ML20212E011
| ML20212E011 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 12/31/1987 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20212D951 | List: |
| References | |
| TASK-2.K.3.03, TASK-TM NUDOCS 8703040253 | |
| Download: ML20212E011 (13) | |
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YANKEE ATDMIC ELECTRIC COMPANY
. YANKEE NUCLEAR POWER STATION (DOCKET No. 50-29) 1986 ANNUAL REPORT OF-CHANGES,. TESTS AND EXPERIMENTS 5
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'h 1986'A'nnual Report
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TABLE OF CONTENTS TITLE. PAGE-
' Table.of Contents .. . . . . . .. . . . . .x.. . . . . . . i Introduction. . . . . . . . . ... . . . . . . . . . . . . . . . ii 1 Changes A. . Engineering' Design. Changes . . . . . . . . . . . . .. 1 B. Plant-Design Changes . . . . . . . . . . . . . . . .- . 3 C. Plant Alterations.=. . . . . . . . . . . . . . . . 4 D. Operational Changes. . . .. . .. . . . .. . . . 7 Tests . . . . . . . . .. . . . . . . . . . . . .. . . .. . . 10 Experiments . . . .. . . . . . . . . . . . . . . . . . . .- 10 Safety and Relief Valve Failures and Challenges . . . . . . 10 n
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h 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.
This. annual report is submitted in accordance with Technical Specification 6.9.2.b. . This report for changes, tests, and esperiments-is submitted in accordance with 10CFR, Chapter 1-Part SO.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).
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; ENGINEERING!DESIG_N CHANGES (EDC's) n
- EDC:76-Ol3, " Meteorological'1 Data-Collection. System Upgrade"-
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J. The plantsmeteorological data collection system has:been upgraded. .The. changes consist of-1the; installation;of a new i 200 foot ~ tower.with' wind. speed and wind _ direction sensorsDat a ~^ nominal elevation of,33 feet ~and 195 feet, temperature.and dew
.pointisensors_at a nominal elevation of'33 feet, differential'
.. temperature between nominal elevations.of.33 feet and-195 feet' ,
and solar radiation and precipitation sensors at the base of ?the-tower. The. data is collected on strip-charts andlon computer. F' The computer averages data over 15 minute intervals with local printout. . The stored' data is available for. transfer to a central computer at VNPS corporate office. e _The system.and its power supply.are classified as non-nuclear
- safety' class._ The-Technical Specifications and the-FSAR have
- ;been revised-to include this' change.
2 LEDC 78-005, " Radiation Monitoring System Upgrade" Th'e radiation monitoring system' upgrade consisted of 1) replacing _ the process radiation monitoring equipment with new equipment that is more sensitive, more accurate, and more reliable, and 2) installing a new area radiation monitoring system. 4 , The process radiation monitors that were upgraded are: 1) Steam j' Benerator. Blowdown (4-channels); 2) Waste Disposal Cover Gas Loop
' Seal;.3) Liquid Rad Waste Effluent:line; 4) Component Cooling System; 5) Main Coolant 1Bleedline; 6) Vapor Container Air Particulate;_7) Primary Vent Stack Radiation Monitor; and the 8)
- Air Ejector. In addition, radiation monitors were installed on
'- -the Steam-Generator Blowdown Tank Effluent line and on the Post Accident Hydrogen Vent line. All monitors have remote readout in the Control Room as well as local readout where appropriate.
- f. The Area Radiation Monitoring System'was installed as follows:
- 1) An Accident Emergency Gamma Guard was installed in the Turbine y _ Hall to monitor the gamma dose emanating from the Vapor Container during accident conditions; 2) A Post-Accident gamma monitor was installed in the Turbine Hall adjacent to the Emergency Gamma 1, Guard to monitor the gamma dose emanating from the Vapor l Container during post-accident conditions; 3) Two post-accident
! qamma monitors were installed in the Vapor Container; 4) A high range noble gas monitor was installed on the_ Primary Vent Stack effluent sample line; 5) Four gamma monitors (one for each steam 4 line) were 1
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-g-1 installed'on th'e Steam Generator steam lines upstream of_-the non- 'l return valves / safety relief valves; 6) Twenty-four monitors ~were-
' installed in:areasEthroughout the plant where high radiation areas could be encountered.
' All monitors have remote readout in the Control Room as well as local' readout where appropriate.
The plant Technical Specifications and the FSAR have been revised to incorporate this' change. 1 EDC 84-326, " Installation Of A Iodine Sampler On The Primary Vent Stack-Effluent Monitoring System" A " grab" sample. system h1 as been installed on the existing Primary Vent Stack (PVS) effluent monitoring-system. It provides for the q collection of PVS effluent samples under post-accident ; conditions. It consists of a. locally operated sample pump, flow indicator,zremovable sample cartridge andEvalving. The previously installed system (a gas stripping bubbler system) installed:under-PDC 01-12 was left in place. The equipment'and--piping modified by this change and the new sample system are classified as non-nuclear safety. The Technical _ Specifications were-not affected by'this change. The-FSAR has been revised to incorporate this change. a I i i 4
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PLANT DESIGN' CHANGES.'(PDC's) PDC 85-00'4, " Extension Of..'The Fire Protection System To The Garage" - Thez" dry" sprinkler fire protection system was expanded'to include the_ vehicle garage which is now.used'as a. storage area-
.for lubricating. oil. The change provides fire protection for::the-area. (The change,does not affect the' Technical _ Specifications.
The FSAR will be revised to incorporate the change. s . 5 i t' 4 6 4 6 -
.s _ h _4_ g Plant Alterations (PA's) PA 84-021, " Replacement Of The Valve Room Door' With' A Lockable Wire Menh Door" The " solid"~ ' door at-the enteance to the Valve Room has been l replaced with.a wire mesh door. This change allows for natural
. circulation of air through_-the valve room providing heat removal-capability.from the valve room, such that the ambient temperature limitation of Vapor Container. isolation solenoids in the room
~will.notiexceed designitemperature during-operation of the Emergency, Core Cooling System.
The door.is a non-nuclear safety class. structure and is not'a fire barrier. The change maintains access exclusion required for radiation protection while. allowing air circulation for heat. removal requirements. It does not affect the plant Technical Specification or'the-FSAR. PA 05-008, " Emergency Diesel Generator Fuel Oil Tank Overflow
- ' Tank Isolation-Valve Addition" The addition of.the isolation valve (FO-V-77) between-Lhe tank 4
and:the-level alarm switch; permits the switch to be isolated from the tank for operability testing and maintenance. 1 The tank and its associated piping and instrumentation are
. classified as.non-nuclear safety class. The Technical Specifications and the FSAR were not affected by this change.
HPA 86-008," Addition Of'A Television Camera To The Security Closed Circuit Television Surveillance System" r; An additional camera was installed in the security closed circuit television surveillance system to provide increased coverage of the plant security area. This change has been incorporated in the plant Security Plan which is referenced in the FSAR. The change does not affect the Technical Specifications.
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s PA 06-009, " Addition Of Cooling Fans To Process Radiation Monitoring Cabinets A and B" Forced draft cooling fans have been added to process radiation monitoring. cabinets "A" and-"B". These fans provide required ,. ' cooling capability.to allow the equipment.in the cabinets to operate within design specification temperature limits. The equipment in the cabinets and the power supply sources.for-the fans are classified as non-nuclear safety. The change does not; affect the Technical Specifications or the FSAR. 9 PA 85-017,." Modification Of The Security Fence-Inertia Guard '
. Alarm System" - +
+ - The security fence inertia guard alarm system has been modified to provide redundant alarm reset capability. This change has-been incorporated in the plant Security Plan-which is referenced in the FSAR. ' Tine details of the change are considered part of physical protection for the plant in accordarce with 10.CFR Part
. . 73. Details are available to authorized persannel. >
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<. l PA B5-020, " Chemistry Laboratory Floor And Work Benches upgrada" ,
The " Primary" and " Secondary" chemistry laboratories were upgraded with the replacement of existing work benches'.11th modern benches. The linoleum floor covering was removed and replaced with a " poured" epoxy floor. The Technical Specifications were not affected by this change. The laboratory floor plan shown in the FSAR will be-revised to incorporate the change. . PA 85-026, " Guardhouse Upgrade" Th'e security quardhouse was modified to conform to the plant Security Plan. Details of the change are considered part of the physical protection of the plant in accordance with 10 CFR Part
- 73. Details of the change are available to authorized personnel.
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j PA.86-OO6, " Wire Mesh _ Partition:In The Primary Auxiliary Building-Cubicle Area"
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* ,. . A Tpe addition of this wire, mesh- partition and door provide control-
, .t o the high radiation exclusion area created by various primary
' coolant auxiliary equipment located in. cubicles along the corridor- of . the :Primar y Auxiliary Building. .This local partition al,lo,wsl ^the main' access door to the corridor to remain unlocked,-
th l allowing quick, easy access to all areas except the high ( radiation exclusion area. l-
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The Technical Specifications and:the FSAR were not affected by
-the change.
PA 86-007, " Fire Protection System Pressure-Maintenance Tank'
' Instrument Column Flush Connections" A supply line'and drain line were added to the plant fire.
. protection. system pressure maintenance tank instrument column _to facilitate-flushing of the column. .The piping is classified'as non-nuclear-safety class. The change did not affect the Technical Specifications or the FSAR.
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y bd s PLANT-OPERATION CHANGES - e.A ,-
- 1. -Lithium: Hydroxide Addition To The Main Coolant l System Via The Corrosion Control System.
Lithium,.in the form of Li.G4, was used.for Main Coolant water chemistry'pH control. The Li+ was'added to the main
~ coolant-using~the; corrosion control. system-feed.to the Charging-Pump; suction header.- The corrosion. control system is comprised of a 32 gallon capacity batch / feed tank of 304 stainless. steel and a transfer'line to the Charging Pump suction header of-316 stainless steel. .The concentration of Li'Gi/HeO in the batch / feed tank was approximately 0.5 molar.
The. potential for corrosion of the tank and piping as a result of contact with this solution (pH>12.0)'is considered minimal. The: tank, the transfer pump and piping were designed and installed to accommodate solutions of this nature. (Previously the system was.used to transfer 60,000 ppm ammonium hydroxide solution'to the main coolant system).
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Any leakage fromnthe system would be contained by the plant
; floor drain' system.
- 2. Heater Drain Tank Level Transmitter Isolation
' ~ The Heater Drain Tank Level Transmitter was isolated to find and correct the cause of erratic operation. -The' transmitter provides signals to the circuitry for pump start /stop and high/ low level alarms. The purpose of this circuitry is to protect the heater. drain pump on low suction and is anticipatory to a boiler feed pump trip on low suction pressure due to heater drain pump cavitation. The maintenance activity disabled only the auto trip feature for r these pumps, leaving the manual trip capability intact and operable. No other protective circuitry is affected. The
-trip function of the level transmitter and its associated L circuitry is non-nuclear safety; however, it is mentioned in the FSAR.
1 During the period that the transmitter was isolated, tuo auxiliary operators were stationed locally; one to monitor tank level and trip the pump if necessary, the other to control tank level. The Control Room operator also has the capability to trip the pump manually. g L The effect of removing the auto trip feature is not significant. If a trip did not occur when required, the l consequences would mean damage to the heater drain pump and ! ultimately result in a plant trip via other protective
- circuitry (i.e. Iow suction pressure to boiler feed pump I would trip the BFP which in turn would Lr ip the plant un low L SG 1evels).
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= - [, af : w - ,, The accident analysis assumptions'and FSAR Section 409, " Main;
' Steam Break"-_were reviewedLto; determine any potential.
consideration. In both cases,;a-heater drain pump trip.on
- low' tank level is not. assumed in the analysis.
~3. . Main Coolant Loop Bypass Valve Operation in Modes'I and 2.
In order _to; reduce Main Coolant System valve stem leak-off: flow,Jandthence stem leak-off temperature, it may become-necessary to back-seat individual Main Coolant' Loop-Bypass: , line valves. If so, the> resultant _ flow-of coolant through. the bypass line/s would cause an equivalent. loss of flow-
.through the reactor core. . Plant operation.in this'
-configuration was evaluated to determine if: 1) it is counter to, or requires a change to, the Technical 1 Specifications, or 2) it creates an unreviewed safety question.
- 1) ThefTechnical Specifications require a minimum flow rate oft 38.3E6 lb m/hr.- Calculations using approved plant procedures show the minimum flow rate with Number 1-Loop bypass' valve open to be 40.54E6 lb m/hr.- The minimum flow ~
rate with Number'4 Loop bypass. valve.open.is; calculated to be 40.40E6 lb m/hr. These calculations use data-from Secondary plant calorimetric-measurements.
- 2) Accident Analysis:
A) Break Analysis: In the large break scenario, the open bypass valve will not significantly affect refill by-the ECCS via the cold legs. This is due primarily to the fact that loop pressures are equal and that the relative' flow area of the cold-leg to the bypass line is roughly 10 to
- 1. Thus, ECC flow should be preferential towards the vessel. During the reflood phase, the open bypass valve
. enhances core reflood for cold leg breaks since steam venting is improved. For hot leg breaks, the open valve would have little, if any, effect on system response.
, For small breaks which remain above the procedural limit j for terminating RCP operation, 1200 psig, the open bypass l valve has no real impact on the event. In small breaks where the RCP's are tripped according to procedure, flow in the open bypass line would tend to stagnate since the line risps in olpvatinn ahnvp thn main ennlant nininn hv about six feet or so. Since the fluid in the bypass line {~ would be at or near Two., the bypass line behavior will be p similar to that of the upper head region. If ECC flow to b Loop 2 were to bypass the vessel via the open line, it
- would be inconsequential since the distribution of ECC in small breaks is not nearly as important as its make up capability.
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B) Core Therma'l Design Margin: An energy balance was developed for each loop using calorimetric data-obtained from Loop I with that loop bypass valve open.
~The energy balance for. Loop 1 tookLinto account the mix of loop bypass flow at To and vessel flow at.Ts that enters No'. I Steam Generator. The results'of the calculation show that, with Loop 1 bypass valve open, the' core-inlet mass flux is not.less than 2.509E6 lb m/hr - ft". This is at least.9% greater than the: Core XV design value of 2.295E6 lb m/hr. - ft" determined by COBRA-III-C calculations. With-two loop bypasses open the calculated core-inlet mass flux is 2.466E6 lb m/hr - ft", foc a design margin of 7%.
In summary, no ler.Snical Specifications applicable to core thermal design limits are violated when one or two loop bypass valves are opened at full power conditions.- Neither the consequences of LOCA-events, _ nor the thermal design margins for non-LOCA transients are affected in significantly adverse ways, when operating in this manner.
- 4. Plant Operation With The Reactor Source Range Channel 2-Detector High Voltage Power Supply Deenergized Due to the potential for loss of the reactor source range channel 2 neutron detector by automatic electrical energization from intermediate range channel 4, the source range channel 2 manual high voltage power supply switch was placed in the off position. The potential for loss was present because intermediate range channel 4 was reading low and drifting lower. It would automatically energize the source range chanr.el 2 at IE-9 amps.
As only one source range channel is required for plant shutdown (channel one was available) and because channel 2 could have been switched on manually when needed, the Technical Specifications were not violated. t
TESTS ~-
-f*3 tests were performed during'1986 which are-reportable pursuant' to 10 CFR 50.59.
EXPERIMENTS-
-There were no experiments conducted'during'1986.
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 rettet valves required to be operable by Technical Specifications.
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