the management structure for the emergency planning group was not an accident initiator, nor did it result in new plant operating modes or configurations which could result in the malfunction of a safety-related component. This change to the management structure did not require a change to the technical specifications, or change a parameter or equipment operating condition specified in the technical specifications.

t There was no increase in the probability of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Procedure OP4100 Appendix B. Revision. Revision 22 ECCS Automatic Inlilation Test General Summary Appendix B to OP 4100 addresses the Diesel Generator Transient Performance Monitoring.

The implementation of this Appendix affects the emergency diesel generator metering and indication circuitry.

I This Appendix installs temporary recorders to the diesel generator metering and instrument

} circuits during the integrated ECCS test to record generator: stator amps, line voltage, field voltage, kW, and frequency. These connections do not jeopardize the ability of the diesel to  !

start or energize emergency loads.

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. Safety Evaluallon 2

The emergency diesel generators are not initiators of any accidents analyzed in the FSAR.

implementation of this appendix did not affect the availability of the diesels to provide power to those systems necessary to limit the radioactive release from accidents analyzed in the FSAR.

This equipment was only in place during a plant shutdown when postulated transients will not occur. The margin of safety was not reduced as the testing provides additional data to assure that the diesels meet the requirements of Technical Specifications.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

i Procedure OP4111&ontrol Rod drive System Surveillance General Summary This procedure change addressed the IST testing of the scram discharge line ball check valves on the control rod Hydraulic Control Units (HCU). This change provided a methodology to safety perform a " closed" test of these valves, each refueling outage. This will assure that reverse flow from the scram discharge volume cannot occur following a scram that is not quickly reset when reactor pressure has decreased following a LOCA or ADS actuation.

Reverse flow could potentially be exerted on the top of a control rod drive piston and drive the rod out of the core if it were unlatched with the drive overtravelled. 1 Safety Evaluation Summary The reactor manual control system and the HCU's are not accident initiators. This test was l performed with the reactor shutdown and the reactor cavity flooded. This test does not change the worth of any control rod and therefore cannot increase the radiological consequences of the control rod accident analyzed in the FSAR. The test established a manual drain path from the reactor to radwaste and is restricted through a 1/2" connection. The volume in the reactor cavity provides assurance that the reactor will remained flooded. A scram was put in place during the test and all rods were verified inserted. The margin of safety did not change as there was no I impact on reactor power limits, clad integrity, pressure limits or boundaries.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present 1 significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a l reasonable assurance that the health and safety of the public was not endangered.

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l Etocedure OP4114. Standby Liquid Control System Surveillance  !

General Summary This procedure change permits using an increased Boron concentration in the Standby Liquid Control (SLC) test tank water during the once-per-cycle flow test directly into the reactor.  ;

The previous minimum concentration allowed was <100 ppb. Because of the difficulty and time j involved to achieve this concentration it was desirable to raise the limit to one that is more readily achievable, yet ensures that no undesirable conditions are created. The new limit is 40 )

ppm.

i Following injection of the total test tank volume at 40 ppm, the resulting boron concentration in l the reactor vessel would be 0.3 ppm. As the specified impurity concentration for boron in l reactor water is less than 1 ppm the transient boron level of 0.3 ppm will not cause any adverse  ;

affects on the Reactor Pressure Vessel or interrtals.

Safety Evaluation Summary The FSAR states that a boron concentration of <50 ppm will have no effect on reactor  ;

operations and therefore will not affect any of the accidents or transients evaluated in the FSAR.

Further, the Shutdown Margin Check procedure now requires that the boron concentration be verified to be <100 ppb or a determination is made regarding the affect of boron concentration '

on critical values used for shutdown margin demonstrations. '

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Procedure OP 4181 Revision 26 Service Water / Alternate Cooling System Surveillanca General Summary To perform In-Service Testing (IST) flow testing of a single SW pump by using the fire pump test station, one side of the SW system must be isolated by closing either the SW-5A or 5B valve (cross tie line 20" SW-3 isolations) and the companion pump secured. This puts the system in a configuration in which both pumps on a header side are out-of-service with a cross tie line 20" SW-3 closed. This system configuration could not have met its mission under limiting design basis assumptions. .

The SW-19A&B low pressure isolation logic must be bypassed during flow testing to ensure that the Diesel Generator on the opposite header is not isolated from SW cooling if a low pressure signal is received. If the SW-19A&B valves remain open when a low header pressure signal is received, it is expected that the SW 20 valve will close to ensure that the non-safety SW loads are shed, thereby directing all SW cooling to safety required functions. To perform the SW pump in-Service Testing (IST) test / post maintenance test (PMT), the procedure section for SW Pump Capacity Test was changed: The SW-19A & 19B valve low pressure isolation logic will be bypassed (such that these valves remain open while the flow test is being performed); pressure on the in-service system will be maintained greater than or equal to 105 psi; a seven day Administrative LCO duration (versus the permitted 15 day LCO duration  !

specified) is being entered when SW-8 is open, SW-5A/B is closed and the SW-19A/B valve low pressure isolation logic is bypassed.

Safatv Evaluationjummary A turbine or generator trip is an initiator of abnormal transients evaluated in the FSAR. These -

l transients could occur by inadvertent isolation of the Service Water in the turbine building.

l Although low pressure isolation logic for the 19A and 19B valves is bypassed, the backup valve, l SW-20, will be operable to isolate nonessential turbine building loads if required. Non-safety l

loads can be isolated so that the service water can be directed to provide necessary equipment cooling with no resulting impact on radiological consequences. Alternate testing required by Technical Specifications will be performed to ensure availability of the Alternate Cooling system j to support mitigation of accidents.

l There was no increase in the probability of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This procedure change did not present ,

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is  !

I reasonable assurance that the health and safety of the public was not endangered.

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l Procedure OP 4378. Revision 24. Excess Flow Check Valve Functional Test General Summary The procedure change revised the acceptable leak rate limit for the instrument sensing line excess flow check valves from 1 gpm to 1.5 gpm based on an engineering analysis that states that this will not impact 10 CFR 100 off-site dosage limits.

Safety Evaluation SMmmaly This equipment is not an initiator of any of the accidents or transients analyzed in the FSAR and the maximum potential flow is bounded by existing analyses. Reactor coolant inventory decrease through these excess flow check would be negligible compared to the current analysis. There is no reduction in the margin of safety and any additional radioactive release is bounded by the current analysis.

There was no incr62c3 in the probability of occurrence or consequences of an accident or malfunction as previoutly evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable 43surance that the health and safety of the public was not endangered.

Emcedure OP4623 Reyision 19 Sampling and Treatment of closed Copling_ Water Systems General Summary Revision 19 added action J," Filtration / demineralization of the Reactor Building Closed Cooling Water (RBCCW) system", to provide for the installation, operation, resin and filter changeout and removal of the cleanup skid. This section was added to eliminate the recurrent use of the Temporary Modification process. The portable skid consists of a mechanical filter, carbon filters and mixed bed deionizers.

Safety Evaluation Summary The filtration system does not interfere with normal operation of the RBCCW system, nor with the ability of the RBCCW system to support the safety design bases. The portions of the RBCCW system required to mitigate the consequences of an equipment malfunction are not affected by the installation of this temporary system.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

e # 1 Procedure OP5230. Revision 10. Refuel Platform faeneral Summary This procedure change allows the physical removal of track switches from the Refueling Platform to allow the platform to be moved out of a Foreign Materials Exclusion zone and allow maintenance to be performed on the Refueling Platform more safety and efficiently.

{

These switches provide control rod block interlocks that prohibit control rod withdrawal during fuel movement in the vessel and also prohibit platform movement over the core in any mode switch position except " REFUEL". This procedure ensures that this work is only performed while the Mode Switch is in "RUN" or the reactor cavity shield blocks are in place when the refuelinterlocks provided by these switches are not required. The procedure also verifies the reinstallation of these switches and checks their operability.

Safety Evaluation Summary The refueling interlocks are not accident initiators. This procedure is only performed with the j reactor vessel assembled, therefore, no refueling accident can take place as a result of this '

change and no transient analysis is affected. This change does not introduce any new failure modes for the refueling interlocks. There is no operability requirements for the refueling l interlocks while the reactor is assembled and the cavity shield blocks are in place. Therefore, there is no effect on the margin of safety.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present i significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered. i P_rocedure OP5245. Revision 4. 4kV Switchgear inspection and Testing General SummAIY l This procedure change provides a means to supply temporary power to the Stack Gas Sample i Pumps during 4kV switchgear Bus 1 and 3 inspection, which results in a loss of the normal power source to the pumps. Technical Specifications require a minimum number of stack gas channels to remain operable. This change maintains the operability of the Stack Gas Sample pumps.

Safetv Evaluation Summary Temporary power to the stack gas sample pumps are not accident or transient initiators or l mitigators. The stack gas sample pumps provide monitoring capabilities of all releases including the time they are supplied by temporary power. The temporary power does not create any different type of malfunction then that previously analyzed. This configuration does not affect any margin of safety.

l There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present A

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

e Etonedure OP5371 Revision 4  !

Gereral Summary Revision 4 of OP5371 added sections U and V, " Refueling Bridge Travel Over the Core with '

Reactor Mode Selector Switch (RMSS) in Shutdown - Alteration and Restoration" This activity I had previously been performed by Temporary Modification. The procedure involved installation of a jumper in the Reactor Manual Control System (RMCS) to allow refueling bridge movement ,

over the core independent of RMSS position. This is done to allow core inspections, LPRM replacement and other non-core alternations to occur with the RMSS in Shutdown. The single rod permissive was still be in place with the RMSS in Refuel under this condition.

Safety Evaluation Summarv >

Failure of the refueling interlocks is neither an initiator nor a mitigator for any accidents. The l

only applicable design basis accident is the Refueling Accident. The initiating cause of the Refueling Accident is the dropping of a fuel assembly. The addition of procedure sections U and V does not increase the chances of a fuel assembly being dropped. No fuel movement is permitted during the implementation of Section U. Before starting fuel movement, the mode switch must be locked in the refuel position. A caution tag is placed on the mode switch to keep the mode switch in SHUTDOWN while Section U of this procedure is in effect; this tag will not i be removed until the jumper is removed. With the mode switch in SHUTDOWN it is physically 4 impossible to withdraw a control rod.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This procedure change did not present ,

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Special Test Procedure 95-001 Turbine Performance Testina_

General Summary This special test procedure delineates the method used to determine the performance or

capacity of the turbine-generator following the installation of new LP turbines and moisture I separator vanes. The test will determine maximum capacity or maximum generator output corrected to the manufacturers design conditions. Test results will be compared to the special

! test done prior to the modifications. Temporary instrumentation was installed to monitor various parameters. ,

Safety Evaluation Summary The special instruments and tubing installed to accommodate this testing are not accident initiators and will not change the operations of any systems.

There was no inciease in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present i

i

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a i

reasonable assurance that the health and safety of the public was not endangered.

1 Sagtdal Test Procedure 95-005 Turbine / Generator Startup from the 1995 Refueling l Qutaan 2

General Summary Special Test Procedure 95-005 was conducted to ensure proper startup and operation of the .

~ Turbine Generator System following replacement of the low pressure turbines during the 1995  !

refueling outage. This startup was performed with new low pressure turbines including rotors, '

diaphragms and inner casings. I With the tight clearances in the retrofit turbines, the possibility of encountering " rubs" during startup and subsequent loading of the machine was increased. Rotor rubbing occurs when the

! turbine rotating and stationary components contact while the turbine is turning. The result of

, rubbing is localized hot spots on the rotor surface at the point of contact. The heat developed during a rub causes the rotor material at the point of contact to expand compared with the rest of the rotor. This non-uniform expansion causes thermal distortions resulting in rotor bowing.

As a rotor bows, its centerline of mass and center of rotation move relative to each other resulting in changes in vibration level. Because of this potential, additional vibration monitoring i instrumentation was temporarily installed to al!ow startup personnel to more accurately monitor i the turbine startup.

Safetv Evaluation Summary The low pressure turbines are non-safety components, and the temporary instrumentation i installed was non-intrusive and did not affect the operation of any plant equipment. This special i test did not operate any systems differently than stated in norma! operating procedures, with i

the exception that the automatic Turbine Supervisory Instrumentation (TSI) trips were not j bypassed during startup. This provided more conservative protection and guidance to quickly i decrease condenser vacuurn to rapidly slow the turbine if necessary.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.  !

Spadallest Procedure 95-012 In-Situ Differential Pressure Testing of Valves HPCl-14.

HP_CI-20. HPCI-21 General Summary This test procedure was performed to monitor and trend the Residual Heat Removal (RHR) pump suction pressures in the torus cooling mode, during the HPCI surveillance test, to demonstrate the ability of the RHR pump suction strainer to remain free of foreign material.

This procedure addressed concerns identified in NRC Bulletin 95-02, " Unexpected Clogging of i

a Residual Heat Removal Pump Suction Strainer While Operating in Suppression Pool Cooling Mode" Safety Evaluation Summary During the testing, the RHR and HPCI systems remained operable and capable of fulfilling their safety function. All automatic actuation signals and responses remained operable. When both RHR "B" loop pumps were operating, RHR-658, the RHR pump discharge valve, remained open to ensure heat exchanger design flow and differential pressure would not be exceeded.

Stringent test criteria ensured the testing would be terminated before any adverse strainer clogging occurred. The temporary test instrumentation had no control function and did not affect any of the automatic features of the RHR or HPCI systems. Installation requirements and procedural controls ensured pressure boundary integrity.

There was no increase in the probability of occurrence or consequences of an accident or  ;

malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or impicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered. I i

Special Test Prosedure 95-012 in-Situ Differential Testing of Valves HPCI-14. HPCI-20  !

and HPCI-21 General Summaty This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. Three valves were tested under this procedure. HPCI-14, steam supply to the HPCI turbine stop valve, HPCI-20, discharge shutoff valve, and HPCI-21, pump test line to Condensate Storage Tank valve. These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment.

Safety Evaluation Summarv 1

The HPCI System is not an accident initiator and serves to limit the release of radioactive materials to the environs. Backup systems were available when this system was taken out of service for the duration of the test, therefore there would be no increase in the consequences of an accident previously evaluated. The system was operated in a manner consistent with the normal HPCI pump and valve surveillance such that there was no increase in the probability of a malfunction. The margin of safety was not reduced as backup systems were available in accordance with Technical Specifications.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

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Special Test Procedure 96-001 In-Situ Differential Testing of Valves RHR-25A and RHR-27A 1 General Summarv l

This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. Two valves were tested under this procedure, RHR-25A and RHR 27A, inboard and outboard injection valves, respectively. Both valves have a safety function to open automatically upon receipt of a LPCI (Low Pressure Coolant injection) injection signal. These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment.  !

Safety Evaluation Summarv Neither the LPCI nor the shutdown cooling portions of the RHR System are accident initiators.

When testing was performed, the plant was in a refueling outage, the "A" RHR Subsystem was l declared inoperable and the "B" system was in the Shutdown Cooling Mode. The components were tested in a manner consistent with the normal surveillance therefore there was no l increase in the probability of a malfunction. The margin of safety was not reduced as backup I systems were available in accordance with Technical Specifications There was no increase in the probability of occurrence or consequences of an accident or l malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Special Test Procedure 96-002 In-Situ Differential Pressure Testina of Valve RHR 258 l and RHR 27B i

General Summarv )

This test was written to gain information on motor operated valve performance when operated i under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. Two valves were tested under this procedure, RHR-25B and RHR 278, inboard and outboard injection valves, respectively. Both valves have a safety function to open automatically upon receipt of a LPCI (Low Pressure Coolant Injection) injection signal. These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment.

Safety Evaluation Summary Neither the LPCI nor the shutdown cooling portions of the RHR System are accident initiators.

When testing was performed, the plant was in a refueling outage, the "B" RHR Subsystem was declared inoperable and the "A" system was in the Shutdown Cooling Mode. The components were tested in a manner consistent with the normal surveillance therefore there was no increase in the probability of a malfunction. The margin of safety was not reduced as backup systems were available in accordance with Technical Specifications There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Special Test Procedure 96-003 in-Situ Differential Pressure Testing of Valves RHR 16A and RHR 39A General Summary This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. Two valves were tested under this procedure, RHR 16A, a minimum flow isolation valve, and RHR 39A, the Suppression Chamber Sprayfrest isolation valve. RHR 39A has a safety function to close either upon receipt of an  !

RHR initiation or a Primary Containment isolation Group ll isolation to assure primary I containment integrity. These tests involved both static and dynamic tests with measurements I taken by temporary MOV diagnostic test equipment.

Safety evaluation Summary Neither the LPCI nor the Containment Cooling portions of the RHR Systems are considered  !

accident initiators. When testing was performed, the plant was in a refueling outage and components were tested in a manner consistent with the normal surveillance; therefore there was no increase in the probability of a malfunction The margin of safety did not change as the Technical Specification requirements, for the ,

removal of this system for testing, were met. l There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present  ;

significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a I reasonable assurance that the health and safety of the public was not endangered.

Special Test Procedure 96-004 In-Situ Differential Pressure Testing of Valves RHR-16B, RHR-34B. and RHR-39B l

GenRIal Summary I This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. Two valves were tested under this procedure, RHR 16A, a minimum flow isolation valve, RHR-34B, the Suppression Chamber Spray Bypass isolation / throttle valve, and RHR 39A, the Suppression Chamber Spray / Test isolation valve. Both RHR 34B and RHR 39A have a safety function to close either upon receipt of an RHR initiation or a Primary Containment Isolation Group ll isolation to assure primary containment integrity, These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment.

4 Safety Evaluation Summary Neither the LPCI nor the Containment Cocling portions of the RHR Systems art asidered accident initiators. When testing was performed, the plant was in a refueling outage and components were tested in a manner consistent with the normal surveillance; therefore there

, was no increase in the probability of a malfunction The margin of safety did not change as the Technical Specification requirements, for the l

removal of this system for testing, were met.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a

f reasonable assurance that the health and safety of the public was not endangered.

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) Special Test Procedure 96-006 In-Situ Differential Testina of Valve CS-12A 3

i General Summary This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to conduct this test was a result of Generic Letter 89-10. One valve was tested under this procedure, CS-12A, an injection isolation valve for the "A" Core Spray System. CS-12A has a I safety function to open, when closed, upon receipt of a Core Spray System initiation signal and to close for a primary containment isolation signal. CS-5A is the minimum flow valve. This valve is required to close for injection and open to provide pump protection. Performance testing of this valve was not required but was assessed to monitor its influence on the l differential pressure imposed on CS-12A as the two valves changed position. These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment.

Safety Evaluation Summarv This test was performed with the Drywell and Reactor Pressure Vessel heads removed. The only transient or accident parameter variations which could occur are coolant inventory increases or decreases. During the test, water level was administratively controlled above the Top of the Active Fuel. This ensured that core submergence was maintained and flooding

! didn't occur. The Core Spray spargers are located above the reactor core, as such, any failure during testing would not result in draining the reactor pressure vessel. Technical Specification compliance for the Core Spray System was maintained throughout the test; therefore the margin of safety was unchanged.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasona ale assurance that the health and safety of the public was not endangered.

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Special Test Procedure 96-007 In-Situ Differential Pressure Testing of SW Valve V70-19B General Summarv '

This test was written to gain information on motor operated valve performance when operated under specific test conditions of differential pressure and flow rate. The commitment made to -

conduct this test was a result of Generic Letter 89-10. One valve was tested under this procedure. Service Water (SW) V70-19B an isolation valve in the Service Water system that is normally open and automatically closes on low header pressure during accident conditions.

Under post-accident conditions, this valve has a safety function to close by remote manual control from the Control Room. This test was conducted during a p! ant refueling outage when Service Water to the Turbine Building was not required. These tests involved both static and dynamic tests with measurements taken by temporary MOV diagnostic test equipment. ,

Safety Evaluation Summary Testing of V70-19B is not an initiator of any of the accidents or transients identified in the 1 FSAR. Core coolant temperature increase, due to a loss of SW flow the RHR SW system was

considered during this test. Alternate testing was performed on alternate isolation valves that isolate the non-safety class portion of the SW system to ensure SW flow to the RHR SW system. During the test, the isolation valves will be exercised consistent with normal valve surveillance. The testing was performed within the requirements of the Technical Specifications I such that the margin of safety did not change.

There was no increase in the probability of occurrence or consequences of an accident or l malfunction as previously evaluated in the FSAR. This special test procedure did not present

! significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered. l Special Test Procedure 96-008 In Situ Differential Pressure Testina of Valve CS-5

)

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! General Summarv  !

This test was written to gain information on motor operated valve performance when operated i under specific test conditions of differential pressure and flow rate. The commitment made to i conduct this test was a result of Generic Letter 89-10. One valve was tested under this

, procedure. CS-5A a minimum flow valve for the "A" Core Spray System, is normally open and is required to close for injection and open to provide pump protection. CS-26A is the full flow j test valve for the "A" Core Spray System. Performance testing of this valve was not required

, but was assessed to monitor the pressure affects produced by CS-5A on CS-26A. These tests 1 involved both static and dynamic tests with measurements taken by temporary MOV diagnostic l test equipment.

Safety Evalttalion Summarv .

The Core Spray Systems are not accident or transient initiators. The testing was conducted in j accordance with approved procedures and only involved the Full Flow Test mode. As such the

, flow was from Torus to Torus. This test was performed with the Drywell and Reactor Pressure Vessel heads removed. The only transient or accident parameter variations which could occur are coolant inventory increases or decreases. However, the Core Spray spar 0ers are located

above the reactor core, as such, any failure during testing would not result in draining the reactor pressure vessel. Technical Specification compliance for the Core Spray System was maintained throughout the test; therefore the margin of safety was unchanged.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Sendal Test Procedure 96-009 East Switchgear Room Enclosure Integrity Test General Summary This test was conducted to collect test data that was used in a subsequent analysis to confirm the integrity of the East Switchgear Room as an Automatic Carbon Dioxide (CO 2) System protected enclosure and test its ability to contain the gas at the required concentration for the prescribed retention time.

Two tests were performed to completed this Special Test Procedure, a door fan test and a tracer gas dilution test. The door fan test and tracer gas test was performed under the same conditions which the Switchgear Room would be subjected to during an actual CO2 discharge.

The door fan was used to pressurize and depressurize the enclosure to determine its Equivalent Leakage Area (ELA)in square inches. The Tracer Gas Dilution Test was used to determine the air exchange rate of the Switchgear Room caused by leakage or pressure differentials.

Safety Evaluation Summary Cables and equipment located in the Switchgear Room are not accident initiators and do not have the potential to cause the occurrence of an accident evaluated in the FSAR. The probability of a malfunction will not be increased as the temperature in the Switchgear Room will only slightly increase, for the duration of the test, while the HVAC is secured. Although the CO2 System will be inoperable for the duration of the test, continuous, trained, fire watches will be present. The SFe gas will not adversely affect the operability of any plant equipment powered or controlled by cables or instrumentation located in the Switchgear Room, therefore the margin of safety is maintained.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

SERCIAI Test Procedure 96-010 West Switchgear RoomEnGlature Integrityleat General Summary This test was conducted to collect test data that was used in a subsequent analysis to confirm the integrity of the West Switchgear Room as an Automatic Carbon Dioxide (CO2) System

. . - .=-- . . -- - - -

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l protected enclosure and test its ability to contain the gas at the required concentration for the l prescribed retention time.

Two tests were performed to completed this Special Test Procedure, a door fan test and a tracer gas dilution test. The door fan test and tracer gas test was performed under the same condKions which the Switchgear Room would be subjected to during an actual CO2 discharge.

The door fan was used to pressurize and depressurize the enclosure to determine its Equivalent Leakage Area (ELA)in square inches. The Tracer Gas Dilution Test was used to determine the air exchange rate of the Switchgear Room caused by leakage or pressure l differentia!s.

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' Safety _Ey;,luation summary Cables ax equipment located in the Switchgear Room are not accident initiators and do not -

have the potential to cause the occurrence of an accident evaluated in the FSAR. The 1

probability of a malfunction will not be increased as the temperature in the Switchgear Room will only slightly increase, for the duration of the test, while the HVAC is secured. Although the CO2 System will be incperable for the duration of the test, continuous trained fire watches will be present. The SF6 gas will not adversely affect the operability of any plant equipment powered or controlled by cables or instrumentation located in the Switchgear Room; therefore the margin ;f safety is maintained.

1 There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This special test procedure did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

1 Temoorary Modification 95-037 General Summary This Temporary Modification installed two freeze seals in the 1.5" cross connect line between  !

12"FP-1 and 12"FP-2. This allowed for isolation of valve V70-26, Service Water crosstie check i valve to Fire System, and the downstream restricting orifice for maintenance. The function of '

the cross connect line is to provide a pressure source from 12" SW-4A to the associated fire system. This static pressure source is required so that fire pumps P-40-1 A/B will not continuously cycle to maintain system pressure.

If the freeze seals had failed, fire water from 12"FP-1 and 12"FP-2 would have exited through the open bonnet of V70-26. The three floor drains located in the service water pump room would have accommodated this leakage. Additional deflection devices were provided. Base!

on the contingency plans in place, gross leakage would not have occurred. Pipe break during a seismic event was evaluated and found not to be a credible failure. l Safety Evaluation Summary This modification was bounded by the arbitrary single pipe rupture, and intake structure flooding study. Total failure of both freeze seals would essentially have resulted in nuisance leakage from the fire protection system at the intake structure, which would have been accommodated i by the floor drains. This modification did not affect station blackout, ATWS, Appendix R, alternate shutdown, or prevent any SSC from performing its intended safety function. The l_ freeze seals could not have failed in a manner which would have affected any of the transients previously evaluated in the FSAR. Controls in place provided a barrier between any possible freeze seal failure and the service water pumps.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered, a

Ismporary Modification 95-048

) General Summarv i This Temporary Modification installed two freeze seals in the 1.5" cross connect line between i 12"FP-1 and 12"FP-2 to allow repair of a pinhole leak. This cross connect line provides a  ;

pressure source from 12" SW-4A to the associated fire system; this static pressure source is  :

I required so that the fire pumps P-40-1 A/B will not continuously cycle to maintain system j pressure. During implementation of this Temporary Modification, an electric fire pump was l running to maintain system pressure. Contingency plans were in place to mitigate any effects of a freeze seal failure. '

1 Safety Evaluation Summary l

! This modification was bounded by the arbitrary single pipe rupture, and intake structure flooding study. Total failure of both freeze seals would essentially have resulted :n nuisance leakage 4

from the fire protection system at the intake structure, which would have been accommodated i by the floor drains. This modification did not affect station blackout, ATWS, Appendix R, alternate shutdown, or prevent any SSC from performing its intended safety function. The freeze seats could not have failed in a manner which would have affected any of the transients previously evaluated in the FSAR. Controls in place provided a barrier between any possible freeze seal failure and the service water pumps.  ;

1 There was no increase in the probability of occurrence or consequences of an accident or  !

malfunction as previously evaluated in the FSAR. This temporary modification did not present l significant hazards not described or implicit in the Vermont Yankee FSAR, and there is  ;

reasonable assurance that the health and safety of the public was not endangered.

1 Temporary Modification 95-049 DJineIALS.umman Central Vermont Pub!!c Service (CVPS) added a second transformer,115/69 kV, at Vernon Road; this required circuit modification to pevent degradation of line protection. Temporary modification 95-049 adc'ed a relay, diode and capacitor to the K-186115 kV line secondary trip circuit. This relay will act Jate on directional overcurrent faults and initiate a transfer trip signal to Keene and Vernon Road substations. This modification provides additional protection the K-186 line and increasts the speed of clearing faults on the line.

i

. . l l

Safety Evaluation Summaly l

This modification added a relay to the secondary trip circuit of the K-186 line to initiate a  ;

transfer trip to Keene and Vernon Road substations for a directional overcurrent fault. This

]

does not create a different type of malfunction since failure of this breaker is backed up by '

breaker failure relays; this would lead to loss of the 115 kV yard, which is already analyzed.

This modification did not reduce the difference between a system failure point and safety limit or j margin of safety, since this modification is not associated with any safety limits as defined by i Technical Specifications. I i

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

i Temporarv Radjfication 96-001 General Summarv This temporary modification removed the 6 digit from the Rod Position Indication System (RPIS) position on control rod 06-15. This was done to allow for clarity for all other rod positions for Control Rod 06-15. Prior to this, Control Rod 06-15 had a 46 superimposed on the four rod display and full core display panel, making it difficult for the operator to determine the position of Control Rod 06-15, and preventing the rod worth minimizer and ERFIS from l recognizing Rod 06-15's position. Initial troubleshooting indicated that the problem is with the l'

Position Indication Probe (PIP), scheduled to be replaced during the 1996 refueling outage.

Removal of the 6 digit eliminated positions 06,16,26,36 and 46; however, this allowed the remaining positions to be clearly seen. Removal of the 6 digit did not affect the full in or full out position indication. Control Rod 06-15 is a peripheral control rod and during startups and shutdowns it is scheduled to be withdrawn continuously to 48 or continuously inserted to 00  ;

when using a "B" sequence. In thc case of "A" sequences, the control rod only stops at position '

12 on withdrawals and inserts; this is also the case for the Rapid Shutdown Sequences. In no i cases do the sequences call for stopping the control rod at any positions ending in "6". l Technical Specification surveillance 4.3.A.2 requires each partially or fully withdrawn operable .

control rod to be cycled "one notch" each week. In lieu of this requirement, exercising for  !

Control Rod 06-15 will be position 48, to 44, to 48.

Safety Evaluation Summary Although this temporary modification could have resulted in the receipt of a rod drift alarm if i Rod 06-15 were moved to any of the 6 digit positions, this situation was of minimal significance because receipt of the alarm would not prompt operators to take a nonconservative action, the i rod was not scheduled to be left at any of these positions during the remainder of the operating cycle, and rod drift detection capability for all other rod positions remained functional. The temporary modification did not impact the overtravel alarm for rod 06-15. Therefore, the ability T . .

for operators to be alerted to a rod drift condition or to periodically verify coupling integrity for this rod was not jeopardized.

There was no increase in the probability of occurrence or consequences of an accident ar malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Iamporary ModHication 96-0Q2 j General Summary t

This temporary modification electrically backseated RCIC-16 to reduce or stop a minor packing steam leak. Wiring changes were made which effectively jumper across the "open" limit swit.ch i contacts (LS 5-6) in the RCIC-16 control logic scheme, and result in specific relays remaining

, energized until the setting of the MOV "open" torque switch is reached.

Sdety Evaluation Summary RCIC-16 in combination with t'ne Primary Containment Isolation System (PCIS) performe an accident mitigation function. This temporary modification does not increase the radiolegical l j consequences of an accident previously evaluated in the FSAR because the bypassing of ,

I "open" limit switch contacts in the RCIC-16 control logic scheme, to allow the valve to be .

backseated utilizing the circuith open torque switch, does not affect the ability of this valve to perform its intended function, i.e. isolate the RCIC steam line if required. The RCIC-16 valve 4 4

"close" control logic scheme and the PCIS input signals to this logic remain unchanged. RCIC- i 16 and its associated controllogic continue to perform the same functions within the bounds of j

the equipment's specifications. The interaction of this equipment with other structures, <

! systems, and components is no different than previously evaluated.

There was no increase in the probability of occurrence or consequences of an accident or

. malfunction as previously evaluated in the FSAR. This temporary modification did not present l significant hazards not described or implicit in the Vermont Yankee FSAR and there is j reasonable assurance that the health and safety of the public was not endangered.

!, Immporarv Modification 96-010

)

) General SummAIX l l This temporary modification adjusted the Diesel Generator (DG) 1-1 A exhaust fan TEF-2 damper linkage so the damper was in the normal full open position when the motor driven 4

mechanism was in the closed position, and electrically and mechanically disabled the damper motor operator. This would have prevented differential pressure from developing across the DG room wal!s during a postulated tornado. A personnel protection screen was provided on the I inside of the building surrounding fan TEF-2 and the exhaust damper. This screen prevented .

1 birds from entering the building while the dampers were opened.

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Safety EvaluatioASummary i I

This temporary modification did not reduce the difference between a system failure point and  ;

accepted safety limit or reduce the margin of safety. The damper was fixed in the position required for operation of fan TEF-2, thus assuring that the ventilation system was availablo to support DG-1-1 A operation in all cases where it is taken credit for, including those involving a

postulated tornado event.

4

' There was no increase in the probability of occurrence or consequences of an accident or  ;

malfunction as previously evaluated in the FSAR. This temporary modification did not present i

significant hazardr. not described or implicit in the Vermont Yankee FSAR, and there is reasonable assumnce that the health and safety of the public was not endangered.

4 Temporary Modification 96-011 Genatal Summary This temporary modification adjusted the Diesel Generator (DG) 1-1B exhaust fan TEF-3 damper linkage so the damper was in the normal full open position when the motor driven mechanism was in the closed position, and electrically and mechanically disabled the damper motor operator. This would have prevented differential pressure from developing across the '

DG room walls during a postulated tomado A personnel protection screen was provided on the inside of the building surrounding fan TEF-3 and the exhaust damper. This screen prevented  ;

birds from entering the building while the dampers were opened. '

Safety Evaluation Summarv i

This temporary modification did not reduce the difference between a system failure point and  ;

accepted safety limit or reduce the margin of safety. The damper was fixed in the position required for operation of fan TEF-3, thus assuring that the ventilation system was available to support DG-1-18 operation in all cases where it is taken credit for, including those involving a postulated tornado event.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This tempontry modification did not present significant hazards not described or implicit in the Vennont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

IRmpatary Modification 96-012 Geaeral Summary This temporary modification disconnected the high temperature alarm in the control room for one control rod drive. The alarm remained operational for the remaining control rods and the local alarm for the effected rod remained operational. The alarm was disconnected due frequent i spurious alarms which were attributed to a bad detection element which could not replaced until l the refueling outage. Since the alarms created unnecessary control room distractions and could mask other valid alarms, the alarm was disconnected.

j - a Safety Evaluation Summaty I

This temporary modification had no effect on the actual cooling supply to the control rod and the local recording of the temperature was available and reviewed at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift. It was noted that inadequate cooling to a single control rod is unlikely and the system continued 1 to monitor the remaining rods. Per the FSAR " Cooling water (to the drives) may be interrupted

] for short periods without damage"

There was no increase in the probability of occurrence or consequences of an accident or +

1 malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a ,

reasonable assurance that the health and safety of the public was not endangered. -

Temporary Modification 96-014 General Summary i

' This temporary modification installed a tent against the turbine building to switchgear wall to 1

support the replacement of two fire dampers. The tent was necessary to provide a ventilation flow path, an EQ barrier and a radiation control area barrier while the existing duct was removed to allow for the replacement of the dampers which are located in the duct at the point it penetrates the wall.

j Safety Evaluation Summary l The ventilation system is NNS. The tent effectively replaced the duct work as an EQ barrier between a mild and potentially harsh environment and to separate the radiologically controlled area from the non-radiologically controlled area. The switchgear room was monitored to ensure l adequate cooling continued to be supplied.

2 There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This temporary modification did not present

, significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

i Temporary Modification 96-025 General Summary This temporary modification installed two freeze seals in the 1.5" cross connect line between the two main firewater headers to facilitate repair of a pinho;e leak on the cross connect line.

The free.:e seals were required to provide adequate isolation to perform the repair.

Safety Evaluation Summary The fire system is not an initiator of any of the accidents or malfunctions analyzed in the FSAR nor do the freeze seals affect the performance of the fire systern. The system is not used to mitigate any of the design basis accidents listed in the FSAR. No Technical Specification bases are affected by this modification, therefore then margin of safety is not reduced.

There was no increase in the probability c currence or consequences of an accident or malfunction as previously evaluated in the OAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

IRmoorary Modification 96-029 GBRelaLSummary Since discharge of water through the hotwell level control system to the CST is not allowed during outages, this temporary modification was used to provide an alternate flow path. The internals and actuator of directional control valve V20-342 were removed, allowing the valve to act as a " tee" and permitting flow from the Reactor Water Cleanup system (RWCU) directly to ,

the waste sample tanks instead of sending the water to the hotwell. Water from RWCU is diverted around the radwgste process system with a valve installed in the inlet of the waste surge tank TK-11-1 A. This modification allows the filtration and demineralization process of the radwaste system to be bypassed; however, there is no threat of releasing unacceptable water to the CST because a Chemistry sample is required to generate a permit to release water from the sample tanks to the CST.

Safety Evaluation Summary This modification does not affect the probability of a malfunction which initiates any of the transients listed in the FSAR. It does not increase the radiological consequences above that analyzed in the FSAR. The liquid radwaste system cannot fail in a manner which will affect or I initiate any transient, and is not used to mitigate the radiological effects of those transients.

There was no increase in the probability of occurrence or consequences of an accident or l malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered. j l

Temoorary Modification 96-036 j

)

i General Summary This temporary modification installed a special tool on the reacar vessel flange for the i

, performance of Reactor Pressure Vessel (RPV) weld inspections, The inspections were conducted during a refuelir ' outage with the vessel depressurized and the head removed. The tool straddled the vessel flange and directed a retractable mast into the vessel to position a 4

weld inspection device on the welds. The device then moved along the welds to perform the  ;

inspection in the area and then the entire tool was repositioned to the next area. The tool i

, constituted a heavy load. The toolitself moved around the flange on wheels designed so that no damage occurred to the flange.

J

_ Safety Evaluation Summary The tool and the equipment used to install and remove it, were designed to meet or exceed the requirements of NUREG-0612 for the handling of heavy loads over the reactor cavity. No fuel moves were allowed whenever the tool was being moved. The tool itself is NNS but was 4

designed as seismically qualified equipment and could only be installed with the reactor shutdown, depressurized and the vessel head removed.

There was no increase in the probability of occurrence or consequences of an accident or

' malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Temporary Endification 96-038 I GanaraLSummary

i. This temporary modification installed an Auxiliary Bridge (AB) on the refuel floor to support the inspection and repair of the core shroud. The modification also addressed repositioning the '

refuel bridge and avoidance of any interference between the two devices. The AB was installed and removed using the Reactor Building (RB) overhead crane. The AB was qualified to handle heavy loads and to withstand a seismic event without " tipping"

) Safety Evaluation Summary i

The Auxiliary Bridge (AB) was designed, constructed, installed and removed following the rules of NUREG-0612. The AB was designed and constructed to meet or exceed the requirements for VY's existing refueling bridge. No fuel movement was allowed while the AB was over the core. The AB is NNS and did not interface with any equipment important to safety nor could it ,

i challenge or degrade the performance of any safety system functions.

There was no increase in the probability of occurrence or consequences of an accident or l

l malfunction as previously evaluated in the FSAR. This temporary modi'ication did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a j reasonable assurance that the health and safety of the public was not endangered.

(  !

Temporary Modification 96-039 '

General Summary This temporary modification removed (and reinstalled) the existing auxiliary hoist and installed (and removed) a temporary marionette assembly from/on the north monorail of the refuel bridge. The marionette assembly was needed to support the use of specialized tools and equipment required to support the inspection and repair of the core shroud.

Safety Evaluation Summarv

This temporary modification was installed during the refueling outage and removed prior to startup thus the only DBA of concern is the Refueling Accident. The refuel bridge plays no part in the mitigation of the Refueling Accident or the initiation or mitigation of the abnormal 4

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

l transients listed in the FSAR or other design events (ATWS, SBO, etc.). The installation was  ;

verified to be within the design basis of the refuel bridge and conducted in compliance with the l i requirements of NUREG-0612.

i~

There was no increase in the probability of occurrence or consequences of an accident or '

malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankae FSAR, and there is a reasonable assurance that the health and safcty of the public was not endangered.

3 Temporary Modification 96-042 Genera 1 Summary

, This temporary modification installed a freeze seal to support a design change for the i

installation of a new 8" supply to the reactor building fire suppression water supply. Without the i 1

freeze seal, the existing configuration required isolation of all fire suppression water to the reactor building to make this change.

Safety Evaluation Summarv

, This temporary modification did not effect any initiation or mitigation modes for the DBA's listed j in the FSAR nor did it effect any of the abnormal transients listed in the FSAR (or other  !

transients such as ATWS, SBO, etc.). Potential flooding resulting from a failure of the seal was found to be bounded by VY's turbine building flooding study and it was noted that there was no

safety related equipment in the vicinity of the opening.

I I There was no increase in the probability of occurrence or consequences of an accident or i

4 malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.  !

i Iemporary Modification 96-043 I l General Summary This temporary modification was installed to change the failure mode of SCW-46A, a temperature control valve, the SAC-1 A and B air handling unit, discharge dampers and the SAC-1B supply damper. Originally these ccmponents failed shut on a loss of air. This change, modified SCW-46A so that it now fails open on loss of air, and installed a manual means to block open the SAC-1 A and B dampers. These changes enable the control room personnel to adjust control room temperature manually in the event of a loss of air to the normal automatic temperature control system.

Safety Evaluation Summary Temperature control of the control room has no effect on the probability of occurrence or radiological consequences of any accident previously evaluated in the FSAR. The ability to operate control room HVAC in the event of a loss of air w;ll enhance mitigation activities for transients etc. since the control room environment can be maintained within FSAR values. This change did not increase the probability of, or radiological consequences from a malfunction which could initiate an abnormal transient as discussed in the FSAR. The change did not introduce a different type of accident or equipment malfunction nor did it reduce any margin of safety.

1 There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This temporary modification did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a i

reasonable assurance that the health and safety of the public was not endangered.

FSAR Change 24 VDC B.atterv Chargers

- General Summary The description in the FSAR Section 8.7.3 was revised to accurately describe how the replacement battery chargers function. There are two 24VDC battery systems and two changers for each system.

i l

Safety Evaluation Summary The 24 VDC battey chargers are not initiators or mitigators of any of the accidents described in the FSAR and do not affect the radiological consequences of any accidents or malfunctions.

l The changes in battery charger type does not create any different type of accident or '

malfunction which is different than that described in the FSAR.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

FSAR Change. Section 7.18Ji i

General Summary This change revises the description of how the ARl/RPT neon indicating lights are used in testing ARl/RPT circuitry and clarifies the ARl/RPT system surveillance program. This change 1 results in no physical changes to the plant as described in the FSAR.

i Safety Evaluation Sumrnaly The ARl/RPT tinstrumentation systems are not accident initiators for any accidents discussed in the FSAR. As thb change results in no physical changes to the plant, there are no increased consequences of an accident or malfunction or an increase in the probability of a malfunction nor does it resu!t in any new failure modes. The margin of safety is not rifected by this change.

There was no increase in the probability of occurrence or consequances of an accident or malfunction as previously evaluated in the FSAR. This change rild not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

FSAR Change. Section_BJ '

General Summary This change updates the diesel loading schedule to incorporate equipment changes installed during the 1995 refueling outage and adds margin in the voltage transient analysis based on outage test data. Additionally, the lighting panelload totals are revised.

Safety Evaluation Summary The diesel generators are not accident initiators. The automatic operation of equipment required to mitigate the consequences of accidents evaluated in the FSAR are not affected by this change. This change does not impact the ability of the diesel generators to start and accept emergency loads. No new accidents are created by the change in generatorloading nor are any different malfunctions created by this change. This change does not impact the margin of safety described in the FSAR or Technical Specifications.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Core Operating Limits Report (COLR) Revision with Modified Maximum Average PlaDAI Linear Heat Generation Rate (MAPLHGR) Limits General SuJDmary Evaluation of a potential Loss of Stator Cooling event indicated that MAPLHGR limits for Cvele 18 needed to be modified to account for the Loss of Stator Cooling (LOSC) event results. The LOSC results in a local thermal overpower (TOP) of 28.51% which exceeded the assumption of 25% used by General Electric in the original MAPLHGR limits.

Safety Evaluation Summary The lower MAPLHGR limits provide more operating margin for the Loss of Coolant Accident (LOCA), the Control Rod Drop Accident (CRDA), and the Main Steam Line Break (MSLB); there is no increase in the consequences of an accident previously evaluated. The new MAPLHGR limits do not affect any initiators of an equipment malfunction or the integrity of any piece of equipment. The mechanical design of all fuel bundles has been evaluated with the revised assumption regarding the Cycle 18 operation. The new 28.51% TOP was accounted for in the generation of the new MAPLHGR limits. The analysis of the LOSC transient assumed various malfunctions of equipment important to safety. Given the new operating limits, the analysis showed that fuel integrity was assured. Therefore, the new MAPLHGR limits do not increase the consequences of a malfunction of equipment important to safety previously evaluated. The new limits did not change any piece of equipment which could be an accident initiator, did not require any equipment to be operated beyond its design envelopes, and did not involve new plant equipment.

4 There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This COLR revision did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

s Cycle 18 Core Operating Limits Report (COLR). Revision 3 - Exclusion Area Change General Summarv l This change revised the equation for determining the thermal-hydraulic stability exclusion region i to reflect the conservative limits previously established for flows <30% as shown on the COLR Figure 2.4.1. This is consistent with the limits previously shown on Figure 2.4.1 (Rev. 7).

l Safety Evaluation Summarv I

The revised figure provides limits for normal operation and maneuvers to avoid thermal hydraulic oscillation and does not impact assumptions made in the analysis supporting the accidents listed in the FSAR or the consequences of such accidents. This change did not effect the function of any equipment and did not increase the probability of a malfunction or the consequences of such a malfunction. This change did not change plant operating strategies, equipment, or operation such that equipment operates beyond its' design limits and therefor did

not create a new type of accident or malfunction. This change provides consistency with the existing operating guidance and ensures maintenance of the existing margins of safety defined by Technical Specifications. '

l There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This COLR revision did not present significant hazards not described or implicit ;n the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Cycle 18 Core Operating Limits Report (COLR). Revision 3 - Change to Safetv Valve _and

+

Safety Re. lief Valve Setpolut Tolerance i

General S.ummarv This change revised the as found setpoint acceptance cnteria for Safety Valves (SVs) and

Safety Relief Valves (SVs) from a tolerance of +/- 1% to +/- 3%.

Safety Evaluation Summary '

This change to the as-found setpoint tolerance does not constitute a changes to the valve design and does not effect the probability or consequences of any of the FSAR evaluated accidents. Potential effects of this change on the transients analyzed in the FSAR and the effects of the transients on the SRV piping / supports and the Torus in particular, were evaluated and no increase in the probability or consequences of a malfunction of equipment important to safety was found. Since this change did not add equipment, change the design of the equipment and will not result in the operation of any system or component beyond their original

. design envelope, this change does create the possibility of an accident or malfunction of a different type from those evaluated by the FSAR. It was also demonstrated that there was no reduction in the margin of safety as a result of this change.

There was no increase in the probab;lity of occurrence or consequences of an accident or j malfunction as previously evaluated in the FSAR. This COLR revision did not present significant hazards not described or imp!icit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

Thermaj Performance of Heat Exchangers  ;

Gene.ral Summary The Vermont Yankee Residual Heat Removal (RHR) heat exchangers are tubular heat exchangers with a parallel-counterflow configuration with two tube passes Each heat exchanger has 1664 tubes (820 inlet tubes and 844 outiet tubes). Some tubes may be plugged to prevent tube failures due to tube degradation. A safety evaluation was written to consider plugging up to five percent of the tubes in one or both heat exchangers.

The original heat exchanger performance considered a total heat exchanger fouling of 0.0025 2

hr-ft - F/ Btu. It was desired to allow a total heat exchanger fouling of 0.0029 hr-ft2- F/ Btu; this increased fouling is consistent with the heat exchanger thermal performance testing currently implernented at Vermont Yankee.

l Safety Evaluation Summary These are passive changes to the heat exchangers; there is a small effect on the heat transfer capability of the heat exchangers. The RHR heat exchangers are isolated from the reactor coolant system during normal power operations; any activity associated with the RHR heat exchangers cannot affect the probability of a reactor coolant system pipe break accident. The RHR system is operated during refueling operations to provide cooling. This is not a limiting mode of operation of the RHR system, and system operating procedures and failure modes are not affected by tube plugging or increase in allowed fouling; therefore refueling accidents are not affected by the modification to the RHR heat exchangers.

The RHR system can be used to provide pool cooling during reactor power operation. This mode of operation can be used to assist in maintaining the torus water temperature below 100 F. This is not a limiting mode of operation of the RHR system, and system operating procedures and failure modes are not affected by tube plugging or increase in allowed fouling; therefore, this mode of operation is not affected by the modification to the RHR heat exchangers.

The RHR system is expected to operate following an accident. Plugging of up to five percent of the RHR heat exchanger tubes and allowing an increase in the total heat exchanger fouling to 2

0.0029 hr-ft ,cF/ Btu have a small effect on the thermal performance of the heat exchangers.

This results in a slight reduction in the heat removal capability of the RHR system when operated in the containment cooling mode following an accident; the heat exchanger remains capable of limiting the temperature of the suppression pool water to 176 F following a loss of coolant accident. A suppression pool temperature of 176 F provides adequate net positive suction head (NPSH) for the LPCS at a flow rate up to 3,750 gpm using conservative 7

J assumptions. Run-out flow of the LPCS system may result in a flow rate of up to 4,400 gpm.

Reactor operators are provided with procedure ON3164 which provides instructions to throttle, 1

secure, or re-align the core spray pump if symptoms of pump cavitation are observed. This

procedure prevents damage to the LPCS pumps due to cavitation whether from strainer

. plugging or high torus pool temperature. For single LPCI pump operation, a torus water temperature of 176 F provides sufficient NPSH for flow rates from 6,000 gpm to 7,300 gpm.

Sufficient NPSH is available for operation of the RHR mode of LPCI operation at the 7,000 gpm

design flow rate.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This modification of the thermal performance of heat exchangers did not present significant hazards not described or implicit in the Vermont

, Yankee FSAR, and there is reasonable assurance that the health and safety of the public was

not endangered.

Saf.ety Evaluatig.n for Pressure Locking of V31ve V23-14 General Summa 1y Pressure locking of the V23-14 valve is a prMntial operating concern during heatup - ...e High Pressure Coolant injection (HPCI) turbine steam supply piping until the valve is cycled at normal temperature and pressure. During system heatup, the valve orientation tends itself to the potential for water to collect in the piping near the valve seat and the water eventually makes its way into the valve bonnet. Once the bonnet is full, further heatup of the piping and valve could potentially result in pressure locking. The valvo is vulnerable during the period from the initial heatup until the valve is cycled at full operating system temperature and pressure. To add.ess concerns raised in NRC Generic Letter 86-10, and Generic Letter 97-07, a safety evaluation was written to address pressure locking of V23-14.

Cycling the valve every 10 degree F increase in reactor temperature after the valve is in its fully closed position provides assurance that the valve will not be pressure locked. The valve motor operator is capable of an indefinite number of cycles provided that the motor is allowed to cool for ten minutes prior to the next complete stroke cycle (open and close). If an accident or abnormal operational transient occurred immediately after the valve was closed, the valve would be capable of performing its opening safety function.

Safety.Eyaluation summary The HPCI system respd:, to mitigate FSAR Chapter 14.6 accidents. More frequent cycling of V23-14 is equivalent te more frequent surveillance testing and does not change the design of the HPCI system or &ct its ability to respond to an abnormal operational transient or accident.

Cycling of V23-14 does not change the probability of an inadvertent HPCI pump start at power.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. The potential of V23-14 pressure locking

. does not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public is not endangered.

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i Cycle 19 Reload General Summarv

' This reload consisted of the replacement of 120 GE-98 fuel bundles with new GE-9B bundles with an average initial enrichment of 3.54 w/o U-235 as opposed to the 3.35 w/o U-235 average initial enrichment of the discharged bundles. This change was thoroughly analyzed and demonstrated to meet the design criteria of Section 1.5 of the FSAR. This change did not effect

' Technical Specifications since the methods used to evaluate the core performance are currently identified in Technical Specifications and resulting operating limits were provided in the Vermont Yankee Cycle 19 Core Performance Analysis Report (CPAR). FSAR Chapters 4,6,7 and 14 will change as a result of this reload.

f Safety Evaluation Summarv

) The evatuations of new bundles found no increase in the probability of occurrence of an i accident evaluated in the FSAR or in the radiological consequences of such accidents. It was

, also found that there was no increase in the probability or occurrence of the malfunction of equipment important to safety or in the ra:liolegical consequences of such malfunctions. The new bundles did not create the possibility of an accident or malfunction of a different type than previously evaluated and did not reduce the margin of safety, i!

l There was no increase in the probability of occurrence or consequences of an accident or 4

malfunction as previously evaluated in the FSAR. This change did not present significant j hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable

} assurance tha' the health and safety of the public was not endangered.

j lastallation and Test Procedure for EQCE_94-411  !

General Summary This I & T was written to provide detailed instructions and define technical requirements for the

addition of seismic supports and fasteners to DC Buses 1,2, and 3 and various instrument i racks. This installation met all the requirements of EDCR 94-411 ECN 2.

Safety Evaluation Summary

, The modifications to equipment being implemented by this I&T do not affect accident initiaters.

The administrative and physical precautions taken during the implementation of these

modifications will ensure that the equipment to which the modifications are being performed will be available to perform their safety functions during the modifications. Evaluations were performed which identified the potential events which could be initiated by inadvertent bumping j of this equipment or breaker tripping or mis-positioning. Any potential event was previously i . evaluated and addressed in existing procedures.

_ There was no increase in the probability of occurrence or consequences of en accident or 5

malfunction as previously evaluated in the FSAR. This test procedure did not present j significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

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4 Qperation with Recirculation Discharge Bypass Valves Oogn General Summary An evaluation was performed to determine the impact of the recircula: ion discharge bypass valves on Loss of Coolant Accident (LOCA) analysis. Neither the transient nor LOCA models used these valves and bypass lines in their models because the valves were assumed closed.

The results of the analysis show that having the bypass valves open or closed have a negligible impact on the results of the LOCA/ Emergency Core Cooling System (ECCS) analysis, and no impact on LOCA/ Containment analysis and LOCA/ Loads.

Safety Evaluation Summary This change did not result in the modification or addition of any plant equipment; only the normal position of the recirculation bypass valves was modified to reflect plant practice. The effect of having the bypass valve open is to provide a small, high resistance flow path in parallel with a small portion of the recirculation loop discharge flow path. The reduction on recirculation loop flow resistance is insignificant. Therefore, the recirculation pump will operate in the same manner as if the bypass valve were closed, resulting in recirculation and total core flow being unchanged. This change had no impact on any other plant equipment or accident or transient analysis. These valves are powered from the UPS units which were specifically designed to supply emergency loads, which included repositioning these valves in the event of an accident.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. Operation with the recirculation discharge i bypass valves open does not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Regulatorv Guide 1.97 General Summary Regulatory Guide 1.97 (RG 1.97) had been last submitted in September 1989. Since that time, changes have taken place which impacted that submittal. Also, corrections due to incomplete / inadequate information identified in the original submittal needed to be addressed.

The RG 1.97 submittal was revised, updated, enhanced, and resubmitted on 3/29/96. Review of the RG 1.97 changes determined that one item potentially impacted the Safety Evaluation Report (SER), issued by the NRC based on information provided in the September 1989 submittal.

The September 1989 RG 1.97 submittal stated that stack gas I and 11 satisfied RG 1.97 category 2 requirements. Category 2 required that these instrurnents be installed and maintained under an approved Quality Assurance program (other RG 1.97 Category 2 requirements were adequately addressed in the submittal). The instrumentation installed to address post-accident effluent discharges is stack gas Ill. Stack gas ill has a wide range, satisfied the RG 1.97 upper range requirement, but did not adequately cover the lower range requirement. The lower range requirement was satisfied by stack gas I and 11 (these monitors are used during normal operation). The commitment to apply RG 1.97 Category 2 requirements to stack gas I and 11 was unintentional. However, the SER assumed they were applied. A safety evaluation was performed on stack gas I and Il to assess the impact to the SER if they were downgraded to RG 1.97 Category 3. Category 3 does not require that an approved QA program be applied.

Safety Evaluation Summary Stack gas I and ll are used to monitor the radioactive effluent discharge up the plant stack. A complete failure of these monitors during a post-accident situation would not cause a malfunction of equipment relied upon to mitigate an accident. These monitors provide indication an annunciation only; there are no automatic control functions of either NNS or safety related systems / components. Stack Gas I and 11 can faillow, high, or intermittent. In any case, when the radioactive release becomes high enough to be a concern, Stack Gas til will be available. As such, failure of Stack Gas I and 11 cannot cause a different type of malfunction of equipment. Section 3.2.6/4.2.6 of the Vermont Yankee Technical Specifications address the requirements for post-accident monitoring. Stack Gas I and 11 is not included and is not reiied on. The High Range Stack Monitor, Stack Gas Ill, is relied on to perform this function.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. Resubmittal of RG 1.97 does not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is reasonable assurance that the health and safety of the public was not endangered.

Residual Heat Removal (RHR) Minimum Flow Valve Change from Normally Closed to Normally Open General Summary The Appendix R project team identified a potential for RHR pump damage, due to lack of minimum flow protection, which potentially affects the safety analysis. The RHR minimum flow valves (V10-16A/B) are normally closed. There is one minimum flow line per train serving the two pumps in each train. The two RHR pumps in each train are cross-powered (i.e., one is powered from Bus 3 and the other from Bus 4). The minimum flow valve in each train is powered from a single power source, either Bus 3 or Bus 4. For a loss of coolant accident (LOCA), the failure of power from one bus, for example from an emergency diesel generator (EDG) failure, would result in one RHR pump in each train being operable but only one minflow valve being operable. For accident scenarios where the RHR pumps automatically start, but must operate at minimum flow conditions until the reactor has depressurized, the loss of minimum flow bypass for the cross powered pump for ra extended time threatens the continued operability of the pump, and may potentially result in pump failure.

A valve lineup change to maintain the minimum flow vdves open during normal operation is proposed as a short term resolution to this concern. With the valves normally open, a loss of power would not prevent them from performing the minimum flow pump protection function. A smallloss of Low Pressure Coolant injection (LPCI) injection flow, with a conservatively assumed 500 gpm flow loss, could result in only an approximate 15 F increase in PCT, with a maximum PCT still below 1800 F.

Safety Evaluation Summary This change to the RHR minimum flow valve lineup during normal operation affects the performance of the RHR system for soma accident scenarios. The RHR system is used to mitigate the consequences of a LOCA in both the short term and long term phases. Short term, the RHR acts in its LPCI mode to inject water into the reactor vessel through the recirculation loops following a LOCA. Long term, the RHR removes sensible and decay heat from the suppression pool to assure containment integrity and long-term core cooling. Maintaining the RHR minimum flow valves in the open position will slightly reduce the total amount of RHR flow delivered to the reactor vessel in the LPCI mode for the RHR pump which is powered by the opposite bus. For long4cim cooling, the operator will need to select the RHR pump that is powered from the same source that provides power to the Emergency Core Cooling System (ECCS) corner room Reactor Recirculation Unit (RRU), torus cooling valves, and the RHR pump's associated minimum flow valves; therefore, long-term cooling will be unaffected because the minimum flow valve will have power and will close when RHR flow exceeds its preset value.

The effect on the results of the LOCA analysis of keeping the minimum flow valve open was conservatively estimated by reducing RHR flow by 500 gpm at all points on the RHR pump curve used in the LOCA analysis. Since the RHR pump curve used in the LOCA analysis is already lower than the Technical Specification minimum flow requirements, the effect of the open minimum flow valve is more than adequately bounded. The analysis was done for the limiting break size and location, and a break size on either side of the limiting size to assure all potential effects were captured in the analysis. The cases assumed a break in the discharge side of a recirculation loop and the single failure of the DC-1 power supply. This results in an ECCS capability of one core spray and two RHR pumps in the LPCI mode, one of which is rendered ineffective in the short term because of spillage through the broken recirculation line.

The analysis resulted in a 15 K increase in peak cladding temperature, which, according to 10CFR50.46(a)(3)(i), is less than the range of increase in peak cladding temperature considered to be significant. Furthermore, the calculated PCT of 1793 F is well within the 10CFR50.46 acceptance criterion of 2200 F.

The margin of safety is improved by the proposed change in that the potential for losing an RHR pump due to overheating in a deadheaded condition is reduced. Opening the minimum flow valves and maintaining them open during normal operation will assure the RHR pumps are adequately cooled under minimum flow conditions unless a minimum flow valve spuriously closes. This will assure a minimum of two low pressure ECCS pumps are always available for short term core cooling in the event of a LOCA. With the current RHR minimum flow valve lineup, the potential exists for the loss of one of two RHR pumps on a given emergency bus to fail if the minimum flow valve powered from the other emergency bus remains closed for a sufficient period of time with the RHR pump operating in a deadheaded condition. If an RHR pump were assumed to fait due to operating deadheaded for a period of time before the LPCI injection valve opens, the resulting peak cladding temperature would exceed the peak cladding temperature for the case in which the minimum flow valve remained open throughout the course of the accident.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated n: the FSAR Changing the RHR rninimum flow valve from normally closed to normally open does not present significant hazards not described or implicit a

l l in the Vermont Yankee FSAR, (4nd there is reasonable assurance that the health and safety of the public was not endangered FSAR change for Flood-up tyjth the Core Spray _Sy11em General Summarv j FSAR Section 10.5 states that the reactor cavity is flooded up using the Condensate Transfer System. This FSAR change allows for an alternate method to flood up the cavity using a Core Spray Pump in addition to the present method stated in the FSAR. Both methods utilize the Condensate Storage tank as a source of water.

Safety Evaluation SuminaIX The core Spray System is not an accident initiator but is used to mitigate accidents. During reactor cavity flood-up the reactor is not operating or refueling. The system is still available for i its safety function by re aligning its suction valve to the torus. Technical Specifications were met for this change. l 4

There was no increave in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant l hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered. l Head Flange _ Leak Detection Instrument Line Safety Classification Upgrade General Summary i During a review of the IST program, questions were raised about the appropriateness of several existing Safety Class 2 (SC2) to Non-Nuclear Safety (NNS) boundaries. Originally the Head Flange Leak Detection instrument line and the instrument lines on the SRV downcomers and ,

associated instrumentation were NNS downstream of the excess flow check valve as allowed  !

by the safety classification manual This was an appropriate classification based on the function of the instruments (no safety functions) and the existence of the excess flow check valve (ECV) as an approved isolation device. During the review of the IST program it was noted that the l ECV was not tested and should not be relied upon as an isolation device and therefor the l instrument lines and instruments should be safety class for pressure boundary purposes. The  ;

classification of f.he instrument 3 was changed and drawing updates initiated. This evaluation i was conducted to determine the sceptability of the installed components.

3 Safety Evaluation SummaIy The evaluation found that the materials used in the instrument lines as well as the instruments themselves were either fully qualified or assessed to be qualified for safety class service.

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable assurance that the health and safety of the public was not endangered.

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I Residual Heat Removal (RHR) and Core Spray (CS) Containment isolation Reauirements General Summary This change revised the designation of the primary containment boundary within the RHR and CS systems. Prior to the change, the primary containment boundary for the RHR and CS supply lines to the reactor vessel were defined as the outboard Primary Containment Isolation Valves (PCIVs). The change established the RHR an CS systems as closed seismic loops i which serve as extensions of primary containment and therefore require a single PCIV.

4 Specifically RHR-25A and B, the inboard injection valves, RHR-46A and B, check valves, CS-i 11 A and B, discharge valves, CS-13A and B, check valves, and CS-30A and B pressure switch ,

isolation valves which were previously considered PCIVs are no longer considered as such.

l RHR-27A and B, outboard injection valves, and CS-12A and B, discharge valves, were and '

l remain PCIVs tested in accordance with 10CFR50 Appendix J.

1 The use of this new approach as an "other defined basis"is allowed by 10CFR50 Appendix A  !

General Design Criteria (GDC) 55 and 56 as discussed in Reg. Guide 1.141 and ANS-56.2. )

i Safetv EvaluatioDJummary

This change did not affect the design, functions or operation of any valves in the RHR or CS systems. This change eliminated Appendix J testing for the valves no longer considered PCIVs 1

but did retain the testing for the remaining valves. The seismic loop provides equivalent protection to the existence of a second PCIV in the effected lines. The loops are also located i within the secondary containment such that any leakage will be contained within the reactor i

building and any release processed through the Standby Gas Treatment System. There was no redesign of the components as a result of this change and no new failure modes were created l by this change. There were no changes to any hardware or logic as a result of this change.

There was no increase in the probability cf occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant .

hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable  !

{ assurance that the health and safety of the public was not endangered.

SRY_Eoaltion Indication Instrument Line_ Safety Classification Upgrade 4

GantraLSMinmary '

During a review of the IST program, questions were raised about the appropriateness of several existing Safety Class 2 (SC2) to Non-Nuclear Safety (NNS) boundaries. Originally the Safety

- Relief Valve (SRV) Position Indication instrument line was NNS downstream of the excess flow check valve as allowed by the safety classification manual. These instruments were installed as a result of Reg Guide 1.97 concerns, The originalinstruments which were environmentally and seismically qualified were later replaced with like equipment qualified to higher temperatures.

l The first installation was considered to be NNS even though the instruments were fully qualified t

to safety class requirements. The replacement instruments were installed as safety class, however the system drawing was not updated to note the change in classification. This j evaluation was completed to support the belated correction to the drawing as required by j

procedure and to evaluate the circumstances to ensure no adverse conditions were introduced during the period of misclassification.

l Safetv Evaluation Summary The materials installed were identified as qualified since installation and therefore the administrative misclassification did not effect the acceptability of the installation.  !

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This change did not present significant

] hazards not described or implicit in the Vermont Yankee FSAR, and there is a reasonable

assurance that the health and safety of the public was not endangered.

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Revision 5 to the Safe Shutdown Capability Analysis (SSCA)

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Gangral Summarv a

As a result of a design verification of the Appendix R Program, various changes were made to the strategies described in this document. The changes included:

1. Taking credit for off-site power for fires which do not require Alternate Shutdown capability per Section Ill.L of Appendix R, provided damage from the fire cannot cause a loss of off-site power.

2. Taking credit for the Vernon Tie under Alternate Shutdown scenarios.

3. Reliance on Automatic Depressurization and Core Spray (ADS /CS) to accomplish core cooling for fires in several different zones of the reactor building.

4. Provision for the use of Safety Relief Valves and Low Pressure Coolant Injection as a method for depressurization and cooling. This method is to be used as a backup to the Alternate Shutdown System.

5. Provision of mitigating strategies for important spurious actions that could adversely impact safe shutdown.

6. Use of new assumptions, plant configurations and analysis techniques in support of the SSCA. This includes:

a. Incorporation of the VY Appendix R, Ill.L Report into the SSCA.

b. Redesignation of the cable vault / control room barrier as a smoke seal and not as a fire barrier.

c. Declassification of the switchgear room as an Appendix R Section Ill.L area.

d. Designation of the upper Reactor Core isolation Cooling room as a separate fire zone.

7. Creation of new coping strategies for;

a. Loss of Containment (Drywell) Cooling.

b. Loss of Fuel Pool Cooling.

c. Loss of Control Room Ventilation.

8. Revision of the Alternate Shutdown Timeline to coincide with the revised analysis which showed 22.5 minutes prior to reactor level reaching the top of active fuel. Strategies to address the new timeline were made possible by the installation of several design changes.

9. New analysis and calculations in support of the design verification including;

a. Core Uncovery Timeline Analysis for Alternate Shutdown

b. ADS /CS Core Cooling Analysis for Appendix R Fire Scenarios. (Supports ADS /CS use in shutdown scenarios)

c. Containment Heat Up Analysis.

10. Submittal / retraction of exemptions to the requirements of Appendix R including:

i a. The use of the Vernon Tie for Alternate Shutdown.

I b. The use of ADS /CS strategy for some reactor building fires.

c. Substitution of a 1-hour enclosure vs. a 3-hour separation for the SRV inhibit switch in i the control room. (This was later determined to be unnecessary.)

d. The use of a 1-hour cable for ECCS Corner Room Coolers.

e. The use of Security Lighting vs. Appendix R Lights for restoration of the SRV N2 supply.

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f. A prior request for exernption from certain aspects of fuse replacement was withdrawn.

While these exemption requests were in review, the previously established compensatory

, actions remain in effect.

Safety Evaluation Smmmary The changes to the strategies have no affects on the systems, structures, or components associated with the DBA's listed in the FSAR nor do the changes affect the performance of any system used to mitigate the DBA's. The new strategies do not increase the probability for initiation of the Abnormal Operational Transient listed in the FSAR. The new strategies do not create the possibility for a new type of accident or malfunction of equipment or reduction in any

margin of safety.

. There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant 4

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The following two summaries were inadvertently omitted from the Cycle 17 Operating Report:

PDCR 94-012 Toxic Gas Retire in Place General Summary This design change converted Temp Mod 92-01 and 92-41 to a permanent change. In addition, the design change removed eight relays and the associated wiring, four connections to the safety class control room HVAC duct, two connections to the NNS control room air intake, some of the switch board wiring abandoned by the TM's and updated drawings to show the current l

, configuration.  !

Safety Evaluation Summary There was no safety evaluation performed for this design change since the changes were previously addressed by the evaluations performed for the TM's and the change was reviewed and approved by the NRC as part of the approval for Amendment No.132 to facility Operating 1 License No. DPR-28. This was confirmed by the performance of a screening evaluation. i There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present significant hazards not described or implicit in the Vermont Yankee FSAR., and there is a reasonable assurance that the health and safety of the public was not endangered.

EDCR 94-411. EDCR 94-411 ECN-1. EDCR 94-411 ECN-2. Unresolved Safety issue USI-A-46 (SQUG - Seismic Qualification Users Group) and IPEEE (Individual Plant Examination ,

for External Events of Severe Accident Vulnerabilities) (Seismic) Structural Modificallona I General Summary This design change implemented structural modifications to plant equipment related to the Safe  ;

Shutdown Equipment List (SSEL). The modifications ensure that seismic requirements  !

applicable to USI-A-46 and seismic IPEEE criteria related to equipment anchorage are met.

These criteria meet or exceed the original seismic design basis in FSAR Section 12 and in applicable original equipment specifications.

Top and bottom integral anchorage was added to Control Room Panels and to Motor Control Centers MCC 88, MCC 98, MCC DC2A, MCC 6A, and MCC 7A. These modifications ensure that the equipment to which they were installed will be able to perform its safety function subsequent to a licensing basis seismic event.

ECN-1 and ECN-2 added seismic anchorage to additional DC busses instrument racks.

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Safety Evaluation Summaty  ;

The modifications implemented by this design change do not interfere with any active features of the equipment to which they are installed. They are classified as safety related and are designed to criteria which meet or exceed the licensing basis seismic requirements. There is no adverse impact on any manual or automatic actions associated with the equipment. The margin of safety is inherently increased relative to the equipment capabilities to physically remain in place, and perform its safety function, subsequent to a seismic event. The  ;

> modifications have no adverse impact on any margin of safety. '

There was no increase in the probability of occurrence or consequences of an accident or malfunction as previously evaluated in the FSAR. This design change did not present ,

, significant hazards not described or implicit in the Vermont Yankee FSAR, and there is I s

. reasonable assurance that the health and safety of the public was not endangered.

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