ML20235V944
| ML20235V944 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/31/1988 |
| From: | Capstick R VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-89-22, NUDOCS 8903100419 | |
| Download: ML20235V944 (16) | |
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. ANNUAL REPORT 1988 6
OPERATIONS *
SUMMARY
Between January 1 and December 31 of 1988, Vermont Yankee completed a number i of changes. The following report describes those changes which constituted a change in the facility as described in the Final Safety Analysis Report (FSAR).
The report includes three (3) Engineering Design Change Requests (EDCR), one Plant Design Change Request (PDCR), one (1) Plant Alteration Request (PAR),
twelve (12) Temporary Mechanical Bypass Requests (MBR), two (2) Temporary Lifted Lead and Jumper Requests (LL/JR), and one (1) Instrument Setpoint Change. There were no Special Test Procedures or Experiments, and no Safety and Relief Valve challenges and/or Failures.
A. Chance in Facility Desion
- 1. During 1988 there were no changes made which required authorization from the Commission.
- 2. The following changes did not require prior Commission approval, they were reviewed by the Plant Operations Review Committee (PORC), and approved by the Plant Manager and Manager of Operations. It was deter-mined that these changes did not involve unreviewed safety questions as defined in 10 CFR 50.59 (a)(2), based on the information pretented.
(a) EDCR 87-405 ECN-1, " DESIGN, FABRICATION AND INSTALLATION OF THE 20x18 NES SPENT FUEL RACK" was completed May 10, 1988.
GENERAL
SUMMARY
The Vermont Yankee Spent Fuel Storage Pool was originally designed and licensed on the basis of a complete, closed fuel cycle. Tnis would require on-site storage of spent fuel for a year or two prior to shipment to a reprocessing facility. As the reactor core at Vermont Yankee contains 368 fuel assemblies, with between 92 and 136 being replaced on a proposed annual refueling schedule, a fuel storage capacity of 600 assemblies was considered adequate.
Once it became clear that reprocessing would not be available,
. Vermont Yankee took steps to provide additional on-site storage.
In September 1977, Amendment No. 37 to the Vermont Yankee Operating License was granted by the NRC allowing installation of new racks to accommodate 2,000 spent fuel assemblies. This woul:
permit Vermont Yankee to operate and maintain full core reserve discharge capability until 1990. However, the last three (3) par Systems spent fuel racks were not installed in the Spent Fuel Pool. This resulted in a fuel storage capacity of 1890 fuel assemblies and a loss of full core discharge capability after the 1989 refueling outage.
8903100419 DR 881231 l ADOCK 05000271 PNU ZH7-10
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, Originally.'this dIsign chang 2 was to provid2 Varmont Yanke2 with the ability to store 2.000 fuel assemblies. This would be
- i. accomplished by installing one (1) 20 x 18 Nuclear Energy Services L . (NES) Spent. Fuel Rack. Due to both, licensing and fabrication delays, this rack was not installed. However, the NES lifting tools and rack uoender were retained in this design change to pro-vide for moving the NES rack when it arrived. Therefore, this design change provides only for the relocationaof the PAR racks and miscellaneous equipment. Also provided.is the procurement of' the NES lifting fixtures to facilitate reracking. This ECN reflects the decrease in scope of the original design change package.
SAFETY EVALUATION
SUMMARY
r To facilitate the installation of the NES Spent Fuel Rack into the fuel pool, two of the existing PAR racks were relocated to the northeast' corner of the pool.
The relocation of PAR Rack Nos. 15 and 18 did not affect the racks center-to-center spacing. Additionally, an evaluation of Boral poison misalignment and rack module spacing was performed on a more conservative design (i.e., smaller center-to-center, higher enrichment) which demonstrated that if two adjacent rack modules with Boral sheets in-phase (i.e., face-to-face) have a clearance-of 1.0 inch between cell walls, the resultant K-effective increase is negligible.
No spent fuel racks were lifted over any spent fuel assemblies or
. rack containing spent fuel assemblies during this relocation.
An evaluation was completed which stated that the floor analysis done for the PAR racks was based on an equivalent uniform load.
The relocation of the PAR racks to a different location in the pool will have no affect on this floor loading analysis. l Therefore, relocating the PAR racks to'the northeast corner of the l fuel pool will not invalidate the existing floor analysis.
This change did not present significant hazards not described or ,
. implicit in the Safety Analysis Report and there is reasonable I assurance that the health and safety of the public was not endangered.
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3 (b) EDCR 87-408 ECN-1 "MOV LIMIT SWITCH REWIRE" was completed on April 7, 1988.
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' GENERAL
SUMMARY
' Routine testing revealed that motor-operated valves can be damaged 4 due to thrusting into the backseat under torque. The backseat torque, during the opening stroke of the valve, can exceed the manufa?CJrer's recommended value, and overstress the valve stem.
The excessive torque is caused by motor-operator inertia at the end'of the open cycle. This is also referred'to as coastdown.
The motor operator continues to run until the "open" torque switch operates to de-energize the motor when the valve disk hits the backseat. The motor, however, has enough inertia to provide excessive torque. In order to prevent the overtorque condition from occurring, a 95 percent open limit switch contact was con-nected in series with the torque switch for each valve. This limit switch contact will operate to de-energize the motor' opera-tor on the open stroke when the valve is approximately 95 percent '
open. Motor coastdown will provide the additional valve travel to fully open the valve without excessive backseating torque.
Additionally, to ensure that the valve opens on demand, an open limit switch was connected in parallel with the open torque switch to bypass the torque switch during the beginning of the open stroke. This will permit maximum torque for unseating the valve during the beginning of the opening cycle.
SAFETY-EVALUATION
SUMMARY
For valves which tre:normally closed, and must open upon an acci-dent signal, this change will help ensure performance of their safety function. The change will also help protect the valves from internal damage and enable the valves to reposition. The addition of the open limit switch in paralle' with the open torque switch will ensure that maximum torque can be applied to unseat.
the valves at the beginning of the open stroke. This change will also modify the present open-circuit wiring which allows the limit switch to bypass the torque switch during the entire stroke, thus preventing the torque switch from protecting the valve for mid-travel obstructions. The addition of the 95 percent open limit switch in series with the open torque switch protects the valves from internal damage by ensuring that excessive torque is not applied to the valve backseat at the end of the open stroke. The 95 percent open limit switch contact will de-energize the motor operator before the valve is fully open, allowing motor coastdown to supply the additional valve travel to fully open the valve.
The 95 percent value is approximate and will be set in the field i to allow full valve opening after motor coastdown. However, even if 100 percent valve opening were not achieved, there is a negli-gible flow difference between 95 and 100 percent valve open.
This alteration did not present significant hazards not described or implicit in the Safety analysis Report and there is reasonable assurance that the health and safety of the public was not endangered.
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,(c) EDCR 87-404 " POST ACCIDENT SAMPLING SYSTEM MODIFICATION" was completed on November 28. 1988 GENERAL
SUMMARY
These modifications provided plant personnel with the capability to monitor reactor coolant liquid flow rate at the Post-Accident Sample System Panaa, This ECDR also provided for the addition of a vacuum pumo check valve and a'Balston Vacuum Pump inlet filter.
This improved system performance by preventing oil / water mixture contamination of the gas sample cylinders and vacuum pump.
SAFETY EVALUATION
SUMMARY
The Post-Accident Sampling System was installed in response to the requirements of NUREG-0737. Item II.B.3. The system changes pre-sented in this EDCR were system improvements based on recommen-dations made during an NRC inspection. The affected system portion is neither safety class nor seismic. The electrical power supply for this modification was designed to be consistent with the balance of the sample panel electrical power supply to be energized by the standby diesel generators on loss of off-site power.
(d) PDCR 87-003 " FUEL GRAPPLE REPLACEMENT" was completed on May 24 IC88 GENERAL
SUMMARY
The purpose of this PDCR was to replace the existing Fuel Grapple with General Electric's Redundant Grapple Hook Modification Kit.
This modification consisted of a new redundant hook fuel grapple head as well as minor electrical changes to the refueling plat-form control panel. This PDCR also covered the welding of a piece of stainless steel flat stock to the back side of each of the two hooks on the new redundant fuel grapple.
The existing single grapple hock assembly was replaced by a redun-dant hook fuel grapple head. The head has two independent grapple books, (a primary hook and a secondary hook) operated by indepen-dent air cylinder actuators. Both hooks are equipped with inde-pendent position sensing micro switches that are wired in series to one position indicator located on the grapple console.
The position sensing micro switches on the redundant hook will eliminate the problems and maintenance associated with the origi-nal proximity switch. On average, two proximity switch failures have occurred during each refueling outage irrespective of the preventive maintenance performed. Each failure causes a minimum of four hours of critical path delay during fuel movements.
A piece of stainless steel flat stock was welded to the back sides of the two new fuel grapple hooks so that a fuel bundle will not be hooked by the U-bolt that is located on the back sides.
The Redundant Hook Modification Kit will assure the operators that they have made a proper alignment and positive engagement of the fuel bundle.
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. t SAFETY EVALUATION
SUMMARY
- i The original Refueling Grapple had a one hook head design with a
. position indication switch, that was found to be unreliable in the past. The same grapple design had been the cause of fuel. bundle drops at other sites. This grapple had been the cause of slowing the refueling process at Vermont Yankee when positive engagement could not be verified by the indicating sensor, thus requiring alternate verification methods. The new Redundant Hook Grapple head design has two hooks, each with a factor of safety equal to or greater than the existing single hook.
The two hooks are configured such that simultaneous hang up on the fuel bundle bail should not be possible. The new grapple is controlled by independent air cylinder actuators. These actuators use the same air flow and pressure as the existing grapple. Each hook has an independent position sensor that is wired in series to a single indicating light on the control panel.
This alteration did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the heath and safety of the public was not endangered.
(e) PAR 88-04 " STATION AIR COMPRESSOR UNLOADER LINE FILTER" was completed on July 20, 1988 GENERAL
SUMMARY
This PAR was written to make permanent, the modifications and alterations previously installed under temporary Mechanical Bypass 86-0042. The Mechanical Bypass installed filters in the Station Air Compressor Unioader supply line. The purpose of the filters was to eliminate rust in the unloader supply lines. This action reduced maintenance requirements and improved compressor reliabi-lity. To make this a permanent change, the filter rack support was slightly modified.
SAFETY EVALUATION
SUMMARY
This PAR affected the Service and Instrument Air System. This system provides the compressed air requirements for general plant l
services. Mechanical Bypass 86-0042 and subsequently the PAR were originated to permanently install air filters in the Station Air compressor Unloader line. The unloader's function is to signal the compressor to start when insufficient air quantity is in the system.
If the line filters were to become clogged, the unloader would not receive any air ir.dication thereby signaling the compressor to continue working. Tnis would be spparent because the compressors would run continuously causing the pressure relief valves to open.
l The change did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the health and safety of the public was not endangered, l
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(f)' MECHANICAL BYPASS 88-0050 was installed December 12, 1988 and removed December 14, 1988 GENERAL
SUMMARY
This Temporary Mechanical Bypass provided Service Water from the ,
"A" Turbine Building. Closed Cooling Water (TBCCW) Heat Exchanger for cleaning the internals of the "B" TBCCW Heat Exchanger. The' "B" TBCCW Heat Exchanger had previously been isolated and disassembled.
Drainage from the cleaning operation was collected in barrels and pumped to the adjacent Turbine Building Clean Equipment Drain ~
Sump. Sump capacity is adequate for this service and its use had been approved by the Chemistry Department.
e SAFETY EVALUATION
SUMMARY
The operation'of the TBCCW system was not affected since one of the two 100% capacity heat exchangers remained in service.- The small volume of Service Water (3-5 gpm) diverted from the operating exchanger did not decrease its cooling capacity.
Service Water system operation was not affected at,this-low bypass flow rate. The Service Water System is designed for 125 psig at 150*F. The red rubber hose used for the bypass, rated for 300 psig, was adequate for this service.
MBR 88-0050 did not present significant hazards not described or implicit'in the Safety Analysis Report, and there is reasonable assurance that the health and safety of the public was not ,
endangered.
(g)- MECHANICAL BYPASS 88-0049 was. installed December 7, 1988 and removed December 7, 1988 GENERAL
SUMMARY
This Temporary Mechanical Bypass provided Service Water from the "B" Turbine Building Closed Cooling Water (TBCCW) Heat Exchanger for cleaning the internals of the "A'"TBCCW Heat Exchanger. The "A" TBCCW Heat Exchanger had previously been isolated and disassembled.
Drainage from the cleaning operation was collected in barrels and pumped to the adjacent Turbine Building Clean Equipment Drain Sump. Sump capacity is adequate for this service and its use had been approved by the Chemistry Department.
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, . SAFETY EVALUATION
SUMMARY
The operation of the TBCCW system was not affected since one of
-the two 100% capacity heat exchangers remained in service. .The small volume of Service Water (3-5 gpm) diverted from the. '
operating exchanger did not decrease its cooling capacity.
i Service Water system operation was not affected at this low bypass flow rate. The Service Water System is designed for 125 psig at (
150*F. The red rubber bose used for the bypass, rated for 300 psig, was adequate for this service.
i MBR 88-0049 did not present-significant hazards not described or implicit in the Safety Analysis. Report, and there is reasonable assurance that the health and safety of the public was not endangered.
(h) MECHANICAL BYPASS 88-0029 was installed August 25, 1966 and removed August 25,.1988 GENERAL
SUMMARY
This Temporary Mechanical. Bypass installed a drain'line from a 55 gallon collection drum to the Reactor Building Roof Drain System.
The collection drum supported the repair effort ascociated with Residual Heat Removal (RHR) Service Water (SW) Valve 89A. A sub-mersible pump was-installed in the collection drum with a flexible hose discharging to the Reactor Building Roof Drain System. The roof drain system, at this location, is approximately 20 feet higher in elevation than the collection drum. ,
A 9 inch loop seal with 2 check valves was installed in the discharge line to prevent backflow from the roof drain.
SAFETY EVALUATION
SUMMARY
The primary concern with this Mechanical Bypass was Secondary Containment integrity. Mechanical Bypass Request 88-0029 provided 2 seals which maintained Secondary-containment.
- 1) The submersible pump was maintained under water whenever l the roof drain valve was open.
- 2) The loop seal was filled with water whenever the roof drain valve was opened.
The water seals were maintained prior to, and throughout the time ;
that Secondary Containment integrity was not maintained by the roof drain isolation valve.
This activity did not present significant hazards not described or implicit in the safety analysis report, and there is reasonable assurance that the health and safety of the public was not endangered.
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L(i)'. MECHANICAL BYPASS 88-0020 was. installed July 7, 1988 and removed j n July:12;:1988_ ~
R GENERAL
SUMMARY
1 This temporary Mechanical Bypass installed' equipment for the.
detection of condenser leakage. Helium injection equipment was attached to the:Off-gas Sampling System. Helium was'then-injected L intoLthe system, and the condenser was examined for 1eaks. This L action was conducted while the reactor was operating at:1ow power-(10-15%).
SAFETY EVALUATION
SUMMARY
The helium detection equipment was connected.to-sampling' inlet and.
outlet _ valves in the existing system. The respective valves were:
. closed _during instal _lation'and removal of this bypass. This bypass did not, in any way, hinder the' operation of the Off-gas.
Monitoring System. The materials used were determined to be-acceptable for this application.
This Temporary Mechanical Bypass did not present significant' hazards not described or implicit in the Safety Analysis Report and there is reasonable' assurance that the health and safety of the public was not endangered.'
(j) MECHANICAL BYPASS 88-0014 was installed July 1, 1988 and removed July 5, 1988 GENERAL
SUMMARY
This temporary Mechanical Bypass' installed equipment.for the detection of condenser leakage. Helium injection equipment was attached to'the Off-gas Sampling System. Helium was then injected-to the system, and the condenser was examined ~for leaks. _Thjs action was conducted while the reactor was' operating at low power (10-15%).
SAFETY EVALUATION
SUMMARY
The helium detection equipment was connected to sampling inlet and outlet valves in the existing system. The respective valves were closed during installation and removal of this bypass. This bypass did not, in any way, hinder the operation of the Off-gas Monitoring System. The materials used were determined to be acceptable for this application.
This Temporary Mechanical Bypass did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the health and safety of the public was not endangered.
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,(k) MECHANICAL BYPASS 88-0011 was insta11cd May 2, 1988 and rcmoved May 22, 1988 v
GENERAL
SUMMARY
This Mechanical Bypass installed a temporary replacement for the
'"B" Diesel Generator HVAC penthouse louvor air cylinder. This installation allowed the removal of the air cylinders for disassembly and inspection. A routine inspection had revealed a broken piece part in one cylinder. It was considered prudent, in light of this information, to disassemble and examine the remaining louver air cylinders. The cylinders were replaced with a section of bar stock. This maintained the louvers in a fully open position. In addition, the pneumatic cylinder air supply header was isolated, and the breaker for the cylinder solenoid was tagged out. This prevented any interaction with other systems.
SAFETY EVALUATION
SUMMARY
The safety function of the air cylinders is to open the inlet louvers to cool the "B" DG Room. Since the louvers will be secured in the open position, the safety function of the air cylinders is not compromised. The reason the louvers are capable of closing is to help keep the room warm in the winter. Since the DG room temperature is not a concern during the summer months, the louver " closed" function is not necessary.
No seismic' concern exists, since the bar stock will weigh less than the air cylinder. The profile of the bar will be narrower than that of the air cylinder. The bar stock will be a rigid unit and less susceptible to co11cpse than the air cylinder, during a seismic event.
The pneumatic cylinder air supply header will be isolated, and the breaker for the solenoid will remain tagged open. This will eli-minate any possible interaction with other systems.
There are no environmental qualification requirements associated with the penthouse louver enclosure.
This installation did not present significant hazards not described or implicit in the Safety Analysis Report, and there is reasonable assurance that the health and safety of the public was !
not endangered.
(1) MECHANICAL BYPASS 88-0008 was installed March 19, 1988 and removed June 27, 1988 GENERAL
SUMMARY
This Temporary Mechanical Bypass installed a pressure gauge at Fuel Pool Cooling Valve 210. This gauge was used to gather data for the design upgrade of the Fuel Pool Cooling System.
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- ~ SAFETY EVALUAT10N
SUMMARY
" Technical. Specifications Section-4;5.D.2' requires that each
. ' Service Water Pump deliver.2700.gpm against a.TOH of'250. feet in order'to provide sufficient heat sink capacity to dissipate resi-dual heat after a shutdown or accident. .The pressure rating of.
the installed: gauge was greater than the rating for the system.
.If'a .line break.were to: occur, it would occur in a Non Nuclear Safety;(NNS) portion of Service' Water. A failure of.th'e NNS-
. position of: the Service Water System has been previously analyzed not to impair the safety. function of the system.
This Mechanical Bypass did not present significant hazards not
-described or. implicit in the Safety Analysis Report and there is reasonable assurance'that the public health and safety was not endangered.
(m) MECHANICAL BYPASS 88-0007 was. installed March-17,1 1988 and' removed June 22, 1988.
GENERAL
SUMMARY
This Temporary Mechanical Bypass installed a pressure gauge at Fuel Pool Cooling Valve 209. This gauge was used to gather data for the design upgrade of the Fuel Pool Cooling System.
SAFETY EVALUATION
SUMMARY
Technical Specifications.Section 4.5.D.2 requires that each Service Water Pump deliver 2700 gpm against a TDH of 250 feet.in order to provide sufficient heat. sink capacity to dissipate resi-dual heat after a shutdown or accident. The. pressure rating of.
the' installed gauge was greater than the rating for the system.
If a line break were to occur, it would occur in a Non Nuclear' Safety (NNS) portion of Service Water. . A failure of the NNS position of the Service Water System has been previously analyzed not to impair the. safety function'of the system.
This Mechanical Bypass did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the public health and safety was not endangered.
(n) MECHANICAL BYPASS 88-0006 was installed April 29, 1988 and removed May 6, 1988 GENERAL
SUMMARY
This Temporary Mechanical Bypass installed a drain line from a 55 gallon collection drum to the Reactor Building Roof Drain System.
The collection drum supported the repair effort associated with Residual Heat Removal (RHR) Service Water (SW) Pump 1A. A sub-mersible pump was installed in the collection drum with a flexible hose discharging to the Reactor Building Roof Drain System. The roof drain system, at this location, is approximately 20 feet higher in elevation than the collection drum.
A 9 inch loop seal with 2 check valves was installed in the discharge line to prevent backflow from the roof drain.
SAFETY EVALUATION
SUMMARY
The primary concern with this Mechanical Bypass was Secondary Containment integrity. Mechanical 'dypass Request 88-0006 provided 2 seals which maintained Secondary Containment.
- 1) The submersible pump was maintained under water whenever the roof drain valve was open.
- 2) The loop seal was filled with water whenever the roof drain valve was opened.
The water seals were maintained prior to, and throughout the time that Secondary Containment integrity was not maintained by the Roof Drain Isolation Valve.
This activity did not present significant hazards not described or implicit in the Safety Analysis Report, and there is reasonable assurance that the health and safety of the public was not encangered.
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(o) MECHANICAL BYPASS88-000( was installed February 10, 1988 and was removed February 11, 1988 l GENERAL
SUMMARY
Hechanical Bypass 88-0004 temporarily provided Service Water from the "A" Reactor Building Closed Cooling 4ater (RBCCW) Heat Exchanger, for cleaning the internals of the "B" RBCCW Heat Exchanger. The "B" RBCCW Heat Exchanger had previously been iso-lated and disassembled. ,
Drainage from the cleaning operation was collected in a barrel.
and pumped to the adjacent Reactor Building Roof Drain Line. This use was approved by the Chemistry Department prior to starting.
SAFETY EVALUATION
SUMMARY
The operation of the RBCCW system was not affected, since one of the two 100% capacity heat exchangers remained in service. The small volume of Service Water (3-5 gpm) diverted from the operating exchanger did not decrease its cooling capacity.
Service Water system operation was not affected by the low bypass flow rate.
The concern with the connection between the collection drum and the Roof Drain dealt with Secondary Containment integrity.
Mechanical Bypass Request 88-0004 provided 2 seals which mein-tained Secondary Containment
- 1) The submersible pump was maintained under water wheneser the roof drain valve was open.
I 2) The loop seal was filled with water whenever the roof I
drain valve was opened.
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j The water seals were maintained prior to, and throughout the time <
that Secondary Containment integrity was not maintained by the
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. Roof Drain Isolation Valve. I
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This activity did not present significant hazards not described or .
implicit in the Safety Analysis Report, and there is reasonable assurance that the health and safety of the public was not endangered.
(p) MECHANICAL BYPASS 88-0001 was installed February 2, 1988 and was removed February 8, 1988 GENERAL
SUMMARY
Mechanical Bypass 88-0001 temporarily provided Service Water from the "B" Reactor Building Closed Cooling Water (RBCCW) Heat Exchanger, for cleaning the internals of the "A" RBCCW Heat Exchanger. The "B" RBCCW Heat Exchanger had previously oeen isc -
lated and disassembled.
Drainage from the cleaning operation was collected in a barrel, and pumped to the adjacent Reactor Building Roof Orain Line. This use was approved by the Chemistry Department prior to starting.
SAFETY EVALUATION
SUMMARY
The operation of the RBCCW system was not affected, since one of the two 100% capacity heat exchangers remained in service. The small volume of Service Water (3-5 gpm) diverted from the operating heat exchanger did not decrease its cooling capacity.
Service Water system operation was not affected by the low bypass flow rate.
The concern with the connection between the collection drum and the Roof Drain dealt with Secondary Containment integrity.
Mechanical Bypass Request 88-0001 provided 2 seals which main-tained Secondary containment
- 1) The submersible pump was maintained under water whenever the roof drain valve was open.
- 2) The loop seal was filled with water whenever the roof drain valve was opened.
The water seals were maintained prior to, and throughout the time that Secondary Containment integrity was not maintained by the Roof Drain Isolation Valve.
This activity did not present significant hazards not described or implicit in the Safety Analysis Report, and there is reasonable assurance that the health and safety of the public was not endangered 4
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(q) MECHANICAL BYPASS 87-0028: was installed November 9, 1987 and removed June 26; 1988 GENERAL
SUMMARY
This~ Mechanical Bypass' temporarily removed the high water level switch on the "B" Main Steam Line. With the switch removed, the high level alarm function was disable in the "B" Main Steam Line.
The piping for the level switch was removed and pipe caps were temporarily installed.
. SAFETY EVALUATION
SUMMARY
The water level _ indicator is classified Non Nuclear Safety. Even though water level in the "B" Main Steam Line could not be deter-mined directly, level could be determined from lines A,C,D due to similar configurations and conditions. Should water fill the "B" Main Steam Line Drain and flow downstream, protection for the Turbine was provided by the low point drain lines between the' Turbine Stop and Control Valves. These drain lines are open to the Main Condenser through a restricting orifice, and therefore provide continuous protection.
This Mechanical Bypass did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the public health and safety was not endangered.
(r) LL/JR 88-0011 (Temporarily installed lifted lead and jumper) was installed May 31, 1988 and removed December 1, 1988.
GENERAL
SUMMARY
This LL/JR installed a Relay Annunciator circuit for the Uppe,r Hi oil reservoir level alarm on the "B" Recire Pump. This was done to isolate the grounded level switch from the D.C. Annunciator System. The relay circuit uses the installed level switch to operate an A.C. relay, the relay. output contact is then connected to the C.P, annunciator. The grounded lead on the level switch was connected to the neutral side of the 120 VAC supply. A power monitor relay was also included to provide an alarm in the event of a power loss to the relay circuit. Power for the relay circuit was from the nearest 120V receptacle to B778.
SAFETY EVALUATION
SUMMARY
The relays were mounted such that they did not vibrate onto any of the energized circuits in the box. There is no safety class wiring located in this box. All wiring deals with the "B" Recire Pump Lube Oil alarms which is non-safety.
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This LL/JR made no connections to any safety circuits. The alarm.
system is non-safety, as are the Recire. Pumps, and the power
. supply.
This LL/JR enhanced operation by providing continued' monitoring of.
the "B" Recire. Pump. upper _ oil reservoir Hi level indication, without degrading the C.R. Annunciator system.
The FSAR, Figure'4.3-3 shows only that the level switches provide
.an annunciator in the Control Room. This relay circuit will allow this alarm capability to be maintained.
This change did not present significant hazards not described or.
implicit in the Safety Analysis Report and there is reasonable assurance that the health and safety of the public was not endangered.
(s) LL/JR 88-0001 was installed January 6, 1988 and removed June 29, ,
1988 GENERAL
SUMMARY
This LL/JR's purpose was to disconnect the leads for the annun-ciator contacts of the Automatic Depressurization System (ADS) bypass switch. The lifting of the leads disabled the alarm which indicated that the switch was in the bypass position. This was done to temporarily correct a short between contacts on the bypass switch.
SAFETY EVALUATION
SUMMARY
All safety functions of the ADS Bypass switch were maintained while this LL/JR was in effect. In the normal position remote operation of the relief valves was possible, in bypass all remote operation was blocked. The switch was placed in the normal posi-tion, and the key was removed to prevent inadvertent actuation.
The key was fastened to the Control Room panel where it was readily available.
The change did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the public health and safety was not endangered.
(t) INSTRUMENT SETPOINT CHANGE 87-26 was completed June 29, 1988.
GENERAL
SUMMARY
The purpose of this Instrument Setpoint Change was to restore the Diesel Generator (DG) Air System Compressor starting setpoint to the original design setting. The DG Air System supplies the compressed air required for starting the two Emergency Diesel Generators. The starting setpoint had been increased to ensure that sufficient compressed air would be available for multiple starts of the DG. A review of the DG safety design basis did not indicate a need for successive starts. It appears that the suc-cessive start criteria was included in the design specification to provide additional assurance that the DG would start when required.
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SAFETY EVALUATION
SUMMARY
, The FSAR'does.not explicitly identify multiple start attempts as a safety objective for the Diesel Generator System.- In any event,.
it has been determined through pre-operational testing that up to four starts are possible from a single receiver. In addition, actual testing performance has demonstrated that the diesels'are very-reliable.
The revised setpoint will still allow the Diesel Generators to start within the analyzed time frame and pick up their required load. As stated in FSAR Section 8.5, The diesels will start with a starting air pressure as low as 150 psig.
, This change did not present significant hazards not described or implicit in the Safety Analysis Report and there is reasonable assurance that the public health and safety was not endangered.1 B. Test Experiments
- 1. None C. Safety and Relief Valve Cha11ences and/cr Failures
- 1. None D. Special Test Procedures
- 1. None EJT/tms:ES9024.1/ES
! VbRMONT YANKEE NUCLEAR. POWER CORPORATION
. RD 5, Box 169, Ferry Road, Brattleboro, VT 05301 g %
ENGINEERING OFFICE 580 MAIN STA2ET
.) March 1, 1989 f3o$"j$#8 ii BVY 89-22
~
United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC. 20555
Reference:
(a). License No. DPR-28 (Docket No. 50-271)
Subject:
Vermont Yankee 1988 Annual Operating Report
Dear Sir:
Enclosed please find one copy of the Vermont Yankee Nuclear Power Corporation Annual Operating Report submitted in accordance with 10CFR50.59(b). This report describes the facility changes, tests, and experiments conducted without prior NRC approval during the. year 1988.
We trust this information is acceptable; however, should you have any questions, please contact this office.
Very truly yours, VERMONT YANKEE NUCL AR POWER CORPORATION R. W. Capstick Licensing Engineer RWC/sv cc: USNRC, Region I USNRC, Resident Inspector - VYNPC
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