ML20235U669
| ML20235U669 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1988 |
| From: | St Laurent N YANKEE ATOMIC ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BYR-89-038, BYR-89-38, NUDOCS 8903090317 | |
| Download: ML20235U669 (29) | |
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YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION (DOCKET No.
50-29) 1988 ANNUAL REPORT hn$f309Esilggggas A
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1988 ANNUAL REPORTE TABLE OF CONTENTS i
TITLE PAGE Table of Contents.....................................
1 Introduction..........................................
ii Changes l
1 A.
Engineering Design Changes...................
1
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B.
Plant Design Changes.........................
14 1
i C.
Plant Alterations............................
21 i
D.
Operational Changes..........................
21 Tests.................................................
24 Experiments.................'..........................
24 Safety and Relief Valve Failures and Challenges.......
24 Specific Activity Analyses............................
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INTRODUCTION The Yankee Nuclear Power Station is a pressurized water reactor plant of 185 MW electrical capacity.
The Nuclear Steam Supply System is a Westinghouse four-loop design.
The architect / engineer and constructor for this plant was Stone &
Webster Engineering Corporation of Boston, Massachusetts.
The main condenser cooling is a.once-through design using the Deerfield River as the cooling medium.
The plant is operated in accordance with Facility Operating License DPR-3, issued July 19, 1960.
The date of initial reactor criticality was August 19, 1960, and commercial operation began July 1,
1961.
Part 1 of this annual report is submitted in accordance with Technical Specification 6.9.2.b.
This report of changes, tests, and experiments is submitted in accordance with 10CFR, Chapter 1,
Part 50.59(b).
The changes, tests, and experiments identified in this report have been reviewed for and were determined not to constitute an unreviewed safety question as described in 10CFR50.59(a)(2).
Part 2 of this annual report is submitted in accordance with Technical Specification 6.9.2.d.
This report of Specific Activity Analyses is submitted to identify primary coolant which has exceeded the limits of Technical Specification 3.4.7.
Part 3 of this annual report is submitted in accordance with TMI Action Plan, Item II.K.3.3 and provides a summary of safety valve and relief valve failures and challenges.
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PLANT CHANGES
-A.
ENGINEERING DESIGN CHANGES (EDC'S) o EDCR 96-302 - SAFE SHUTDOWN SYSTEM PIPING MODIFICATIONS INSIDE THE VAPOR CONTAINER PHASE III The support modifications installed by this EDCR are as follows:
1.
The pipe support qualification of the piping is listed below.
The piping meets the requirements of the design criteria when subjected to seismic loads based on the NRC spectra, a.
Main Steam Piping, all four loops b.
Feedwater Piping all four loops c.
Steam Generator Blowdown Piping, Loops 1 and 4 d.
Main Coolant Loops and Pressurizer Drains e.
Pressurizer Sample and Vent Lines f.
Charging Line Through the Drain Box g.
Component Cooling Piping-h.
Shutdown Cooling Piping 2.
One support on the tubing from the Loop 2 safety injection piping to Transmitter SI-PT-2.
This j
tubing support meets the requirements of the design criteria when subjected to seismic loads based on the NRC spectra.
l The affected piping was in a stress-analyzed condition j
prior to plant startup.
The design criteria, together with l
the conservative relative seismic building displacements, l
provides assurance that the piping, supports, valves, and j
equipment will maintain their integrity and perform their intended function during a seismic event equivalent to an i
upper level Safe Shutdown Earthquake (SSE).
ANSI B31.1 provides assurance that the piping, supports, valves, and
)
equipment will perform their intended function during
]
normal and emergency conditions.
The modifications j
required by this design change will not affect the environment inside the Vapor Container.
Therefore, the environmental qualification of safety-related equipment
. located therein remains valid.
The piping systems affected by this EDCR are classified as Safety C1 esses 1,
2, and 3.
Table 3.7-4 of Technical l
Specification 3.7.9 has been revised by the addition of a snubber.
Technical Specification 3.7.9 allows the addition of snubbers to safety class piping without prior License Amendment provided that a revision to Table 3.7-4 is included with the next License Amendment request.
Therefore, Proposed Change No. 218, including the revised table, will be submitted to the NRC after installation of the snubber.
There are no FSAR pages affected by this i
EDCR.
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o EDCR 87-302 - BATTERY NO. 3 REPLACEMENT On April 30, 1988, the existing 125 de, 560 Ah, Battery No.
3, purchased in 1970 with an expected life of 16 years, was removed from service.
The removal of the existing battery was performed by first paralleling a new 125 V dc, 1,304 Ah i
battery (new Battery No. 3) to the existing battery, and l
subsequently disconnecting the existing battery.
This battery swapping was performed with the plant in hot standby operation Mode 3.
The existing Battery Charger No. 3 and Switchboards were left in place to perform the same function as before but in conjunction with the new Battery No.
3.
During the 1988 refueling outage, a new upgraded 125 V de Switchboard No. 3 and a new 3OOA Battery Charger No. 3 were installed, both located in the Battery No. 3 Room.
The new battery, switchboard, and charger have been provided with larger capacity to allow for future expansion and greater flexibility for powering additional loads.
The new 125 V dc, 1,304 Ah, Battery No.
3, is located in the Battery No. 3 Room, adjacent to the Post-Incident Cooling System (PICS) Room.
The new 3OOA Battery Charger No.
3, manufactured by Power Conversion Products, Inc.,
identical to the existing Battery Charger Nos. 1 and 2, was installed adjacent to the new Battery No.
3.
The new charger and spare charger have adequate capability to power the normal dc Bus 3 loads, and fully recharge the new batte'ry from a complete discharge state in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and 10.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, respectively.
The DC System performance will be enhanced, since on loss of AC, all safety and nonsafety dc loads supplied from the new Battery No. 3 can be powered for a substantially longer time, due to the increased size of the new battery.
The safety classification of the components affected by this design change is designated as follows:
Battery No.
3, 3OOA Charger No.
3, 75A Charger No.
3, Existing 125 V dc Switchboard No.
3, SOOA Dattery Charger No. 3 ac Power Supply, and New 125 V de Switchboard No. 3 are Safety Class; the Vent Fan and Flow Switch are Non-Nuclear Safety Class.
The removal of the existing battery and its associated equipment from a harsh environment and its replacement with new equipment located in a mild environment, reduces the probability of malfunction.
The new battery, its charger, and dc switchboard perform the same function as in the previous configuration.
The protection of the new battery is assured by a new set of fuses which, similar to the existing fuses, are equipped with blown fuse alarms to alert the operators of battery 2
I a
inoperability as a result of a blown fuse.
The new Battery Charger No. 3 power supply is located in a mild environment and has the capability to adequately feed the new charger.
A change to the Technical Specifications was required to allow startup with the new Battery No. 3 and its associated l
equipment.
A proposed Technical Specification change letter was issued to, and approved by, the Commission as Amendment No. 114 to the Technical Specifications.
Sections 226 and 235 of the Final Safety Analysis Report (FSAR) will be revised as a result of the modifications made by this EDCR.
o EDCR 87-306 - PROTECTIVE RELAY IMPROVEMENTS This EDCR improved the capability of the plant's Overcurrent Relay Protective System to isolate faults closest to their sources and to minimize equipment damage and unnecessary loss of power to electrical equipment not directly affected by a fault.
This change also allows ACB 448 and 424 to be closed following a generator or SST 1 fault only if ACB 124 is open to isolate 2400 V Bus 1 from the fault.
This will speed up the restoration of power to 2400 V Bus 1 and 480 V Bus 4-1, by not requiring trip signals to be reset prior to closing the breakers.
This modification prevents ACB 424 and 448 from tripping on generator or SST 1 faults during bus tie modes with ACB 124 open.
The safety classification of the components affected by this modification is designated as follows: ACB 424, ACB 448, ACB 124, MCC 1, Bus 2, Feeder Breaker (Bus 6-3) MCC 3, Bus 2, Feeder Breaker (Bus 4-1), MCC 4, Bus 2, Feeder Breaker (Bus 4-1), MCC 1,
Bus 1,
Feeder Breaker (Bus 5-2),
MCC 4,
Bus 1,
Feeder Breaker (Bus 5-2), BT1A, EBus 1,
Feeder Breaker (Bus 6-3), BTEA, EBus 2, Feeder Breaker (Bus 4-1), BT3A, EBus 3, Feeder Breaker (Bus 5-2), Emergency Turning Gear Oil Pump, 7.5 KVA XFMR "B"
for Gaitronics are Non-Nuclear Safety Class.
EMCC 1,
Normal Feeder Breaker (EBus 1), EMCC 1,
Alternate Feeder Breaker (EBus 2), EMCC 5,
Feeder Breaker (EMCC 3), EMCC 6, Feeder Breaker (EMCC' 4), BT1B, Feeder Breaker (EBus 1), Transformer "B" Primary Breaker are Safety Class.
The coordination of the breakers will minimize damage to the system and its components, and limit the extent and duration of service interruption whenever equipment failure occurs on the system.
The changes created by this EDCR do not modify the electrical system as described in the FSAR.
The modification of Air Circuit Breakers 448 and 424 controls allows the re-energization of 480 V Bus 4-1 following its isolation from the generator or SST 1 fault and enhances the existing control circuit design, speeding up the plant 3
4 s
electrical system operation following' clearing of a fault.
Moreover, the plant Protective System is enhanced since the improvements in the breaker coordination will quickly isolate the affected portions of the system, while maintaining normal service for the rest of the system.
The equipment affected by this modification will perform the same function as at the present time.
The breaker function remains unchanged while the flexibility of being manually operated following a fault will be increased.
The coordination improvement does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR, since the new setpoints will enhance the protection of the electrical system, isolating only the portions affected by equipment failure, faults, or human errors.
o EDCR 87-310 - REACTOR CAVITY DRAIN LINE MODIFICATIONS Vertical standpipes were installed at the existing reactor cavity vent lines.
Each standpipe was fitted with an inlet filter screen.
The existing reactor cavity drain piping inside containment was modified as described belows a.
Stainless steel plugs were installed in each of two existing 3" diameter cavity drain lines which penetrate vertically through the concrete reactor support structure, b.
A removable flanged spool piece was added to each drain line below the reactor support structure to facilitate installation and removal of the plugs.
c.
The existing 1" overflow lines connected between the cavity drain lines and the Vapor Container (VC) sump ring header were tied together upstream of a new normally closed isolation valve to facilitate manual draining to the VC drain tank.
This is required to preclude draining the reactor cavity to the VC sump via the ECCS overflow tees following neutron shield tank failure.
d.
Thermal relief valves were installed on each telltale drain line.
These are required to preclude containment barrier piping ruptures caused by thermal expansion of j
trapped fluid following the postulated accident.
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The neutron shield tank telltale drain line piping outside containment was modified as described below:
l a.
A single level switch was installed to monitor unidentified leakage which may collect in the cavity during normal plant operation.
The switch annunciated in the Control Room.
This was necessitated by removal of the 1" over flow lines inside the containment discussed in the paragraph above (Item c).
Thus, the l
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L level switch addition and associated piping chances described provide a method for leakage detection and quantification equivalent to that provided by the current system design.
b.
The neutron shield tank telltale lines were tied together upstream of the level switch to provide continuous monitoring of the inner and outer cavity drains.
The safety classification of the systems or portions of the systems affected by this change are as follows:
a.
The neutron shield tank telltale drain piping outside the vapor container, including the level switch float cage, up to and including the containment isolation valve, is Safety Class 2.
b.
The relief valves located inside the vapor container on the nonsafety-related portion of the drain - line have been classified as Safety Class 2.
This classification is based on the function of the relief valves in protecting the containment boundary during the hypothetical core melt accident.
Note that this is a conservative classification since the postulated accident falls outside the plant design basis for mitigating such events.
c.
The Magnetrol level switch is classified as Requiring Quality Assurance (RDA), commensurate with its function as defined in proposed Technical Specification Change No.
217.
d.
The reactor cavity standpipes and supports are non-nuclear safety.
The neutron shield tank drain piping and supports inside the vapor container up to the first weld inboard of the containment penetrations are classified non-nuclear safety.
The modified containment penetration design enhances the integrity of the containment pressure boundary by elimination of a potential leakage path.
All modifications have been evaluated for potential impact on intended functional capabilities and in no way alter the safety-related function of any system, structure, or components installed to mitigate any accidents.
In addition, all modifications implemented by this change have been designed to be entirely consistent with the existing plant design bases.
This process of evaluating the potential impact on installed safety-related systems, combined with the use of consistent design criteria, precludes the possibility of an accident which has not been previously evaluated.
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.e This modification requires revision to Technical Specification 3/4.6, " Containment Systems."
Table 3.6.1, Category B.1, was revised to delete VD-V-752, as this valve was being removed and the outer telltale drain line was tied into the inner drain line upstream of the new level switch.
This' modification also: revised Technical Specification 3/4.4.5, " Main Coolant System Leakage Detection Systems,"'
to add the operability requirements of the new level switch as part of the Main Coolant Leakage Detection System.
Surveillance requirements include periodic (once per 18 months) channel functional testing and channel calibration.
Technical Specification action requirements include opening the normally closed isolation valve to allow potential leakage to be directed to the VC drain tank if the level switch is inoperable.
Proposed Change 217 was generated to address these changes.
This change was approved and incorporated as Amendment No. 116.
The radioactive drains affected by this modification are discussed in FSAR Section 211, " Primary Plant Vent and Drain System."
The required revisions will'be incorporated in the next annual FSAR update scheduled following the 1988 refueling outage, o
EDCR 88-302 - NRV PRESSURE SWITCH HEAT TRACE POWER SUPPLY MODIFICATION This design change provides an independent power supply to the redundant pressure switch cabinets and associated pressure switch sensing line heat trace.
Therefore, a loss of a single power supply will not disable all of the pressure switch heat trace.
In the past, all of this equipment was powered from a single power supply.
The functionality of the system remains the same as before.
The independent power supply and better monitoring capabilities enhance the reliability of the system.
Other considerations such as separation of safety and nonsafety-related cables and the functionality of the pressure switch cabinets remain unchanged.
This modification also relocates controls to a new and larger cabinet and provides additional monitoring capability to the operator.
This change is, therefore, an enhancement to the existing design, and will not increase the probability of malfunction of any associated systems or components.
The. safety classification of the components affected by this modification is designated as follows:
L 1.
Main steam line pressure switch heat trace panel and associated components heat tracing, distribution cabinet C4, distribution cabinet T7, and pressure switch termination cabinet (PSTC) are Non-Nuclear Safety Class.
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2.
'The pressure switch cabinet (PSC1)' pressure switch cabinet (PSC2), and pressure switch cabinet-(PSC3) are. Safety Class.
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The Plant Technical Specifications and FSAR are not affected by the implementation of this modification.
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EDCR 89-303 TURBINE HALL HIGH RADIATION MONITOR REPLACEMENT q
l The old post-accident monitor (Nuclear Research Corporation) was replaced with a Victoreen high range radiation monitoring channel, Victoreen Model No. 877.
This replacement channel is similar to the instrument.
channels which are currently in use for the Vapor Container High Range Radiation Monitoring System.
The instrument change consisted of an ionization chamber detector, a i
readout module, rack assembly, coax cable, and the I
1 necessary connectors for wiring between the detector and the module.
This modification provides an enhancement of the present Turbine Building high range monitoring channel by ensuring continuous, reliable monitoring.
The new equipment has
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proven reliability by its. operation in other plant monitoring channels for long periods of time.
i The new channel includes an improved fusing scheme by which the external wiring is further orotected.by a 1 ampere fuse.
The new equipment does not increase the electrical j
loading of the nonessential UPS, the power source for this channel.
The modifications in this Design Change will not adversely impact any safety-related systems.
The new equipment will provide continuous reliability and improved range ability.
The human factors enhancements which result from this modification provide additional aids for the plant operators during abnormal plant operation; the status l
lights of the monitored parameter / equipment are in direct operator view, and the operators are familiar with the readout meter / status light arrangement.
The ratemeter, detector, and cabling do not perform any i
engineered safeguard system functions, nor does the cabinet j
in which this equipment is housed contain any nuclear safety-related equipment.
The equipment is classified as i
non-nuclear safety.
Changes to the Technical Specifications are not required as a result of this modification.
No additional function has been added or deleted as a result of this change; it is simply an equipment replacement of existing equipment which is not presently referenced in the Technical Specifications.
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Section'215 of the Final Safety Analysis Report (FSAR),
j Instrumentation Systems, requires a~ revision as a result of the modifications made by this EDCR.
o EDCR 88-306
-REPLACEMENT OF SW-TV-408 The-replacement of CIS Valve SW-TV-408 in the Service Water q
System does not alter the function of the system in any I
way.
The existing globe valve, which had a history of leakage problems, has been replaced with a ball. valve which 1
is better suited for the isolation function.
The operation of the new valve is identical to the globe valve with the only difference being the leak tight shut-off capability of~
l the ball valve.
By solving the leakage problem of SW-TV-1 408, the new valve improves the present CIS.
L Additional changes improve the leakage testing capabilities.
of SW-TV-408.
The previous test method involved pressurizing an entire manifold and four valves which were never intended to be test barrier valves.
Any leakage past these four valves had to be attributed to SW-TV-408.
This design change reduces the test volume ar,d thereby creates a more controlled and accurate test setup.
Replacement of SW-V-872 with a manual ball valve moves it from just downstream of SW-TV-408 to just upstream of SW-TV-408.
A new spool piece with test tap was installed between SW-V-l i
872 and SW-TV-408.
These changes make the 6" spool piece the test volume and allow for a very controlled leak test.
In addition ta these modifications, a vent. connection was installed downstream of SW-TV-408.
This connection provides venting during the leakage testing.
The changes 4
described above do not detract from any of the existing i
system functions but do provide the optimum leak test setup for SW-TV-408.
The safety classification of the components affected by this EDCR are as follows:
Isolation Valve SW-V-872, the CIS Valve SW-TV-408, and the spool piece (including test valve) are classified Safety Class 2: The test connection and slip-on flange downstream of SW-TV-408 are classified non-nuclear safety:
The Safety Class 2 to non-nuclear safety interface remains at SW-TV-408.
The system description in FSAR Section 222, " Water Supply Systems" remains unchanged.
This design change does not impact any sections or bases of the Technical i
Specifications, o
EDCR 88-307 - NI UPGRADE-MECHANICAL MODIFICATIONS This design change had two purposes.
The first was to provide conduit slab penetrations in the Control Room floor for electrical conduit to be run to the (future) NI cabinet (EDCR 86-310).
The second purpose was to install the anchor bolts for this new cabinet.
The Control Room 8
s concrete floor and structural steel supporting it is a fire barrier accepted by the NRC.
The new penetrations through the floor are sealed with a three hour fire-rated material p
to maintain the fire rating of this barrier.
This modification ensures that the Control Room floor retains its structural integrity after the conduit penetrations and new NI panel are in place.
Additionally, the fire rating of the Control Room floor slab and steel is maintained at its rating.
Therefore, neither the probability of a structural failure nor the consequences of a fire in the Switchgear Room have been increased.
Equipment above, below, and in the vicinity of this design change are unaffected.
The consequences of a fire in the Switchgear Room has been previously analyzed.
The Control Room floor and structural steel are an acceptable fire barrier.
This modification maintains the Control Room floor and structural steel as an equally acceptable fire barrier.
It also maintains the structural integrity of the Control Room floor.
Therefore, it does not cause an increase in the consequences of an accident of a different type than any previously analyzed.
The new structural steel, existing steel, and the reinforced concrete Control Room floor are classified as non-nuclear safety.
The anchor bolts for the new NI cabinet are safety class.
The Plant Technical Specifications and FSAR are not affected by the implementation of this change.
o EDCR 98-308 - CONTROL ROOM EMERGENCY AIR CLEANING SYSTEM (CREACS) ISOLATION DAMPERS This design change adds compressed air tubing and air-operated isolation dampers to the CREACS.
These isolation dampers were installed both upstream and downstream of the normal Control Room Ventilation System where it joins with the CREACS.
Unlike the outdoor air intake damper and the normal ventilation recirculation damper, the isolation dampers are low-leakage type dampers certified under the Air Movement and Control Association (AMCA) certifi:.d ratings program.
These isolation dampers will minimize the air leakage between the CREACS and the normal Ventilation System.
To provide for improved CREACS flow balancing, blade material is being added to the CREACS volume control damper.
The resulting increase in surface area will increase the volume damper throttling capability and allow for CREACS operation at its rated flow of 3,000 cfm.
The original surface area of the CREACS volume control damper provided only enough system throttling capacity to maintain the CREACS at slightly less than its 3,300 cfm upper-flow limit.
The new surface area allows the CREACS 9
to'be throttled to its design flow of 3,000 cfm.
The design intent, as well as the operation of the'CREACS, remains unchanged by this design change.
Rather, the CREACS ability to perform its intended function has been.
enhanced.
This design change affects equipment which is used following an accident to filter the Control Room atmosphere in order to minimize radiation exposure to Control Room personnel.
This design change adds new isolation dampers-with more superior low-leakage characteristics than the existing normal ventilation isolation dampers.
Since the original isolation dampers remain functional, the new isolation dampers provide increased assurance of the CREACS isolation capability.
Therefore, this design change serves to diminish the probability or consequences of equipment malfunction and reduces the exposure of Control Room operators due to air leakage into the CREACS.
This design change serves to improve the performance of the CREACS without affecting it operationally.
Mechanically, air-operated isolation dampers continue to be-used - the notable characteristic of these new dampers being their improved low-leakage performance.
The' fail-safe or inadvertent positioning of these dampers is consistent with the existing normal ventilation dampers.
That is, all dampers fail to a closed position to assure operation of the CREACS.
Therefore, this design change cannot create an accident or equipment malfunction of a different type which has not been previously analyzed.
The CREACS isolation dampers, volume control damper, duct work, compressed air tubing, and supports are all classified as Non-Nuclear Safety (NNS).
FSAR Sections ii, vii, and xii, " Drawing Index," will be revised to add,
" Flow Diagram, Control Room Ventilation," Drawing No.
9699-FM-93A to Section 228.
FSAR Section 228, " Ventilation Systems," will be revised to add reference to the " Flow Diagram, Control Room Ventilation," Drawing No. 9699-FM-93A.
This EDCR does not affect any Plant Technical Specifications.
o EDCR 88-309 - REPLACEMENT OF MAIN STEAM LINE PRESSURE SWITCHES This design change replaced main steam line ASCO diaphragm-type pressure switches with Barksdale bourdon tube pressure switches.
The Barksdale bourdon tube pressure switches should not experience the failure modes observed in the ASCO pressure switches since these pressure switches are i
constructed of an all-metal pressure boundary wetted sensor as opposed to the temperature / pressure sensitive organic diaphragm sensor of the ASCO switches.
The function of the microswitch actuation to operate the l
NRV control circuit logic remains unchanged using a 10 s
d
normally closed contact (fail safe) that closes on decreasing pressure.
The Barksdale pressure switches (Model No. BIT-M12SS-TC) were procured commercial grade and were upgraded to Safety Class 2 far the pressure retaining parts (bourdon tube) and to Safety Class for the electrical interface (Microswitch).
The upgrade to Safety Class 2 and Safety Class is attributed to the analysis of operational and trending data obtained from two similar switches (Barksdale Model No.
BIT-M12SS) installed on the non-nuclear side of the main steam line per a Temporary Change Request (TCR), and the preoperational test and critical characteristic verifications of the pressure switches prior to installation.
Additionally, the pressure switches were mounted in an inverted configuration, pressure port at 12 o' clock, per the vendor recommendation and installed per TCR, to ensure a static head of water is formed in the bourdon tube.
This static head of water will maintain the ambient temperature surrounding the Microswitch below 165'F.
Operational and trending data from the pressure switches installed by a TCR indicates the ambient temperature to be approximately 85'F.
The equipment affected by this design change is classified as follows; The microswitches are Safety Class, and the bourdon tubes (pressure boundary wetted parts) are Safety Class 2.
The Plant Technical Specifications and FSAR are not affected by the implementation of this change, o
EDCR 88-310 - WATER CLEAN-UP SYSTEM TIE-IN CONNECTIONS AND VC PENETRATIONS The design changes incorporated within this EDCR were a direct result of the ongoing engineering design of the Water Cleanup System (WCS).
The ECCS LPSI header tie-in installed under this EDCR functions as the WCS suction line.
The two vapor containment penetrations installed under the EDCR act (1) as a spare penetration and (2) as the WCS return line to containment.
The component cooling feed and return line tie-ins installed within this EDCR act as WCS heat exchanger cooling water inlet and outlet system connections.
During the interim period from tie-in installation during the refueling to eventual WCS l
completion, each of the above piping connections consists l
of independent locked closed manual isolation valves and l
welded pipe caps.
These modifications with their respective valves will not be utilized until the subsequent installation of the WCS is complete.
The component cooling flow channel (CC-FIT-201) was also modified and upgraded by this EDCR.
This upgrade consisted of the installation of a new flow indicating transmitter, converter, and power supply.
The installation of the above 11
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g equipment provides reliable and accurate Component Cooling System (CCS) flow-measurement'to support the future WCS heat exchanger component cooling water flow adjustment.
In
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support of a completed'WCG design, additional 1
instrumentation / control modifications were performe' within the. Control Room and Switchgear. Room.
These modifications include indicator cutouts on the recirculation control panel (RECIRC), routing a conduit slab from the Nonessential Uninterruptible Power Supply (NEUPS) cabinet, and removal of the obsolete security cutout switch in the switchgear, installed during the refueling to minimize the possibility of affecting plant operations during normal operation.
The modifications encompassed do not degrade the safety function of any existing safety class' system or instrument channel.
Each modification installed by this EDCR results either in no change to the existing system operation or provides for enhanced operation of an existing system.
The systems affected by these modifications are classified as follows: WCS Containment Penetration and ECCS LPSI Header Tie-Ins are Safety Class 2; Component Cooling System Tie-Ins and Flow Channel (FIT-201) Pressure Boundary Wetted Parts are Safety Class 3; Component Cooling Flow channel (FIT-201) Electronics, Moore I/P Converter (CC-I/P-201),
Power Supply (CC-P/S-201), and 1 Amp Fuse are Non-Nuclear Safety; 15 Amp Breaker is Safety Class.
The following Technical Specification changes were required, and approved due to this modification.
1) add a new Section C.2, check valves (subject to Type C testing), to Table 3.6-1, containing Cher.k Valve WC-V-622 and 2) add Manual Isolation Valves WC-V-621, VD-V-1170, and VD-V-1171 to Section B.1 of Table 3.6-1.
The component cooling piping modification and component cooling flow channel upgrade do not affect Section 206 of the FSAR.
The ECCS piping modification does not affect the written portion of Section 212.
The new piping penetrations do not affect any section of the FSAR.
o EDCR 88-313 - MAIN COOLANT LOOP ISOLATION VALVES: DISC AND STEM REPLACEMENT This design change provides the evaluation and l
justification for the installation of new valve disc assemblies and stems in any of the four (4) hot leg main coolant stop valves.
The new discs and stems have been redesigned to improve their long-term reliability.
The changes in the design of the valve disc assemblies and valve stem have been evaluated.
The new valve discs and stems are direct replacements with respect to form, fit, 12
w and function.-.The closure time of the valve.will not be affected by the installation of the new discs.
. The new discs and stems do not affect the structural
. integrity'of the pressure boundary (valve body and bonnet),
since they are valve internals and not part of the valve body.
.The new discs and stems are dimensionally equivalent to the original design.
Therefore, operation of the valve is not affected by this design change.
The changes in material selection, spring configuration, and hard-facing process were made to enhance long-term reliability and to incorporate the latest technology in manufacturing of valve' discs.
Westinghouse has successfully applied these changes to larger 24" gate valve discs.
To date, no problems have been reported resulting from the changes in the material, spring configuration, or hard-facing process.
The installation of this change also evaluated the use of the temporary blind flanges which may be installed on the valve following tne removal of the bonnet and internals if the valve is disassembled for a long period of. time.
Installation of the flanges, using the design parameters and specifications contained in this EDCR, ensures that the MCS remains filled during refueling operations.
The blind' flanges have been designed for the static pressure of. water above the valve as well as the peak pressure and temperature that could occur in the unlikely event that shutdown cooling is lost.
For the 1998 refueling, the flanges were used during Mode 5 only.
For future outages, the blind flanges may be used during Modes 5 and 6.
The blind flanges must be upgraded to Safety Class 2 requirements for use in Mode 6.
There will be no permanent changes to any plant system due to their installation.
The 20" main coolant isolation valves are classified as Safety Class 1.
The temporary blind flange and gasket are classified non-nuclear safety.
No changes to the Plant Technical Specifications and FSAR are required to upgrade the four hot leg isolation valves.
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B.
PLANT DESIGN CHANGES (PDC'S) o PDCR 86-010 - REPLACEMENT OF ACCESS CONTROL DOORS This change replaced access control doors which.were damaged or under UL rated throughout the plant.
The doors are classified as Non-Nuclear Safety with some of the doors being classified as Requiring Quality Assurance.
Since the-design change is identified as " Safeguards Information" additional information will be provided upon request.
The Technical Specifications and the text of the FSAR are-not affected by the implementation of this change.
Compensatory (Security and Fire Protection) measures were taken to ensure that the systems remained functional at all times.
o PDCR 87-013 - REPLACEMENT OF UH-13-1 & UH-13-2 IN WATER TREATMENT This design change consists of replacing the heaters in the Water-Treatment Room with heaters of greater capacity.
The new heaters will supply adequate warmth for the secondary auxiliary operator during cold weather.
The operation and function of the heating system for the Water Treatment Room remains. unchanged.
The old (UH-13-1 and UH-13-2) heaters supplied 80,000 BTU /HR each (design) while the new ones supply 177,100 BTU /HR.
Minor changes to the 15 PSIG steam supply and condensate return piping were required to accommodate.the-larger heaters.
Also, the existing support brackets welded to the structural steel of the building were replaced with longer brackets.
The heating system in the Water Treatment Room is classified as Non-nuclear Safety.
It is not an Environmentally Qualified (EQ) system nor impacts on any EQ systems or components.
The piping is not seismic-related.
This design change has no effect on the FSAR or any Plant Technical Specification.
o PDCR 87-015 - CONTROL ROOM L EOF DATA COMMUNICATIONS IMPROVEMENTS This design change affects the SPDS, MET NOVA computer, PDP-8 and Yankee Plant Network (YPN).
The EOF MET terminal performance was improved by having current data directly transferred to the EOF, Plant operations were not affected during the implementation of the changes in this PDCR.
The i
equipment involved with this PDCR is not involved in any plant process necessary for operation.
Further, the changes do not affect the* basic functions of the computers and data systems, but improve their overall reliability by improving the data channels and terminals.
Software i
changes were only necessary to the YPN.
This equipment is classified as Non-Nuclear Safety Related.
Technical Specifications are not affected by this design 14
4 change.
FSAR Section 301, " Meteorology" is affected by this design change, however the function of the MET system was not changed from the FSAR description.
o PDCR 87-016 - REPLACEMENT OF THE NON-RETURN VALVE SOLENOIDS This modification is a one-for-one replacement of the exercise and dump solenoid valves on the existing Hockwell f
International model A260 valve actuator on the Plant's Non-l Return Valves (NRV).
This one-for-one replacement with Rockwell qualified solenoids in no way altered the function or intent of the system as it existed.
All wiring, separation and redundancy of the original system design remains the same.
The Solenoid valves for the NRV actuators are Safety Class equipment.
The Plant Technical Specifications and FSAR are not affected by the implementation of this change.
o PDCR 87-017 - COMPONENT COOLING WATER SURGE TANK SEISMIC MODIFICATIONS (CCWST)
This design change added seven diagonal braces to the CCWST legs.
The diagonal braces enhance the seismic resistance of the tank and ensure that the tank legs and anchorage meet the requirements of DC-1, " Seismic Re-Evaluation and Retrofit Criteria" Revision 4.
There was no welding performed on the tank pressure boundary.
Two diagonal braces are installed in an "X"
pattern on the north, east, and went sides of the tank.
Due to the piping interferences only one diagonal brace was installed on the tank's south side.
After installation, the braces were labeled as " Seismic Installation, No Modification Without Engineering Approval" to ensure the braces remain in an "as analyzed" condition.
Prior to the change, if the CCWST were to collapse during a seismic event it could have potentially affected piping and/or equipment which is part of the Safe Shutdown System.
Specifically, the Emergency Boiler Feedwater piping and Containment Isolation System solenoid valve rack could be affected.
The tank legs and anchorages were evaluated using seismic loads based on the NRC site-specific spectrum.
The Component Cooling Water Surge Tank, and therefore the tank legs and anchorage, are classified SC-3.
This change has no impact on the normal or emergency operation of the Component Cooling system.
The modification is for structural purposes only.
This change does not affect the FSAR text or plant Technical Specification.
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o PDCR G7-018 - INSTALLATION OF STEAM GENERATOR BLOWDOWN ON-LINE CHEMISTRY MONITORING INSTRUMENTATION This-modification adds on-line chemical analysis instrumentation to the steam generator blowdown. sample lines. The local monitoring panel is' located in the upper Primary Auxiliary Building Sample Room.
Local digital readout is now available at.this panel for the convenience of the chemist.
Remote readout, alarm and recording was added in the Water Treatment plant at.the Secondary Chemistry Monitoring Panel.
The inlet sample filter currently installed on the temporary blowdown sample panel was retained and sample line and filter line purge capability was added.
In addition to the cooling provided by the steam generator blowdown sample coolers, supplemental cooling is provided by a new coil type sample cooler.
This modification was done at the request of the Chemistry Department.- This modification improves the Chemistry and Operadions Departments abilities to monitor and evaluate changes in the secondary systems water quality.
This change was done to a Non-Nuclear Safety Sample System.
The modification does not affect the text of FSAR Section 206.
The Plant Technical Specifications, including the basis for them, have been reviewed and are not affected by this modification.
o PDCR 88-002 - REMOVAL OF AS-TV-405 AS-TV-405 was removed frem the auxiliary steam system to prevent a loss of condenser vacuum after a loss of instrument air.
This valve was an air operated valve which would close upon loss of instrument air, shutting off steam flow to the air ejector.
Eliminating the possibility for a loss of condenser vacuum eliminates one of the required j
operator actions following a loss of instrument air.
The air line to the valve operator was also removed and the port capped.
l Manual isolation of this header is now provided by l
AS-V-608, upstream of the trip valve which was removed.
l The trip valve was replaced by a straight length of 3" Sch.
l 80 pipe, A106 Gr.
B material, butt welded in place.
AS-l TV-405 was a non-nuclear safety, non-seismic, non-environmentally qualified component.
The auxiliary steam and instrument air systems are not the subjects of any Technical Specifications.
The FSAR text is not affected by the implementation of this change.
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o PDCR 88-003 - MODIFICATION OF FEEDWATER HEATER & VENTING SYSTEMS To improve the efficiency of the Feedwater Heater Vent System, _the following changes were made'to the vent piping:
a)
Change'the size of the existing vent. orifices, b)
Increase the size of the vent header.
c)
Install an isolation valve on both sides of each vent orifice, and place the orifice in the by-pass piping of the vent line.
d)
Install a test tap in the vent piping.
The reason for this change was external. corrosion of the heater tube revealed as a result of an evaluation of Eddy Current tests of the #2 Feedwater Heater during the 1981, 1984, and 1987 refueling outages, and subsequent examination of a section of tube removed from the heater during the 1987 outage.
The cause of this damage has been attributed to ammonia induced corrosion.
This-type of corrosion is typically associated with gas pocketing or low flow in air removal sections of feedwater heaters.
Based on this assumed cause, a study of the feedwater heater venting system was undertaken.
This study compared the existing vent system against recommended vent system design, as detailed in EPRI Report NP-4057.
A review of this study was performed and several recommendations were made concerning the design and operation of the feedwater heater vent system.
The operational changes recommended have already been implemented and are as follows.
1)
Vent each feedwater heater individually and alternately during start-up using all available vent connections (start-up and operating vents).
2).
By-pass the vent orifices during plant utart-up.
3)
Operate the vents in a non-cascading mode with the loop seal drain line valve, VD-V-774, open.
These design and operational changes should ensure adequate venting of corrosive non-condensibles from the shell side of the feedwater heaters, thus reducing the corrosion of the heater tubes.
The Feedwater Heater Vent System is a non-nuclear safety, non-seismic, non-environmentally qualified system.
This change increased the ability of the Feedwater Heater Vent System to remove corrosive non-condensibles from the shell j
side of the feedwater heaters, thus reducing corrosion of the heater tubes.
The Feedwater Heater Vent System is not the subject of any FSAR accident analysis.
This change does not have any effect on the plant Technical Specifications; however, Section 219 of the FSAR will be modified.
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o PDCR 88-006 - ADDITION OF VENT VALVES FOR TESTING OF
- 1 FEEDWATER HIGH AND LOW LEVEL SWITCHES (FW-HLS-401 AND FW-LLS-401) AND REMOVAL OF ABANDONED IN-PLACE LEVEL CONTROLLER (FW-LC-401A)
This modification added vent valves to the FW-HLS-401 and FW-LLS-401 piping configurations located on the No.
l' Feedwater Heater.
Additionally, valves VD-V-901, -802,
-803, -804, -805, -GO6, -954, -955 and -1097 which leaked by, were replaced.
Also, an abandoned in-place level controller, FW-LC-401A, which was mounted to the existing piping by a flange connection, was removed.
Both piping I
configurations were prefabricated and welded such that the level switches are mounted closer to the feedwater heater l
shell to help eliminate excessive vibrations that might set off the switches.
This modification allows the Instrument and Control Department to perform testing on level switches FW-HLS-401.
and FW-LLS-401 and allows the switches to be flushed of sediment.
The installation of valves VD-V-801, -802,
-1097, -805, -954 and -955 allows sections of the level control piping to be isolated safely without leak-by.
The FW-LC-401A was abandoned per PA 85-15 and was replaced with a precision, high grain Moore Products relay (Model GC-67RIOO).
PA 85-15 states that the FW-LC-401A will be removed at some later time.
The #1 Feedwater Heater and its associated piping configurations which hold the high and low level alarms are classified as non-nuclear safety.
The Plant's Technical Specifications and the FSAR text are not affected by this design change.
o PDCR 88-009 - SUPPORT AND SHIELDING OF PAB DRAIN LINE TO THE GRAVITY DRAIN TANK This PDCR adds a pipe support to the PAB Sump Drain Line 2" DRL 151-16 and makes the existing lead foil shielding on the line permanent.
The pipe, shielding, and supports are NNS.
The shielding or support if they were to fall off or fail, would not affect any operating system.
Technical Specifications nor FSAR are affected by this change.
o PDCR 88-015 - RP CLEAN ROOM AND STAIR MODIFICATIONS INSIDE VAPOR CONTAINER The new Clean Room was requested by the Radiation Protection (RP) Department to allow primarily for " cold weather" implementation of protective clothing requirement.
The Clean Room, along with the stair modifications, enhanced the traffic flow within the VC and alleviated the need for going to the RP checkpoint for clothing changes.
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n; A new packaged modular Clean Room, sized.14'-O" x 10'-6" by 7'-O" high, complete with two (2) airlocks, was temporarily installed inside the Vapor Container (VC) on the circumferential platform (commonly referred to as
" Broadway") at El.
1101'-O".
The floor within the Clean Room was the existing 3/8" checkered plate flooring.
The hardsided Clean Room was constructed of lightweight clear-
"LEXAN" and stainless steel framework.
'It was installed for the Clean Room lighting, outlets, and 125 cfm ventilation unit required to provide filtered inlet air.
The total continuous current of all loads will not exceed 15 A.
A temporary frisking booth, sized 7'-O" x 3'-O" x
6'-6" high, was installed in the Clean Room.
The booth consisted of lead blankets hung from four (4) prefabricated steel racks supported on casters.
Each rack will support three (3) 90 lb lead blankets.
To accommodate the new Clean Room, modifications and additions to the existing stairs and platforms were required.
Engineering Services Group (ESG) evaluated the fire safety issue regarding the possibility of.the stainless steel /"LEXAN" Clean Room as a permanent installation.
Because the "LEXAN" (polycarbonate) material, introduced the potential for additional combustion heat gain, the Clean Room installation for this PDCR is acceptable for temporary installation in the VC during outage durations.. The temporary increase in fire loading during an outage was reviewed'and determined acceptable.
Two (2) battery-operated smoke alarms were also installed in the Clean Room.
The affected structural steel is classified as non-nuclear safety (NNS). This change has no impact on the normal operation of the plant since the structure was dismantled and removed from the VC prior to plant startup.
The Plant Technical Specifications and FSAR are not affected by the implementation of this change.
o PDCR 88-016 - VC VENTILATION COOLING SYSTEM MODIFICATIONS The following modifications were made to the vapor container cooling system and associated service water piping as follows:
- 1) Air filters were added to the inlet face of the existing VC air cooling coils.
- 2) Differential pressure gages were installed in the VC ventilation fan housing to monitor unit differential pressure.
3)The temporary Class 218 copper piping installed by plant Temporary Change Request was permanently incorporated into the design of the sersice water system.
This new piping will enable future chemical cleaning of the suction piping for SW-P-51-2.
This configuration is similar to that currently existing for SW-P-51-1.
- 4) Flanges were installed on the service water inlet and outlet nozzles for each VC air cooling coil on cooling units 1,
3 and 4.
- 5) Vent valves were installed on the outlet high points for coolers 19
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1, 3 and 4.
6)Three cooling coils for No. 3 VC air cooling unit and two cooling coils for No. 1 unit were replaced with spare coils which are identical to those installed in No. 2 ventilation unit except that the nominal tube length is 6 inches shorter in'the new coils.
This different tube length results in a reduction of approximately 6.6% in the overall heat transfer area of the replacement coil units.
- 7) Removable spool pieces were installed in the supply and i
return piping to each set of cooling coils to allow the coolers to be bypassed during periodic chemical cleaning of the associated service water piping.
The VC cooling system performs no active or passive function in shutting down the reactor or in mitigating the effects of any accidents.
These units provide for the active containment heat removal during normal plant operation.
The maximum average containment ambient temperature is limited by Technical Specification 3.6.1.5 to 120'F.
Should heat removal capability degrade sufficiently to allow ambient temperature to exceed the T.S.
limit, then the plant must be shut down and be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Addition of the air filters and differential pressure gages limit system maintenance and down time during normal operation and provide useful data on system performance.
The filters have minimal impact on system cooling capability and obviate the need to periodically clean the cooling coil surfaces.
Modification of the coil supply and return piping by addition of flanges will have no adverse impact on the pressure boundary integrity of the system.
Installation of the five new spare cooling coils in No. 1 and 3 ventilation units should improve overall current system performance.
The VC air coolers and associated service water piping are classified non-nuclear safety.
The containment ventilation system is described in FSAR Section 216 and is affected by this change.
This modification has no impact on the plant Technical Specifications.
o PDCR 89-017 - SI-LS-6 TUBING MODIFICATIONS The tubing configuration for the safety injection
(
I accumulator reference leg tell-tale alarm switch was modified to facilitate better venting of the switch housing.
The reason for this change is accumulation of water can cause erroneous accumulator level indication.
The reference leg tell tale alarm annunciated in the Control Room if water accumulates in the accumulator dry reference leg.
The old configuration did not provide for adequate switch housing venting which resulted in delayed switch actuation during periodic functional testing.
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- O The switch and associated tubing is classified Safety _
Class 2.
The functional requirements for the switch are unchanged by this PDCR.
Remote safety injection accumulator tank level indication is required by T.S.
3.5.1.b.
This modification does not affect the Technical Specifications or the FSAR text.
The safety injection accumulator is discussed in FSAR Section 212.
C.
PLANT ALTERATIONS (PA'S)
No Plant Alterations were completed this year.
D.
OPERATIONAL CHANGES 1.
NORMAL OPERATION OF THE EMERf(NCY ATMOSPHERIC STEAM DUMP (EASD) VALVES During Mode 3, hot standby "pe ation, Main Coolant System (MCS) temperature is t.ormallr..eaintained by venting steam to atmosphere utilizing MS-TV-411 when the main condenser is not available or the non-return valves are closed.
Be:ause of the availability of MS-TV-411, the EASD valves have not been needed during mode 3 operation.
If the plant is in made 3, with MS-TV-411 unavailable, EASD valves could be used in place of MS-TV-411.
The EASD System was specifically designed for throttling steam service, and to provide capacity greater than MS-TV-411.
Their use in place of MS-TV-411 is acceptable since they will operate within their design capability.
In fact, they were designed for a more severe post-accident condition.
1 The EASD System up to valves MS-MOV-659, 670, 681 and 692 are classified as Safety Class 2.
The electrical power and control circuitry for the motor operators are classified as Safety Class.
2.
BLEEDLINE VARI ORIFICE FIXED TO 50 GPM Three orifices, 20 gpm, 75 gpm and vari orifice are connected in parallel in the bleedline to set the bleed flow.
Due to a malfunction in the vari orifice, flow could no longer be varied from O-60 gpm and was fixed at 50 gpm.
With the vari orifice in a fixed position, operation was l
inconsistent with that described in FSAR section 203.
The vari orifice is not mentioned in any Accident Analysis listed in the FSAR.
Operation with the vari orifice in its fixed position does not increase the probability of an accident occurring nor an accident not previously evaluated.
During normal operation, if a reduced flow rate is wanted; the 20 gpm orifice may be utilized.
Also, the l
20 gpm orifice can be used in conjunction with the vari orifice.
1 i
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l 21 L
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'The vari orifice.is Safety Class I for the pressure retaining portion only.
Repairs were completed during the 1988 Refueling outage; operation of the vari orifice is now as described in the FSAR.
3.
USE OF A LAYDOWN PAD IN THE SPENT FUEL PIT (SFP)
The cask lift process is controlled by plant procedure OP-5000.229. _ Clearances in the laydown area at the northern end of the pool are very restrictive.
This problem is compounded by.the limited length of the building roof. hatch opening.- The primary obstructions in the pool are the upender hydraulic jack base plate and-its alignment lug, the concrete pedestal for the track support, and the support angle for the temporary gate.
During the initial attempt to set a shipping cask into the SFP on August 10, 1988, it was determined that there was insufficient clearance on the SFP floor to do so.
The preferred method to address this situation was to place a laydown pad onto the bottom of the SFP.
The pad is fabricated from structural shapes of ASTM A-36 steel.
These beams are arranged in a closely spaced grid pattern strategically positioned to avoid the various obstructions on the pool floor.
It allows use of space above the hydraulic jack base plate (but not above the piston itself) and other areas which would otherwise be unavailable.
The stress to individual beams was minimum, about 15 percent of allowable, when subjected to the full load of the 35 ton cask.
Likewise, beam deflection was very small, less than 0.03 inch.
Stresses in the stainless steel SFP liner and reinforced concrete was imperceptibly small.
The laydown pad was temporary.
It was needed only for the period in which cask lifting operations occurred.
As such, it is considered to be part of the cask lift equipment.
It did not alter any part of the permanent plant configuration, consequently use of the pad does not constitute a design change.
The cischarge and both suction lines to the SFP cooler are Safety Class 3.
The SFP liner, spent fuel racks, and all fuel handling equipment within the pool are classified Non-Nuclear Safety.
The discharge and one of the suction lines is located at the opposite end of the SFP from the location of the laydown pad and cask lift operations.
The other suction line is located in the northeast corner of the pool.
Placement of the laydown pad into the SFP is both an operational improvement and an enhancement to safety.
The safe load path for cask lift operations over the SFP was unchanged by the introduction of the laydown pad.
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.c Likewise, the lifted load and total number of lifts was the same.
There is no impact on the Yard Area Crane Operations.
The top surface of the pad provides a level set-down area which is smoother and larger than the otherwise available area on the SFP floor.
The steel beam gridwork of the laydown pad creates an energy absorbing barrier.
In the unlikely event of a load drop, the SFP floor was protected slightly more than what was credited in the drop analysis for cask CNS 3-55.
The only safety-related equipment within the vicinity of the laydown pad and cask lift operations is one of the two suction lines to the SFP cooler.
This line is located in the corner of the pool away from the lifting operations required for the laydown pad and the cask itself.
The only other equipme: t of concern is the Yard Area Crane.
The operational characteristics of this crane were evaluated under NUREG-0612 and found to be acceptable.
Its operation was unaffected by the laydown pad.
There will be no changes to any existing plant system.
No new materials, previously unanalyzed, were placed into the pool.
The cask base plate was set 20 inches above the SFP floor.
This was higher than originally planned but no operational or radiological consequences result from this 20 inch change.
The pad sub-assemblies all weigh less than the Technical Specification limit of 900 pounds.
The presence of the steel beams in the pool had no effect on the SFP cooling system or any other primary or backup system.
4.
INCREASE THE FUSE SIZE FOR THE CONTROL ROD #3 INDICATION CIRCUIT The resistance of the primary coils on the Control Rod Indicating Coil Stacks is normally 1250 + 150 ohms (DC).
The coils are arranged in three parallel groups of five coils in series, taken twice.
Each group is protected by a O.3 AMP fuse.
Deterioration of the coil wiring and/or the wiring connector has caused the resistance of the groups to decrease to 250-300 ohms (DC).
This causes enough increase in current to blow the 0.3 AMP fuses after approximately l
five minutes.
An inplant test performed on December 28, 1987 concluded that at normal coil stack operating temperature, *200 F, a 0.6 AMP fuse provided adequate protection for the primary winding of an indicating coil assembly.
The slightly higher current from Rod #3 coil stack will not adversely affect the AC supply to the other coil stacks or affect their ability to indicate Rod position.
The AC supply is protected by a 20-AMP breaker at Distribution Cabinet "A"
in the Switchgear Room.
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The safety classification for this equipment is Non Nuclear Safety.
The Plant Technical Specification and the FSAR text are not affected by the implementation of this change.
5.
FEEDWATER CONTROL SYSTEM POWER SUPPLY REPLACEMENT The need to replace a power supply in the feedwater control system was established.
The existing unit was replaced with an in-kind powe-supply, however it did not have the battery back-up capability as did the original.
Whereas with the original power supply, a loss of AC (e.g.
loss of vital bus #2) would have resulted in an automatic switching to the battery, the new configuration without the battery backup will require an immediate need to establish manual feedwater control with a higher potential of reactor scram due to steam generator level swings.
The battery back-up capability was a feature offered by the manufacturer and accepted as a " nice-to-have" feature.
The accident analyses do not credit the battery backup capability.
A loss of AC supply would result in a loss of automatic and remote manual control to the #2 and #4 feedwater control valves.
Both the loss of feedwater and loss of main coolant analyses were reviewed and are conservatively bounding.
Both the FSAR and EDCR (#83-3) have been reviewed for design considerations.
Neither documents describe this feature or prescribe a requirement for its installation.
The battery back-up capability is not required by Technical Specifications.
TESTS No tests were performed during 1988 which are reportable pursuant to 10CFR50.59.
EXPERIMENTS There were no experiments conducted during 1988.
l SAFETY AND RELIEF VALVE FAILURES AND CHALLENGES There were no challenges to the pressurizer or steam generator's
)
safety and relief valves, nor were there any failures of those safety and relief valves required to be operable by the Plant i
Technical Specifications.
The pressurizer code safety valve (PR-SV-182) setpoint was found to be 74 psig over it's nameplate setpoint of 2560 psig.
This pressure exceeds the Technical Specifications Limit allowable range by 0.008 percent.
This event was reported in Licensee Event Report 50-29/88-013.
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1
.a SPECIFIC ACTIVITY ANALYSES Plant Technical Specification 6.9.2.d requires that the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.7.
No such Limits have been exceeded for 1988.
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i YANKEE ATOMIC ELECTRIC COMPANY r.,.m n.<m m u m Star Route, Rowe, Massachusetts 01367
.Yauxes rebruerv ee, 1989
~
BYR 89-038 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555
Reference:
License No. DPR-3 (Docket No. 50-29)
Subject:
1988 Annual Report
Dear Sir:
Enclosed in the Yankee Atomic Electric Company Annual Report for 1988.
This report is submitted in accordance with 10CFR50.59 (b) and the requirements of YNPS Technical Specification 6.9.2.b and 6.9.2.d.
The repor t briefly describes (1) facility and procedure changes, tests, and experiments implemented without prior NRC approval under the provisions of 10CFR50.59, and (2) specific activity analyses in which coolant exceeded the limits of Technical Specification 3.4.7.
Also included is a summary of safety valve and relief valve failures and challenges, as required by TMI Action Plan, Item II.K.3.3.
We trust thir. information is satisfactory.
If you have any questions, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY 0
J Wtl Normand N.
St. Laurent Plant Superintendent NNS/pkg cc:
USNRC, Region I USNRC, Resident Inspector YNPS
/
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